Thermionic switched self-actuating reactor shutdown system

DOEpatents

Barrus, Donald M.; Shires, Charles D.; Brummond, William A.

1989-01-01

A self-actuating reactor shutdown system incorporating a thermionic switched electromagnetic latch arrangement which is responsive to reactor neutron flux changes and to reactor coolant temperature changes. The system is self-actuating in that the sensing thermionic device acts directly to release (scram) the control rod (absorber) without reference or signal from the main reactor plant protective and control systems. To be responsive to both temperature and neutron flux effects, two detectors are used, one responsive to reactor coolant temperatures, and the other responsive to reactor neutron flux increase. The detectors are incorporated into a thermionic diode connected electrically with an electromagnetic mechanism which under normal reactor operating conditions holds the the control rod in its ready position (exterior of the reactor core). Upon reaching either a specified temperature or neutron flux, the thermionic diode functions to short-circuit the electromagnetic mechanism causing same to lose its holding power and release the control rod, which drops into the reactor core region under gravitational force.

Thermionic reactors for space nuclear power

NASA Technical Reports Server (NTRS)

Homeyer, W. G.; Merrill, M. H.; Holland, J. W.; Fisher, C. R.; Allen, D. T.

1985-01-01

Thermionic reactor designs for a variety of space power applications spanning the range from 5 kWe to 3 MWe are described. In all of these reactors, nuclear heat is converted directly to electrical energy in thermionic fuel elements (TFEs). A circulating reactor coolant carries heat from the core of TFEs directly to a heat rejection radiator system. The recent design of a thermionic reactor to meet the SP-100 requirements is emphasized. Design studies of reactors at other power levels show that the same TFE can be used over a broad range in power, and that design modifications can extend the range to many megawatts. The design of the SP-100 TFE is similar to that of TFEs operated successfully in test reactors, but with design improvements to extend the operating lifetime to seven years.

Spaceborne power systems preference analyses. Volume 2: Decision analysis

NASA Technical Reports Server (NTRS)

Smith, J. H.; Feinberg, A.; Miles, R. F., Jr.

1985-01-01

Sixteen alternative spaceborne nuclear power system concepts were ranked using multiattribute decision analysis. The purpose of the ranking was to identify promising concepts for further technology development and the issues associated with such development. Four groups were interviewed to obtain preference. The four groups were: safety, systems definition and design, technology assessment, and mission analysis. The highest ranked systems were the heat-pipe thermoelectric systems, heat-pipe Stirling, in-core thermionic, and liquid-metal thermoelectric systems. The next group contained the liquid-metal Stirling, heat-pipe Alkali Metal Thermoelectric Converter (AMTEC), heat-pipe Brayton, liquid-metal out-of-core thermionic, and heat-pipe Rankine systems. The least preferred systems were the liquid-metal AMTEC, heat-pipe thermophotovoltaic, liquid-metal Brayton and Rankine, and gas-cooled Brayton. The three nonheat-pipe technologies selected matched the top three nonheat-pipe systems ranked by this study.

Thermionic converter temperature controller

DOEpatents

Shaner, Benjamin J [McMurray, PA; Wolf, Joseph H [Pittsburgh, PA; Johnson, Robert G. R. [Trafford, PA

2001-04-24

A method and apparatus for controlling the temperature of a thermionic reactor over a wide range of operating power, including a thermionic reactor having a plurality of integral cesium reservoirs, a honeycomb material disposed about the reactor which has a plurality of separated cavities, a solid sheath disposed about the honeycomb material and having an opening therein communicating with the honeycomb material and cavities thereof, and a shell disposed about the sheath for creating a coolant annulus therewith so that the coolant in the annulus may fill the cavities and permit nucleate boiling during the operation of the reactor.

Preliminary plan for testing a thermionic reactor in the Plum Brook Space Power Facility

NASA Technical Reports Server (NTRS)

Haley, F. A.

1972-01-01

A preliminary plan is presented for testing a thermionic reactor in the Plum Brook Space Power Facility (SPF). A technical approach, cost estimate, manpower estimate, and schedule are presented to cover a 2 year full power reactor test.

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.

Analysis of closed cycle megawatt class space power systems with nuclear reactor heat sources

NASA Technical Reports Server (NTRS)

Juhasz, A. J.; Jones, B. I.

1987-01-01

The analysis and integration studies of multimegawatt nuclear power conversion systems for potential SDI applications is presented. A study is summarized which considered 3 separate types of power conversion systems for steady state power generation with a duty requirement of 1 yr at full power. The systems considered are based on the following conversion cycles: direct and indirect Brayton gas turbine, direct and indirect liquid metal Rankine, and in core thermionic. A complete mass analysis was performed for each system at power levels ranging from 1 to 25 MWe for both heat pipe and liquid droplet radiator options. In the modeling of common subsystems, reactor and shield calculations were based on multiparameter correlation and an in-house analysis for the heat rejection and other subsystems.

High-temperature, high-power-density thermionic energy conversion for space

NASA Technical Reports Server (NTRS)

Morris, J. F.

1977-01-01

Theoretic converter outputs and efficiencies indicate the need to consider thermionic energy conversion (TEC) with greater power densities and higher temperatures within reasonable limits for space missions. Converter-output power density, voltage, and efficiency as functions of current density were determined for 1400-to-2000 K emitters with 725-to-1000 K collectors. The results encourage utilization of TEC with hotter-than-1650 K emitters and greater-than-6W sq cm outputs to attain better efficiencies, greater voltages, and higher waste-heat-rejection temperatures for multihundred-kilowatt space-power applications. For example, 1800 K, 30 A sq cm TEC operation for NEP compared with the 1650 K, 5 A/sq cm case should allow much lower radiation weights, substantially fewer and/or smaller emitter heat pipes, significantly reduced reactor and shield-related weights, many fewer converters and associated current-collecting bus bars, less power conditioning, and lower transmission losses. Integration of these effects should yield considerably reduced NEP specific weights.

Thermionic reactor power conditioner design for nuclear electric propulsion.

NASA Technical Reports Server (NTRS)

Jacobsen, A. S.; Tasca, D. M.

1971-01-01

Consideration of the effects of various thermionic reactor parameters and requirements upon spacecraft power conditioning design. A basic spacecraft is defined using nuclear electric propulsion, requiring approximately 120 kWe. The interrelationships of reactor operating characteristics and power conditioning requirements are discussed and evaluated, and the effects on power conditioner design and performance are presented.

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

NASA Astrophysics Data System (ADS)

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

2001-02-01

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

Venting of fission products and shielding in thermionic nuclear reactor systems

NASA Technical Reports Server (NTRS)

Salmi, E. W.

1972-01-01

Most thermionic reactors are designed to allow the fission gases to escape out of the emitter. A scheme to allow the fission gases to escape is proposed. Because of the low activity of the fission products, this method should pose no radiation hazards.

Mathematical Modeling Of A Nuclear/Thermionic Power Source

NASA Technical Reports Server (NTRS)

Vandersande, Jan W.; Ewell, Richard C.

1992-01-01

Report discusses mathematical modeling to predict performance and lifetime of spacecraft power source that is integrated combination of nuclear-fission reactor and thermionic converters. Details of nuclear reaction, thermal conditions in core, and thermionic performance combined with model of swelling of fuel.

Developmental status of thermionic materials.

NASA Technical Reports Server (NTRS)

Yang, L.; Chin, J.

1972-01-01

Description of the reference materials selected for the major components of the unit cell of a thermionic pile element (TFE), the out-of-pile and in-pile test results, and current efforts for improving the life and performance of thermionic fuel elements. The component materials are required to withstand the fuel burnup and fast neutron fluence dictated by the thermionic reactor system. Tungsten was selected as the cladding material because of its compatibility with both the carbide and the oxide fuel materials. Niobium was selected as the collector material because its thermal expansion coefficient matches closely with that of the thin aluminum oxide layer used to electrically insulate the collector from the TFE sheath. An unfueled converter has performed stably over 41,000 hr. Accelerated irradiation tests have attained burnups equivalent to that for 40,000 hr of the thermionic reactor under consideration.

High power densities from high-temperature material interactions. [in thermionic energy conversion and metallic fluid heat pipes

NASA Technical Reports Server (NTRS)

Morris, J. F.

1981-01-01

Thermionic energy conversion (TEC) and metallic-fluid heat pipes (MFHPs), offering unique advantages in terrestrial and space energy processing by virtue of operating on working-fluid vaporization/condensation cycles that accept great thermal power densities at high temperatures, share complex materials problems. Simplified equations are presented that verify and solve such problems, suggesting the possibility of cost-effective applications in the near term for TEC and MFHP devices. Among the problems discussed are: the limitation of alkali-metal corrosion, protection against hot external gases, external and internal vaporization, interfacial reactions and diffusion, expansion coefficient matching, and creep deformation.

Experimental detailed power distribution in a fast spectrum thermionic reactor fuel element at the core/BeO reflector interface region

NASA Technical Reports Server (NTRS)

Klann, P. G.; Lantz, E.

1973-01-01

A zero-power critical assembly was designed, constructed, and operated for the prupose of conducting a series of benchmark experiments dealing with the physics characteristics of a UN-fueled, Li-7-cooled, Mo-reflected, drum-controlled compact fast reactor for use with a space-power conversion system. The critical assembly was modified to simulate a fast spectrum advanced thermionics reactor by: (1) using BeO as a reflector in place of some of the existing molybdenum, (2) substituting Nb-1Zr tubing for some of the existing Ta tubing, and (3) inserting four full-scale mockups of thermionic type fuel elements near the core and BeO reflector boundary. These mockups were surrounded with a buffer zone having the equivalent thermionic core composition. In addition to measuring the critical mass of this thermionic configuration, a detailed power distribution in one of the thermionic element stages in the mixed spectrum region was measured. A power peak to average ratio of two was observed for this fuel stage at the midplane of the core and adjacent to the reflector. Also, the power on the outer surface adjacent to the BeO was slightly more than a factor of two larger than the power on the inside surface of a 5.08 cm (2.0 in.) high annular fuel segment with a 2.52 cm (0.993 in. ) o.d. and a 1.86 cm (0.731 in.) i.d.

High power densities from high-temperature materials interactions. [thermionic energy conversion and metallic fluid heat pipes

NASA Technical Reports Server (NTRS)

Morris, J. F.

1981-01-01

Thermionic energy converters and metallic-fluid heat pipes are well suited to serve together synergistically. The two operating cycles appear as simple and isolated as their material problems seem forebodingly deceptive and complicated. Simplified equations verify material properties and interactions as primary influences on the operational effectiveness of both. Each experiences flow limitations in thermal emission and vaporization because of temperature restrictions redounding from thermophysicochemical stability considerations. Topics discussed include: (1) successful limitation of alkali-metal corrosion; (2) protection against external hot corrosive gases; (3) coping with external and internal vaporization; (4) controlling interfacial reactions and diffusion; and (5) meeting other thermophysical challenges; expansion matches and creep.

Thermionic energy conversion technology - Present and future

NASA Technical Reports Server (NTRS)

Shimada, K.; Morris, J. F.

1977-01-01

Aerospace and terrestrial applications of thermionic direct energy conversion and advances in direct energy conversion (DEC) technology are surveyed. Electrode materials, the cesium plasma drop (the difference between the barrier index and the collector work function), DEC voltage/current characteristics, conversion efficiency, and operating temperatures are discussed. Attention is centered on nuclear reactor system thermionic DEC devices, for in-core or out-of-core operation. Thermionic fuel elements, the radiation shield, power conditions, and a waste heat rejection system are considered among the thermionic DEC system components. Terrestrial applications include topping power systems in fossil fuel and solar power generation.

Radiatively coupled thermionic and thermoelectric power system concept

NASA Technical Reports Server (NTRS)

Shimada, K.; Ewell, R.

1981-01-01

The study presented showed that the large power systems (about 100 kW) utilizing radiatively coupled thermionic or thermoelectric converters could be designed so that the power subsystem could be contained in a Space Shuttle bay as a part of an electrically propelled spacecraft. The radiatively coupled system requires a large number of individual converters since the transferred heat is smaller than with the conductively coupled system, but the advantages of the new system indicates merit for further study. The advantages are (1) good electrical isolation between converters and the heat source, (2) physical separation of converters from the heat source (making the system fabrication manageable), and (3) elimination of radiator heat pipes, which are required in an all-heat-pipe power system. In addition, the specific weight of the radiatively coupled power systems favorably compares with that of the all-heat-pipe systems.

Multiphoton laser ionization for energy conversion in barium vapor

NASA Astrophysics Data System (ADS)

Makdisi, Y.; Kokaj, J.; Afrousheh, K.; Mathew, J.; Nair, R.; Pichler, G.

2013-03-01

We have studied the ion detection of barium atoms in special heated ovens with a tungsten rod in the middle of the stainless steel tube. The tungsten rod was heated indirectly by the oven body heaters. A bias voltage between the cell body and the tungsten rod of 9 V was used to collect electrons, after the barium ions had been created. However, we could collect the electrons even without the bias voltage, although with ten times less efficiency. We studied the conditions for the successful bias-less thermionic signal detection using excimer/dye laser two-photon excitation of Rydberg states below and above the first ionization limit (two-photon wavelength at 475.79 nm). We employed a hot-pipe oven and heat-pipe oven (with inserted mesh) in order to generate different barium vapor distributions inside the oven. The thermionic signal increased by a factor of two under heat-pipe oven conditions.

NUCLEAR REACTOR AND THERMIONIC FUEL ELEMENT THEREFOR

DOEpatents

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

1963-12-01

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

High efficiency thermionic converter studies

NASA Technical Reports Server (NTRS)

Huffman, F. N.; Sommer, A. H.; Balestra, C. L.; Briere, T. R.; Lieb, D.; Oettinger, P. E.; Goodale, D. B.

1977-01-01

Research in thermionic energy conversion technology is reported. The objectives were to produce converters suitable for use in out of core space reactors, radioisotope generators, and solar satellites. The development of emitter electrodes that operate at low cesium pressure, stable low work function collector electrodes, and more efficient means of space charge neutralization were investigated to improve thermionic converter performance. Potential improvements in collector properties were noted with evaporated thin film barium oxide coatings. Experiments with cesium carbonate suggest this substance may provide optimum combinations of cesium and oxygen for thermionic conversion.

Fuel elements of thermionic converters

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

Hunter, R.L.; Gontar, A.S.; Nelidov, M.V.

1997-01-01

Work on thermionic nuclear power systems has been performed in Russia within the framework of the TOPAZ reactor program since the early 1960s. In the TOPAZ in-core thermionic convertor reactor design, the fuel element`s cladding is also the thermionic convertor`s emitter. Deformation of the emitter can lead to short-circuiting and is the primary cause of premature TRC failure. Such deformation can be the result of fuel swelling, thermocycling, or increased unilateral pressure on the emitter due to the release of gaseous fission products. Much of the work on TRCs has concentrated on preventing or mitigating emitter deformation by improving themore » following materials and structures: nuclear fuel; emitter materials; electrical insulators; moderator and reflector materials; and gas-exhaust device. In addition, considerable effort has been directed toward the development of experimental techniques that accurately mimic operational conditions and toward the creation of analytical and numerical models that allow operational conditions and behavior to be predicted without the expense and time demands of in-pile tests. New and modified materials and structures for the cores of thermionic NPSs and new fabrication processes for the materials have ensured the possibility of creating thermionic NPSs for a wide range of powers, from tens to several hundreds of kilowatts, with life spans of 5 to 10 years.« less

Calculated power distribution of a thermionic, beryllium oxide reflected, fast-spectrum reactor

NASA Technical Reports Server (NTRS)

Mayo, W.; Lantz, E.

1973-01-01

A procedure is developed and used to calculate the detailed power distribution in the fuel elements next to a beryllium oxide reflector of a fast-spectrum, thermionic reactor. The results of the calculations show that, although the average power density in these outer fuel elements is not far from the core average, the power density at the very edge of the fuel closest to the beryllium oxide is about 1.8 times the core avearge.

An out-of-core thermionic-converter system for nuclear space power

NASA Technical Reports Server (NTRS)

Breitwieser, R.

1972-01-01

Design of the nuclear thermionic space power system, 40 50 70 Kw(e) power range, are given. The design configuration (1) meets the constraints of readily available launch vehicles; (2) allows for off-design operation including startup, shutdown, and possible emergency conditions; (3) provides tolerance of failure by extensive use of modular, redundant elements; (4) incorporates and uses heat pipes in a fashion that reduces the need for extensive in-pile testing of system components; and (5) uses thermionic converters, nuclear fuel elements, and heat transfer devices in a geometrical form adapted from existing incore thermionic system designs. Designs and in some cases performance data for elements and groups of the elements of the system are included. Benefits of the highly modular system approach to reliability, safety, economy of development, and flexibility are discussed.

Design of an external-fueled thermionic diode for in-pile testing.

NASA Technical Reports Server (NTRS)

Ernst, D. M.; Peelgren, M. L.

1971-01-01

Description of an external-fueled thermionic diode suitable for in-pile testing in a research reactor. The active electrode area is 94 sq cm. The 10-in. long, 1.5-in.-OD emitter body is tungsten 2% thoria. The fuel is contained in six 0.4-in.-diam holes equally spaced about the 0.5-in. central emitter hole. The collector is niobium-1% zirconium. The expected diode performance is 6 W/sq cm at 2000 K. In addition to following the constraints imposed by the in-pile testing and the electrically heated performance mapping prior to insertion in-pile, the diode will have end configurations prototypical of those anticipated for a flow-through, NaK-cooled, external-fuel thermionic reactor.

A Heat Pipe Coupled Planar Thermionic Converter: Performance Characterization, Nondestructive Testing, and Evaluation.

DTIC Science & Technology

1992-03-15

Pipes, Computer Modelling, Nondestructive Testing. Tomography , Planar Converter, Cesium Reservoir 19. ABSTRACT (Continue on reverse if necessary and...Investigation ........................ 32 4.3 Computed Tomography ................................ 33 4.4 X-Ray Radiography...25 3.4 LEOS generated output data for Mo-Re converter ................ 26 4.1 Distance along converter imaged by the computed tomography

IECEC '83; Proceedings of the Eighteenth Intersociety Energy Conversion Engineering Conference, Orlando, FL, August 21-26, 1983. Volume 1 - Thermal energy systems

NASA Astrophysics Data System (ADS)

Among the topics discussed are the nuclear fuel cycle, advanced nuclear reactor designs, developments in central status power reactors, space nuclear reactors, magnetohydrodynamic devices, thermionic devices, thermoelectric devices, geothermal systems, solar thermal energy conversion systems, ocean thermal energy conversion (OTEC) developments, and advanced energy conversion concepts. Among the specific questions covered under these topic headings are a design concept for an advanced light water breeder reactor, energy conversion in MW-sized space power systems, directionally solidified cermet electrodes for thermionic energy converters, boron-based high temperature thermoelectric materials, geothermal energy commercialization, solar Stirling cycle power conversion, and OTEC production of methanol. For individual items see A84-30027 to A84-30055

Design, fabrication, and testing of an external fuel (UO2), full-length thermionic converter

NASA Technical Reports Server (NTRS)

Schock, A.; Raab, B.

1971-01-01

The development of a full-length external-fuel thermionic converter for in-pile testing is described. The development program includes out-of-pile performance testing of the fully fueled-converter, using RF-induction heating, before its installation in the in-pile test capsule. The external-fuel converter is cylindrical in shape, and consists of an inner, centrally cooled collector, and an outer emitter surrounded by nuclear fuel. The term full-length denotes that the converter is long enough to extend over the full height of the reactor core. Thus, the converter is not a scaled-down test device, but a full-scale fuel element of the thermionic reactor. The external-fuel converter concept permits a number of different design options, particularly with respect to the fuel composition and shape, and the collector cooling arrangement. The converter described was developed for the Jet Propulsion Laboratory, and is based on their concept for a thermionic reactor with uninsulated collector cooling as previously described. The converter is double-ended, with through-flow cooling, and with ceramic seals and emitter and collector power take-offs at both ends. The design uses a revolver-shaped tungsten emitter body, with the central emitter hole surrounded by six peripheral fuel holes loaded with cylindrical UO2 pellets.

A pilot level decision analysis of thermionic reactor development strategy for nuclear electric propulsion

NASA Technical Reports Server (NTRS)

Menke, M. M.; Judd, B. R.

1973-01-01

The development policy for thermionic reactors to provide electric propulsion and power for space exploration was analyzed to develop a logical procedure for selecting development alternatives that reflect the technical feasibility, JPL/NASA project objectives, and the economic environment of the project. The partial evolution of a decision model from the underlying philosophy of decision analysis to a deterministic pilot phase is presented, and the general manner in which this decision model can be employed to examine propulsion development alternatives is illustrated.

Thermionic fuel element for the S-prime reactor

NASA Astrophysics Data System (ADS)

Van Hagan, Thomas H.; Drees, Elizabeth A.

1993-01-01

Technical aspects of the thermionic fuel element (TFE) design proposed for the S-PRIME space nuclear power system are discussed. Topics covered include the rational for selecting a multicell TFE approach, a technical description of the S-PRIME TFE and its estimated performance, and the technology readiness of the design, which emphasizes techology maturity and low risk.

Nuclear thermionic power plants in the 50-300 kWe range.

NASA Technical Reports Server (NTRS)

Van Hoomissen, J. E.; Sawyer, C. D.; Prickett, W. Z.

1972-01-01

This paper reviews the results of recent studies performed by General Electric on in-core thermionic reactor power plants in the 50-300 kWe range. In particular, a 100 kWe manned Space Base mission and a 240 kWe unmanned electric propulsion mission are singled out as representative design points for this concept.

Design and optimization of the heat rejection system for a liquid cooled thermionic space nuclear reactor power system

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

Moriarty, M.P.

1993-01-15

The heat transport subsystem for a liquid metal cooled thermionic space nuclear power system was modelled using algorithms developed in support of previous nuclear power system study programs, which date back to the SNAP-10A flight system. The model was used to define the optimum dimensions of the various components in the heat transport subsystem subjected to the constraints of minimizing mass and achieving a launchable package that did not require radiator deployment. The resulting design provides for the safe and reliable cooling of the nuclear reactor in a proven lightweight design.

Design and optimization of the heat rejection system for a liquid cooled thermionic space nuclear reactor power system

NASA Astrophysics Data System (ADS)

Moriarty, Michael P.

1993-01-01

The heat transport subsystem for a liquid metal cooled thermionic space nuclear power system was modelled using algorithms developed in support of previous nuclear power system study programs, which date back to the SNAP-10A flight system. The model was used to define the optimum dimensions of the various components in the heat transport subsystem subjected to the constraints of minimizing mass and achieving a launchable package that did not require radiator deployment. The resulting design provides for the safe and reliable cooling of the nuclear reactor in a proven lightweight design.

Thermionic reactor ion propulsion system /TRIPS/ - Its multi-mission capability.

NASA Technical Reports Server (NTRS)

Peelgren, M. L.

1972-01-01

The unmanned planetary exploration to be conducted in the last two decades of this century includes many higher energy missions which tax all presently available propulsion systems beyond their limit. One candidate with the versatility and performance to meet these mission objectives is nuclear electric propulsion (NEP). Additionally, the NEP System is feasible in orbit raising operations with the Shuttle or Shuttle/Tug combination. A representative planetary mission is described (Uranus-Neptune flyby with probe), and geocentric performance and tradeoffs are discussed. The NEP System is described in more detail with particular emphasis on the power subsystem consisting of the thermionic reactor, heat rejection subsystem, and neutron shield.

Pulsed thermionic converter study

NASA Technical Reports Server (NTRS)

1976-01-01

A nuclear electric propulsion concept using a thermionic reactor inductively coupled to a magnetoplasmadynamic accelerator (MPD arc jet) is described, and the results of preliminary analyses are presented. In this system, the MPD thruster operates intermittently at higher voltages and power levels than the thermionic generating unit. A typical thrust pulse from the MPD arc jet is characterized by power levels of 1 to 4 MWe, a duration of 1 msec, and a duty cycle of approximately 20%. The thermionic generating unit operates continuously but with a lower power level of approximately 0.4 MWe. Energy storage between thrust pulses is provided by building up a large current in an inductor using the output of the thermionic converter array. Periodically, the charging current is interrupted, and the energy stored in the magnetic field of the inductor is utilized for a short duration thrust pulse. The results of the preliminary analysis show that a coupling effectiveness of approximately 85 to 90% is feasible for a nominal 400 KWe system with an inductive unit suitable for a flight vehicle.

Reactor pressure vessel head vents and methods of using the same

DOEpatents

Gels, John L; Keck, David J; Deaver, Gerald A

2014-10-28

Internal head vents are usable in nuclear reactors and include piping inside of the reactor pressure vessel with a vent in the reactor upper head. Piping extends downward from the upper head and passes outside of the reactor to permit the gas to escape or be forcibly vented outside of the reactor without external piping on the upper head. The piping may include upper and lowers section that removably mate where the upper head joins to the reactor pressure vessel. The removable mating may include a compressible bellows and corresponding funnel. The piping is fabricated of nuclear-reactor-safe materials, including carbon steel, stainless steel, and/or a Ni--Cr--Fe alloy. Methods install an internal head vent in a nuclear reactor by securing piping to an internal surface of an upper head of the nuclear reactor and/or securing piping to an internal surface of a reactor pressure vessel.

New Concepts for Compact Space Reactor Power Systems for Space Based Radar Applications: A Feasibility Study

DTIC Science & Technology

1989-12-01

SPENT FUEL REPROCESSING COULD ALSO BE EMPLOYED IRRADIATION EXPERIENCE - EXTREMELY LIMITED - JOINT US/UK PROGRAM (ONGOING) - TUI/KFK PROGRAM (CANCELED...only the use of off-the-shelf technologies. For example, conventional fuel technology (uranium dioxide), conventional thermionic conversion...advanced fuel (Americium oxide, A1TI2O3) and advanced thermionic conversion. Concept C involves use of an advanced fuel (Americium oxide, Arri203

Irradiation test of tungsten clad uranium carbide-zirconium carbide ((U,Zr)C) specimens for thermionic reactor application at conditions conductive to long-term performance

NASA Technical Reports Server (NTRS)

Creagh, J. W. R.; Smith, J. R.

1973-01-01

Uranium carbide fueled, thermionic emitter configurations were encapsulated and irradiated. One capsule contained a specimen clad with fluoride derived chemically vapor deposited (CVD) tungsten. The other capsule used a duplex clad specimen consisting of chloride derived on floride derived CVD tungsten. Both fuel pins were 16 millimeters in diameter and contained a 45.7-millimeter length of fuel.

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.

Comparison of Tungsten and Molybdenum Based Emitters for Advanced Thermionic Space Nuclear Reactors

NASA Astrophysics Data System (ADS)

Lee, Hsing H.; Dickinson, Jeffrey W.; Klein, Andrew C.; Lamp, Thomas R.

1994-07-01

Variations to the Advanced Thermionic Initiative thermionic fuel element are analyzed. Analysis included neutronic modeling with MCNP for criticality determination and thermal power distribution, and thermionic performance modeling with TFEHX. Changes to the original ATI configuration include the addition of W-HfC wire to the emitter for high temperature creep resistance improvement and substitution of molybdenum for the tungsten base material. Results from MCNP showed that all the tungsten used in the coating and base material must be 100% W-184 to obtain criticality. The presence of molybdenum in the emitter base affects the neutronic performance of the TFE by increasing the emitter neutron absorption cross section. Due to the reduced thermal conductivity for the molybdenum based emitter, a higher temperature is obtained resulting in a greater electrical power production. The thermal conductivity and resistivity of the composite emitter region were derived for the W-Mo composite and used in TFEHX.

Coupled reactor kinetics and heat transfer model for heat pipe cooled reactors

NASA Astrophysics Data System (ADS)

Wright, Steven A.; Houts, Michael

2001-02-01

Heat pipes are often proposed as cooling system components for small fission reactors. SAFE-300 and STAR-C are two reactor concepts that use heat pipes as an integral part of the cooling system. Heat pipes have been used in reactors to cool components within radiation tests (Deverall, 1973); however, no reactor has been built or tested that uses heat pipes solely as the primary cooling system. Heat pipe cooled reactors will likely require the development of a test reactor to determine the main differences in operational behavior from forced cooled reactors. The purpose of this paper is to describe the results of a systems code capable of modeling the coupling between the reactor kinetics and heat pipe controlled heat transport. Heat transport in heat pipe reactors is complex and highly system dependent. Nevertheless, in general terms it relies on heat flowing from the fuel pins through the heat pipe, to the heat exchanger, and then ultimately into the power conversion system and heat sink. A system model is described that is capable of modeling coupled reactor kinetics phenomena, heat transfer dynamics within the fuel pins, and the transient behavior of heat pipes (including the melting of the working fluid). This paper focuses primarily on the coupling effects caused by reactor feedback and compares the observations with forced cooled reactors. A number of reactor startup transients have been modeled, and issues such as power peaking, and power-to-flow mismatches, and loading transients were examined, including the possibility of heat flow from the heat exchanger back into the reactor. This system model is envisioned as a tool to be used for screening various heat pipe cooled reactor concepts, for designing and developing test facility requirements, for use in safety evaluations, and for developing test criteria for in-pile and out-of-pile test facilities. .

Nuclear thermionic converter. [tungsten-thorium oxide rods

NASA Technical Reports Server (NTRS)

Phillips, W. M.; Mondt, J. F. (Inventor)

1977-01-01

Efficient nuclear reactor thermionic converter units are described which can be constructed at low cost and assembled in a reactor which requires a minimum of fuel. Each converter unit utilizes an emitter rod with a fluted exterior, several fuel passages located in the bulges that are formed in the rod between the flutes, and a collector receiving passage formed through the center of the rod. An array of rods is closely packed in an interfitting arrangement, with the bulges of the rods received in the recesses formed between the bulges of other rods, thereby closely packing the nuclear fuel. The rods are constructed of a mixture of tungsten and thorium oxide to provide high power output, high efficiency, high strength, and good machinability.

Solid-State Thermionic Nuclear Power for Megawatt Propulsion, Planetary Surface and Commercial Power Project

NASA Technical Reports Server (NTRS)

George, Jeffrey

2014-01-01

Thermionic (TI) power conversion is a promising technology first investigated for power conversion in the 1960's, and of renewed interest due to modern advances in nanotechnology, MEMS, materials and manufacturing. Benefits include high conversion efficiency (20%), static operation with no moving parts and potential for high reliability, greatly reduced plant complexity, and the potential for reduced development costs. Thermionic emission, credited to Edison in 1880, forms the basis of vacuum tubes and much of 20th century electronics. Heat can be converted into electricity when electrons emitted from a hot surface are collected across a small gap. For example, two "small" (6 kWe) Thermionic Space Reactors were flown by the USSR in 1987-88 for ocean radar reconnaissance. Higher powered Nuclear-Thermionic power systems driving Electric Propulsion (Q-thruster, VASIMR, etc.) may offer the breakthrough necessary for human Mars missions of < 1 yr round trip. Power generation on Earth could benefit from simpler, moe economical nuclear plants, and "topping" of more fuel and emission efficient fossil-fuel plants.

Heat Pipe Solar Receiver for Oxygen Production of Lunar Regolith

NASA Astrophysics Data System (ADS)

Hartenstine, John R.; Anderson, William G.; Walker, Kara L.; Ellis, Michael C.

2009-03-01

A heat pipe solar receiver operating in the 1050° C range is proposed for use in the hydrogen reduction process for the extraction of oxygen from the lunar soil. The heat pipe solar receiver is designed to accept, isothermalize and transfer solar thermal energy to reactors for oxygen production. This increases the available area for heat transfer, and increases throughput and efficiency. The heat pipe uses sodium as the working fluid, and Haynes 230 as the heat pipe envelope material. Initial design requirements have been established for the heat pipe solar receiver design based on information from the NASA In-Situ Resource Utilization (ISRU) program. Multiple heat pipe solar receiver designs were evaluated based on thermal performance, temperature uniformity, and integration with the solar concentrator and the regolith reactor(s). Two designs were selected based on these criteria: an annular heat pipe contained within the regolith reactor and an annular heat pipe with a remote location for the reactor. Additional design concepts have been developed that would use a single concentrator with a single solar receiver to supply and regulate power to multiple reactors. These designs use variable conductance or pressure controlled heat pipes for passive power distribution management between reactors. Following the design study, a demonstration heat pipe solar receiver was fabricated and tested. Test results demonstrated near uniform temperature on the outer surface of the pipe, which will ultimately be in contact with the regolith reactor.

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.

Thermionic system evaluated test (TSET) facility description

NASA Astrophysics Data System (ADS)

Fairchild, Jerry F.; Koonmen, James P.; Thome, Frank V.

1992-01-01

A consortium of US agencies are involved in the Thermionic System Evaluation Test (TSET) which is being supported by the Strategic Defense Initiative Organization (SDIO). The project is a ground test of an unfueled Soviet TOPAZ-II in-core thermionic space reactor powered by electrical heat. It is part of the United States' national thermionic space nuclear power program. It will be tested in Albuquerque, New Mexico at the New Mexico Engineering Research Institute complex by the Phillips Laboratoty, Sandia National Laboratories, Los Alamos National Laboratory, and the University of New Mexico. One of TSET's many objectives is to demonstrate that the US can operate and test a complete space nuclear power system, in the electrical heater configuration, at a low cost. Great efforts have been made to help reduce facility costs during the first phase of this project. These costs include structural, mechanical, and electrical modifications to the existing facility as well as the installation of additional emergency systems to mitigate the effects of utility power losses and alkali metal fires.

Shielded fluid stream injector for particle bed reactor

DOEpatents

Notestein, John E.

1993-01-01

A shielded fluid-stream injector assembly is provided for particle bed reactors. The assembly includes a perforated pipe injector disposed across the particle bed region of the reactor and an inverted V-shaped shield placed over the pipe, overlapping it to prevent descending particles from coming into direct contact with the pipe. The pipe and shield are fixedly secured at one end to the reactor wall and slidably secured at the other end to compensate for thermal expansion. An axially extending housing aligned with the pipe and outside the reactor and an in-line reamer are provided for removing deposits from the inside of the pipe. The assembly enables fluid streams to be injected and distributed uniformly into the particle bed with minimized clogging of injector ports. The same design may also be used for extraction of fluid streams from particle bed reactors.

Mission operations for unmanned nuclear electric propulsion outer planet exploration with a thermionic reactor spacecraft.

NASA Technical Reports Server (NTRS)

Spera, R. J.; Prickett, W. Z.; Garate, J. A.; Firth, W. L.

1971-01-01

Mission operations are presented for comet rendezvous and outer planet exploration NEP spacecraft employing in-core thermionic reactors for electric power generation. The selected reference missions are the Comet Halley rendezvous and a Jupiter orbiter at 5.9 planet radii, the orbit of the moon Io. The characteristics of the baseline multi-mission NEP spacecraft are presented and its performance in other outer planet missions, such as Saturn and Uranus orbiters and a Neptune flyby, are discussed. Candidate mission operations are defined from spacecraft assembly to mission completion. Pre-launch operations are identified. Shuttle launch and subsequent injection to earth escape by the Centaur D-1T are discussed, as well as power plant startup and the heliocentric mission phases. The sequence and type of operations are basically identical for all missions investigated.

Analyzing the thermionic reactor critical experiments. [thermal spectrum of uranium 235 core

NASA Technical Reports Server (NTRS)

Niederauer, G. F.

1973-01-01

The Thermionic Reactor Critical Experiments (TRCE) consisted of fast spectrum highly enriched U-235 cores reflected by different thicknesses of beryllium or beryllium oxide with a transition zone of stainless steel between the core and reflector. The mixed fast-thermal spectrum at the core reflector interface region poses a difficult neutron transport calculation. Calculations of TRCE using ENDF/B fast spectrum data and GATHER library thermal spectrum data agreed within about 1 percent for the multiplication factor and within 6 to 8 percent for the power peaks. Use of GAM library fast spectrum data yielded larger deviations. The results were obtained from DOT R Theta calculations with leakage cross sections, by region and by group, extracted from DOT RZ calculations. Delineation of the power peaks required extraordinarily fine mesh size at the core reflector interface.

Sodium Based Heat Pipe Modules for Space Reactor Concepts: Stainless Steel SAFE-100 Core

NASA Technical Reports Server (NTRS)

Martin, James J.; Reid, Robert S.

2004-01-01

A heat pipe cooled reactor is one of several candidate reactor cores being considered for advanced space power and propulsion systems to support future space exploration applications. Long life heat pipe modules, with designs verified through a combination of theoretical analysis and experimental lifetime evaluations, would be necessary to establish the viability of any of these candidates, including the heat pipe reactor option. A hardware-based program was initiated to establish the infrastructure necessary to build heat pipe modules. This effort, initiated by Los Alamos National Laboratory and referred to as the Safe Affordable Fission Engine (SAFE) project, set out to fabricate and perform non-nuclear testing on a modular heat pipe reactor prototype that can provide 100 kilowatt from the core to an energy conversion system at 700 C. Prototypic heat pipe hardware was designed, fabricated, filled, closed-out and acceptance tested.

Nuclear electric propulsion reactor control systems status

NASA Technical Reports Server (NTRS)

Ferg, D. A.

1973-01-01

The thermionic reactor control system design studies conducted over the past several years for a nuclear electric propulsion system are described and summarized. The relevant reactor control system studies are discussed in qualitative terms, pointing out the significant advantages and disadvantages including the impact that the various control systems would have on the nuclear electric propulsion system design. A recommendation for the reference control system is made, and a program for future work leading to an engineering model is described.

Test development for the thermionic system evaluation test (TSET) project

NASA Astrophysics Data System (ADS)

Morris, D. Brent; Standley, Vaughn H.; Schuller, Michael J.

1992-01-01

The arrival of a Soviet TOPAZ-II space nuclear reactor affords the US space nuclear power (SNP) community the opportunity to study an assembled thermionic conversion power system. The TOPAZ-II will be studied via the Thermionic System Evaluation Test (TSET) Project. This paper is devoted to the discussion of TSET test development as related to the objectives contained in the TSET Project Plan (Standley et al. 1991). The objectives contained in the Project Plan are the foundation for scheduled TSET tests on TOPAZ-II and are derived from the needs of the Air Force Thermionic SNP program. Our ability to meet the objectives is bounded by unique constraints, such as procurement requirements, operational limitations, and necessary interaction between US and Soviet Scientists and engineers. The fulfillment of the test objectives involves a thorough methodology of test scheduling and data managment. The overall goals for the TSET program are gaining technical understanding of a thermionic SNP system and demonstrating the capabilities and limitations of such a system while assisting in the training of US scientist and engineers in preparation for US SNP system testing. Tests presently scheduled as part of TSET include setup, demonstration, and verification tests; normal and off-normal operating test, and system and component performance tests.

Pipe connector

DOEpatents

Sullivan, Thomas E.; Pardini, John A.

1978-01-01

A safety test facility for testing sodium-cooled nuclear reactor components includes a reactor vessel and a heat exchanger submerged in sodium in the tank. The reactor vessel and heat exchanger are connected by an expansion/deflection pipe coupling comprising a pair of coaxially and slidably engaged tubular elements having radially enlarged opposed end portions of which at least a part is of spherical contour adapted to engage conical sockets in the ends of pipes leading out of the reactor vessel and in to the heat exchanger. A spring surrounding the pipe coupling urges the end portions apart and into engagement with the spherical sockets. Since the pipe coupling is submerged in liquid a limited amount of leakage of sodium from the pipe can be tolerated.

158. ARAIII Reactor building (ARA608) Secondary cooling loop and piping ...

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

158. ARA-III Reactor building (ARA-608) Secondary cooling loop and piping plan. This drawing was selected as a typical example of piping arrangements within reactor building. Aerojet/general 880-area/GCRE-608-P-16. Date: February 1958. INeel index code no. 063-0608-50-013-102641. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

In-pile and out-of-pile testing of a molybdenum-uranium dioxide cermet fueled themionic diode

NASA Technical Reports Server (NTRS)

Diianni, D. C.

1972-01-01

The behavior of Mo-UO2 cermet fuel in a diode for thermionic reactor application was studied. The diode had a Mo-0.5 Ti emitter and niobium collector. Output power ranged from 1.4 to 2.8 W/cm squared at emitter and collector temperatures of 1500 deg and 540 C. Thermionic performance was stable within the limits of the instrumentation sensitivity. Through 1000 hours of in-pile operation the emitter was dimensionally stable. However, some fission gases (15 percent) leaked through an inner clad imperfection that occurred during fuel fabrication.

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

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

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

2005-02-06

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

Advanced design concepts in nuclear electric propulsion. [and spacecraft configurations

NASA Technical Reports Server (NTRS)

Peelgren, M. L.; Mondt, J. F.

1974-01-01

Conceptual designs of the nuclear propulsion programs are reported. Major areas of investigation were (1) design efforts on spacecraft configuration and heat rejection subsystem, (2) high-voltage thermionic reactor concepts, and (3) dual-mode spacecraft configuration study.

Processing of thermionic power on an electrically propelled spacecraft

NASA Technical Reports Server (NTRS)

Macie, T. W.

1973-01-01

A study to define the power processing equipment required between a thermionic reactor and an array of mercury-ion thrusters for a nuclear electric propulsion system is reported. Observations and recommendations that resulted from this study were: (1) the preferred thermionic-fuel-element source voltages are 23 V or higher; (2) transistor characteristics exert a strong effect on power processor mass; (3) the power processor mass could be considerably reduced should the magnetic materials that exhibit low losses at high frequencies, that have a high Curie point, and that can operate at 15 to 20 kG become avaliable; (4) electrical component packaging on the radiator could reduce the area that is sensitive to meteoroid penetration, thereby reducing the meteoroid shielding mass requirement; (5) an experimental model of the power processor design should be built and tested to verify the efficiencies, masses, and all the automatic operational aspects of the design.

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

Fast reactor power plant design having heat pipe heat exchanger

DOEpatents

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

1984-08-30

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

Fast reactor power plant design having heat pipe heat exchanger

DOEpatents

Huebotter, Paul R.; McLennan, George A.

1985-01-01

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

Passive heat transfer means for nuclear reactors

DOEpatents

Burelbach, James P.

1984-01-01

An improved passive cooling arrangement is disclosed for maintaining adjacent or related components of a nuclear reactor within specified temperature differences. Specifically, heat pipes are operatively interposed between the components, with the vaporizing section of the heat pipe proximate the hot component operable to cool it and the primary condensing section of the heat pipe proximate the other and cooler component operable to heat it. Each heat pipe further has a secondary condensing section that is located outwardly beyond the reactor confinement and in a secondary heat sink, such as air ambient the containment, that is cooler than the other reactor component. Means such as shrouding normally isolated the secondary condensing section from effective heat transfer with the heat sink, but a sensor responds to overheat conditions of the reactor to open the shrouding, which thereby increases the cooling capacity of the heat pipe. By having many such heat pipes, an emergency passive cooling system is defined that is operative without electrical power.

Passive heat-transfer means for nuclear reactors. [LMFBR

DOEpatents

Burelbach, J.P.

1982-06-10

An improved passive cooling arrangement is disclosed for maintaining adjacent or related components of a nuclear reactor within specified temperature differences. Specifically, heat pipes are operatively interposed between the components, with the vaporizing section of the heat pipe proximate the hot component operable to cool it and the primary condensing section of the heat pipe proximate the other and cooler component operable to heat it. Each heat pipe further has a secondary condensing section that is located outwardly beyond the reactor confinement and in a secondary heat sink, such as air ambient the containment, that is cooler than the other reactor component. By having many such heat pipes, an emergency passive cooling system is defined that is operative without electrical power.

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

Wichman, K.; Tsao, J.; Mayfield, M.

The regulatory application of leak before break (LBB) for operating and advanced reactors in the U.S. is described. The U.S. Nuclear Regulatory Commission (NRC) has approved the application of LBB for six piping systems in operating reactors: reactor coolant system primary loop piping, pressurizer surge, safety injection accumulator, residual heat removal, safety injection, and reactor coolant loop bypass. The LBB concept has also been applied in the design of advanced light water reactors. LBB applications, and regulatory considerations, for pressurized water reactors and advanced light water reactors are summarized in this paper. Technology development for LBB performed by the NRCmore » and the International Piping Integrity Research Group is also briefly summarized.« less

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

DOEpatents

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

1993-11-16

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

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

DOEpatents

Ekeroth, Douglas E.; Corletti, Michael M.

1993-01-01

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

Application of a Systems Engineering Approach to Support Space Reactor Development

NASA Astrophysics Data System (ADS)

Wold, Scott

2005-02-01

In 1992, approximately 25 Russian and 12 U.S. engineers and technicians were involved in the transport, assembly, inspection, and testing of over 90 tons of Russian equipment associated with the Thermionic System Evaluation Test (TSET) Facility. The entire Russian Baikal Test Stand, consisting of a 5.79 m tall vacuum chamber and related support equipment, was reassembled and tested at the TSET facility in less than four months. In November 1992, the first non-nuclear operational test of a complete thermionic power reactor system in the U.S. was accomplished three months ahead of schedule and under budget. A major factor in this accomplishment was the application of a disciplined top-down systems engineering approach and application of a spiral development model to achieve the desired objectives of the TOPAZ International Program (TIP). Systems Engineering is a structured discipline that helps programs and projects conceive, develop, integrate, test and deliver products and services that meet customer requirements within cost and schedule. This paper discusses the impact of Systems Engineering and a spiral development model on the success of the TOPAZ International Program and how the application of a similar approach could help ensure the success of future space reactor development projects.

Development concept for a small, split-core, heat-pipe-cooled nuclear reactor

NASA Technical Reports Server (NTRS)

Lantz, E.; Breitwieser, R.; Niederauer, G. F.

1974-01-01

There have been two main deterrents to the development of semiportable nuclear reactors. One is the high development costs; the other is the inability to satisfy with assurance the questions of operational safety. This report shows how a split-core, heat-pipe cooled reactor could conceptually eliminate these deterrents, and examines and summarizes recent work on split-core, heat-pipe reactors. A concept for a small reactor that could be developed at a comparatively low cost is presented. The concept would extend the technology of subcritical radioisotope thermoelectric generators using 238 PuO2 to the evolution of critical space power reactors using 239 PuO2.

Design of megawatt power level heat pipe reactors

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

Mcclure, Patrick Ray; Poston, David Irvin; Dasari, Venkateswara Rao

An important niche for nuclear energy is the need for power at remote locations removed from a reliable electrical grid. Nuclear energy has potential applications at strategic defense locations, theaters of battle, remote communities, and emergency locations. With proper safeguards, a 1 to 10-MWe (megawatt electric) mobile reactor system could provide robust, self-contained, and long-term power in any environment. Heat pipe-cooled fast-spectrum nuclear reactors have been identified as a candidate for these applications. Heat pipe reactors, using alkali metal heat pipes, are perfectly suited for mobile applications because their nature is inherently simpler, smaller, and more reliable than “traditional” reactors.more » The goal of this project was to develop a scalable conceptual design for a compact reactor and to identify scaling issues for compact heat pipe cooled reactors in general. Toward this goal two detailed concepts were developed, the first concept with more conventional materials and a power of about 2 MWe and a the second concept with less conventional materials and a power level of about 5 MWe. A series of more qualitative advanced designs were developed (with less detail) that show power levels can be pushed to approximately 30 MWe.« less

High efficiency thermionic converter studies

NASA Technical Reports Server (NTRS)

Huffman, F. N.; Sommer, A. H.; Balestra, C. L.; Briere, D. P.; Oettinger, P. E.

1976-01-01

The objective is to improve thermionic converter performance by means of reduced interelectrode losses, greater emitter capabilities, and lower collector work functions until the converter performance level is suitable for out-of-core space reactors and radioisotope generators. Electrode screening experiments have identified several promising collector materials. Back emission work function measurements of a ZnO collector in a thermionic diode have given values less than 1.3 eV. Diode tests were conducted over the range of temperatures of interest for space power applications. Enhanced mode converter experiments have included triodes operated in both the surface ionization and plasmatron modes. Pulsed triodes were studied as a function of pulse length, pulse potential, inert gas fill pressure, cesium pressure, spacing, emitter temperature and collector temperature. Current amplifications (i.e., mean output current/mean grid current) of several hundred were observed up to output current densities of one amp/sq cm. These data correspond to an equivalent arc drop less than 0.1 eV.

Effect on Non-Uniform Heat Generation on Thermionic Reactions

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

Schock, Alfred

The penalty resulting from non-uniform heat generation in a thermionic reactor is examined. Operation at sub-optimum cesium pressure is shown to reduce this penalty, but at the risk of a condition analogous to burnout. For high pressure diodes, a simple empirical correlation between current, voltage and heat flux is developed and used to analyze the performance penalty associated with two different heat flux profiles, for series-and parallel-connected converters. The results demonstrate that series-connected converters require much finer power flattening than parallel converters. For example, a ±10% variation in heat generation across a series array can result in a 25 tomore » 50% power penalty.« less

Transient Response to Rapid Cooling of a Stainless Steel Sodium Heat Pipe

NASA Technical Reports Server (NTRS)

Mireles, Omar R.; Houts, Michael G.

2011-01-01

Compact fission power systems are under consideration for use in long duration space exploration missions. Power demands on the order of 500 W, to 5 kW, will be required for up to 15 years of continuous service. One such small reactor design consists of a fast spectrum reactor cooled with an array of in-core alkali metal heat pipes coupled to thermoelectric or Stirling power conversion systems. Heat pipes advantageous attributes include a simplistic design, lack of moving parts, and well understood behavior. Concerns over reactor transients induced by heat pipe instability as a function of extreme thermal transients require experimental investigations. One particular concern is rapid cooling of the heat pipe condenser that would propagate to cool the evaporator. Rapid cooling of the reactor core beyond acceptable design limits could possibly induce unintended reactor control issues. This paper discusses a series of experimental demonstrations where a heat pipe operating at near prototypic conditions experienced rapid cooling of the condenser. The condenser section of a stainless steel sodium heat pipe was enclosed within a heat exchanger. The heat pipe - heat exchanger assembly was housed within a vacuum chamber held at a pressure of 50 Torr of helium. The heat pipe was brought to steady state operating conditions using graphite resistance heaters then cooled by a high flow of gaseous nitrogen through the heat exchanger. Subsequent thermal transient behavior was characterized by performing an energy balance using temperature, pressure and flow rate data obtained throughout the tests. Results indicate the degree of temperature change that results from a rapid cooling scenario will not significantly influence thermal stability of an operating heat pipe, even under extreme condenser cooling conditions.

65. ARAII. Interior view of SL1 reactor building control piping ...

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

65. ARA-II. Interior view of SL-1 reactor building control piping for water purification system. On operating floor of building. March 21, 1958. Ineel photo no. 58-1360. Photographer: Jack L. Anderson. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

152. ARAIII Reactor building (ARA608) Details of heater and piping ...

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

152. ARA-III Reactor building (ARA-608) Details of heater and piping pits, including instrumentation plan. Aerojet-general 880-area/GCRE-608-T-18. Date: November 1958. Ineel index code no. 063-0608-25-013-102677. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

Promethus Hot Leg Piping Concept

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

AM Girbik; PA Dilorenzo

2006-01-24

The Naval Reactors Prime Contractor Team (NRPCT) recommended the development of a gas cooled reactor directly coupled to a Brayton energy conversion system as the Space Nuclear Power Plant (SNPP) for NASA's Project Prometheus. The section of piping between the reactor outlet and turbine inlet, designated as the hot leg piping, required unique design features to allow the use of a nickel superalloy rather than a refractory metal as the pressure boundary. The NRPCT evaluated a variety of hot leg piping concepts for performance relative to SNPP system parameters, manufacturability, material considerations, and comparison to past high temperature gas reactormore » (HTGR) practice. Manufacturability challenges and the impact of pressure drop and turbine entrance temperature reduction on cycle efficiency were discriminators between the piping concepts. This paper summarizes the NRPCT hot leg piping evaluation, presents the concept recommended, and summarizes developmental issues for the recommended concept.« less

Status report on nuclear electric propulsion systems

NASA Technical Reports Server (NTRS)

Stearns, J. W.

1975-01-01

Progress in nuclear electric propulsion (NEP) systems for a multipayload multimission vehicle needed in both deep-space missions and a variety of geocentric missions is reviewed. The space system power level is a function of the initial launch vehicle mass, but developments in out-of-core nuclear thermionic direct conversion have broadened design options. Cost, design, and performance parameters are compared for reusable chemical space tugs and NEP reusable space tugs. Improvements in heat pipes, ion engines, and magnetoplasmadynamic arc jet thrust subsystems are discussed.

Experiments on rehabilitation of radioactive metallic waste (RMW) of reactor stainless steels of Siberian chemical plant

NASA Astrophysics Data System (ADS)

Kolpakov, G. N.; Zakusilov, V. V.; Demyanenko, N. V.; Mishin, A. S.

2016-06-01

Stainless steel pipes, used to cool a reactor plant, have a high cost, and after taking a reactor out of service they must be buried together with other radioactive waste. Therefore, the relevant problem is the rinse of pipes from contamination, followed by returning to operation.

Transient modeling of the thermohydraulic behavior of high temperature heat pipes for space reactor applications

NASA Technical Reports Server (NTRS)

Hall, Michael L.; Doster, Joseph M.

1986-01-01

Many proposed space reactor designs employ heat pipes as a means of conveying heat. Previous researchers have been concerned with steady state operation, but the transient operation is of interest in space reactor applications due to the necessity of remote startup and shutdown. A model is being developed to study the dynamic behavior of high temperature heat pipes during startup, shutdown and normal operation under space environments. Model development and preliminary results for a hypothetical design of the system are presented.

Experimental study of Siphon breaker about size effect in real scale reactor design

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

Kang, S. H.; Ahn, H. S.; Kim, J. M.

2012-07-01

Rupture accident within the pipe of a nuclear reactor is one of the main causes of a loss of coolant accident (LOCA). Siphon-breaking is a passive method that can prevent a LOCA. In this study, either a line or a hole is used as a siphon-breaker, and the effect of various parameters, such as the siphon-breaker size, pipe rupture point, pipe rupture size, and the presence of an orifice, are investigated using an experimental facility similar in size to a full-scale reactor. (authors)

Pilot-plant studies of the pipe- and pipe-cross reactors in production of granular polyphosphate fertilizers. TVA Circular Z-148. [Methods of increasing polyphosphate content of suspension fertilizers

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

Parker, B.R.; Norton, M.M.; Stumpe, T.R.

1982-01-01

Improvements have been made in the pipe-reactor or pipe-cross reactor/drum-granulator process to increase the polyphosphate content of the granular product. The goal of producing a granular APP product containing 20% of P/sub 2/O/sub 5/ as polyphosphate without adding external heat or sulfuric acid to the process has not yet been realized; however, products containing slightly more than 10% of the P/sub 2/O/sub 5/ as polyphosphate have been made without the need for external heat. Test results indicate that additions of small amounts of sulfuric acid, use of reactant NH/sub 3/:H/sub 3/PO/sub 4/ed mole ratios greater than 1.05, or use ofmore » some acid preheat may be required to consistently obtain 12% of the P/sub 2/O/sub 5/ as polyphosphate as is desired for use of the product in the preparation of suspension fertilizers. However, continued testing is being done to determine how high a mole ratio may be used successfully and to determine the effect of sulfate addition on use of the granular products for producing suspension fertilizers. The effort to obtain higher polyphosphate levels from the pipe-reactor and drum and the pipe-cross reactor and drum systems is being continued.« less

Solid0Core Heat-Pipe Nuclear Batterly Type Reactor

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

Ehud Greenspan

This project was devoted to a preliminary assessment of the feasibility of designing an Encapsulated Nuclear Heat Source (ENHS) reactor to have a solid core from which heat is removed by liquid-metal heat pipes (HP).

Three dimensional contact/impact methodology

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

Kulak, R.F.

1987-01-01

The simulation of three-dimensional interface mechanics between reactor components and structures during static contact or dynamic impact is necessary to realistically evaluate their structural integrity to off-normal loads. In our studies of postulated core energy release events, we have found that significant structure-structure interactions occur in some reactor vessel head closure designs and that fluid-structure interactions occur within the reactor vessel. Other examples in which three-dimensional interface mechanics play an important role are: (1) impact response of shipping casks containing spent fuel, (2) whipping pipe impact on reinforced concrete panels or pipe-to-pipe impact after a pipe break, (3) aircraft crashmore » on secondary containment structures, (4) missiles generated by turbine failures or tornados, and (5) drops of heavy components due to lifting accidents. The above is a partial list of reactor safety problems that require adequate treatment of interface mechanics and are discussed in this paper.« less

High Power LaB6 Plasma Source Performance for the Lockheed Martin Compact Fusion Reactor Experiment

NASA Astrophysics Data System (ADS)

Heinrich, Jonathon

2016-10-01

Lockheed Martin's Compact Fusion Reactor (CFR) concept is a linear encapsulated ring cusp. Due to the complex field geometry, plasma injection into the device requires careful consideration. A high power thermionic plasma source (>0.25MW; >10A/cm2) has been developed with consideration to phase space for optimal coupling. We present the performance of the plasma source, comparison with alternative plasma sources, and plasma coupling with the CFR field configuration. ©2016 Lockheed Martin Corporation. All Rights Reserved.

Transient Approximation of SAFE-100 Heat Pipe Operation

NASA Technical Reports Server (NTRS)

Bragg-Sitton, Shannon M.; Reid, Robert S.

2005-01-01

Engineers at Los Alamos National Laboratory (LANL) have designed several heat pipe cooled reactor concepts, ranging in power from 15 kWt to 800 kWt, for both surface power systems and nuclear electric propulsion systems. The Safe, Affordable Fission Engine (SAFE) is now being developed in a collaborative effort between LANL and NASA Marshall Space Flight Center (NASA/MSFC). NASA is responsible for fabrication and testing of non-nuclear, electrically heated modules in the Early Flight Fission Test Facility (EFF-TF) at MSFC. In-core heat pipes must be properly thawed as the reactor power starts. Computational models have been developed to assess the expected operation of a specific heat pipe design during start-up, steady state operation, and shutdown. While computationally intensive codes provide complete, detailed analyses of heat pipe thaw, a relatively simple. concise routine can also be applied to approximate the response of a heat pipe to changes in the evaporator heat transfer rate during start-up and power transients (e.g., modification of reactor power level) with reasonably accurate results. This paper describes a simplified model of heat pipe start-up that extends previous work and compares the results to experimental measurements for a SAFE-100 type heat pipe design.

Self-actuating reactor shutdown system

DOEpatents

Barrus, Donald M.; Brummond, Willian A; Peterson, Leslie F.

1988-01-01

A control system for the automatic or self-actuated shutdown or "scram" of a nuclear reactor. The system is capable of initiating scram insertion by a signal from the plant protection system or by independent action directly sensing reactor conditions of low-flow or over-power. Self-actuation due to a loss of reactor coolant flow results from a decrease of pressure differential between the upper and lower ends of an absorber element. When the force due to this differential falls below the weight of the element, the element will fall by gravitational force to scram the reactor. Self-actuation due to high neutron flux is accomplished via a valve controlled by an electromagnet and a thermionic diode. In a reactor over-power, the diode will be heated to a change of state causing the electromagnet to be shorted thereby actuating the valve which provides the changed flow and pressure conditions required for scramming the absorber element.

DETECTION OF COATING FAILURES IN A NEUTRONIC REACTOR

DOEpatents

Snell, A.H.; Allison, S.K.

1958-02-11

This patent relates to water-cooled reactor systems and discloses a means to detect leaks in the jackets of jacketed fuel elements comprising a neutron detector located in the cooling water discharge pipe,the pipe being provided with an enlarged portion for housing the detector so that the latter is completely surrounded by the water in its passage through the pipe, said enlarged portion and detector being shielded from the reactor for the purpose of detecting only those delayed neutrons emitted in the cooling water and due to the latter picking up fission fragments from the defective fuel elements.

Survey of Current and Next Generation Space Power Technologies

DTIC Science & Technology

2006-06-26

different thermodynamic cycles, such as the Brayton, Rankine, and Stirling cycles, alkali metal thermal electric converters ( AMTEC ) and thermionic...efficiencies @ 1700K. The primary issue with this system is the integration of the converter technology into the nuclear reactor core. AMTEC (static...Alkali metal thermal to electric converters ( AMTECs ) are thermally powered electrochemical concentration cells that convert heat energy directly to DC

Vented nuclear fuel element

DOEpatents

Grossman, Leonard N.; Kaznoff, Alexis I.

1979-01-01

A nuclear fuel cell for use in a thermionic nuclear reactor in which a small conduit extends from the outside surface of the emitter to the center of the fuel mass of the emitter body to permit escape of volatile and gaseous fission products collected in the center thereof by virtue of molecular migration of the gases to the hotter region of the fuel.

Development and optimization of water treatment reactors using TiO2-modified polymer beads with a refractive index identical to that of water

NASA Astrophysics Data System (ADS)

Myoga, Arata; Iwashita, Ryutaro; Unno, Noriyuki; Satake, Shin-ichi; Taniguchi, Jun; Yuki, Kazuhisa; Seki, Yohji

2018-03-01

Various water purification reactors were constructed using beads of TiO2-coated MEXFLON, which is a fluoropolymer exhibiting a refractive index identical to that of water. The performance of these reactors was evaluated in a recirculation experiment utilizing an aqueous solution of methylene blue. Reactor pipes (length = 150 mm, internal diameter = 10 mm) were made of a fluorinated ethylene polymer with a refractive index of 1.338 and contained 206-bead clusters. A UV lamp was used to irradiate eight reactor pipes surrounding it. The above-mentioned eight bead-packed pipes were connected both in series and in parallel, and the performances of these two reactor types were compared. A pseudo-first-order rate constant of 0.70 h- 1 was obtained for the series connection, whereas the corresponding value for the parallel connection was 1.5 times smaller, confirming the effectiveness of increasing the reaction surface by employing a larger number of beads.

Development and optimization of water treatment reactors using TiO2-modified polymer beads with a refractive index identical to that of water

NASA Astrophysics Data System (ADS)

Myoga, Arata; Iwashita, Ryutaro; Unno, Noriyuki; Satake, Shin-ichi; Taniguchi, Jun; Yuki, Kazuhisa; Seki, Yohji

2018-06-01

Various water purification reactors were constructed using beads of TiO2-coated MEXFLON, which is a fluoropolymer exhibiting a refractive index identical to that of water. The performance of these reactors was evaluated in a recirculation experiment utilizing an aqueous solution of methylene blue. Reactor pipes (length = 150 mm, internal diameter = 10 mm) were made of a fluorinated ethylene polymer with a refractive index of 1.338 and contained 206-bead clusters. A UV lamp was used to irradiate eight reactor pipes surrounding it. The above-mentioned eight bead-packed pipes were connected both in series and in parallel, and the performances of these two reactor types were compared. A pseudo-first-order rate constant of 0.70 h- 1 was obtained for the series connection, whereas the corresponding value for the parallel connection was 1.5 times smaller, confirming the effectiveness of increasing the reaction surface by employing a larger number of beads.

Development costs for a nuclear electric propulsion stage.

NASA Technical Reports Server (NTRS)

Mondt, J. F.; Prickett, W. Z.

1973-01-01

Development costs are presented for an unmanned nuclear electric propulsion (NEP) stage based upon a liquid metal cooled, in-core thermionic reactor. A total of 120 kWe are delivered to the thrust subsystem which employs mercury ion engines for electric propulsion. This study represents the most recent cost evaluation of the development of a reactor power system for a wide range of nuclear space power applications. These include geocentric, and outer planet and other deep space missions. The development program is described for the total NEP stage, based upon specific development programs for key NEP stage components and subsystems.

ETR BUILDING, TRA642, INTERIOR. BASEMENT. LIQUID SODIUM PIPING INSIDE CUBICLE ...

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

ETR BUILDING, TRA-642, INTERIOR. BASEMENT. LIQUID SODIUM PIPING INSIDE CUBICLE SHOWN IN ID-33-G-101. INL NEGATIVE NO. HD24-3-4. Mike Crane, Photographer, 11/2000 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Research of thermionic converter collector properties in model experiments with surface control

NASA Astrophysics Data System (ADS)

Agafonov, Valerii R.; Vizgalov, Anatolii V.; Iarygin, Valerii I.

Consideration was given to a possible scheme of phenomena on electrodes leading to changes in emission properties (EP) of a thermionic converter (TEC) collector. It was based on technology and materials typical of the TOPAZ-type reactor-converter (TRC). The element composition (EC), near-surface layer (NSL) structure, and work function (WF) of a collector made from niobium-based polycrystal alloy were studied within this scheme experimentally. The influence of any media except for the interelectrode gap (IEG) medium was excluded when investigating the effect of thermovacuum treatment (TVT) as well as the influence of carbon monoxide, hydrogen, and methane on the NSL characteristics. Experimental data and analytical estimates of the impact of fission products of the nuclear fuel on collector EP are presented. The calculation of possible TRC electrical power decrease was also carried out.

A PC-based high temperature gas reactor simulator for Indonesian conceptual HTR reactor basic training

NASA Astrophysics Data System (ADS)

Syarip; Po, L. C. C.

2018-05-01

In planning for nuclear power plant construction in Indonesia, helium cooled high temperature reactor (HTR) is favorable for not relying upon water supply that might be interrupted by earthquake. In order to train its personnel, BATAN has cooperated with Micro-Simulation Technology of USA to develop a 200 MWt PC-based simulation model PCTRAN/HTR. It operates in Win10 environment with graphic user interface (GUI). Normal operation of startup, power maneuvering, shutdown and accidents including pipe breaks and complete loss of AC power have been conducted. A sample case of safety analysis simulation to demonstrate the inherent safety features of HTR was done for helium pipe break malfunction scenario. The analysis was done for the variation of primary coolant pipe break i.e. from 0,1% - 0,5 % and 1% - 10 % helium gas leakages, while the reactor was operated at the maximum constant power of 10 MWt. The result shows that the highest temperature of HTR fuel centerline and coolant were 1150 °C and 1296 °C respectively. With 10 kg/s of helium flow in the reactor core, the thermal power will back to the startup position after 1287 s of helium pipe break malfunction.

Optimization of a heat-pipe-cooled space radiator for use with a reactor-powered Stirling engine

NASA Technical Reports Server (NTRS)

Moriarty, Michael P.; French, Edward P.

1987-01-01

The design optimization of a reactor-Stirling heat-pipe-cooled radiator is presented. The radiator is a self-deploying concept that uses individual finned heat pipe 'petals' to reject waste heat from a Stirling engine. Radiator optimization methodology is presented, and the results of a parametric analysis of the radiator design variables for a 100-kW(e) system are given. The additional steps of optiminzing the radiator resulted in a net system mass savings of 3 percent.

Probabilistic pipe fracture evaluations for leak-rate-detection applications

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

Rahman, S.; Ghadiali, N.; Paul, D.

1995-04-01

Regulatory Guide 1.45, {open_quotes}Reactor Coolant Pressure Boundary Leakage Detection Systems,{close_quotes} was published by the U.S. Nuclear Regulatory Commission (NRC) in May 1973, and provides guidance on leak detection methods and system requirements for Light Water Reactors. Additionally, leak detection limits are specified in plant Technical Specifications and are different for Boiling Water Reactors (BWRs) and Pressurized Water Reactors (PWRs). These leak detection limits are also used in leak-before-break evaluations performed in accordance with Draft Standard Review Plan, Section 3.6.3, {open_quotes}Leak Before Break Evaluation Procedures{close_quotes} where a margin of 10 on the leak detection limit is used in determining the crackmore » size considered in subsequent fracture analyses. This study was requested by the NRC to: (1) evaluate the conditional failure probability for BWR and PWR piping for pipes that were leaking at the allowable leak detection limit, and (2) evaluate the margin of 10 to determine if it was unnecessarily large. A probabilistic approach was undertaken to conduct fracture evaluations of circumferentially cracked pipes for leak-rate-detection applications. Sixteen nuclear piping systems in BWR and PWR plants were analyzed to evaluate conditional failure probability and effects of crack-morphology variability on the current margins used in leak rate detection for leak-before-break.« less

PBF Cooling Tower. View from highbay roof of Reactor Building ...

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

PBF Cooling Tower. View from high-bay roof of Reactor Building (PER-620). Camera faces northwest. East louvered face has been installed. Inlet pipes protrude from fan deck. Two redwood vents under construction at top. Note piping, control, and power lines at sub-grade level in trench leading to Reactor Building. Photographer: Kirsh. Date: June 6, 1969. INEEL negative no. 69-3466 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

Examination of UC-ZrC after long term irradiation at thermionic temperature

NASA Technical Reports Server (NTRS)

Yang, L.; Johnson, H. O.

1972-01-01

Two fluoride tungsten clad UC-ZrC fueled capsules, designated as V-2C and V-2D, were examined a hot cell after irradiation in NASA Plum Brook Reactor at a maximum cladding temperature of 1930 K for 11,089 and 12,031 hours to burnups of 3.0 x 10 to the 20th power and 2.1 x 10 to the 20th power fission/c.c. respectively. Percentage of fission gas release from the fuel material was measured by radiochemical means. Cladding deformation, fuel-cladding interaction and microstructures of fuel, cladding, and fuel-cladding interface were studied metallographically. Compositions of dispersions in fuel, fuel matrix and fuel-cladding interaction layer were analyzed by electron microprobe techniques. Axial and radial distributions of burnup were determined by gamma-scan, autoradiography and isotopic burnup analysis. The results are presented and discussed in conjunction with the requirements of thermionic fuel elements for space power application.

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

DOEpatents

Hampel, V.E.

1988-05-17

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

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

DOEpatents

Hampel, Viktor E.

1989-01-01

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

Effects of phosphate addition on biofilm bacterial communities and water quality in annular reactors equipped with stainless steel and ductile cast iron pipes.

PubMed

Jang, Hyun-Jung; Choi, Young-June; Ro, Hee-Myong; Ka, Jong-Ok

2012-02-01

The impact of orthophosphate addition on biofilm formation and water quality was studied in corrosion-resistant stainless steel (STS) pipe and corrosion-susceptible ductile cast iron (DCI) pipe using cultivation and culture-independent approaches. Sample coupons of DCI pipe and STS pipe were installed in annular reactors, which were operated for 9 months under hydraulic conditions similar to a domestic plumbing system. Addition of 5 mg/L of phosphate to the plumbing systems, under low residual chlorine conditions, promoted a more significant growth of biofilm and led to a greater rate reduction of disinfection by-products in DCI pipe than in STS pipe. While the level of THMs (trihalomethanes) increased under conditions of low biofilm concentration, the levels of HAAs (halo acetic acids) and CH (chloral hydrate) decreased in all cases in proportion to the amount of biofilm. It was also observed that chloroform, the main species of THM, was not readily decomposed biologically and decomposition was not proportional to the biofilm concentration; however, it was easily biodegraded after the addition of phosphate. Analysis of the 16S rDNA sequences of 102 biofilm isolates revealed that Proteobacteria (50%) was the most frequently detected phylum, followed by Firmicutes (10%) and Actinobacteria (2%), with 37% of the bacteria unclassified. Bradyrhizobium was the dominant genus on corroded DCI pipe, while Sphingomonas was predominant on non-corroded STS pipe. Methylobacterium and Afipia were detected only in the reactor without added phosphate. PCR-DGGE analysis showed that the diversity of species in biofilm tended to increase when phosphate was added regardless of the pipe material, indicating that phosphate addition upset the biological stability in the plumbing systems.

Effect of pipe corrosion scales on chlorine dioxide consumption in drinking water distribution systems.

PubMed

Zhang, Zhe; Stout, Janet E; Yu, Victor L; Vidic, Radisav

2008-01-01

Previous studies showed that temperature and total organic carbon in drinking water would cause chlorine dioxide (ClO(2)) loss in a water distribution system and affect the efficiency of ClO(2) for Legionella control. However, among the various causes of ClO(2) loss in a drinking water distribution system, the loss of disinfectant due to the reaction with corrosion scales has not been studied in detail. In this study, the corrosion scales from a galvanized iron pipe and a copper pipe that have been in service for more than 10 years were characterized by energy dispersive spectroscopy (EDS) and X-ray diffraction (XRD). The impact of these corrosion scale materials on ClO(2) decay was investigated in de-ionized water at 25 and 45 degrees C in a batch reactor with floating glass cover. ClO(2) decay was also investigated in a specially designed reactor made from the iron and copper pipes to obtain more realistic reaction rate data. Goethite (alpha-FeOOH) and magnetite (Fe(3)O(4)) were identified as the main components of iron corrosion scale. Cuprite (Cu(2)O) was identified as the major component of copper corrosion scale. The reaction rate of ClO(2) with both iron and copper oxides followed a first-order kinetics. First-order decay rate constants for ClO(2) reactions with iron corrosion scales obtained from the used service pipe and in the iron pipe reactor itself ranged from 0.025 to 0.083 min(-1). The decay rate constant for ClO(2) with Cu(2)O powder and in the copper pipe reactor was much smaller and it ranged from 0.0052 to 0.0062 min(-1). Based on these results, it can be concluded that the corrosion scale will cause much more significant ClO(2) loss in corroded iron pipes of the distribution system than the total organic carbon that may be present in finished water.

Heat pipes for use in a magnetic field

DOEpatents

Werner, Richard W.; Hoffman, Myron A.

1983-01-01

A heat pipe configuration for use in a magnetic field environment of a fusion reactor. Heat pipes for operation in a magnetic field when liquid metal working fluids are used are optimized by flattening of the heat pipes having an unobstructed annulus which significantly reduces the adverse side region effect of the prior known cylindrically configured heat pipes. The flattened heat pipes operating in a magnetic field can remove 2--3 times the heat as a cylindrical heat pipe of the same cross sectional area.

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.

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

Kupca, L.; Beno, P.

A very brief summary is provided of a primary circuit piping material properties analysis. The analysis was performed for the Bohunice V-1 reactor and the Kola-1 and -2 reactors. Assessment was performed on Bohunice V-1 archive materials and primary piping material cut from the Kola units after 100,000 hours of operation. Main research program tasks included analysis of mechanical properties, corrosion stability, and microstructural properties. Analysis results are not provided.

Heat-transfer analysis of double-pipe heat exchangers for indirect-cycle SCW NPP

NASA Astrophysics Data System (ADS)

Thind, Harwinder

SuperCritical-Water-cooled Reactors (SCWRs) are being developed as one of the Generation-IV nuclear-reactor concepts. SuperCritical Water (SCW) Nuclear Power Plants (NPPs) are expected to have much higher operating parameters compared to current NPPs, i.e., pressure of about 25 MPa and outlet temperature up to 625 °C. This study presents the heat transfer analysis of an intermediate Heat exchanger (HX) design for indirect-cycle concepts of Pressure-Tube (PT) and Pressure-Vessel (PV) SCWRs. Thermodynamic configurations with an intermediate HX gives a possibility to have a single-reheat option for PT and PV SCWRs without introducing steam-reheat channels into a reactor. Similar to the current CANDU and Pressurized Water Reactor (PWR) NPPs, steam generators separate the primary loop from the secondary loop. In this way, the primary loop can be completely enclosed in a reactor containment building. This study analyzes the heat transfer from a SCW primary (reactor) loop to a SCW and Super-Heated Steam (SHS) secondary (turbine) loop using a double-pipe intermediate HX. The numerical model is developed with MATLAB and NIST REFPROP software. Water from the primary loop flows through the inner pipe, and water from the secondary loop flows through the annulus in the counter direction of the double-pipe HX. The analysis on the double-pipe HX shows temperature and profiles of thermophysical properties along the heated length of the HX. It was found that the pseudocritical region has a significant effect on the temperature profiles and heat-transfer area of the HX. An analysis shows the effect of variation in pressure, temperature, mass flow rate, and pipe size on the pseudocritical region and the heat-transfer area of the HX. The results from the numerical model can be used to optimize the heat-transfer area of the HX. The higher pressure difference on the hot side and higher temperature difference between the hot and cold sides reduces the pseudocritical-region length, thus decreases the heat-transfer surface area of the HX.

63. REACTOR CHAMBER (BASF) FROM NORTH SHOWING NEUTRON SHIELD TANK ...

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

63. REACTOR CHAMBER (BASF) FROM NORTH SHOWING NEUTRON SHIELD TANK AND REACTOR PIPING (LOCATION RRR) - Shippingport Atomic Power Station, On Ohio River, 25 miles Northwest of Pittsburgh, Shippingport, Beaver County, PA

Heat pipes for use in a magnetic field

DOEpatents

Werner, R.W.; Hoffman, M.A.

1983-07-19

A heat pipe configuration for use in a magnetic field environment of a fusion reactor is disclosed. Heat pipes for operation in a magnetic field when liquid metal working fluids are used are optimized by flattening of the heat pipes having an unobstructed annulus which significantly reduces the adverse side region effect of the prior known cylindrically configured heat pipes. The flattened heat pipes operating in a magnetic field can remove 2--3 times the heat as a cylindrical heat pipe of the same cross sectional area. 4 figs.

Heat dissipating nuclear reactor

DOEpatents

Hunsbedt, A.; Lazarus, J.D.

1985-11-21

Disclosed is a nuclear reactor containment adapted to retain and cool core debris in the unlikely event of a core meltdown and subsequent breach in the reactor vessel. The reactor vessel is seated in a cavity which has a thick metal sidewall that is integral with a thick metal basemat at the bottom of the cavity. The basemat extends beyond the perimeter of the cavity sidewall. Underneath the basemat is a porous bed with water pipes and steam pipes running into it. Water is introduced into the bed and converted into steam which is vented to the atmosphere. A plurality of metal pilings in the form of H-beams extend from the metal base plate downwardly and outwardly into the earth.

Heat dissipating nuclear reactor

DOEpatents

Hunsbedt, Anstein; Lazarus, Jonathan D.

1987-01-01

Disclosed is a nuclear reactor containment adapted to retain and cool core debris in the unlikely event of a core meltdown and subsequent breach in the reactor vessel. The reactor vessel is seated in a cavity which has a thick metal sidewall that is integral with a thick metal basemat at the bottom of the cavity. The basemat extends beyond the perimeter of the cavity sidewall. Underneath the basemat is a porous bed with water pipes and steam pipes running into it. Water is introduced into the bed and converted into steam which is vented to the atmosphere. A plurality of metal pilings in the form of H-beams extends from the metal base plate downwardly and outwardly into the earth.

Piping support system for liquid-metal fast-breeder reactor

DOEpatents

Brussalis, Jr., William G.

1984-01-01

A pipe support consisting of a rigid link pivotally attached to a pipe and an anchor, adapted to generate stress or strain in the link and pipe due to pipe thermal movement, which stress or strain can oppose further pipe movement and generally provides pipe support. The pipe support can be used in multiple combinations with other pipe supports to form a support system. This support system is most useful in applications in which the pipe is normally operated at a constant elevated or depressed temperature such that desired stress or strain can be planned in advance of pipe and support installation. The support system is therefore especially useful in steam stations and in refrigeration equipment.

75 FR 36698 - Draft Regulatory Guide: Issuance, Availability

Federal Register 2010, 2011, 2012, 2013, 2014

2010-06-28

... information based on the likelihood of pipe breaks of different sizes. The rule would divide all coolant... to and including a ``transition break size,'' and breaks larger than the transition size up to the largest pipe in the reactor coolant system. Selection of the transition size was based upon pipe break...

High Temperature Water Heat Pipes Radiator for a Brayton Space Reactor Power System

NASA Astrophysics Data System (ADS)

El-Genk, Mohamed S.; Tournier, Jean-Michel

2006-01-01

A high temperature water heat pipes radiator design is developed for a space power system with a sectored gas-cooled reactor and three Closed Brayton Cycle (CBC) engines, for avoidance of single point failures in reactor cooling and energy conversion and rejection. The CBC engines operate at turbine inlet and exit temperatures of 1144 K and 952 K. They have a net efficiency of 19.4% and each provides 30.5 kWe of net electrical power to the load. A He-Xe gas mixture serves as the turbine working fluid and cools the reactor core, entering at 904 K and exiting at 1149 K. Each CBC loop is coupled to a reactor sector, which is neutronically and thermally coupled, but hydraulically decoupled to the other two sectors, and to a NaK-78 secondary loop with two water heat pipes radiator panels. The segmented panels each consist of a forward fixed segment and two rear deployable segments, operating hydraulically in parallel. The deployed radiator has an effective surface area of 203 m2, and when the rear segments are folded, the stowed power system fits in the launch bay of the DELTA-IV Heavy launch vehicle. For enhanced reliability, the water heat pipes operate below 50% of their wicking limit; the sonic limit is not a concern because of the water, high vapor pressure at the temperatures of interest (384 - 491 K). The rejected power by the radiator peaks when the ratio of the lengths of evaporator sections of the longest and shortest heat pipes is the same as that of the major and minor widths of the segments. The shortest and hottest heat pipes in the rear segments operate at 491 K and 2.24 MPa, and each rejects 154 W. The longest heat pipes operate cooler (427 K and 0.52 MPa) and because they are 69% longer, reject more power (200 W each). The longest and hottest heat pipes in the forward segments reject the largest power (320 W each) while operating at ~ 46% of capillary limit. The vapor temperature and pressure in these heat pipes are 485 K and 1.97 MPa. By contrast, the shortest water heat pipes in the forward segments operate much cooler (427 K and 0.52 MPa), and reject a much lower power of 45 W each. The radiator with six fixed and 12 rear deployable segments rejects a total of 324 kWth, weights 994 kg and has an average specific power of 326 Wth/kg and a specific mass of 5.88 kg/m2.

Safe Affordable Fission Engine-(SAFE-) 100a Heat Exchanger Thermal and Structural Analysis

NASA Technical Reports Server (NTRS)

Steeve, B. E.

2005-01-01

A potential fission power system for in-space missions is a heat pipe-cooled reactor coupled to a Brayton cycle. In this system, a heat exchanger (HX) transfers the heat of the reactor core to the Brayton gas. The Safe Affordable Fission Engine- (SAFE-) 100a is a test program designed to thermally and hydraulically simulate a 95 Btu/s prototypic heat pipe-cooled reactor using electrical resistance heaters on the ground. This Technical Memorandum documents the thermal and structural assessment of the HX used in the SAFE-100a program.

Reactor Materials Program - Baseline Material Property Handbook - Mechanical Properties of 1950's Vintage Stainless Steel Weldment Components, Task Number 89-23-A-1

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

Stoner, K.J.

1999-11-05

The Process Water System (primary coolant) piping of the nuclear production reactors constructed in the 1950''s at Savannah River Site is comprised primarily of Type 304 stainless steel with Type 308 stainless steel weld filler. A program to measure the mechanical properties of archival PWS piping and weld materials (having approximately six years of service at temperatures between 25 and 100 degrees C) has been completed. The results from the mechanical testing has been synthesized to provide a mechanical properties database for structural analyses of the SRS piping.

Effect of reactor coolant radioactivity upon configuration feasibility for a nuclear electric propulsion vehicle

NASA Technical Reports Server (NTRS)

Soffer, L.; Wright, G. N.

1973-01-01

A preliminary shielding analysis was carried out for a conceptual nuclear electric propulsion vehicle designed to transport payloads from low earth orbit to synchronous orbit. The vehicle employed a thermionic nuclear reactor operating at 1575 kilowatts and generated 120 kilowatts of electricity for a round-trip mission time of 2000 hours. Propulsion was via axially directed ion engines employing 3300 pounds of mercury as a propellant. The vehicle configuration permitted a reactor shadow shield geometry using LiH and the mercury propellant for shielding. However, much of the radioactive NaK reactor coolant was unshielded and in close proximity to the power conditioning electronics. An estimate of the radioactivity of the NaK coolant was made and its unshielded dose rate to the power conditioning equipment calculated. It was found that the activated NaK contributed about three-fourths of the gamma dose constraint. The NaK dose was considered a sufficiently high fraction of the allowable gamma dose to necessitate modifications in configuration.

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

Kornfeldt, H.; Bjoerk, K.O.; Ekstroem, P.

The protection against dynamic effects in connection with potential pipe breaks has been implemented in different ways in the development of BWR reactor designs. First-generation plant designs reflect code requirements in effect at that time which means that no piping restraint systems were designed and built into those plants. Modern designs have, in contrast, implemented full protection against damage in connection with postulated pipe breaks, as required in current codes and regulations. Moderns standards and current regulatory demands can be met for the older plants by backfitting pipe whip restraint hardware. This could lead to several practical difficulties as thesemore » installations were not anticipated in the original plant design and layout. Meeting the new demands by analysis would in this situation have great advantages. Application of leak-before-break criteria gives an alternative opportunity of meeting modem standards in reactor safety design. Analysis takes into account data specific to BWR primary system operation, actual pipe material properties, piping loads and leak detection capability. Special attention must be given to ensure that the data used reflects actual plant conditions.« less

TOPAZ II Anti-Criticality Device Rapid Prototype

NASA Astrophysics Data System (ADS)

Campbell, Donald R.; Otting, William D.

1994-07-01

The Ballistic Missile Defense Organization (BMDO) has been working on a Nuclear Electric Propulsion Space Test Project (NEPSTP) using an existing Russian Topaz II reactor system to power the NEPSTP satellite. Safety investigations have shown that it will be possible to safely launch the Topaz II system in the United States with some modification to preclude water flooded criticality. A ``fuel-out'' water subcriticality concept was selected by the Los Alamos National Laboratory (LANL) as the baseline concept. A fuel-out anti-criticality device (ACD) conceptual design was developed by Rockwell. The concept functions to hold the fuel from the four centermost thermionic fuel elements (TFEs) outside the reactor during launch and reliably inserts the fuel into the reactor once the operational orbit is achieved. A four-tenths scale ACD rapid prototype model, fabricated from the CATIA solids design model, clearly shows in three dimensions the relative size and spatial relationship of the ACD components.

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.

The use of dual mode thermionic reactors in supporting Earth orbital and space exploration missions

NASA Astrophysics Data System (ADS)

Zubrin, Robert M.; Sulmeisters, Tal K.

1993-01-01

Missions requiring large amounts of electric power to support their payload functions can be enabled through the employment of nuclear electric power reactors, which in some cases can also assist the mission by making possible the employment of high specific impulse electric propulsion. However it is found that the practicality and versality of using a power reactor to provide advanced propulsion is enormously enhanced if the reactor is configured in such a way to allow it to generate a certain amount of direct thrust as well. The use of such a system allows the creation of a common bus upper stage that can provide both high power and high impulse (with short orbit transfer times). It is shown that such a system, termed an Integral Power and Propulsion Stage (IPAPS), is optimal for supporting many Earth, Lunar, planetary and asteroidal observation, exploration, and communication support missions, and it is therefore recommended that the nuclear power reactor ultimately selected by the government for development and production be one that can be configured for such a function.

Megawatt Class Nuclear Space Power Systems (MCNSPS) conceptual design and evaluation report. Volume 3, technologies 2: Power conversion

NASA Technical Reports Server (NTRS)

Wetch, J. R.

1988-01-01

The major power conversion concepts considered for the Megawatt Class Nuclear Space Power System (MCNSPS) are discussed. These concepts include: (1) Rankine alkali-metal-vapor turbine alternators; (2) in-core thermionic conversion; (3) Brayton gas turbine alternators; and (4) free piston Stirling engine linear alternators. Considerations important to the coupling of these four conversion alternatives to an appropriate nuclear reactor heat source are examined along with the comparative performance characteristics of the combined systems meeting MCNSPS requirements.

Static and dynamic high power, space nuclear electric generating systems

NASA Technical Reports Server (NTRS)

Wetch, J. R.; Begg, L. L.; Koester, J. K.

1985-01-01

Space nuclear electric generating systems concepts have been assessed for their potential in satisfying future spacecraft high power (several megawatt) requirements. Conceptual designs have been prepared for reactor power systems using the most promising static (thermionic) and the most promising dynamic conversion processes. Component and system layouts, along with system mass and envelope requirements have been made. Key development problems have been identified and the impact of the conversion process selection upon thermal management and upon system and vehicle configuration is addressed.

Transient thermohydraulic heat pipe modeling

NASA Astrophysics Data System (ADS)

Hall, Michael L.; Doster, Joseph M.

Many space based reactor designs employ heat pipes as a means of conveying heat. In these designs, thermal radiation is the principle means for rejecting waste heat from the reactor system, making it desirable to operate at high temperatures. Lithium is generally the working fluid of choice as it undergoes a liquid-vapor transformation at the preferred operating temperature. The nature of remote startup, restart, and reaction to threats necessitates an accurate, detailed transient model of the heat pipe operation. A model is outlined of the vapor core region of the heat pipe which is part of a large model of the entire heat pipe thermal response. The vapor core is modeled using the area averaged Navier-Stokes equations in one dimension, which take into account the effects of mass, energy and momentum transfer. The core model is single phase (gaseous), but contains two components: lithium gas and a noncondensible vapor. The vapor core model consists of the continuity equations for the mixture and noncondensible, as well as mixture equations for internal energy and momentum.

Application of Molten Salt Reactor Technology to Nuclear Electric Propulsion Mission

NASA Technical Reports Server (NTRS)

Patton, Bruce; Sorensen, Kirk; Rodgers, Stephen L. (Technical Monitor)

2002-01-01

Nuclear electric propulsion (NEP) and planetary surface power missions require reactors that are lightweight, operationally robust, and scalable in power for widely varying scientific mission objectives. Molten salt reactor technology meets all of these requirements and offers an interesting alternative to traditional gas cooled, liquid metal, and heat pipe space reactors.

Space Nuclear Reactor Engineering

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

Poston, David Irvin

We needed to find a space reactor concept that could be attractive to NASA for flight and proven with a rapid turnaround, low-cost nuclear test. Heat-pipe-cooled reactors coupled to Stirling engines long identified as the easiest path to near-term, low-cost concept.

Sodium Heat Pipe Module Processing For the SAFE-100 Reactor Concept

NASA Technical Reports Server (NTRS)

Martin, James; Salvail, Pat

2003-01-01

To support development and hardware-based testing of various space reactor concepts, the Early Flight Fission-Test Facility (EFF-TF) team established a specialized glove box unit with ancillary systems to handle/process alkali metals. Recently, these systems have been commissioned with sodium supporting the fill of stainless steel heat pipe modules for use with a 100 kW thermal heat pipe reactor design. As part of this effort, procedures were developed and refined to govern each segment of the process covering: fill, leak check, vacuum processing, weld closeout, and final "wet in". A series of 316 stainless steel modules, used as precursors to the actual 321 stainless steel modules, were filled with 35 +/- 1 grams of sodium using a known volume canister to control the dispensed mass. Each module was leak checked to less than10(exp -10) std cc/sec helium and vacuum conditioned at 250 C to assist in the removal of trapped gases. A welding procedure was developed to close out the fill stem preventing external gases from entering the evacuated module. Finally the completed modules were vacuum fired at 750 C allowing the sodium to fully wet the internal surface and wick structure of the heat pipe module.

Sodium Heat Pipe Module Processing For the SAFE-100 Reactor Concept

NASA Astrophysics Data System (ADS)

Martin, James; Salvail, Pat

2004-02-01

To support development and hardware-based testing of various space reactor concepts, the Early Flight Fission-Test Facility (EFF-TF) team established a specialized glove box unit with ancillary systems to handle/process alkali metals. Recently, these systems have been commissioned with sodium supporting the fill of stainless steel heat pipe modules for use with a 100 kW thermal heat pipe reactor design. As part of this effort, procedures were developed and refined to govern each segment of the process covering: fill, leak check, vacuum processing, weld closeout, and final ``wet in''. A series of 316 stainless steel modules, used as precursors to the actual 321 stainless steel modules, were filled with 35 +/-1 grams of sodium using a known volume canister to control the dispensed mass. Each module was leak checked to <10-10 std cc/sec helium and vacuum conditioned at 250 °C to assist in the removal of trapped gases. A welding procedure was developed to close out the fill stem preventing external gases from entering the evacuated module. Finally the completed modules were vacuum fired at 750 °C allowing the sodium to fully wet the internal surface and wick structure of the heat pipe module.

Nuclear reactor heat transport system component low friction support system

DOEpatents

Wade, Elman E.

1980-01-01

A support column for a heavy component of a liquid metal fast breeder reactor heat transport system which will deflect when the pipes leading coolant to and from the heavy component expand or contract due to temperature changes includes a vertically disposed pipe, the pipe being connected to the heavy component by two longitudinally spaced cycloidal dovetail joints wherein the distal end of each of the dovetails constitutes a part of the surface of a large diameter cylinder and the centerlines of these large diameter cylinders intersect at right angles and the pipe being supported through two longitudinally spaced cycloidal dovetail joints wherein the distal end of each of the dovetails constitutes a part of the surface of a large diameter cylinder and the centerlines of these large diameter cylinders intersect at right angles, each of the cylindrical surfaces bearing on a flat and horizontal surface.

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

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

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

150. ARAIII Reactor building (ARA608) Sections. Show highbay section, heater ...

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

150. ARA-III Reactor building (ARA-608) Sections. Show high-bay section, heater stack, and depth of reactor, piping, and heater pits. Aerojet-general 880-area/GCRE-608-A-3. Date: February 1958. Ineel index code no. 063-0608-00-013-102613. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

Engineering design aspects of the heat-pipe power system

NASA Technical Reports Server (NTRS)

Capell, B. M.; Houts, M. G.; Poston, D. I.; Berte, M.

1997-01-01

The Heat-pipe Power System (HPS) is a near-term, low-cost space power system designed at Los Alamos that can provide up to 1,000 kWt for many space nuclear applications. The design of the reactor is simple, modular, and adaptable. The basic design allows for the use of a variety of power conversion systems and reactor materials (including the fuel, clad, and heat pipes). This paper describes a project that was undertaken to develop a database supporting many engineering aspects of the HPS design. The specific tasks discussed in this paper are: the development of an HPS materials database, the creation of finite element models that will allow a wide variety of investigations, and the verification of past calculations.

Heat Pipe Technology: A bibliography with abstracts

NASA Technical Reports Server (NTRS)

1974-01-01

This bibliography lists 149 references with abstracts and 47 patents dealing with applications of heat pipe technology. Topics covered include: heat exchangers for heat recovery; electrical and electronic equipment cooling; temperature control of spacecraft; cryosurgery; cryogenic, cooling; nuclear reactor heat transfer; solar collectors; laser mirror cooling; laser vapor cavitites; cooling of permafrost; snow melting; thermal diodes variable conductance; artery gas venting; and venting; and gravity assisted pipes.

Alkali metal vapors - Laser spectroscopy and applications

NASA Technical Reports Server (NTRS)

Stwalley, W. C.; Koch, M. E.

1980-01-01

The paper examines the rapidly expanding use of lasers for spectroscopic studies of alkali metal vapors. Since the alkali metals (lithium, sodium, potassium, rubidium and cesium) are theoretically simple ('visible hydrogen'), readily ionized, and strongly interacting with laser light, they represent ideal systems for quantitative understanding of microscopic interconversion mechanisms between photon (e.g., solar or laser), chemical, electrical and thermal energy. The possible implications of such understanding for a wide variety of practical applications (sodium lamps, thermionic converters, magnetohydrodynamic devices, new lasers, 'lithium waterfall' inertial confinement fusion reactors, etc.) are also discussed.

90. ARAIII. GCRE reactor building (ARA608) mechanical loop pit. Shows ...

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

90. ARA-III. GCRE reactor building (ARA-608) mechanical loop pit. Shows nitrogen gas compressor in foreground, piping installations on walls of pit, and other details. February 24, 1959. Ineel photo no. 59-880. Photographer: Ken Mansfield. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

Exploring the engineering limit of heat flux of a W/RAFM divertor target for fusion reactors

NASA Astrophysics Data System (ADS)

Mao, X.; Fursdon, M.; Chang, X. B.; Zhang, J. W.; Liu, P.; Ellwood, G.; Qian, X. Y.; Qin, S. J.; Peng, X. B.; Barrett, T. R.; Liu, P.

2018-06-01

The design and development of a fusion reactor divertor plasma facing component (PFC) is one of the many challenging issues on the road to commercial use of fusion energy. The divertor PFC is expected to exhaust steady state heat loads in the region of 10 MW m‑2 while keeping temperatures and thermo-mechanical stresses in its structure within the allowable limits. For ITER (International Thermo-Nuclear Experimental Reactor) a water cooled W/CuCrZr divertor PFC concept has been developed. However, this concept is not necessarily assured for use in future fusion reactors mainly because the neutron radiation dose would be at least an order magnitude higher, resulting in limited thermo-mechanical performance and considerably more activated waste products. In the present study, a water cooled divertor PFC using reduced activation ferritic-martensitic (RAFM) steel as the heat sink pipe has been designed with pressurised water reactor-like cooling conditions (pressure of 15.5 MPa, velocity of 10–20 m s‑1 and temperature of 300 °C). The PFC is made up of a number of rectangular tungsten tiles, each with an inner circular hole (so-called monoblocks), joined onto a RAFM steel pipe with copper interlayers. The thermo-mechanical performance of the PFC has been studied in detail. The heat transfer coefficient between the RAFM pipe inner surface and the water was calculated using published correlations. Geometric parameters and water velocity were optimized with finite element (FE) thermal analysis, to achieve acceptable temperatures in the structure given the target exhaust heat load of 10 MW m‑2. Under this heat load and the optimised thermal design parameters, the structure of the PFC was further assessed by mechanical analysis. We find that under these conditions the RAFM steel pipe experiences cyclic plasticity, and fails the common linear elastic ratchetting (3 Sm) rule. Nevertheless, the designed W/RAFM divertor PFU can withstand 10 MW m‑2 heat load, albeit with a fatigue life of approximately 0.55 years based on the expected operation scenario of a prototype or test reactor. This study extends the state of knowledge of the technological limit of a divertor based on a RAFM steel pipe structure.

Screening reactor steam/water piping systems for water hammer

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

Griffith, P.

1997-09-01

A steam/water system possessing a certain combination of thermal, hydraulic and operational states, can, in certain geometries, lead to a steam bubble collapse induced water hammer. These states, operations, and geometries are identified. A procedure that can be used for identifying whether an unbuilt reactor system is prone to water hammer is proposed. For the most common water hammer, steam bubble collapse induced water hammer, six conditions must be met in order for one to occur. These are: (1) the pipe must be almost horizontal; (2) the subcooling must be greater than 20 C; (3) the L/D must be greatermore » than 24; (4) the velocity must be low enough so that the pipe does not run full, i.e., the Froude number must be less than one; (5) there should be void nearby; (6) the pressure must be high enough so that significant damage occurs, that is the pressure should be above 10 atmospheres. Recommendations on how to avoid this kind of water hammer in both the design and the operation of the reactor system are made.« less

Nuclear reactor system study for NASA/JPL

NASA Technical Reports Server (NTRS)

Palmer, R. G.; Lundberg, L. B.; Keddy, E. S.; Koenig, D. R.

1982-01-01

Reactor shielding, safety studies, and heat pipe development work are described. Monte Carlo calculations of gamma and neutron shield configurations show that substantial weight penalties are incurred if exposure at 25 m to neutrons and gammas must be limited to 10 to the 12th power nvt and 10 to the 6th power rad, instead of the 10 to the 13th power nvt and 10 to the 7th power rad values used earlier. For a 1.6 MW sub t reactor, the required shield weight increases from 400 to 815 kg. Water immersion critically calculations were extended to study the effect of water in fuel void spaces as well as in the core heat pipes. These show that the insertion into the core of eight blades of B4C with a mass totaling 2.5 kg will guarantee subcriticality. The design, fabrication procedure, and testing of a 4m long molybdenum/lithium heat pipe are described. It appears that an excess of oxygen in the wick prevented the attainment of expected performance capability.

Uranium nitride behavior at thermionic temperatures

NASA Technical Reports Server (NTRS)

Phillips, W. M.

1973-01-01

The feasibility of using uranium nitride for in-core thermionic applications was evaluated in electrically heated thermal gradient tests and in flat plate thermionic converters. These tests indicated that grain boundary penetration of uranium nitride into both tungsten and rhenium will occur under thermal gradient conditions. In the case of the tungsten thermionic converter, this led to grain boundary rupture of the emitter and almost total loss of electrical output from the converter. It appears that uranium nitride is unsuitable for thermionic applications at the 2000 K temperatures used in these tests.

Small Reactor for Deep Space Exploration

ScienceCinema

none,

2018-06-06

This is the first demonstration of a space nuclear reactor system to produce electricity in the United States since 1965, and an experiment demonstrated the first use of a heat pipe to cool a small nuclear reactor and then harvest the heat to power a Stirling engine at the Nevada National Security Site's Device Assembly Facility confirms basic nuclear reactor physics and heat transfer for a simple, reliable space power system.

Preliminary Design of a SP-100/Stirling Radiatively Coupled Heat Exchanger

NASA Technical Reports Server (NTRS)

Schmitz, Paul; Tower, Leonard; Dawson, Ronald; Blue, Brian; Dunn, Pat

1995-01-01

Several methods for coupling the SP-100 space nuclear reactor to the NASA Lewis Research Center's Free Piston Stirling Power Convertor (FPSPC) are presented. A 25 kWe, dual opposed Stirling convertor configuration is used in these designs. The concepts use radiative coupling between the SP-100 lithium loop and the sodium heat pipe of the Stirling convertor to transfer the heat from the reactor to the convertor. Four separate configurations are presented. Masses for the four designs vary from 41 to 176 kgs. Each design's structure, heat transfer characteristics, and heat pipe performance are analytically modeled.

Preliminary design of a SP-100/Stirling radiatively coupled heat exchanger

NASA Astrophysics Data System (ADS)

Schmitz, Paul; Tower, Leonard; Dawson, Ronald; Blue, Brian; Dunn, Pat

1995-10-01

Several methods for coupling the SP-100 space nuclear reactor to the NASA Lewis Research Center's Free Piston Stirling Power Convertor (FPSPC) are presented. A 25 kWe, dual opposed Stirling convertor configuration is used in these designs. The concepts use radiative coupling between the SP-100 lithium loop and the sodium heat pipe of the Stirling convertor to transfer the heat from the reactor to the convertor. Four separate configurations are presented. Masses for the four designs vary from 41 to 176 kgs. Each design's structure, heat transfer characteristics, and heat pipe performance are analytically modeled.

Thermionic photovoltaic energy converter

NASA Technical Reports Server (NTRS)

Chubb, D. L. (Inventor)

1985-01-01

A thermionic photovoltaic energy conversion device comprises a thermionic diode mounted within a hollow tubular photovoltaic converter. The thermionic diode maintains a cesium discharge for producing excited atoms that emit line radiation in the wavelength region of 850 nm to 890 nm. The photovoltaic converter is a silicon or gallium arsenide photovoltaic cell having bandgap energies in this same wavelength region for optimum cell efficiency.

Peach Bottom Atomic Power Station recirc pipe dose rates with zinc injection and condenser replacement

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

DiCello, D.C.; Odell, A.D.; Jackson, T.J.

1995-03-01

Peach Bottom Atomic Power Station (PBAPS) is located near the town of Delta, Pennsylvania, on the west bank of the Susquehanna River. It is situated approximately 20 miles south of Lancaster, Pennsylvania. The site contains two boiling water reactors of General Electric design and each rated at 3,293 megawatts thermal. The units are BWR 4s and went commercial in 1977. There is also a decommissioned high temperature gas-cooled reactor on site, Unit 1. PBAPS Unit 2 recirc pipe was replaced in 1985 and Unit 3 recirc pipes replaced in 1988 with 326 NGSS. The Unit 2 replacement pipe was electropolished,more » and the Unit 3 pipe was electropolished and passivated. The Unit 2 brass condenser was replaced with a Titanium condenser in the first quarter of 1991, and the Unit 3 condenser was replaced in the fourth quarter of 1991. The admiralty brass condensers were the source of natural zinc in both units. Zinc injection was initiated in Unit 2 in May 1991, and in Unit 3 in May 1992. Contact dose rate measurements were made in standard locations on the 28-inch recirc suction and discharge lines to determine the effectiveness of zinc injection and to monitor radiation build-up in the pipe. Additionally, HPGe gamma scans were performed to determine the isotopic composition of the oxide layer inside the pipe. In particular, the specific ({mu}Ci/cm{sup 2}) of Co-60 and Zn-65 were analyzed.« less

Nuclear Design of the HOMER-15 Mars Surface Fission Reactor

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

Poston, David I.

2002-07-01

The next generation of robotic missions to Mars will most likely require robust power sources in the range of 3 to 20 kWe. Fission systems are well suited to provide safe, reliable, and economic power within this range. The goal of this study is to design a compact, low-mass fission system that meets Mars surface power requirements, while maintaining a high level of safety and reliability at a relatively low cost. The Heat pipe Power System (HPS) is one possible approach for producing near-term, low-cost, space fission power. The goal of the HPS project is to devise an attractive spacemore » fission system that can be developed quickly and affordably. The primary ways of doing this are by using existing technology and by designing the system for inexpensive testing. If the system can be designed to allow highly prototypic testing with electrical heating, then an exhaustive test program can be carried out quickly and inexpensively, and thorough testing of the actual flight unit can be performed - which is a major benefit to reliability. Over the past 4 years, three small HPS proof-of-concept technology demonstrations have been conducted, and each has been highly successful. The Heat pipe-Operated Mars Exploration Reactor (HOMER) is a derivative of the HPS designed especially for producing power on the surface of Mars. The HOMER-15 is a 15-kWt reactor that couples with a 3-kWe Stirling engine power system. The reactor contains stainless-steel (SS)-clad uranium nitride (UN) fuel pins that are structurally and thermally bonded to SS/sodium heat pipes. Fission energy is conducted from the fuel pins to the heat pipes, which then carry the heat to the Stirling engine. This paper describes conceptual design and nuclear performance the HOMER-15 reactor. (author)« less

ETR HEAT EXCHANGER BUILDING, TRA644. WORKERS ARE INSTALLING HEAT EXCHANGER ...

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

ETR HEAT EXCHANGER BUILDING, TRA-644. WORKERS ARE INSTALLING HEAT EXCHANGER PIPING. INL NEGATIVE NO. 56-3122. Jack L. Anderson, Photographer, 9/21/1956 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

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

DTIC Science & Technology

2009-06-01

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

Cesium vapor thermionic converter anomalies arising from negative ion emission

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

Rasor, Ned S., E-mail: ned.rasor@gmail.com

2016-08-14

Compelling experimental evidence is given that a longstanding limit encountered on cesium vapor thermionic energy converter performance improvement and other anomalies arise from thermionic emission of cesium negative ions. It is shown that the energy that characterizes thermionic emission of cesium negative ions is 1.38 eV and, understandably, is not the electron affinity 0.47 eV determined for the photodetachment threshold of the cesium negative ion. The experimental evidence includes measurements of collector work functions and volt-ampere characteristics in quasi-vacuum cesium vapor thermionic diodes, along with reinterpretation of the classic Taylor-Langmuir S-curve data on electron emission in cesium vapor. The quantitative effects ofmore » negative ion emission on performance in the ignited, unignited, and quasi-vacuum modes of cesium vapor thermionic converter operation are estimated.« less

Modeling of transient heat pipe operation

NASA Technical Reports Server (NTRS)

Colwell, Gene T.

1987-01-01

The use of heat pipes is being considered as a means of reducing the peak temperature and large thermal gradients at the leading edges of reentry vehicles and hypersonic aircraft and in nuclear reactors. In the basic cooling concept, the heat pipe covers the leading edge, a portion of the lower wing surface, and a portion of the upper wing surface. Aerodynamic heat is mainly absorbed at the leading edge and transported through the heat pipe to the upper and lower wing surface, where it is rejected by thermal radiation and convection. Basic governing equations are written to determine the startup, transient, and steady state performance of a haet pipe which has initially frozen alkali-metal as the working fluid.

The TEF modeling and analysis approach to advance thermionic space power technology

NASA Astrophysics Data System (ADS)

Marshall, Albert C.

1997-01-01

Thermionics space power systems have been proposed as advanced power sources for future space missions that require electrical power levels significantly above the capabilities of current space power systems. The Defense Special Weapons Agency's (DSWA) Thermionic Evaluation Facility (TEF) is carrying out both experimental and analytical research to advance thermionic space power technology to meet this expected need. A Modeling and Analysis (M&A) project has been created at the TEF to develop analysis tools, evaluate concepts, and guide research. M&A activities are closely linked to the TEF experimental program, providing experiment support and using experimental data to validate models. A planning exercise has been completed for the M&A project, and a strategy for implementation was developed. All M&A activities will build on a framework provided by a system performance model for a baseline Thermionic Fuel Element (TFE) concept. The system model is composed of sub-models for each of the system components and sub-systems. Additional thermionic component options and model improvements will continue to be incorporated in the basic system model during the course of the program. All tasks are organized into four focus areas: 1) system models, 2) thermionic research, 3) alternative concepts, and 4) documentation and integration. The M&A project will provide a solid framework for future thermionic system development.

Solid-State Thermionic Power Generators: An Analytical Analysis in the Nonlinear Regime

NASA Astrophysics Data System (ADS)

Zebarjadi, M.

2017-07-01

Solid-state thermionic power generators are an alternative to thermoelectric modules. In this paper, we develop an analytical model to investigate the performance of these generators in the nonlinear regime. We identify dimensionless parameters determining their performance and provide measures to estimate an acceptable range of thermal and electrical resistances of thermionic generators. We find the relation between the optimum load resistance and the internal resistance and suggest guidelines for the design of thermionic power generators. Finally, we show that in the nonlinear regime, thermionic power generators can have efficiency values higher than the state-of-the-art thermoelectric modules.

A SPACESHIP WITH NUCLEAR PROPULSION

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

Polorny, J.

1962-01-01

ABS>A proposed space vehicle with nuclear propulsion for a round-trip Martian mission is described. It would be powered by a 270-Mw graphite- moderated, U-fueled nuclear reactor with a core 1 m high by 1 m in diameter, and use gas as propellant. The gas would be heated to the maximum temperature in the reactor and additionally accelerated by an electromagnetic field. To this end, small quantities of K would be injected into the gas stream to increase its electric conductivity. The required electrical energy would be produced by liquid-Na-cooled thermionic converters. The vehicle would weigh 115000 kg, including 43000 kgmore » of H propellant with tankage, and 7000 kg of sustenance material for one year. Chemical rockets would launch the vehicle with a crew of three men into an earth orbit where nuclear propulsion would take over. Upon reactor start-up, three heat exchangers (minimum dimensions 30 x 18 m) would be fanned out. A shielded well with a diameter of 2.5 m would protect the crew from radiation during reactor operation, passage through the earth radiation belts, and at periods of solar flares. (OTS)« less

Thermionic/AMTEC cascade converter concept for high-efficiency space power

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

Hagan, T.H. van; Smith, J.N. Jr.; Schuller, M.

1996-12-31

This paper presents trade studies that address the use of the thermionic/AMTEC cell--a cascaded, high-efficiency, static power conversion concept that appears well-suited to space power applications. Both the thermionic and AMTEC power conversion approaches have been shown to be promising candidates for space power. Thermionics offers system compactness via modest efficiency at high heat rejection temperatures, and AMTEC offers high efficiency at modest heat rejection temperature. From a thermal viewpoint the two are ideally suited for cascaded power conversion: thermionic heat rejection and AMTEC heat source temperatures are essentially the same. In addition to realizing conversion efficiencies potentially as highmore » as 35--40%, such a cascade offers the following perceived benefits: survivability; simplicity; technology readiness; and technology growth. Mechanical approaches and thermal/electric matching criteria for integrating thermionics and AMTEC into a single conversion device are described. Focusing primarily on solar thermal space power applications, parametric trends are presented to show the performance and cost potential that should be achievable with present-day technology in cascaded thermionic/AMTEC systems.« less

Biofilm thickness measurement using an ultrasound method in a liquid phase.

PubMed

Maurício, R; Dias, C J; Jubilado, N; Santana, F

2013-10-01

In this report, the development of an online, noninvasive, measurement method of the biofilm thickness in a liquid phase is presented. The method is based in the analysis of the ultrasound wave pulse-echo behavior in a liquid phase reproducing the real reactor conditions. It does not imply the removal of the biomass from the support or any kind of intervention in the support (pipes) to detect and perform the measurements (non-invasiveness). The developed method allows for its sensor to be easily and quickly mounted and unmounted in any location along a pipe or reactor wall. Finally, this method is an important innovation because it allows the thickness measurement of a biofilm, in liquid phase conditions that can be used in monitoring programs, to help in scheduling cleaning actions to remove the unwanted biofilm, in several application areas, namely in potable water supply pipes.

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

LaSalle, F.R.; Golbeg, P.R.; Chenault, D.M.

For reactor and nuclear facilities, both Title 10, Code of Federal Regulations, Part 50, and US Department of Energy Order 6430.1A require assessments of the interaction of non-Safety Class 1 piping and equipment with Safety Class 1 piping and equipment during a seismic event to maintain the safety function. The safety class systems of nuclear reactors or nuclear facilities are designed to the applicable American Society of Mechanical Engineers standards and Seismic Category 1 criteria that require rigorous analysis, construction, and quality assurance. Because non-safety class systems are generally designed to lesser standards and seismic criteria, they may become missilesmore » during a safe shutdown earthquake. The resistance of piping, tubing, and equipment to seismically generated missiles is addressed in the paper. Gross plastic and local penetration failures are considered with applicable test verification. Missile types and seismic zones of influence are discussed. Field qualification data are also developed for missile evaluation.« less

Heat Pipe Powered Stirling Conversion for the Demonstration Using Flattop Fission (DUFF) Test

NASA Technical Reports Server (NTRS)

Gibson, Marc A.; Briggs, Maxwell H.; Sanzi, James L.; Brace, Michael H.

2013-01-01

Design concepts for small Fission Power Systems (FPS) have shown that heat pipe cooled reactors provide a passive, redundant, and lower mass option to transfer heat from the fuel to the power conversion system, as opposed to pumped loop designs typically associated with larger FPS. Although many systems have been conceptually designed and a few making it to electrically heated testing, none have been coupled to a real nuclear reactor. A demonstration test named DUFF Demonstration Using Flattop Fission, was planned by the Los Alamos National Lab (LANL) to use an existing criticality experiment named Flattop to provide the nuclear heat source. A team from the NASA Glenn Research Center designed, built, and tested a heat pipe and power conversion system to couple to Flattop with the end goal of making electrical power. This paper will focus on the design and testing performed in preparation for the DUFF test.

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

Vins, M.

This contribution overviews neutron spectrum measurement, which was done on training reactor VR-1 Sparrow with a new nuclear fuel. Former nuclear fuel IRT-3M was changed for current nuclear fuel IRT-4M with lower enrichment of 235U (enrichment was reduced from former 36% to 20%) in terms of Reduced Enrichment for Research and Test Reactors (RERTR) Program. Neutron spectrum measurement was obtained by irradiation of activation foils at the end of pipe of rabit system and consecutive deconvolution of obtained saturated activities. Deconvolution was performed by computer iterative code SAND-II with 620 groups' structure. All gamma measurements were performed on Canberra HPGe.more » Activation foils were chosen according physical and nuclear parameters from the set of certificated foils. The Resulting differential flux at the end of pipe of rabit system agreed well with typical spectrum of light water reactor. Measurement of neutron spectrum has brought better knowledge about new reactor core C1 and improved methodology of activation measurement. (author)« less

Experimental investigation of high temperature high voltage thermionic diode for the space power nuclear reactor

NASA Astrophysics Data System (ADS)

Onufriyev, Valery. V.

2001-02-01

It is well known that the rise of arc from the dense glow discharge is connected with the thermion and secondary processes on the cathode surface (Granovsky, 1971; Leob, 1953; Engel, 1935). First model of breakdown of the cathode layer is connected with the increase of the cathode temperature in consequence of the ion bombardment that leads to the grows its thermo-emissive current. Other model shows the main role of the secondary effects on the cathode surface-the increase of the secondary ion emission coefficient-γi with the grows of glow discharge voltage. But the author of this investigation work of breakdown in Cs vapor (a transmission the glow discharge into self-maintaining arc discharge) discovered the next peculiarity: the value of breakdown voltage is constant when the values of vapor temperature (its pressure pcs) and cathode temperature Tk is constant too (Ub=constant with Tk=constant and pcs=constant) and it is not a statistical value (Onufryev, Grishin, 1996) (that was observed in gas glow discharges other authors (Granovsky, 1971; Leob, 1953; Engel, 1935)). The investigations of thermion high voltage high temperature diode (its breakdown characteristics in closed state and voltage-current characteristics in disclosed state) showed that the value of the breakdown voltage is depended on the vapor pressure in inter-electrode gap (IEG)-pcs and cathode temperature-Tk and is independent on IEG length-Δieg. On this base it was settled that the main role in transition of glow discharge to self-maintaining arc discharge plays an ion cathode layer but more exactly-the region of excited atoms-``Aston glow.'' .

Long titanium heat pipes for high-temperature space radiators

NASA Technical Reports Server (NTRS)

Girrens, S. P.; Ernst, D. M.

1982-01-01

Titanium heat pipes are being developed to provide light weight, reliable heat rejection devices as an alternate radiator design for the Space Reactor Power System (SP-100). The radiator design includes 360 heat pipes, each of which is 5.2 m long and dissipates 3 kW of power at 775 K. The radiator heat pipes use potassium as the working fluid, have two screen arteries for fluid return, a roughened surface distributive wicking system, and a D shaped cross section container configuration. A prototype titanium heat pipe, 5.5 m long, was fabricated and tested in space simulating conditions. Results from startup and isothermal operation tests are presented. These results are also compared to theoretical performance predictions that were used to design the heat pipe initially.

Long titanium heat pipes for high-temperature space radiators

NASA Technical Reports Server (NTRS)

Girrens, S. P.; Ernst, D. M.

1982-01-01

Titanium heat pipes are being developed to provide light weight, reliable heat rejection devices as an alternate radiator design for the Space Reactor Power System (SP-100). The radiator design includes 360 heat pipes, each of which is 5.2 m long and dissipates 3 kW of power at 775 K. The radiator heat pipes use potassium as the working fluid, have two screen arteries for fluid return, a roughened surface distributive wicking system, and a D-shaped cross-section container configuration. A prototype titanium heat pipe, 5.5-m long, has been fabricated and tested in space-simulating conditions. Results from startup and isothermal operation tests are presented. These results are also compared to theoretical performance predictions that were used to design the heat pipe initially.

Modeling of a CeO2 thermochemistry reduction process for hydrogen production by solar concentrated energy

NASA Astrophysics Data System (ADS)

Valle-Hernández, Julio; Romero-Paredes, Hernando; Arancibia-Bulnes, Camilo A.; Villafan-Vidales, Heidi I.; Espinosa-Paredes, Gilberto

2016-05-01

In this paper the simulation of the thermal reduction for hydrogen production through the decomposition of cerium oxide is presented. The thermochemical cycle for hydrogen production consists of the endothermic reduction of CeO2 at high temperature, where concentrated solar energy is used as a source of heat; and of the subsequent steam hydrolysis of the resulting cerium oxide to produce hydrogen. For the thermochemical process, a solar reactor prototype is proposed; consisting of a cubic receptacle made of graphite fiber thermally insulated. Inside the reactor a pyramidal arrangement with nine tungsten pipes is housed. The pyramidal arrangement is made respect to the focal point where the reflected energy is concentrated. The solar energy is concentrated through the solar furnace of high radiative flux. The endothermic step is the reduction of the cerium oxide to lower-valence cerium oxide, at very high temperature. The exothermic step is the hydrolysis of the cerium oxide (III) to form H2 and the corresponding initial cerium oxide made at lower temperature inside the solar reactor. For the modeling, three sections of the pipe where the reaction occurs were considered; the carrier gas inlet, the porous medium and the reaction products outlet. The mathematical model describes the fluid mechanics; mass and energy transfer occurring therein inside the tungsten pipe. Thermochemical process model was simulated in CFD. The results show a temperature distribution in the solar reaction pipe and allow obtaining the fluid dynamics and the heat transfer within the pipe. This work is part of the project "Solar Fuels and Industrial Processes" from the Mexican Center for Innovation in Solar Energy (CEMIE-Sol).

Modeling thermionic emission from laser-heated nanoparticles

DOE PAGES

Mitrani, J. M.; Shneider, M. N.; Stratton, B. C.; ...

2016-02-01

An adjusted form of thermionic emission is applied to calculate emitted current from laser-heated nanoparticles and to interpret time-resolved laser-induced incandescence (TR-LII) signals. This adjusted form of thermionic emission predicts significantly lower values of emitted current compared to the commonly used Richardson-Dushman equation, since the buildup of positive charge in a laser-heated nanoparticle increases the energy barrier for further emission of electrons. Thermionic emission influences the particle's energy balance equation, which can influence TR-LII signals. Additionally, reports suggest that thermionic emission can induce disintegration of nanoparticle aggregates when the electrostatic Coulomb repulsion energy between two positively charged primary particles ismore » greater than the van der Waals bond energy. Furthermore, since the presence and size of aggregates strongly influences the particle's energy balance equation, using an appropriate form of thermionic emission to calculate emitted current may improve interpretation of TR-LII signals.« less

Multiple Restart Testing of a Stainless Steel Sodium Heat Pipe Module

NASA Astrophysics Data System (ADS)

Martin, James; Mireles, Omar; Reid, Robert

2005-02-01

A heat pipe cooled reactor is one of several candidate reactor concepts being considered for space power and propulsion systems to support future space exploration activities. Long life heat pipe modules, with concepts verified through a combination of theoretical analysis and experimental evaluations, would be necessary to establish the viability of this option. A number of stainless steel/sodium heat pipe modules have been designed and fabricated to support experimental testing of a Safe Affordable Fission Engine (SAFE) project, a 100-kWt core design pursued jointly by the Marshall Space Flight Center and the Los Alamos National Laboratory. One of the SAFE heat pipe modules was successfully subjected to over 200 restarts, examining the behavior of multiple passive freeze/thaw operations. Typical operation included a 1-hour startup to an average evaporator temperature of 1000 K followed by a 15-minute hold at temperature. Nominal maximum input power to the evaporator (measured at the power supply) during the hold period was 1.9 kW, with approximately 1.6 kW calculated as the axial power transfer to the condenser (the 300W difference was lost to environment at the evaporator surface). Between heating cycles the module was cooled to less than 325 K, returning the sodium to a frozen state in preparation for the next startup cycle.

Multiple Restart Testing of a Stainless Steel Sodium Heat Pipe Module

NASA Technical Reports Server (NTRS)

Martin, James; Mireles, Omar; Reid, Robert

2005-01-01

A heat pipe cooled reactor is one of several candidate reactor cores being considered for space power and propulsion systems to support future space exploration activities. Long life heat pipe modules. with designs verified through a combination of theoretical analysis and experimental evaluations. would be necessary to establish the viability of this option. A hardware-based program was initiated to begin experimental testing of components to verify compliance of proposed designs. To this end, a number of stainless steel/sodium heat pipe modules have been designed and fabricated to support experimental testing of a Safe Affordable Fission Engine (SAFE) project, a 100-kWt core design pursued jointly by the Marshall Space Flight Center and the Los Alamos National Laboratory. One of the SAFE heat pipe modules was successfully subjected to over 200 restarts. examining the behavior of multiple passive freeze/thaw operations. Typical operation included a 1-hour startup to an average evaporator temperature of 1000 K followed by a 15 minute hold at temperature. Nominal maximum input power during the hold period was 1.9 kW. Between heating cycles the module was cooled to less than 325 K, returning the sodium to a frozen state in preparation fop the next startup cycle.

MTR WING, TRA604, INTERIOR. BASEMENT. INTERIOR VIEW FROM SAME LOCATION ...

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

MTR WING, TRA-604, INTERIOR. BASEMENT. INTERIOR VIEW FROM SAME LOCATION IN WEST CORRIDOR AS PHOTO ID-33-G-42 BUT CAMERA FACES SOUTH. SIGN ON DOOR FOR "PIPE TUNNEL" WARNS OF RADIOLOGICAL AND ASBESTOS HAZARDS. DOOR HAS METAL HASPS. SIGN ON OVERHEAD WASTE HEAT RECOVERY PIPES SAYS THEY CONTAIN "ASBESTOS FREE INSULATION." FIRE DOOR AT LEFT LEADS TO STAIRWAY TO FIRST FLOOR. DOOR AT RIGHT LEADS TO ROOM WHICH ONCE CONTAINED MTR LIBRARY. INL NEGATIVE NO. HD46-13-4. Mike Crane, Photographer, 2/2005 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Closeout Report for the Refractory Metal Accelerated Heat Pipe Life Test Activity

NASA Technical Reports Server (NTRS)

Martin, J.; Reid, R.; Stewart, E.; Hickman, R.; Mireles, O.

2013-01-01

With the selection of a gas-cooled reactor, this heat pipe accelerated life test activity was closed out and its resources redirected. The scope of this project was to establish the long-term aging effects on Mo-44.5%Re sodium heat pipes when subjected to space reactor temperature and mass fluences. To date, investigators have demonstrated heat pipe life tests of alkali metal systems up to .50,000 hours. Unfortunately, resources have not been available to examine the effect of temperature, mass fluence, or impurity level on corrosion or to conduct post-test forensic examination of heat pipes. The key objective of this effort was to establish a cost/time effective method to systematically test alkali metal heat pipes with both practical and theoretical benefits. During execution of the project, a heat pipe design was established, a majority of the laboratory test equipment systems specified, and operating and test procedures developed. Procurements for the heat pipe units and all major test components were underway at the time the stop work order was issued. An extremely important outcome was the successful fabrication of an annular wick from Mo-5%Re screen (the single, most difficult component to manufacture) using a hot isostatic pressing technique. This Technical Publication (TP) includes specifics regarding the heat pipe calorimeter water-cooling system, vendor design for the radio frequency heating system, possible alternative calorimeter designs, and progress on the vanadium equilibration technique. The methods provided in this TP and preceding project documentation would serve as a good starting point to rapidly implement an accelerated life test. Relevant test data can become available within months, not years, and destructive examination of the first life test heat pipe might begin within 6 months of test initiation. Final conclusions could be drawn in less than a quarter of the mission duration for a long-lived, fission-powered, deep space probe.

Electrically Heated Testing of the Kilowatt Reactor Using Stirling Technology (KRUSTY) Experiment Using a Depleted Uranium Core

NASA Technical Reports Server (NTRS)

Briggs, Maxwell H.; Gibson, Marc A.; Sanzi, James

2017-01-01

The Kilopower project aims to develop and demonstrate scalable fission-based power technology for systems capable of delivering 110 kW of electric power with a specific power ranging from 2.5 - 6.5 Wkg. This technology could enable high power science missions or could be used to provide surface power for manned missions to the Moon or Mars. NASA has partnered with the Department of Energys National Nuclear Security Administration, Los Alamos National Labs, and Y-12 National Security Complex to develop and test a prototypic reactor and power system using existing facilities and infrastructure. This technology demonstration, referred to as the Kilowatt Reactor Using Stirling TechnologY (KRUSTY), will undergo nuclear ground testing in the summer of 2017 at the Nevada Test Site. The 1 kWe variation of the Kilopower system was chosen for the KRUSTY demonstration. The concept for the 1 kWe flight system consist of a 4 kWt highly enriched Uranium-Molybdenum reactor operating at 800 degrees Celsius coupled to sodium heat pipes. The heat pipes deliver heat to the hot ends of eight 125 W Stirling convertors producing a net electrical output of 1 kW. Waste heat is rejected using titanium-water heat pipes coupled to carbon composite radiator panels. The KRUSTY test, based on this design, uses a prototypic highly enriched uranium-molybdenum core coupled to prototypic sodium heat pipes. The heat pipes transfer heat to two Advanced Stirling Convertors (ASC-E2s) and six thermal simulators, which simulate the thermal draw of full scale power conversion units. Thermal simulators and Stirling engines are gas cooled. The most recent project milestone was the completion of non-nuclear system level testing using an electrically heated depleted uranium (non-fissioning) reactor core simulator. System level testing at the Glenn Research Center (GRC) has validated performance predictions and has demonstrated system level operation and control in a test configuration that replicates the one to be used at the Device Assembly Facility (DAF) at the Nevada National Security Site. Fabrication, assembly, and testing of the depleted uranium core has allowed for higher fidelity system level testing at GRC, and has validated the fabrication methods to be used on the highly enriched uranium core that will supply heat for the DAF KRUSTY demonstration.

Special Purpose Nuclear Reactor (5 MW) for Reliable Power at Remote Sites Assessment Report

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

Sterbentz, James William; Werner, James Elmer; McKellar, Michael George

The Phenomena Identification and Ranking Table (PIRT) technique was conducted on the Special Purpose Reactor nuclear plant design. The PIRT is a structured process to identify safety-relevant/safety-significant phenomena and assess the importance and knowledge base by ranking the phenomena. The Special Purpose Reactor is currently in the conceptual design stage. The candidate reactor has a solid monolithic stainless steel core with an array of heat pipes and fuel pellets embedded in the monolith. The heat pipes are used to remove heat from the core using simple, reliable, and well-characterized physics (capillarity, boiling, and condensation). In the initial design, one heatmore » exchanger is used for the working fluid that produces energy, and a second heat exchanger is used to remove decay heat in emergency or shutdown conditions. In addition, a power conversion cycle such as an open-air Brayton system is available as an option for power conversion and process heat. This report summarizes and documents the process and scope of the four PIRT reviews, noting the major activities and conclusions. The identified phenomena, analyses, rationales, and associated ratings are presented along with a summary of the findings from the four individual PIRTs, namely (1) Reactor Accident and Normal Operations, (2) Heat Pipes, (3) Materials, and (4) Power Conversion. The PIRT reports for these four major system areas evaluated are attached as appendixes to this report and provide considerably more detail about each assessment as well as a more complete listing of the phenomena that were evaluated.« less

ETR HEAT EXCHANGER BUILDING, TRA644. FLOOR PLAN AND SECTIONS. PUMP ...

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

ETR HEAT EXCHANGER BUILDING, TRA-644. FLOOR PLAN AND SECTIONS. PUMP CUBICLES WITH PUMP MOTORS OUTSIDE CUBICLES. HEAT EXCHANGER EQUIPMENT. COOLANT PIPE TUNNEL ENTERS FROM REACTOR BUILDING. KAISER ETR-5582-MTR-644-A-3, 2/1956. INL INDEX NO. 532-0644-00-486-101294, REV. 6. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

ETR, TRA642. NORTHSOUTH SECTION, LOOKING WEST. STEELFRAME ROOF, CRANE RAIL, ...

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

ETR, TRA-642. NORTH-SOUTH SECTION, LOOKING WEST. STEEL-FRAME ROOF, CRANE RAIL, AND CRANES. COOLANT PIPE TUNNEL LEADING TO REACTOR FROM EAST. (THIS WAS A PRELIMINARY CONCEPT DRAWING.) KAISER ETR-5528-MTR-642-A-4, 11/1955. INL INDEX NO. 532-0642-00-486-100912, REV. 1. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

APPARATUS FOR DETECTING AND LOCATING PRESENCE OF FLUIDS

DOEpatents

Williamson, R.R.

1958-09-16

A system is described fur detecting water leaks in water-cooled neutronic reactors by utilizing an electrical hygrometer having a resistance element variable with the moisture content. The graphite blocks, forming the moderator in many types of reactors, coniain ducts in which helium gas is circulated. When a leak occurs in a coolant tube, the water will seep through the graphite until it oozes into one of the helium ducts, where it will be swept along with the helium into a system of pipes that connect each of the helium ducts. By inserting an electric hygrometer in each of these pipes and connecting it to an alarm system, the moisture content of the helium will cause a change in the electrical resistance of the hygrometer which will initiate a signal alarm indicating the presence and position of the leaky water tube in the reactor.

SP-100 system thaw and start-up strategies

NASA Astrophysics Data System (ADS)

Kirpich, A.; Choe, H.

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

The status of power supplies for primary electric propulsion in the U.S.A.

NASA Technical Reports Server (NTRS)

Jones, R. M.; Scott-Monck, J. A.

1984-01-01

This paper reviews the status of and requirements on solar electric and nuclear electric power supplies for primary electric propulsion missions. The power supply requirements of power level, specific mass (kg/kWe) and lifetime are defined as a function of the mission and electric propulsion system characteristics for planetary missions. The technology status of planar and concentrator arrays is discussed. Nuclear reactors and thermoelectric, thermionic, Brayton and Rankine conversion technologies are reviewed, as well as recent nuclear power system design concepts and program activity. Technology projections for power supplies applicable to primary electric propulsion missions are included.

Gravity flow rate of solids through orifices and pipes

NASA Technical Reports Server (NTRS)

Gardner, J. F.; Smith, J. E.; Hobday, J. M.

1977-01-01

Lock-hopper systems are the most common means for feeding solids to and from coal conversion reactor vessels. The rate at which crushed solids flow by gravity through the vertical pipes and valves in lock-hopper systems affects the size of pipes and valves needed to meet the solids-handling requirements of the coal conversion process. Methods used to predict flow rates are described and compared with experimental data. Preliminary indications are that solids-handling systems for coal conversion processes are over-designed by a factor of 2 or 3.

Diamond Thin-Film Thermionic Generator

NASA Astrophysics Data System (ADS)

Clewell, J. M.; Ordonez, C. A.; Perez, J. M.

1997-03-01

Since the eighteen-hundreds scientists have sought to develop the highest thermal efficiency in heat engines such as thermionic generators. Modern research in the emerging diamond film industry has indicated the work functions of diamond thin-films can be much less than one electron volt, compelling fresh investigation into their capacity as thermionic generators and inviting new methodology for determining that efficiency. Our objective is to predict the efficiency of a low-work-function, degenerate semiconductor (diamond film) thermionic generator operated as a heat engine between two constant-temperature thermal reservoirs. Our presentation will focus on a theoretical model which predicts the efficiency of the system by employing a Monte Carlo computational technique from which we report results for the thermal efficiency and the thermionic current densities of diamond thin-films.

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

PBF Reactor Building (PER620). Camera faces southeast. Concrete placement will ...

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

PBF Reactor Building (PER-620). Camera faces southeast. Concrete placement will leave opening for neutron camera to be installed later. Note vertical piping within rebar. Photographer: John Capek. Date: July 6, 1967. INEEL negative no. 67-3514 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

Pressure suppression containment system for boiling water reactor

DOEpatents

Gluntz, Douglas M.; Nesbitt, Loyd B.

1997-01-01

A system for suppressing the pressure inside the containment of a BWR following a postulated accident. A piping subsystem is provided which features a main process pipe that communicates the wetwell airspace to a connection point downstream of the guard charcoal bed in an offgas system and upstream of the main bank of delay charcoal beds which give extensive holdup to offgases. The main process pipe is fitted with both inboard and outboard containment isolation valves. Also incorporated in the main process pipe is a low-differential-pressure rupture disk which prevents any gas outflow in this piping whatsoever until or unless rupture occurs by virtue of pressure inside this main process pipe on the wetwell airspace side of the disk exceeding the design opening (rupture) pressure differential. The charcoal holds up the radioactive species in the noncondensable gas from the wetwell plenum by adsorption, allowing time for radioactive decay before the gas is vented to the environs.

Atmospheric reentry of the in-core thermionic SP-100 reactor system

NASA Technical Reports Server (NTRS)

Stamatelatos, M. G.; Barsell, A. W.; Harris, P. A.; Francisco, J.

1987-01-01

Presumed end-of-life atmospheric reentry of the GA SP-100 system was studied to assess dispersal feasibility and associated hazards. Reentry was studied by sequential use of an orbital trajectory and a heat analysis computer program. Two heating models were used. The first model assumed a thermal equilibrium condition between the stagnation point aerodynamic heating and the radiative cooling of the skin material surface. The second model allowed for infinite conductivity of the skin material. Four reentering configurations were studied representing stages of increased SP-100 breakup: (1) radiator, shield and reactor, (2) shield and reactor, (3) reactor with control drums, and (4) reactor without control drums. Each reentering configuration was started from a circular orbit at 116 km having an inertial velocity near Mach 25. The assumed failing criterion was the attainment of melting temperature of a critical system component. The reentry analysis revealed breakup of the vessel in the neighborhood of 61 km altitude and scattering of the fuel elements. Subsequent breakup of the fuel elements was not predicted. Oxidation of the niobium skin material was calculated to cause an increase in surface temperature of less than ten percent. The concept of thermite analogs for enhancing reactor reentry dispersal was assessed and found to be feasible in principle. A conservative worst-case hazards analysis was performed for radioactive and nonradioactive toxic SP-100 materials assumed to be dispersed during end-of-life reentry. The hazards associated with this phase of the SP-100 mission were calculated to be insignificant.

The advanced thermionic converter with microwave power as an auxiliary ionization source

NASA Technical Reports Server (NTRS)

Manikopoulos, C. N.; Hatziprocopiou, M.; Chiu, H. S.; Shaw, D. T.

1978-01-01

In the search for auxiliary sources of ionization for the advanced thermionic converter plasma, as required for terrestial applications, the use of externally applied microwave power is considered. The present work is part of the advanced model thermionic converter development research currently performed at the laboratory for Power and Environmental Studies at SUNY Buffalo. Microwave power in the frequency range 1-3 GHz is used to externally pump a thermionic converter and the results are compared to the theoretical model proposed by Lam (1976) in describing the thermionic converter plasma. The electron temperature of the plasma is found to be raised considerably by effective microwave heating which results in the disappearance of the double sheath ordinarily erected in front of the emitter. The experimental data agree satisfactorily with theory in the low current region.

Thermionic emission current in a single barrier varactor

NASA Technical Reports Server (NTRS)

Hjelmgren, Hans; East, Jack; Kollberg, Erik

1992-01-01

From I-V measurements on Single Barrier Varactors (SBV) at different temperatures we concluded that thermionic emission across the barrier of the actual device is mainly due to transport through the X band. The same structure was also modeled with a one-dimensional drift-diffusion model, including a 'boundary condition' for thermionic emission across the heterojunction interface. By including thermionic field emission through the top of the triangular barrier of a biased diode and the effect of a non-abrupt interface at the heterojunction, we obtained good agreement between the modeled and measured I-V characteristics.

The mechanism of explosive emission excitation in thermionic energy conversion processes

NASA Astrophysics Data System (ADS)

Bulyga, A. V.

A study has been made of the mechanism of explosive electron emission in vacuum thermionic converters induced by thermionic currents in the case of the anomalous Richardson effect. The latter is associated with a spotted emitting surface and temperature fluctuations. In order to account for one of the components of the electrode potential difference, it is proposed that allowance be made for the difference between the polarization signal velocity in a dense metal electron gas and that in the electron-ion gas of the electrode gap. Ways to achieve explosive emission in real thermionic converters are discussed.

Advanced thermionic converter developments with microwave external pumping

NASA Technical Reports Server (NTRS)

Chiu, H. S.; Shaw, D. T.; Manikopulos, C. N.; Lee, C. H.

1977-01-01

This work reports ion generation in a cesium thermionic converter as part of advanced-model thermionic converter development research. A microwave with frequency in the range between 1-2 GHz is used to externally pump a thermionic converter as part of our effort in the verification of Lam's theory. It is found that the motive peak as predicted in the theory disappears whenever microwave power is used to excite the cesium plasma of the converter. The electron temperature is effectively heated by the microwave and the experimental data agrees with theory in the low-power output region.

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

Chattopadhyay, J.; Dutta, B.K.; Kushwaha, H.S.

Leak-Before-Break (LBB) is being used to design the primary heat transport piping system of 500 MWe Indian Pressurized Heavy Water Reactors (IPHWR). The work is categorized in three directions to demonstrate three levels of safety against sudden catastrophic break. Level 1 is inherent in the design procedure of piping system as per ASME Sec.III with a well defined factor of safety. Level 2 consists of fatigue crack growth study of a postulated part-through flaw at the inside surface of pipes. Level 3 is stability analysis of a postulated leakage size flaw under the maximum credible loading condition. Developmental work relatedmore » to demonstration of level 2 and level 3 confidence is described in this paper. In a case study on fatigue crack growth on PHT straight pipes for level 2, negligible crack growth is predicted for the life of the reactor. For level 3 analysis, the R6 method has been adopted. A database to evaluate SIF of elbows with throughwall flaws under combined internal pressure and bending moment has been generated to provide one of the inputs for R6 method. The methodology of safety assessment of elbow using R6 method has been demonstrated for a typical pump discharge elbow. In this analysis, limit load of the cracked elbow has been determined by carrying out elasto-plastic finite element analysis. The limit load results compared well with those given by Miller. However, it requires further study to give a general form of limit load solution. On the experimental front, a set of small diameter pipe fracture experiments have been carried out at room temperature and 300{degrees}C. Two important observations of the experiments are - appreciable drop in maximum load at 300{degrees}C in case of SS pipes and out-of-plane crack growth in case of CS pipes. Experimental load deflection curves are finally compared with five J-estimation schemes predictions. A material database of PHT piping materials is also being generated for use in LBB analysis.« less

An Experiment on Thermionic Emission: Back to the Good Old Triode

ERIC Educational Resources Information Center

Azooz, A. A.

2007-01-01

A simple experiment to study thermionic emission, the Richardson-Dushman equation and the energy distribution function of thermionic electrons emitted from a hot cathode using a triode vacuum tube is described. It is pointed out that such a distribution function is directly proportional to the first derivative of the Edison anode current with…

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

Twenty years of experience in monitoring 41Ar in a research reactor and decrease of its discharge into the environment.

PubMed

Fukui, M

2004-04-01

The radioactive gas 41Ar has been produced at high concentration by neutron activation near the reactor core in the Kyoto University Research Reactor. A pipe line for an exhaust stream, so-called sweep gas, was fabricated at the construction of the reactor in 1964 in order to exhale 41Ar from the facilities above to the environment. Other exhaust lines with decay tanks were established separately from the sweep line for both the cold neutron source in 1986 and the heavy-water tank in 1996, respectively, because a higher amount of 41Ar was thought to be produced from these facilities due to the improvement. As a result, a slight change in the flow rate of the exhaust was found to have a great deal of influence on both the 41Ar concentration in the reactor room and the rate of emission from the stack. By monitoring the exhaust air from the decay tanks, the mechanism for decreasing the emission was clarified together with identifying an obstacle, i.e., the condensate against the steady state flow, formed in the exhaust pipe. By setting the flow rate suitably in the exhaust line, the rate of 41Ar emission from the biological shielding into both the work place in the reactor room and the environment has been controlled as low as reasonably achievable.

Space nuclear system expansion joints

NASA Technical Reports Server (NTRS)

Whitaker, W. D.; Shimazki, T. T.

1973-01-01

The engineering, design, and fabrication status of the expansion joint unit (EJU) to be employed in the NaK primary coolant piping loop of the 5-kwe Reactor thermoelectric system are described. Four EJU's are needed in the NaK primary coolant piping loop. The four EJU's which will be identical, utilize bellows as the flexing member, are hermetically sealed, and provide double containment. The bellows are of a nested-formed design, and are to be constructed of 1-ply thickness of 0.010-in. Inconel 718. The EJU's provide a minimum piping load margin of safety of +0.22.

Dynamic System Simulation of the KRUSTY Experiment

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

Klein, Steven Karl; Kimpland, Robert Herbert

2016-05-09

The proposed KRUSTY experiment is a demonstration of a reactor operating at power. The planned experimental configuration includes a highly enriched uranium (HEU) reflected core, cooled by multiple heat pipes leading to Stirling engines for primary heat rejection. Operating power is expected to be approximately four (4) to five (5) kilowatts with a core temperature above 1,000 K. No data is available on any historical reactor employing HEU metal that operated over the temperature range required for the KRUSTY experiment. Further, no reactor has operated with heat pipes as the primary cooling mechanism. Historic power reactors have employed either naturalmore » or forced convection so data on their operation is not directly applicable to the KRUSTY experiment. The primary purpose of the system model once developed and refined by data from these component experiments, will be used to plan the KRUSTY experiment. This planning will include expected behavior of the reactor from start-up, through various transient conditions where cooling begins to become present and effective, and finally establishment of steady-state. In addition, the model can provide indicators of anticipated off-normal events and appropriate operator response to those conditions. This information can be used to develop specific experiment operating procedures and aids to guide the operators in conduct of the experiment.« less

Dexter - A one-dimensional code for calculating thermionic performance of long converters.

NASA Technical Reports Server (NTRS)

Sawyer, C. D.

1971-01-01

This paper describes a versatile code for computing the coupled thermionic electric-thermal performance of long thermionic converters in which the temperature and voltage variations cannot be neglected. The code is capable of accounting for a variety of external electrical connection schemes, coolant flow paths and converter failures by partial shorting. Example problem solutions are given.

HEAT Sensor: Harsh Environment Adaptable Thermionic Sensor

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

Limb, Scott J.

2016-05-31

This document is the final report for the “HARSH ENVIRONMENT ADAPTABLE THERMIONIC SENSOR” project under NETL’s Crosscutting contract DE-FE0013062. This report addresses sensors that can be made with thermionic thin films along with the required high temperature hermetic packaging process. These sensors can be placed in harsh high temperature environments and potentially be wireless and self-powered.

Observation of increased space-charge limited thermionic electron emission current by neutral gas ionization in a weakly-ionized deuterium plasma

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

Hollmann, E. M.; Yu, J. H.; Doerner, R. P.

2015-09-14

The thermionic electron emission current emitted from a laser-produced hot spot on a tungsten target in weakly-ionized deuterium plasma is measured. It is found to be one to two orders of magnitude larger than expected for bipolar space charge limited thermionic emission current assuming an unperturbed background plasma. This difference is attributed to the plasma being modified by ionization of background neutrals by the emitted electrons. This result indicates that the allowable level of emitted thermionic electron current can be significantly enhanced in weakly-ionized plasmas due to the presence of large neutral densities.

Improving the Parametric Method of Cost Estimating Relationships of Naval Ships

DTIC Science & Technology

2014-06-01

tool since the total cost of the ship is broken down into smaller parts as defined by the WBS. The Navy currently uses the Expanded Ship Work Breakdown...Includes boilers , reactors, turbines, gears, shafting, propellers, steam piping, lube oil piping, and radiation 300 Electric Plant Includes ship...spaces, ladders, storerooms, laundry, and workshops 700 Armament Includes guns, missile launchers, ammunition handling and stowage, torpedo tubes , depth

Thermionic emission from monolayer graphene, sheath formation and its feasibility towards thermionic converters

NASA Astrophysics Data System (ADS)

Misra, Shikha; Upadhyay Kahaly, M.; Mishra, S. K.

2017-02-01

A formalism describing the thermionic emission from a single layer graphene sheet operating at a finite temperature and the consequent formation of the thermionic sheath in its proximity has been established. The formulation takes account of two dimensional densities of state configuration, Fermi-Dirac (f-d) statistics of the electron energy distribution, Fowler's treatment of electron emission, and Poisson's equation. The thermionic current estimates based on the present analysis is found to be in reasonably good agreement with experimental observations (Zhu et al., Nano Res. 07, 1 (2014)). The analysis has further been simplified for the case where f-d statistics of an electron energy distribution converges to Maxwellian distribution. By using this formulation, the steady state sheath features, viz., spatial dependence of the surface potential and electron density structure in the thermionic sheath are derived and illustrated graphically for graphene parameters; the electron density in the sheath is seen to diminish within ˜10 s of Debye lengths. By utilizing the graphene based cathode in configuring a thermionic converter (TC), an appropriate operating regime in achieving the efficient energy conversion has been identified. A TC configured with the graphene based cathode (operating at ˜1200 K/work function 4.74 V) along with the metallic anode (operating at ˜400 K/ work function 2.0 V) is predicted to display ˜56% of the input thermal flux into the electrical energy, which infers approximately ˜84% of the Carnot efficiency.

Planetary exploration with electrically propelled vehicles.

NASA Technical Reports Server (NTRS)

Stuhlinger, E.

1972-01-01

The characteristics of propulsion systems required for carrying out flight missions within the solar system, as desired by planetary physicists and astronomers, are reviewed. It is shown that an encouraging answer to these requirements is available in the form of electrostatic or ion propulsion systems. The design and performance characteristics of an electrostatic thrustor employing an ion source, accelerating electrode, beam neutralizer, and power source are discussed, together with those of the Kaufmann engine (electrostatic thrustor employing bombardment type ionization). More demanding missions which will become feasible with the advent of nuclear-electric power sources (such as the incore thermionic reactor) may include close orbiters around all the planets, and asteroid and cometary missions.

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

Pressurizer tank upper support

DOEpatents

Baker, Tod H.; Ott, Howard L.

1994-01-01

A pressurizer tank in a pressurized water nuclear reactor is mounted between structural walls of the reactor on a substructure of the reactor, the tank extending upwardly from the substructure. For bearing lateral loads such as seismic shocks, a girder substantially encircles the pressurizer tank at a space above the substructure and is coupled to the structural walls via opposed sway struts. Each sway strut is attached at one end to the girder and at an opposite end to one of the structural walls, and the sway struts are oriented substantially horizontally in pairs aligned substantially along tangents to the wall of the circular tank. Preferably, eight sway struts attach to the girder at 90.degree. intervals. A compartment encloses the pressurizer tank and forms the structural wall. The sway struts attach to corners of the compartment for maximum stiffness and load bearing capacity. A valve support frame carrying the relief/discharge piping and valves of an automatic depressurization arrangement is fixed to the girder, whereby lateral loads on the relief/discharge piping are coupled directly to the compartment rather than through any portion of the pressurizer tank. Thermal insulation for the valve support frame prevents thermal loading of the piping and valves. The girder is shimmed to define a gap for reducing thermal transfer, and the girder is free to move vertically relative to the compartment walls, for accommodating dimensional variation of the pressurizer tank with changes in temperature and pressure.

Analysis of space reactor system components: Investigation through simulation and non-nuclear testing

NASA Astrophysics Data System (ADS)

Bragg-Sitton, Shannon M.

The use of fission energy in space power and propulsion systems offers considerable advantages over chemical propulsion. Fission provides over six orders of magnitude higher energy density, which translates to higher vehicle specific impulse and lower specific mass. These characteristics enable ambitious space exploration missions. The natural space radiation environment provides an external source of protons and high energy, high Z particles that can result in the production of secondary neutrons through interactions in reactor structures. Applying the approximate proton source in geosynchronous orbit during a solar particle event, investigation using MCNPX 2.5.b for proton transport through the SAFE-400 heat pipe cooled reactor indicates an incoming secondary neutron current of (1.16 +/- 0.03) x 107 n/s at the core-reflector interface. This neutron current may affect reactor operation during low power maneuvers (e.g., start-up) and may provide a sufficient reactor start-up source. It is important that a reactor control system be designed to automatically adjust to changes in reactor power levels, maintaining nominal operation without user intervention. A robust, autonomous control system is developed and analyzed for application during reactor start-up, accounting for fluctuations in the radiation environment that result from changes in vehicle location or to temporal variations in the radiation field. Development of a nuclear reactor for space applications requires a significant amount of testing prior to deployment of a flight unit. High confidence in fission system performance can be obtained through relatively inexpensive non-nuclear tests performed in relevant environments, with the heat from nuclear fission simulated using electric resistance heaters. A series of non-nuclear experiments was performed to characterize various aspects of reactor operation. This work includes measurement of reactor core deformation due to material thermal expansion and implementation of a virtual reactivity feedback control loop; testing and thermal hydraulic characterization of the coolant flow paths for two space reactor concepts; and analysis of heat pipe operation during start-up and steady state operation.

DEXTER: A one-dimensional code for calculating thermionic performance of long converters

NASA Technical Reports Server (NTRS)

Sawyer, C. D.

1971-01-01

A versatile code is described for computing the coupled thermionic electric-thermal performance of long thermionic converters in which the temperature and voltage variations cannot be neglected. The code is capable of accounting for a variety of external electrical connection schemes, coolant flow paths and converter failures by partial shorting. Example problem solutions are included along with a user's manual.

Fabrication and surface characterization of composite refractory compounds suitable for thermionic converters

NASA Technical Reports Server (NTRS)

Davis, P. R.; Swanson, L. W.

1979-01-01

The techniques of fabricating and characterizing the surface properties of electrode materials were investigated. The basic surface properties of these materials were studied with respect to their utilization as thermionic energy converter electrodes. Emphasis was placed on those factors (e.g, cesium disorption kinetic and mechanisms of low work function production) which are of primary concern to thermionic converter performance.

Thermionic converter

DOEpatents

Fitzpatrick, G.O.

1987-05-19

A thermionic converter is set forth which includes an envelope having an electron collector structure attached adjacent to a wall. An electron emitter structure is positioned adjacent the collector structure and spaced apart from opposite wall. The emitter and collector structures are in a common chamber. The emitter structure is heated substantially only by thermal radiation. Very small interelectrode gaps can be maintained utilizing the thermionic converter whereby increased efficiency results. 10 figs.

ETR HEAT EXCHANGER BUILDING, TRA644. EAST SIDE. CAMERA FACING WEST. ...

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

ETR HEAT EXCHANGER BUILDING, TRA-644. EAST SIDE. CAMERA FACING WEST. NOTE COURSE OF PIPE FROM GROUND AND FOLLOWING ROOF OF BUILDING. MTR BUILDING IN BACKGROUND AT RIGHT EDGE OF VIEW. INL NEGATIVE NO. HD46-36-3. Mike Crane, Photographer, 4/2005 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Low cost solar array project: Experimental process system development unit for producing semiconductor-grade silicon using the silane-to-silicon process

NASA Technical Reports Server (NTRS)

1981-01-01

The results of the free space reactor experimental work are summarized. Overall, the objectives were achieved and the unit can be confidently scaled to the EPSDU size based on the experimental work and supporting theoretical analyses. The piping and instrumentation of the fluidized bed reactor was completed.

PBF Reactor Building (PER620). Cubicle 10. Camera facing southeast. Loop ...

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

PBF Reactor Building (PER-620). Cubicle 10. Camera facing southeast. Loop pressurizer on right. Other equipment includes loop strained, control valves, loop piping, pressurizer interchanger, and cleanup system cooler. High-density shielding brick walls. Photographer: Kirsh. Date: November 2, 1970. INEEL negative no. 70-4908 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

Emergency heat removal system for a nuclear reactor

DOEpatents

Dunckel, Thomas L.

1976-01-01

A heat removal system for nuclear reactors serving as a supplement to an Emergency Core Cooling System (ECCS) during a Loss of Coolant Accident (LOCA) comprises a plurality of heat pipes having one end in heat transfer relationship with either the reactor pressure vessel, the core support grid structure or other in-core components and the opposite end located in heat transfer relationship with a heat exchanger having heat transfer fluid therein. The heat exchanger is located external to the pressure vessel whereby excessive core heat is transferred from the above reactor components and dissipated within the heat exchanger fluid.

Pressure suppression containment system for boiling water reactor

DOEpatents

Gluntz, D.M.; Nesbitt, L.B.

1997-01-21

A system is disclosed for suppressing the pressure inside the containment of a BWR following a postulated accident. A piping subsystem is provided which features a main process pipe that communicates the wetwell airspace to a connection point downstream of the guard charcoal bed in an offgas system and upstream of the main bank of delay charcoal beds which give extensive holdup to offgases. The main process pipe is fitted with both inboard and outboard containment isolation valves. Also incorporated in the main process pipe is a low-differential-pressure rupture disk which prevents any gas outflow in this piping whatsoever until or unless rupture occurs by virtue of pressure inside this main process pipe on the wetwell airspace side of the disk exceeding the design opening (rupture) pressure differential. The charcoal holds up the radioactive species in the noncondensable gas from the wetwell plenum by adsorption, allowing time for radioactive decay before the gas is vented to the environs. 3 figs.

A NEW THERMIONIC RF ELECTRON GUN FOR SYNCHROTRON LIGHT SOURCES

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

Kutsaev, Sergey; Agustsson, R.; Hartzell, J

A thermionic RF gun is a compact and efficient source of electrons used in many practical applications. RadiaBeam Systems and the Advanced Photon Source at Argonne National Laboratory collaborate in developing of a reliable and robust thermionic RF gun for synchrotron light sources which would offer substantial improvements over existing thermionic RF guns and allow stable operation with up to 1A of beam peak current at a 100 Hz pulse repetition rate and a 1.5 μs RF pulse length. In this paper, we discuss the electromagnetic and engineering design of the cavity and report the progress towards high power testsmore » of the cathode assembly of the new gun.« less

Potential Effects of Leak-Before-Break on Light Water Reactor Design.

DTIC Science & Technology

1985-08-26

Boiler and Pressure Vessel Code . In fact, section 3 of that code was created for nuclear applications. This... Boiler and Pressure Vessel Code . The only major change which leak-before-break would require in these analyses would be that all piping to be considered...XI of the ASME Boiler and Pressure Vessel Code , and is already required for all Class I piping systems in the plant. Class I systems are those

Reactor design and integration into a nuclear electric spacecraft

NASA Technical Reports Server (NTRS)

Phillips, W. M.; Koenig, D. R.

1978-01-01

One of the well-defined applications for nuclear power in space is nuclear electric propulsion (NEP). Mission studies have identified the optimum power level (400 kWe). A single Shuttle launch requirement and science-package integration have added additional constraints to the design. A reactor design which will meet these constraints has been studied. The reactor employs 90 fuel elements, each heat pipe cooled. Reactor control is obtained with BeO/B4C drums in a BeO reflector. The balance of the spacecraft is shielded from the reactor with LiH. Power conditioning and reactor control drum drives are located behind the LiH with the power conditioning. Launch safety, mechanical design and integration with the power conversion subsystem are discussed.

Modeling of transient heat pipe operation

NASA Technical Reports Server (NTRS)

Colwell, G. T.; Hartley, J. G.

1986-01-01

Mathematical models and associated solution procedures which can be used to design heat pipe cooled structures for use on hypersonic vehicles are being developed. The models should also have the capability to predict off-design performance for a variety of operating conditions. It is expected that the resulting models can be used to predict startup behavior of liquid metal heat pipes to be used in reentry vehicles, hypersonic aircraft, and space nuclear reactors. Work to date related to numerical solutions of governing differential equations for the outer shell and the combination capillary structure and working fluid is summarized. Finite element numerical equations using both implicit, explicit, and combination methods were examined.

Heat Pipe Reactor Dynamic Response Tests: SAFE-100 Reactor Core Prototype

NASA Technical Reports Server (NTRS)

Bragg-Sitton, Shannon M.

2005-01-01

The SAFE-I00a test article at the NASA Marshall Space Flight Center was used to simulate a variety of potential reactor transients; the SAFEl00a is a resistively heated, stainless-steel heat-pipe (HP)-reactor core segment, coupled to a gas-flow heat exchanger (HX). For these transients the core power was controlled by a point kinetics model with reactivity feedback based on core average temperature; the neutron generation time and the temperature feedback coefficient are provided as model inputs. This type of non-nuclear test is expected to provide reasonable approximation of reactor transient behavior because reactivity feedback is very simple in a compact fast reactor (simple, negative, and relatively monotonic temperature feedback, caused mostly by thermal expansion) and calculations show there are no significant reactivity effects associated with fluid in the HP (the worth of the entire inventory of Na in the core is .

DynMo: Dynamic Simulation Model for Space Reactor Power Systems

NASA Astrophysics Data System (ADS)

El-Genk, Mohamed; Tournier, Jean-Michel

2005-02-01

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

Evidence for cluster shape effects on the kinetic energy spectrum in thermionic emission.

PubMed

Calvo, F; Lépine, F; Baguenard, B; Pagliarulo, F; Concina, B; Bordas, C; Parneix, P

2007-11-28

Experimental kinetic energy release distributions obtained for the thermionic emission from C(n) (-) clusters, 10< or =n< or =20, exhibit significant non-Boltzmann variations. Using phase space theory, these different features are analyzed and interpreted as the consequence of contrasting shapes in the daughter clusters; linear and nonlinear isomers have clearly distinct signatures. These results provide a novel indirect structural probe for atomic clusters associated with their thermionic emission spectra.

An overview of thermionic power conversion technology

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

White, Morgan C.

1996-12-01

Thermionic energy conversion is one of the many concepts which make up the direct power conversion technologies. Specifically, thermionics is the process of changing heat directly into electricity via a material`s ability to emit electrons when heated. This thesis presents a broad overview of the engineering and physics necessary to make thermionic energy conversion (TEC) a practical reality. It begins with an introduction to the technology and the history of its development. This is followed by a discussion of the physics and engineering necessary to develop practical power systems. Special emphasis is placed on the critical issues which are stillmore » being researched. Finally, there is a discussion of the missions which this technology may fulfill.« less

Combination pipe-rupture mitigator and in-vessel core catcher. [LMFBR

DOEpatents

Tilbrook, R.W.; Markowski, F.J.

1982-03-09

A device is described which mitigates against the effects of a failed coolant loop in a nuclear reactor by restricting the outflow of coolant from the reactor through the failed loop and by retaining any particulated debris from a molten core which may result from coolant loss or other cause. The device reduces the reverse pressure drop through the failed loop by limiting the access of coolant in the reactor to the inlet of the failed loop. The device also spreads any particulated core debris over a large area to promote cooling.

Combination pipe rupture mitigator and in-vessel core catcher

DOEpatents

Tilbrook, Roger W.; Markowski, Franz J.

1983-01-01

A device which mitigates against the effects of a failed coolant loop in a nuclear reactor by restricting the outflow of coolant from the reactor through the failed loop and by retaining any particulated debris from a molten core which may result from coolant loss or other cause. The device reduces the reverse pressure drop through the failed loop by limiting the access of coolant in the reactor to the inlet of the failed loop. The device also spreads any particulated core debris over a large area to promote cooling.

Sensitivity to VSR failure: K pipe break accident

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

Meichle, R.H.

1969-09-12

Reactor effects of failure of a safety rod to scram can be considered in two major respects: The reduction in total safety system strength which will affect the amount of ``prompt drop`` and subsequent flux decay rate of the average neutron flux-level; and the change in local flux distribution due to the absence of the particular rod which fails to enter the reactor. The purpose of this memorandum is to describe the physical effects involved and to indicate the approximate magnitude of both reactor-wide and localized changes in event of failure of a VSR simultaneous with a K Reactor risermore » accident.« less

Materials technology for an advanced space power nuclear reactor concept: Program summary

NASA Technical Reports Server (NTRS)

Gluyas, R. E.; Watson, G. K.

1975-01-01

The results of a materials technology program for a long-life (50,000 hr), high-temperature (950 C coolant outlet), lithium-cooled, nuclear space power reactor concept are reviewed and discussed. Fabrication methods and compatibility and property data were developed for candidate materials for fuel pins and, to a lesser extent, for potential control systems, reflectors, reactor vessel and piping, and other reactor structural materials. The effects of selected materials variables on fuel pin irradiation performance were determined. The most promising materials for fuel pins were found to be 85 percent dense uranium mononitride (UN) fuel clad with tungsten-lined T-111 (Ta-8W-2Hf).

Thermally actuated thermionic switch

DOEpatents

Barrus, Donald M.; Shires, Charles D.

1988-01-01

A thermally actuated thermionic switch which responds to an increase of temperature by changing from a high impedance to a low impedance at a predictable temperature set point. The switch has a bistable operation mode switching only on temperature increases. The thermionic material may be a metal which is liquid at the desired operation temperature and held in matrix in a graphite block reservoir, and which changes state (ionizes, for example) so as to be electrically conductive at a desired temperature.

Thermionic converter

DOEpatents

Fitzpatrick, Gary O.

1987-05-19

A thermionic converter (10) is set forth which includes an envelope (12) having an electron collector structure (22) attached adjacent to a wall (16). An electron emitter structure (24) is positioned adjacent the collector structure (22) and spaced apart from opposite wall (14). The emitter (24) and collector (22) structures are in a common chamber (20). The emitter structure (24) is heated substantially only by thermal radiation. Very small interelectrode gaps (28) can be maintained utilizing the thermionic converter (10) whereby increased efficiency results.

Thermally actuated thermionic switch

DOEpatents

Barrus, D.M.; Shires, C.D.

1982-09-30

A thermally actuated thermionic switch which responds to an increase of temperature by changing from a high impedance to a low impedance at a predictable temperature set point. The switch has a bistable operation mode switching only on temperature increases. The thermionic material may be a metal which is liquid at the desired operation temperature and held in matrix in a graphite block reservoir, and which changes state (ionizes, for example) so as to be electrically conductive at a desired temperature.

Maintenance method and its critical issues for a fast-ignition laser fusion reactor based on a dry wall chamber

NASA Astrophysics Data System (ADS)

Someya, Y.; Matsumoto, T.; Okano, K.; Asaoka, Y.; Hiwatari, R.; Goto, T.; Ogawa, Y.

2008-05-01

The neutronics analysis has been carried out for feasibility study of the FALCON-D concept by Monte Carlo N-paticle transport code (MCNP), in order to inspect the cooling performance of in-vessel and ex-vessel components, and a connection pipe between Vacuum Vessel and reactor room. The nuclear heating rate in the Vacuum Vessel was at the same level as that of NBI duct of the ITER. The temperature of the connection pipe was found to be 345·, ·which was smaller than the melting point of structure materials (F82H). Moreover, the radiation damage of the final optics was also investigated. We propose a sliding changer concept for replacement. This method could be adapted for the replacement of one FPY cycle in the final optics system.

Pressurizer tank upper support

DOEpatents

Baker, T.H.; Ott, H.L.

1994-01-11

A pressurizer tank in a pressurized water nuclear reactor is mounted between structural walls of the reactor on a substructure of the reactor, the tank extending upwardly from the substructure. For bearing lateral loads such as seismic shocks, a girder substantially encircles the pressurizer tank at a space above the substructure and is coupled to the structural walls via opposed sway struts. Each sway strut is attached at one end to the girder and at an opposite end to one of the structural walls, and the sway struts are oriented substantially horizontally in pairs aligned substantially along tangents to the wall of the circular tank. Preferably, eight sway struts attach to the girder at 90[degree] intervals. A compartment encloses the pressurizer tank and forms the structural wall. The sway struts attach to corners of the compartment for maximum stiffness and load bearing capacity. A valve support frame carrying the relief/discharge piping and valves of an automatic depressurization arrangement is fixed to the girder, whereby lateral loads on the relief/discharge piping are coupled directly to the compartment rather than through any portion of the pressurizer tank. Thermal insulation for the valve support frame prevents thermal loading of the piping and valves. The girder is shimmed to define a gap for reducing thermal transfer, and the girder is free to move vertically relative to the compartment walls, for accommodating dimensional variation of the pressurizer tank with changes in temperature and pressure. 10 figures.

High-Performance Solid-State Thermionic Energy Conversion Based on 2D van der Waals Heterostructures: A First-Principles Study.

PubMed

Wang, Xiaoming; Zebarjadi, Mona; Esfarjani, Keivan

2018-06-18

Two-dimensional (2D) van der Waals heterostructures (vdWHs) have shown multiple functionalities with great potential in electronics and photovoltaics. Here, we show their potential for solid-state thermionic energy conversion and demonstrate a designing strategy towards high-performance devices. We propose two promising thermionic devices, namely, the p-type Pt-G-WSe 2 -G-Pt and n-type Sc-WSe 2 -MoSe 2 -WSe 2 -Sc. We characterize the thermionic energy conversion performance of the latter using first-principles GW calculations combined with real space Green's function (GF) formalism. The optimal barrier height and high thermal resistance lead to an excellent performance. The proposed device is found to have a room temperature equivalent figure of merit of 1.2 which increases to 3 above 600 K. A high performance with cooling efficiency over 30% of the Carnot efficiency above 450 K is achieved. Our designing and characterization method can be used to pursue other potential thermionic devices based on vdWHs.

High throughput semiconductor deposition system

DOEpatents

Young, David L.; Ptak, Aaron Joseph; Kuech, Thomas F.; Schulte, Kevin; Simon, John D.

2017-11-21

A reactor for growing or depositing semiconductor films or devices. The reactor may be designed for inline production of III-V materials grown by hydride vapor phase epitaxy (HVPE). The operating principles of the HVPE reactor can be used to provide a completely or partially inline reactor for many different materials. An exemplary design of the reactor is shown in the attached drawings. In some instances, all or many of the pieces of the reactor formed of quartz, such as welded quartz tubing, while other reactors are made from metal with appropriate corrosion resistant coatings such as quartz or other materials, e.g., corrosion resistant material, or stainless steel tubing or pipes may be used with a corrosion resistant material useful with HVPE-type reactants and gases. Using HVPE in the reactor allows use of lower-cost precursors at higher deposition rates such as in the range of 1 to 5 .mu.m/minute.

ETR, TRA642. EASTWEST SECTION, LOOKING NORTH. PATH OF COOLING WATER ...

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

ETR, TRA-642. EAST-WEST SECTION, LOOKING NORTH. PATH OF COOLING WATER PIPE TUNNEL. WORKING AND STORAGE CANAL. SUB-PILE ROOM. CONTROL ROD ACCESS ROOM. FLOOR NAMES. (THIS WAS A CONCEPT DRAWING.) KAISER ETR-5528-MTR-642-A-5, 11/1955. INL INDEX NO. 532-0642-00-486-100913. REV. 0. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Thermionic Power Cell To Harness Heat Energies for Geothermal Applications

NASA Technical Reports Server (NTRS)

Manohara, Harish; Mojarradi, Mohammad; Greer, Harold F.

2011-01-01

A unit thermionic power cell (TPC) concept has been developed that converts natural heat found in high-temperature environments (460 to 700 C) into electrical power for in situ instruments and electronics. Thermionic emission of electrons occurs when an emitter filament is heated to gwhite hot h temperatures (>1,000 C) allowing electrons to overcome the potential barrier and emit into the vacuum. These electrons are then collected by an anode, and transported to the external circuit for energy storage.

Thermodynamics of photon-enhanced thermionic emission solar cells

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

Reck, Kasper, E-mail: kasper.reck@nanotech.dtu.dk; Hansen, Ole, E-mail: ole.hansen@nanotech.dtu.dk; CINF Center for Individual Nanoparticle Functionality, Technical University of Denmark, Kgs. Lyngby 2800

2014-01-13

Photon-enhanced thermionic emission (PETE) cells in which direct photon energy as well as thermal energy can be harvested have recently been suggested as a new candidate for high efficiency solar cells. Here, we present an analytic thermodynamical model for evaluation of the efficiency of PETE solar cells including an analysis of the entropy production due to thermionic emission of general validity. The model is applied to find the maximum efficiency of a PETE cell for given cathode and anode work functions and temperatures.

Post-Test Analysis of a 10-Year Sodium Heat Pipe Life Test

NASA Technical Reports Server (NTRS)

Rosenfeld, John H.; Locci, Ivan E.; Sanzi, James L.; Hull, David R.; Geng, Steven M.

2011-01-01

High-temperature heat pipes are being evaluated for use in energy conversion applications such as fuel cells, gas turbine re-combustors, Stirling cycle heat sources; and with the resurgence of space nuclear power both as reactor heat removal elements and as radiator elements. Long operating life and reliable performance are critical requirements for these applications. Accordingly, long-term materials compatibility is being evaluated through the use of high-temperature life test heat pipes. Thermacore, Inc., has carried out a sodium heat pipe 10-year life test to establish long-term operating reliability. Sodium heat pipes have demonstrated favorable materials compatibility and heat transport characteristics at high operating temperatures in air over long time periods. A representative one-tenth segment Stirling Space Power Converter heat pipe with an Inconel 718 envelope and a stainless steel screen wick has operated for over 87,000 hr (10 years) at nearly 700 C. These life test results have demonstrated the potential for high-temperature heat pipes to serve as reliable energy conversion system components for power applications that require long operating lifetime with high reliability. Detailed design specifications, operating history, and post-test analysis of the heat pipe and sodium working fluid are described. Lessons learned and future life test plans are also discussed.

Ten Year Operating Test Results and Post-Test Analysis of a 1/10 Segment Stirling Sodium Heat Pipe, Phase III

NASA Technical Reports Server (NTRS)

Rosenfeld, John, H; Minnerly, Kenneth, G; Dyson, Christopher, M.

2012-01-01

High-temperature heat pipes are being evaluated for use in energy conversion applications such as fuel cells, gas turbine re-combustors, Stirling cycle heat sources; and with the resurgence of space nuclear power both as reactor heat removal elements and as radiator elements. Long operating life and reliable performance are critical requirements for these applications. Accordingly, long-term materials compatibility is being evaluated through the use of high-temperature life test heat pipes. Thermacore, Inc., has carried out a sodium heat pipe 10-year life test to establish long-term operating reliability. Sodium heat pipes have demonstrated favorable materials compatibility and heat transport characteristics at high operating temperatures in air over long time periods. A representative one-tenth segment Stirling Space Power Converter heat pipe with an Inconel 718 envelope and a stainless steel screen wick has operated for over 87,000 hr (10 yr) at nearly 700 C. These life test results have demonstrated the potential for high-temperature heat pipes to serve as reliable energy conversion system components for power applications that require long operating lifetime with high reliability. Detailed design specifications, operating history, and post-test analysis of the heat pipe and sodium working fluid are described.

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

DOEpatents

Gluntz, Douglas M.

1996-01-01

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

A CFD Model for High Pressure Liquid Poison Injection for CANDU-6 Shutdown System No. 2

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

Bo Wook Rhee; Chang Jun Jeong; Hye Jeong Yun

2002-07-01

In CANDU reactor one of the two reactor shutdown systems is the liquid poison injection system which injects the highly pressurized liquid neutron poison into the moderator tank via small holes on the nozzle pipes. To ensure the safe shutdown of a reactor it is necessary for the poison curtains generated by jets provide quick, and enough negative reactivity to the reactor during the early stage of the accident. In order to produce the neutron cross section necessary to perform this work, the poison concentration distribution during the transient is necessary. In this study, a set of models for analyzingmore » the transient poison concentration induced by this high pressure poison injection jet activated upon the reactor trip in a CANDU-6 reactor moderator tank has been developed and used to generate the poison concentration distribution of the poison curtains induced by the high pressure jets injected into the vacant region between the pressure tube banks. The poison injection rate through the jet holes drilled on the nozzle pipes is obtained by a 1-D transient hydrodynamic code called, ALITRIG, and this injection rate is used to provide the inlet boundary condition to a 3-D CFD model of the moderator tank based on CFX4.3, a CFD code, to simulate the formation of the poison jet curtain inside the moderator tank. For validation, an attempt was made to validate this model against a poison injection experiment performed at BARC. As conclusion this set of models is judged to be appropriate. (authors)« less

Effect of Silver or Copper Nanoparticles-Dispersed Silane Coatings on Biofilm Formation in Cooling Water Systems

PubMed Central

Ogawa, Akiko; Kanematsu, Hideyuki; Sano, Katsuhiko; Sakai, Yoshiyuki; Ishida, Kunimitsu; Beech, Iwona B.; Suzuki, Osamu; Tanaka, Toshihiro

2016-01-01

Biofouling often occurs in cooling water systems, resulting in the reduction of heat exchange efficiency and corrosion of the cooling pipes, which raises the running costs. Therefore, controlling biofouling is very important. To regulate biofouling, we focus on the formation of biofilm, which is the early step of biofouling. In this study, we investigated whether silver or copper nanoparticles-dispersed silane coatings inhibited biofilm formation in cooling systems. We developed a closed laboratory biofilm reactor as a model of a cooling pipe and used seawater as a model for cooling water. Silver or copper nanoparticles-dispersed silane coating (Ag coating and Cu coating) coupons were soaked in seawater, and the seawater was circulated in the laboratory biofilm reactor for several days to create biofilms. Three-dimensional images of the surface showed that sea-island-like structures were formed on silane coatings and low concentration Cu coating, whereas nothing was formed on high concentration Cu coatings and low concentration Ag coating. The sea-island-like structures were analyzed by Raman spectroscopy to estimate the components of the biofilm. We found that both the Cu coating and Ag coating were effective methods to inhibit biofilm formation in cooling pipes. PMID:28773758

Modeling a Packed Bed Reactor Utilizing the Sabatier Process

NASA Technical Reports Server (NTRS)

Shah, Malay G.; Meier, Anne J.; Hintze, Paul E.

2017-01-01

A numerical model is being developed using Python which characterizes the conversion and temperature profiles of a packed bed reactor (PBR) that utilizes the Sabatier process; the reaction produces methane and water from carbon dioxide and hydrogen. While the specific kinetics of the Sabatier reaction on the RuAl2O3 catalyst pellets are unknown, an empirical reaction rate equation1 is used for the overall reaction. As this reaction is highly exothermic, proper thermal control is of the utmost importance to ensure maximum conversion and to avoid reactor runaway. It is therefore necessary to determine what wall temperature profile will ensure safe and efficient operation of the reactor. This wall temperature will be maintained by active thermal controls on the outer surface of the reactor. Two cylindrical PBRs are currently being tested experimentally and will be used for validation of the Python model. They are similar in design except one of them is larger and incorporates a preheat loop by feeding the reactant gas through a pipe along the center of the catalyst bed. The further complexity of adding a preheat pipe to the model to mimic the larger reactor is yet to be implemented and validated; preliminary validation is done using the smaller PBR with no reactant preheating. When mapping experimental values of the wall temperature from the smaller PBR into the Python model, a good approximation of the total conversion and temperature profile has been achieved. A separate CFD model incorporates more complex three-dimensional effects by including the solid catalyst pellets within the domain. The goal is to improve the Python model to the point where the results of other reactor geometry can be reasonably predicted relatively quickly when compared to the much more computationally expensive CFD approach. Once a reactor size is narrowed down using the Python approach, CFD will be used to generate a more thorough prediction of the reactors performance.

PBF Reactor Building (PER620). Camera faces south along west wall. ...

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

PBF Reactor Building (PER-620). Camera faces south along west wall. Gap between native lava rock and concrete basement walls is being backfilled and compacted. Wire mesh protects workers from falling rock. Note penetrations for piping that will carry secondary coolant water to Cooling Tower. Photographer: Holmes. Date: June 15, 1967. INEEL negative no. 67-3665 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

PROCESS WATER BUILDING, TRA605. FLOOR AND ROOF PLANS FOR SECOND ...

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

PROCESS WATER BUILDING, TRA-605. FLOOR AND ROOF PLANS FOR SECOND FLOOR. DETAILS OF CONCRETE ROOF SLABS. FLASH EVAPORATOR SUPPORTS AND PIPE OPENINGS TO TANKS BELOW. NOTE SPECIFIES THAT EQUIPMENT IS TO BE INSTALLED BEFORE ERECTION OF ROOF AND WALLS. BLAW-KNOX 3150-805-4, 1/1951. INL INDEX NO. 531-0605-62-098-100660, REV. 2. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

PROCESS WATER BUILDING, TRA605. FLOOR PLAN AND SECTION OF FLASH ...

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

PROCESS WATER BUILDING, TRA-605. FLOOR PLAN AND SECTION OF FLASH EVAPORATOR ROOM SHOWING ITS LOCATION ABOVE THE SEAL AND SUMP TANKS. PIPING TAKES WATER FROM SEAL TANK UPWARD TO FLASH EVAPORATORS AND THEN BACK DOWN TO SUMP TANK. BLAW-KNOX 3150-5-6, 8/1950. INL INDEX NO. 531-605-00-098-100011, REV. 3. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Simulations of thermionic suppression during tungsten transient melting experiments

NASA Astrophysics Data System (ADS)

Komm, M.; Tolias, P.; Ratynskaia, S.; Dejarnac, R.; Gunn, J. P.; Krieger, K.; Podolnik, A.; Pitts, R. A.; Panek, R.

2017-12-01

Plasma-facing components receive enormous heat fluxes under steady state and especially during transient conditions that can even lead to tungsten (W) melting. Under these conditions, the unimpeded thermionic current density emitted from the W surfaces can exceed the incident plasma current densities by several orders of magnitude triggering a replacement current which drives melt layer motion via the {\\boldsymbol{J}}× {\\boldsymbol{B}} force. However, in tokamaks, the thermionic current is suppressed by space-charge effects and prompt re-deposition due to gyro-rotation. We present comprehensive results of particle-in-cell modelling using the 2D3V code SPICE2 for the thermionic emissive sheath of tungsten. Simulations have been performed for various surface temperatures and selected inclinations of the magnetic field corresponding to the leading edge and sloped exposures. The surface temperature dependence of the escaping thermionic current and its limiting value are determined for various plasma parameters; for the leading edge geometry, the results agree remarkably well with the Takamura analytical model. For the sloped geometry, the limiting value is observed to be proportional to the thermal electron current and a simple analytical expression is proposed that accurately reproduces the numerical results.

Removal of metal from acid mine drainage using a hybrid system including a pipes inserted microalgae reactor.

PubMed

Park, Young-Tae; Lee, Hongkyun; Yun, Hyun-Shik; Song, Kyung-Guen; Yeom, Sung-Ho; Choi, Jaeyoung

2013-12-01

In this study, the microalgae culture system to combined active treatment system and pipe inserted microalgae reactor (PIMR) was investigated. After pretreated AMD in active treatment system, the effluent load to PIMR in order to Nephroselmis sp. KGE 8 culture. In experiment, effect of iron on growth and lipid accumulation in microalgae were inspected. The 2nd pretreatment effluent was economic feasibility of microalgae culture and lipid accumulation. The growth kinetics of the microalgae are modeled using logistic growth model and the model is primarily parameterized from data obtained through an experimental study where PIMR were dosed with BBM, BBM added 10 mg L(-1) iron and 2nd pretreatment effluent. Moreover, the continuous of microalgae culture in PIMR can be available. Overall, this study indicated that the use of pretreated AMD is a viable method for culture microalgae and lipid accumulation. Copyright © 2013 Elsevier Ltd. All rights reserved.

The IRIS Spool-Type Reactor Coolant Pump

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

Kujawski, J.M.; Kitch, D.M.; Conway, L.E.

2002-07-01

IRIS (International Reactor Innovative and Secure) is a light water cooled, 335 MWe power reactor which is being designed by an international consortium as part of the US DOE NERI Program. IRIS features an integral reactor vessel that contains all the major reactor coolant system components including the reactor core, the coolant pumps, the steam generators and the pressurizer. This integral design approach eliminates the large coolant loop piping, and thus eliminates large loss-of-coolant accidents (LOCAs) as well as the individual component pressure vessels and supports. In addition, IRIS is being designed with a long life core and enhanced safetymore » to address the requirements defined by the US DOE for Generation IV reactors. One of the innovative features of the IRIS design is the adoption of a reactor coolant pump (called 'spool' pump) which is completely contained inside the reactor vessel. Background, status and future developments of the IRIS spool pump are presented in this paper. (authors)« less

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

Advanced thermionic converter development

NASA Technical Reports Server (NTRS)

Huffman, F. N.; Lieb, D.; Briere, T. R.; Sommer, A. H.; Rufeh, F.

1976-01-01

Recent progress at Thermo Electron in developing advanced thermionic converters is summarized with particular attention paid to the development of electrodes, diodes, and triodes. It is found that one class of materials (ZnO, BaO and SrO) provides interesting cesiated work functions (1.3-1.4 eV) without additional oxygen. The second class of materials studied (rare earth oxides and hexaborides) gives cesiated/oxygenated work functions of less than 1.2 eV. Five techniques of oxygen addition to thermionic converters are discussed. Vapor deposited tungsten oxide collector diodes and the reflux converter are considered.

Potentialities of TEC topping: A simplified view of parametric effects

NASA Technical Reports Server (NTRS)

Morris, J. F.

1980-01-01

An examination of the benefits of thermionic-energy-conversion (TEC)-topped power plants and methods of increasing conversion efficiency are discussed. Reductions in the cost of TEC modules yield direct decreases in the cost of electricity (COE) from TEC-topped central station power plants. Simplified COE, overall-efficiency charts presented illustrate this trend. Additional capital-cost diminution results from designing more compact furnaces with considerably increased heat transfer rates allowable and desirable for high temperature TEC and heat pipes. Such improvements can evolve of the protection from hot corrosion and slag as well as the thermal expansion compatibilities offered by silicon-carbide clads on TEC-heating surfaces. Greater efficiencies and far fewer modules are possible with high-temperature, high-power-density TEC: This decreases capital and fuel costs much more and substantially increases electric power outputs for fixed fuel inputs. In addition to more electricity, less pollution, and lower costs, TEC topping used directly in coal-combustion products contributes balance-of-payment gains.

Optimization of 200 MWth and 250 MWt Ship Based Small Long Life NPP

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

Fitriyani, Dian; Su'ud, Zaki

2010-06-22

Design optimization of ship-based 200 MWth and 250 MWt nuclear power reactors have been performed. The neutronic and thermo-hydraulic programs of the three-dimensional X-Y-Z geometry have been developed for the analysis of ship-based nuclear power plant. Quasi-static approach is adopted to treat seawater effect. The reactor are loop type lead bismuth cooled fast reactor with nitride fuel and with relatively large coolant pipe above reactor core, the heat from primary coolant system is directly transferred to watersteam loop through steam generators. Square core type are selected and optimized. As the optimization result, the core outlet temperature distribution is changing withmore » the elevation angle of the reactor system and the characteristics are discussed.« less

Photon-enhanced thermionic emission for solar concentrator systems.

PubMed

Schwede, Jared W; Bargatin, Igor; Riley, Daniel C; Hardin, Brian E; Rosenthal, Samuel J; Sun, Yun; Schmitt, Felix; Pianetta, Piero; Howe, Roger T; Shen, Zhi-Xun; Melosh, Nicholas A

2010-09-01

Solar-energy conversion usually takes one of two forms: the 'quantum' approach, which uses the large per-photon energy of solar radiation to excite electrons, as in photovoltaic cells, or the 'thermal' approach, which uses concentrated sunlight as a thermal-energy source to indirectly produce electricity using a heat engine. Here we present a new concept for solar electricity generation, photon-enhanced thermionic emission, which combines quantum and thermal mechanisms into a single physical process. The device is based on thermionic emission of photoexcited electrons from a semiconductor cathode at high temperature. Temperature-dependent photoemission-yield measurements from GaN show strong evidence for photon-enhanced thermionic emission, and calculated efficiencies for idealized devices can exceed the theoretical limits of single-junction photovoltaic cells. The proposed solar converter would operate at temperatures exceeding 200 degrees C, enabling its waste heat to be used to power a secondary thermal engine, boosting theoretical combined conversion efficiencies above 50%.

Low work function materials for microminiature energy conversion and recovery applications

DOEpatents

Zavadil, Kevin R.; Ruffner, Judith A.; King, Donald B.

2003-05-13

Low work function materials are disclosed together with methods for their manufacture and integration with electrodes used in thermionic conversion applications (specifically microminiature thermionic conversion applications). The materials include a mixed oxide system and metal in a compositionally modulated structure comprised of localized discontinuous structures of material that are deposited using techniques suited to IC manufacture, such as rf sputtering or CVD. The structures, which can include layers are then heated to coalescence yielding a thin film that is both durable and capable of electron emission under thermionic conversion conditions used for microminiature thermionic converters. Using the principles of the invention, thin film electrodes (emitters and collectors) required for microconverter technology are manufactured using a single process deposition so as to allow for full fabrication integration consistent with batch processing, and tailoring of emission/collection properties. In the preferred embodiment, the individual layers include mixed BaSrCaO, scandium oxide and tungsten.

Fatigue crack growth of 316NG austenitic stainless steel welds at 325 °C

NASA Astrophysics Data System (ADS)

Li, Y. F.; Xiao, J.; Chen, Y.; Zhou, J.; Qiu, S. Y.; Xu, Q.

2018-02-01

316NG austenitic stainless steel is a commonly-used material for primary coolant pipes of pressurized water reactor systems. These pipes are usually joined together by automated narrow gap welding process. In this study, welds were prepared by narrow gap welding on 316NG austenitic stainless steel pipes, and its microstructure of the welds was characterized. Then, fatigue crack growth tests were conducted at 325 °C. Precipitates enriched with Mn and Si were found in the fusion zone. The fatigue crack path was out of plane and secondary cracks initiated from the precipitate/matrix interface. A moderate acceleration of crack growth was also observed at 325°Cair and water (DO = ∼10 ppb) with f = 2 Hz.

First-Principles Studies of Structure-Property Relationships: Enabling Design of Functional Materials

NASA Astrophysics Data System (ADS)

Zhou, Qunfei

First-principles calculations based on quantum mechanics have been proved to be powerful for accurately regenerating experimental results, uncovering underlying myths of experimental phenomena, and accelerating the design of innovative materials. This work has been motivated by the demand to design next-generation thermionic emitting cathodes and techniques to allow for synthesis of photo-responsive polymers on complex surfaces with controlled thickness and patterns. For Os-coated tungsten thermionic dispenser cathodes, we used first-principles methods to explore the bulk and surface properties of W-Os alloys in order to explain the previously observed experimental phenomena that thermionic emission varies significantly with W-Os alloy composition. Meanwhile, we have developed a new quantum mechanical approach to quantitatively predict the thermionic emission current density from materials perspective without any semi-empirical approximations or complicated analytical models, which leads to better understanding of thermionic emission mechanism. The methods from this work could be used to accelerate the design of next-generation thermionic cathodes. For photoresponsive materials, we designed a novel type of azobenzene-containing monomer for light-mediated ring-opening metathesis polymerization (ROMP) toward the fabrication of patterned, photo-responsive polymers by controlling ring strain energy (RSE) of the monomer that drives ROMP. This allows for unprecedented remote, noninvasive, instantaneous spatial and temporal control of photo-responsive polymer deposition on complex surfaces.This work on the above two different materials systems showed the power of quantum mechanical calculations on predicting, understanding and discovering the structures and properties of both known and unknown materials in a fast, efficient and reliable way.

Development of a New Design Procedure for Overland Flow System.

DTIC Science & Technology

1982-06-18

reactor kinetics, a concept familiar to most environmental engi- neers. In the case of overland flow, the reactor is the soil surface where various physical...site during the entire study. Perforated plastic pipe was used to distri- bute wastewater along the top of each section, and a bed of crushed stone...particulate BOD. The soluble BOD is oxidized by microorganisms which are probably similar to the attached biomass found in trickling filters. However, some

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

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

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

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

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

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

1993-01-01

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

Heat Pipe Space Nuclear Reactor Design Assessment. Volume 2. Feasibility Study of Upgrading the SP-100 Heat Pipe Space Nuclear Power System.

DTIC Science & Technology

1985-08-01

system thermal -to-electric energy conversion efficiency quite high (-20-25 percent). c. Fuel system--Figure 6 compares the fuel swelling of U02 , UN... high temperature reservoir. Figure 28 shows a schematic of an AMTEC operation. The efficiency of ANTEC is calculated by: IV IV + 1 (L + C AT) + Qloss... thermal energy to electrical energy. The high efficiency and low specific mass (for large systems) are the common advantages of these active systems. In

Rupture loop annex ion exchange RLAIX vault deactivation

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

Ham, J.E.; Harris, D.L., Westinghouse Hanford

This engineering report documents the deactivation, stabilization and final conditions of the Rupture Loop Annex Ion Exchange (RLAIX) Vault located northwest of the 309 Building`s Plutonium Recycle Test Reactor (PRTR). Twelve ion exchange columns, piping debris, and column liquid were removed from the vault, packaged and shipped for disposal. The vault walls and floor were decontaminated, and portions of the vault were painted to fix loose contamination. Process piping and drains were plugged, and the cover blocks and rain cover were installed. Upon closure,the vault was empty, stabilized, isolated.

An evolution strategy for lunar nuclear surface power

NASA Technical Reports Server (NTRS)

Mason, Lee S.

1992-01-01

The production and transmission of electric power for a permanently inhabited lunar base poses a significant challenge which can best be met through an evolution strategy. Nuclear systems offer the best opportunity for evolution in terms of both life and performance. Applicable nuclear power technology options include isotope systems (either radioisotope thermoelectric generators or dynamic isotope power systems) and reactor systems with either static (thermoelectric or thermionic) or dynamic (Brayton, Stirling, Rankine) conversion. A power system integration approach that takes evolution into account would benefit by reduced development and operations cost, progressive flight experience, and simplified logistics, and would permit unrestrained base expansion. For the purposes of defining a nuclear power system evolution strategy, the lunar base development shall consist of four phases: precursor, emplacement, consolidation, and operations.

Nuclear electric propulsion stage requirements and description

NASA Technical Reports Server (NTRS)

Mondt, J. F.; Peelgren, M. L.; Nakashima, A. M.; Nsieh, T. M.; Phillips, W. M.; Kikin, G. M.

1974-01-01

The application of a nuclear electric propulsion (NEP) stage in the exploration of near-earth, cometary, and planetary space was discussed. The NEP stage is powered by a liquid-metal-cooled, fast spectrum thermionic reactor capable of providing 120 kWe for 20,000 hours. This power is used to drive a number of mercury ion bombardment thrusters with specific impulse in the range of 4000-5000 seconds. The NEP description, characteristics, and functional requirements are discussed. These requirements are based on a set of five coordinate missions, which are described in detail. These five missions are a representative part of a larger set of missions used as a basic for an advanced propulsion comparison study. Additionally, the NEP stage development plan and test program is outlined and a schedule presented.

Microminiature thermionic converters

DOEpatents

King, Donald B.; Sadwick, Laurence P.; Wernsman, Bernard R.

2001-09-25

Microminiature thermionic converts (MTCs) having high energy-conversion efficiencies and variable operating temperatures. Methods of manufacturing those converters using semiconductor integrated circuit fabrication and micromachine manufacturing techniques are also disclosed. The MTCs of the invention incorporate cathode to anode spacing of about 1 micron or less and use cathode and anode materials having work functions ranging from about 1 eV to about 3 eV. Existing prior art thermionic converter technology has energy conversion efficiencies ranging from 5-15%. The MTCs of the present invention have maximum efficiencies of just under 30%, and thousands of the devices can be fabricated at modest costs.

An analysis of the temperature dependence of the gate current in complementary heterojunction field-effect transistors

NASA Technical Reports Server (NTRS)

Cunningham, Thomas J.; Fossum, Eric R.; Baier, Steven M.

1992-01-01

The temperature dependence of the gate current versus the gate voltage in complementary heterojunction field-effect transistors (CHFET's) is examined. An analysis indicates that the gate conduction is due to a combination of thermionic emission, thermionic-field emission, and conduction through a temperature-activated resistance. The thermionic-field emission is consistent with tunneling through the AlGaAs insulator. The activation energy of the resistance is consistent with the ionization energy associated with the DX center in the AlGaAs. Methods reducing the gate current are discussed.

Real-time first-principles simulations of thermionic emission from N-doped diamond surfaces

NASA Astrophysics Data System (ADS)

Shinozaki, Tomoki; Hagiwara, Satoshi; Morioka, Naoya; Kimura, Yuji; Watanabe, Kazuyuki

2018-06-01

We investigate thermionic emission from N-doped C(100) surfaces terminated with H or Li atoms using finite-temperature real-time density functional theory simulations. The current–temperature characteristics are found to follow the Richardson–Dushman (RD) equation, which was derived from a semiclassical theory. However, the Richardson constants are two orders of magnitude smaller than the ideal values from the RD theory. This considerable reduction is attributed primarily to the extremely low transmission probability of electrons from the surfaces toward the vacuum. The present method enables straightforward evaluation of the ideal efficiency of a thermionic energy converter.

Thermionic Energy Conversion (TEC) topping thermoelectrics

NASA Technical Reports Server (NTRS)

Morris, J. F.

1981-01-01

Performance expectations for thermionic and thermoelectric energy conversion systems are reviewed. It is noted that internal radiation effects diminish thermoelectric figures of merit significantly at 1000 K and substantially at 2000 K; the effective thermal conductivity contribution of intrathermoelectric radiative dissipation increases with the third power of temperature. It is argued that a consideration of thermoelectric power generation with high temperature heat sources should include utilization of thermionic energy conversion (TEC) topping thermoelectrics. However TEC alone or TEC topping more efficient conversion systems like steam or gas turbines, combined cycles, or Stirling engines would be more desirable generally.

Analysis of failed nuclear plant components

NASA Astrophysics Data System (ADS)

Diercks, D. R.

1993-12-01

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

Using heat pipe to make isotherm condition in catalytic converters of sulfuric acid plants

NASA Astrophysics Data System (ADS)

Yousefi, M.; Pahlavanzadeh, H.; Sadrameli, S. M.

2017-08-01

In this study, for the first time, it is tried to construct a pilot reactor, for surveying the possibility of creating isothermal condition in the catalytic convertors where SO2 is converted to SO3 in the sulfuric acid plants by heat pipe. The thermodynamic and thermo-kinetic conditions were considered the same as the sulfuric acid plants converters. Also, influence of SO2 gas flow rate on isothermal condition, has been studied. A thermo-siphon type heat pipe contains the sulfur + 5% iodine as working fluid, was used for disposing the heat of reaction from catalytic bed. Our results show that due to very high energy-efficiency, isothermal and passive heat transfer mechanism of heat pipe, it is possible to reach more than 95% conversion in one isothermal catalytic bed. As the results, heat pipe can be used as a certain piece of equipment to create isothermal condition in catalytic convertors of sulphuric acid plants. With this work a major evaluation in design of sulphuric acid plants can be taken place.

The design and fabrication of a Stirling engine heat exchanger module with an integral heat pipe

NASA Technical Reports Server (NTRS)

Schreiber, Jeffrey G.

1988-01-01

The conceptual design of a free-piston Stirling Space Engine (SSE) intended for space power applications has been generated. The engine was designed to produce 25 kW of electric power with heat supplied by a nuclear reactor. A novel heat exchanger module was designed to reduce the number of critical joints in the heat exchanger assembly while also incorporating a heat pipe as the link between the engine and the heat source. Two inexpensive verification tests are proposed. The SSE heat exchanger module is described and the operating conditions for the module are outlined. The design process of the heat exchanger modules, including the sodium heat pipe, is briefly described. Similarities between the proposed SSE heat exchanger modules and the LeRC test modules for two test engines are presented. The benefits and weaknesses of using a sodium heat pipe to transport heat to a Stirling engine are discussed. Similarly, the problems encountered when using a true heat pipe, as opposed to a more simple reflux boiler, are described. The instruments incorporated into the modules and the test program are also outlined.

Non-Intrusive Velocity Measurements with MTV During DCC Event in the HTTF

NASA Technical Reports Server (NTRS)

Andre, M. A.; Bardet, P. M.; Cadell, S. R.; Woods, B.; Burns, R. A.; Danehy, P. M.

2017-01-01

Velocity profiles are measured using molecular tagging velocimetry (MTV) in the high temperature test facility (HTTF) at Oregon State University during a depressurized conduction cooldown (DCC) event. The HTTF is a quarter scale electrically heated nuclear reactor simulator designed to replicate various accident scenarios. During a DCC, a double ended guillotine break results in the reactor pressure vessel (RPV) depressurizing into the reactor cavity and ultimately leading to air ingress in the reactor core (lock-exchange and gas diffusion). It is critical to understand the resulting buoyancy-driven flow to characterize the reactor self-cooling capacity through natural circulation. During tests conducted at ambient pressure and temperature, the RPV containing helium is opened (via the hot and cold legs) to a large vessel filled with nitrogen to simulate the atmosphere. The velocity profile on the hot leg pipe centerline is recorded at 10 Hz with MTV based on NO tracers. The precision of the velocimetry was measured to be 0.02 m/s in quiescent flow prior to the tests. A helium flow from the RPV is initially observed in the top quarter of the pipe. During the first 20 seconds of the event, helium flows out of the RPV with a maximum velocity below 2 m/s. The velocity profile transitions from parabolic to linear in character and decays slowly over the rest of the recording; peak velocities of 0.2 m/s are observed after 30 min. A counter-flow of nitrogen is also observed intermittently, which occurs at lower velocities (>0.1 m/s).

3D-FE Modeling of 316 SS under Strain-Controlled Fatigue Loading and CFD Simulation of PWR Surge Line

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

Mohanty, Subhasish; Barua, Bipul; Listwan, Joseph

In financial year 2017, we are focusing on developing a mechanistic fatigue model of surge line pipes for pressurized water reactors (PWRs). To that end, we plan to perform the following tasks: (1) conduct stress- and strain-controlled fatigue testing of surge-line base metal such as 316 stainless steel (SS) under constant, variable, and random fatigue loading, (2) develop cyclic plasticity material models of 316 SS, (3) develop one-dimensional (1D) analytical or closed-form model to validate the material models and to understand the mechanics associated with 316 SS cyclic hardening and/or softening, (4) develop three-dimensional (3D) finite element (FE) models withmore » implementation of evolutionary cyclic plasticity, and (5) develop computational fluid dynamics (CFD) model for thermal stratification, thermal-mechanical stress, and fatigue of example reactor components, such as a PWR surge line under plant heat-up, cool-down, and normal operation with/without grid-load-following. This semi-annual progress report presents the work completed on the above tasks for a 316 SS laboratory-scale specimen subjected to strain-controlled cyclic loading with constant, variable, and random amplitude. This is the first time that the accurate 3D-FE modeling of the specimen for its entire fatigue life, including the hardening and softening behavior, has been achieved. We anticipate that this work will pave the way for the development of a fully mechanistic-computer model that can be used for fatigue evaluation of safety-critical metallic components, which are traditionally evaluated by heavy reliance on time-consuming and costly test-based approaches. This basic research will not only help the nuclear reactor industry for fatigue evaluation of reactor components in a cost effective and less time-consuming way, but will also help other safety-related industries, such as aerospace, which is heavily dependent on test-based approaches, where a single full-scale fatigue test can cost millions of dollars and require years of effort to conduct. Toward our goal of demonstration of fully mechanistic fatigue evaluation of reactor components, we also started work on developing a component-level computer model of reactor components, such as 316 SS surge line pipe. This requires developing a thermal-mechanical stress analysis model of the reactor surge line, which, in turn, requires time-dependent temperature and stratification information along the boundary of the pipe. Toward that goal, CFD models of surge lines are being developed. In this report, we also present some preliminary results showing the temperature conditions along the surge line wall under reactor heat-up, cool-down, and steady-state power operation.« less

Kilowatt Reactor Using Stirling TechnologY (KRUSTY) Demonstration. CEDT Phase 1 Preliminary Design Documentation

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

Sanchez, Rene Gerardo; Hutchinson, Jesson D.; Mcclure, Patrick Ray

2015-08-20

The intent of the integral experiment request IER 299 (called KiloPower by NASA) is to assemble and evaluate the operational performance of a compact reactor configuration that closely resembles the flight unit to be used by NASA to execute a deep space exploration mission. The reactor design will include heat pipes coupled to Stirling engines to demonstrate how one can generate electricity when extracting energy from a “nuclear generated” heat source. This series of experiments is a larger scale follow up to the DUFF series of experiments1,2 that were performed using the Flat-Top assembly.

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

DOEpatents

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

1993-01-01

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

THREADED ADAPTOR FOR LUGGED PIPE ENDS

DOEpatents

Robb, J.E.

1962-06-01

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

Power Management and Distribution System Developed for Thermionic Power Converters

NASA Technical Reports Server (NTRS)

Baez, Anastacio N.

1998-01-01

A spacecraft solar, bimodal system combines propulsion and power generation into a single integrated system. An Integrated Solar Upper Stage (ISUS) provides orbital transfer capabilities, power generation for payloads, and onboard propulsion to the spacecraft. A key benefit of a bimodal system is a greater payload-to-spacecraft mass ratio resulting in lower launch vehicle requirements. Scaling down to smaller launch vehicles increases space access by reducing overall mission cost. NASA has joined efforts with the Air Force Phillips Laboratory to develop enabling technologies for such a system. The NASA/Air Force bimodal concept uses solar concentrators to focus energy into an integrated power plant. This power plant consists of a graphite core that stores thermal energy within a cavity. An array of thermionic converters encircles the graphite cavity and provides electrical energy conversion functions. During the power generation phase of the bimodal system, the thermionic converters are exposed to the heated cavity and convert the thermal energy to electricity. Near-term efforts of the ISUS bimodal program are focused on a ground demonstration of key technologies in order to proceed to a full space flight test. Thermionic power generation is one key technology of the bimodal concept. Thermionic power converters impose unique operating requirements upon a power management and distribution (PMAD) system design. Single thermionic converters supply large currents at very low voltages. Operating voltages can vary over a range of up to 3 to 1 as a function of operating temperature. Most spacecraft loads require regulated 28-volts direct-current (Vdc) power. A combination of series-connected converters and powerprocessing boosters is required to deliver power to the spacecraft's payloads at this level.

WATER PUMP HOUSE, TRA619. VIEW OF PUMP HOUSE UNDER CONSTRUCTION. ...

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

WATER PUMP HOUSE, TRA-619. VIEW OF PUMP HOUSE UNDER CONSTRUCTION. CAMERA IS ON WATER TOWER AND FACES NORTHWEST. TWO RESERVOIR TANKS ALREADY ARE COMPLETED. NOTE EXCAVATIONS FOR PIPE LINES EXITING FROM BELOW GROUND ON SOUTH SIDE OF PUMP HOUSE. BUILDING AT LOWER RIGHT IS ELECTRICAL CONTROL BUILDING, TRA-623. SWITCHYARD IS IN LOWER RIGHT CORNER OF VIEW. INL NEGATIVE NO. 2753. Unknown Photographer, ca. 6/1951 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Development of a Rational Design Procedure for Overland Flow Systems,

DTIC Science & Technology

1982-02-01

to pre- report is based on reactor kinetics, a concept dict the hydraulic detention time. familiar to most environmental engineers. In the 2...Determine the removal kinetics for BOD, TSS, case of overland flow, the reactor is the soil surface NH3 -N and total P. where various physical, biological and...water along the top of each section, and a bed of down. The amount of time needed to reach hydraulic crushed stone placed beneath the pipe helped to

U.S. Proliferation Policy and the Campaign Against Transnational Terror: Linking the U.S. Non-Proliferation Regime to Homeland Security Efforts

DTIC Science & Technology

2013-12-01

device (IED) in a public park is more difficult than setting off an IED in a secured government building . Alternatively, constructing a pipe bomb is...against nuclear and power industry installations” intended to “seize nuclear materials and use them to build WMD for their own political use.” 5...reactor under construction .469 Damascus faces unresolved allegations that it illicitly tried to build a plutonium production reactor at a site

Synfuel production in nuclear reactors

DOEpatents

Henning, C.D.

Apparatus and method for producing synthetic fuels and synthetic fuel components by using a neutron source as the energy source, such as a fusion reactor. Neutron absorbers are disposed inside a reaction pipe and are heated by capturing neutrons from the neutron source. Synthetic fuel feedstock is then placed into contact with the heated neutron absorbers. The feedstock is heated and dissociates into its constituent synfuel components, or alternatively is at least preheated sufficiently to use in a subsequent electrolysis process to produce synthetic fuels and synthetic fuel components.

Catalytic cartridge SO/sub 3/ decomposer

DOEpatents

Galloway, T.R.

1980-11-18

A catalytic cartridge surrounding a heat pipe driven by a heat source is utilized as a SO/sub 3/ decomposer for thermochemical hydrogen production. The cartridge has two embodiments, a cross-flow cartridge and an axial flow cartridge. In the cross-flow cartridge, SO/sub 3/ gas is flowed through a chamber and incident normally to a catalyst coated tube extending through the chamber, the catalyst coated tube surrounding the heat pipe. In the axial-flow cartridge, SO/sub 3/ gas is flowed through the annular space between concentric inner and outer cylindrical walls, the inner cylindrical wall being coated by a catalyst and surrounding the heat pipe. The modular cartridge decomposer provides high thermal efficiency, high conversion efficiency, and increased safety. A fusion reactor may be used as the heat source.

DISTRIBUTION SYSTEMS AS RESERVOIRS AND REACTORS FOR INORGANIC CONTAMINANTS

EPA Science Inventory

This paper provides a review of numerous drinking water and geochemical investigations and recent studies of pipe deposits and water treatment materials. This analysis shows that there is growing evidence from analogous natural water systems and some analytical studies that many ...

Inherently Safe Fission Power System for Lunar Outposts

NASA Astrophysics Data System (ADS)

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

2013-09-01

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

Deployment history and design considerations for space reactor power systems

NASA Astrophysics Data System (ADS)

El-Genk, Mohamed S.

2009-05-01

The history of the deployment of nuclear reactors in Earth orbits is reviewed with emphases on lessons learned and the operation and safety experiences. The former Soviet Union's "BUK" power systems, with SiGe thermoelectric conversion and fast neutron energy spectrum reactors, powered a total of 31 Radar Ocean Reconnaissance Satellites (RORSATs) from 1970 to 1988 in 260 km orbit. Two of the former Soviet Union's TOPAZ reactors, with in-core thermionic conversion and epithermal neutron energy spectrum, powered two Cosmos missions launched in 1987 in ˜800 km orbit. The US' SNAP-10A system, with SiGe energy conversion and a thermal neutron energy spectrum reactor, was launched in 1965 in 1300 km orbit. The three reactor systems used liquid NaK-78 coolant, stainless steel structure and highly enriched uranium fuel (90-96 wt%) and operated at a reactor exit temperature of 833-973 K. The BUK reactors used U-Mo fuel rods, TOPAZ used UO 2 fuel rods and four ZrH moderator disks, and the SNAP-10A used moderated U-ZrH fuel rods. These low power space reactor systems were designed for short missions (˜0.5 kW e and ˜1 year for SNAP-10A, <3.0 kW e and <6 months for BUK, and ˜5.5 kW e and up to 1 year for TOPAZ). The deactivated BUK reactors at the end of mission, which varied in duration from a few hours to ˜4.5 months, were boosted into ˜800 km storage orbit with a decay life of more than 600 year. The ejection of the last 16 BUK reactor fuel cores caused significant contamination of Earth orbits with NaK droplets that varied in sizes from a few microns to 5 cm. Power systems to enhance or enable future interplanetary exploration, in-situ resources utilization on Mars and the Moon, and civilian missions in 1000-3000 km orbits would generate significantly more power of 10's to 100's kW e for 5-10 years, or even longer. A number of design options to enhance the operation reliability and safety of these high power space reactor power systems are presented and discussed.

Thermionic gas switch

DOEpatents

Hatch, G.L.; Brummond, W.A.; Barrus, D.M.

1984-04-05

The present invention is directed to an improved temperature responsive thermionic gas switch utilizing a hollow cathode and a folded emitter surface area. The folded emitter surface area of the thermionic switch substantially increases the on/off ratio by changing the conduction surface area involved in the two modes thereof. The improved switch of this invention provides an on/off ratio of 450:1 compared to the 10:1 ratio of the prior known thermionic switch, while providing for adjusting the on current. In the improved switch of this invention the conduction area is made small in the off mode, while in the on mode the conduction area is made large. This is achieved by utilizing a folded hollow cathode configuration and utilizing a folded emitter surface area, and by making the dimensions of the folds small enough so that a space charge will develop in the convolutions of the folds and suppress unignited current, thus limiting the current carrying surface in the off mode.

Photo-assisted electron emission from illuminated monolayer graphene

NASA Astrophysics Data System (ADS)

Upadhyay Kahaly, M.; Misra, Shikha; Mishra, S. K.

2017-05-01

We establish a formalism to address co-existing and complementing thermionic and photoelectric emission from a monolayer graphene sheet illuminated via monochromatic laser radiation and operating at a finite temperature. Taking into account the two dimensional Fermi-Dirac statistics as is applicable for a graphene sheet, the electron energy redistribution due to thermal agitation via laser irradiation, and Fowler's approach of the electron emission, along with Born's approximation to evaluate the tunneling probability, the expressions for the photoelectric and thermionic emission flux have been derived. The cumulative emission flux is observed to be sensitive to the parametric tuning of the laser and material specifications. Based on the parametric analysis, the photoemission flux is noticed to dominate over its coexisting counterpart thermionic emission flux for smaller values of the material work function, surface temperature, and laser wavelength; the analytical estimates are in reasonably good agreement with the recent experimental observations [Massicotte et al., Nat. Commun. 7, 12174 (2016)]. The results evince the efficient utilization of a graphene layer as a photo-thermionic emitter.

Thermionic cogeneration burner design

NASA Astrophysics Data System (ADS)

Miskolczy, G.; Goodale, D.; Moffat, A. L.; Morgan, D. T.

Since thermionic converters receive heat at very high temperatures (approximately 1800 K) and reject heat at moderately high temperatures (approximately 800 K), they are useful for cogeneration applications involving high temperature processes. The electric power from thermionic converters is produced as a high amperage, low-voltage direct current. An ideal cogeneration application would be to utilize the reject heat at the collector temperature and the electricity without power conditioning. A cogeneration application in the edible oil industry fulfills both of these requirements since both direct heat and hydrogen gas are required in the hydrogenation of the oils. In this application, the low-voltage direct current would be used in a hydrogen electrolyzer.

Theory and simulation of backbombardment in single-cell thermionic-cathode electron guns

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

Edelen, J.  P.; Biedron, S.  G.; Harris, J.  R.

This paper presents a comparison between simulation results and a first principles analytical model of electron back-bombardment developed at Colorado State University for single-cell, thermionic-cathode rf guns. While most previous work on back-bombardment has been specific to particular accelerator systems, this work is generalized to a wide variety of guns within the applicable parameter space. The merits and limits of the analytic model will be discussed. This paper identifies the three fundamental parameters that drive the back-bombardment process, and demonstrates relative accuracy in calculating the predicted back-bombardment power of a single-cell thermionic gun.

Theory and simulation of backbombardment in single-cell thermionic-cathode electron guns

DOE PAGES

Edelen, J.  P.; Biedron, S.  G.; Harris, J.  R.; ...

2015-04-01

This paper presents a comparison between simulation results and a first principles analytical model of electron back-bombardment developed at Colorado State University for single-cell, thermionic-cathode rf guns. While most previous work on back-bombardment has been specific to particular accelerator systems, this work is generalized to a wide variety of guns within the applicable parameter space. The merits and limits of the analytic model will be discussed. This paper identifies the three fundamental parameters that drive the back-bombardment process, and demonstrates relative accuracy in calculating the predicted back-bombardment power of a single-cell thermionic gun.

Design, construction and measurements of an alpha magnet as a solution for compact bunch compressor for the electron beam from Thermionic RF Gun

NASA Astrophysics Data System (ADS)

Rajabi, A.; Jazini, J.; Fathi, M.; Sharifian, M.; Shokri, B.

2018-03-01

The beam produced by a thermionic RF gun has wide energy spread that makes it unsuitable for direct usage in photon sources. Here in the present work, we optimize the extracted beam from a thermionic RF gun by a compact economical bunch compressor. A compact magnetic bunch compressor (Alpha magnet) is designed and constructed. A comparison between simulation results and experimental measurements shows acceptable conformity. The beam dynamics simulation results show a reduction of the energy spread as well as a compression of length less than 1 ps with 2.3 mm-mrad emittance.

Reactivity Studies of Inconel 625 with Sodium, and Lunar Regolith Stimulant

NASA Technical Reports Server (NTRS)

Gillies, Donald; Salvail, Pat; Reid, Bob; Colebaugh, James; Easterling, Greg

2008-01-01

In the event of the need for nuclear power in exploration, high flux heat pipes will be needed for heat transfer from space nuclear reactors to various energy conversion devices, and to safely dissipate excess heat. Successful habitation will necessitate continuous operation of alkali metal filled heat pipes for 10 or-more years in a hostile environment with little maintenance. They must be chemical and creep resistant in the high vacuum of space (lunar), and they must operate reliably in low gravity conditions with intermittent high radiation fluxes. One candidate material for the heat pipe shell, namely Inconel 625, has been tested to determine its compatibility with liquid sodium. Any reactivity could manifest itself as a problem over the long time periods anticipated. In addition, possible reactions with the lunar regolith will take place, as will evaporation of selected elements at the external surfaces of the heat pipes, and so there is a need for extensive long-term testing under simulated lunar conditions.

Feasibility study of a passive aeration reactor equipped with vertical pipes for compost stabilization of cow manure.

PubMed

Sylla, Youssouf Boundou; Kuroda, Masao; Yamada, Masayuki; Matsumoto, Naoko

2006-10-01

Pilot-scale composting was carried out with cow manure to evaluate the performances of two passive aeration systems: a conventional passive aeration system equipped with horizontal pipes and an unusual passive aeration method based on air delivery by means of vertical pipes. The effects of both types of passive aeration apparatus were investigated in order to determine the degree of composting rate by continuously monitoring temperature, moisture content, organic matter, electrical conductivity, pH and C/N ratio in the piles. Temperatures in the range of thermophily (55-65 degrees C) were reached in all runs within 1-2 days then lasting for about 1 week, a span long enough for pathogen abatement. Results suggest that passive aeration carried out by vertical pipes is more effective for air delivery into compost piles than conventional passive aeration of air adduction with horizontal pipes. The variation in the number of vertical pipes was revealed to be an important parameter for the control of composting rate and temperature. Composting rates estimated from the heat balance equation were substantially in agreement with those computed through the conversion ratio of total organic matter decrement. The conversion ratios and composting rates obtained in this study using passive aeration with vertical pipes were well aligned with those found using forced air delivery systems.

Phase 1 Space Fission Propulsion Energy Source Design

NASA Technical Reports Server (NTRS)

Houts, Mike; VanDyke, Melissa; Godfroy, Tom; Pedersen, Kevin; Martin, James; Dickens, Ricky; Salvail, Pat; Hrbud, Ivana; Carter, Robert; Rodgers, Stephen L. (Technical Monitor)

2002-01-01

Fission technology can enable rapid, affordable access to any point in the solar system. If fission propulsion systems are to be developed to their full potential; however, near-term customers must be identified and initial fission systems successfully developed, launched, and operated. Studies conducted in fiscal year 2001 (IISTP, 2001) show that fission electric propulsion (FEP) systems with a specific mass at or below 50 kg/kWjet could enhance or enable numerous robotic outer solar system missions of interest. At the required specific mass, it is possible to develop safe, affordable systems that meet mission requirements. To help select the system design to pursue, eight evaluation criteria were identified: system integration, safety, reliability, testability, specific mass, cost, schedule, and programmatic risk. A top-level comparison of four potential concepts was performed: a Testable, Passive, Redundant Reactor (TPRR), a Testable Multi-Cell In-Core Thermionic Reactor (TMCT), a Direct Gas Cooled Reactor (DGCR), and a Pumped Liquid Metal Reactor.(PLMR). Development of any of the four systems appears feasible. However, for power levels up to at least 500 kWt (enabling electric power levels of 125-175 kWe, given 25-35% power conversion efficiency) the TPRR has advantages related to several criteria and is competitive with respect to all. Hardware-based research and development has further increased confidence in the TPRR approach. Successful development and utilization of a "Phase I" fission electric propulsion system will enable advanced Phase 2 and Phase 3 systems capable of providing rapid, affordable access to any point in the solar system.

Amplified Thermionic Cooling Using Arrays of Nanowires

NASA Technical Reports Server (NTRS)

Yang, Eui-Hyeok; Choi, Daniel; Shcheglov, Kirill; Hishinuma, Yoshikazu

2007-01-01

A class of proposed thermionic cooling devices would incorporate precise arrays of metal nanowires as electron emitters. The proposed devices could be highly miniaturized, enabling removal of heat from locations, very close to electronic devices, that have previously been inaccessible for heat-removal purposes. The resulting enhancement of removal of heat would enable operation of the devices at higher power levels and higher clock speeds. Moreover, the mass, complexity, and bulk of electronic circuitry incorporating these highly miniaturized cooling devices could be considerably reduced, relative to otherwise equivalent circuitry cooled by conventional electromechanical, thermoelectric, and fluidic means. In thermionic cooling, one exploits the fact that because only the highest-energy electrons are thermionically emitted, collecting those electrons to prevent their return to the emitting electrode results in the net removal of heat from that electrode. Collection is effected by applying an appropriate positive bias potential to another electrode placed near the emitting electrode. The concept underlying the proposal is that the thermionic-emission current and, hence, the cooling effect attainable by use of an array of nanowires could be significantly greater than that attainable by use of a single emitting electrode or other electron- emitting surface. The wires in an array according to the proposal would protrude perpendicularly from a planar surface and their heights would be made uniform to within a sub-nanometer level of precision

Application of USNRC NUREG/CR-6661 and draft DG-1108 to evolutionary and advanced reactor designs

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

Chang 'Apollo', Chen

2006-07-01

For the seismic design of evolutionary and advanced nuclear reactor power plants, there are definite financial advantages in the application of USNRC NUREG/CR-6661 and draft Regulatory Guide DG-1108. NUREG/CR-6661, 'Benchmark Program for the Evaluation of Methods to Analyze Non-Classically Damped Coupled Systems', was by Brookhaven National Laboratory (BNL) for the USNRC, and Draft Regulatory Guide DG-1108 is the proposed revision to the current Regulatory Guide (RG) 1.92, Revision 1, 'Combining Modal Responses and Spatial Components in Seismic Response Analysis'. The draft Regulatory Guide DG-1108 is available at http://members.cox.net/apolloconsulting, which also provides a link to the USNRC ADAMS site to searchmore » for NUREG/CR-6661 in text file or image file. The draft Regulatory Guide DG-1108 removes unnecessary conservatism in the modal combinations for closely spaced modes in seismic response spectrum analysis. Its application will be very helpful in coupled seismic analysis for structures and heavy equipment to reduce seismic responses and in piping system seismic design. In the NUREG/CR-6661 benchmark program, which investigated coupled seismic analysis of structures and equipment or piping systems with different damping values, three of the four participants applied the complex mode solution method to handle different damping values for structures, equipment, and piping systems. The fourth participant applied the classical normal mode method with equivalent weighted damping values to handle differences in structural, equipment, and piping system damping values. Coupled analysis will reduce the equipment responses when equipment, or piping system and structure are in or close to resonance. However, this reduction in responses occurs only if the realistic DG-1108 modal response combination method is applied, because closely spaced modes will be produced when structure and equipment or piping systems are in or close to resonance. Otherwise, the conservatism in the current Regulatory Guide 1.92, Revision 1, will overshadow the advantage of coupled analysis. All four participants applied the realistic modal combination method of DG-1108. Consequently, more realistic and reduced responses were obtained. (authors)« less

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

Kot, C.A.; Srinivasan, M.G.; Hsieh, B.J.

As part of the Phase II testing at the HDR Test Facility in Kahl/Main, FRG, two series of high-level seismic/vibrational experiments were performed. In the first of these (SHAG) a coast-down shaker, mounted on the reactor operating floor and capable of generating 1000 tonnes of force, was used to investigate full-scale structural response, soil-structure interaction (SSI), and piping/equipment response at load levels equivalent to those of a design basis earthquake. The HDR soil/structure system was tested to incipient failure exhibiting highly nonlinear response. In the load transmission from structure to piping/equipment significant response amplifications and shifts to higher frequencies occurred.more » The performance of various pipe support configurations was evaluated. This latter effort was continued in the second series of tests (SHAM), in which an in-plant piping system was investigated at simulated seismic loads (generated by two servo-hydraulic actuators each capable of generating 40 tonnes of force), that exceeded design levels manifold and resulted in considerable pipe plastification and failure of some supports (snubbers). The evaluation of six different support configurations demonstrated that proper system design (for a given spectrum) rather than number of supports or system stiffness is essential to limiting pipe stresses. Pipe strains at loads exceeding the design level eightfold were still tolerable, indicating that pipe failure even under extreme seismic loads is unlikely inspite of multiple support failures. Conservatively, an excess capacity (margin) of at least four was estimated for the piping system, and the pipe damping was found to be 4%. Comparisons of linear and nonlinear computational results with measurements showed that analytical predictions have wide scatter and do not necessarily yield conservative responses, underpredicting, in particular, peak support forces.« less

NUCLEAR REACTOR

DOEpatents

Christy, R.F.

1958-07-15

A nuclear reactor of the homogeneous liquid fuel type is described wherein the fissionable isotope is suspended or dissolved in a liquid moderator such as water. The reactor core is comprised essentially of a spherical vessel for containing the reactive composition surrounded by a reflector, preferably of beryllium oxide. The reactive composition may be an ordinary water solution of a soluble salt of uranium, the quantity of fissionable isotope in solution being sufficient to provide a critical mass in the vessel. The liquid fuel is stored in a tank of non-crtttcal geometry below the reactor vessel and outside of the reflector and is passed from the tank to the vessel through a pipe connecting the two by air pressure means. Neutron absorbing control and safety rods are operated within slots in the reflector adjacent to the vessel.

Thermal synthesis apparatus

DOEpatents

Fincke, James R [Idaho Falls, ID; Detering, Brent A [Idaho Falls, ID

2009-08-18

An apparatus for thermal conversion of one or more reactants to desired end products includes an insulated reactor chamber having a high temperature heater such as a plasma torch at its inlet end and, optionally, a restrictive convergent-divergent nozzle at its outlet end. In a thermal conversion method, reactants are injected upstream from the reactor chamber and thoroughly mixed with the plasma stream before entering the reactor chamber. The reactor chamber has a reaction zone that is maintained at a substantially uniform temperature. The resulting heated gaseous stream is then rapidly cooled by passage through the nozzle, which "freezes" the desired end product(s) in the heated equilibrium reaction stage, or is discharged through an outlet pipe without the convergent-divergent nozzle. The desired end products are then separated from the gaseous stream.

Heavy-section steel technology program. Semiannual progress report for period ending February 28, 1973

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

None

1974-02-01

The materials investigations under the HSST program are divided into studies of unirradiated materials and studies of irradiation effects. The studies of unirradiated materials, which include inspection, characterization, metallurgy, variability determinations, transition temperature investigations, fracture mechanics studies, and fatigue-crack propagation tests, are discussed. The investigations of irradiated materials include studies of radiation effects on A-533-B steel. Results of studies on thick pressure vessels and pipes of ASTM A508 steel are also reported along with results of studies on Mode III crack extension in reactor piping. (JRD)

SPERTI/PBF. Contextual aerial view after PBF had begun operating, but ...

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

SPERT-I/PBF. Contextual aerial view after PBF had begun operating, but prior to expansion of southwest corner of Reactor Building (PER-620). Camera facing northeast. Reactor Building in center of view. Cooling Tower (PER-720) to its left. Warehouse (PER-625) at lower left was built in 1966. SPERT-I Reactor Building (PER-605) and Instrument Cell Building (PER-604) at right of view. Buried cables and piping proceed from PBF toward lower edge of view to Control Building further south and out of view. Photographer: Farmer. Date: March 26, 1976. INEEL negative no. 76-1344 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

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

DOEpatents

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

1993-12-14

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

CLOSURE DEVICE

DOEpatents

Linzell, S.M.; Dorcy, D.J.

1958-08-26

A quick opening type of stuffing box employing two banks of rotatable shoes, each of which has a caraming action that forces a neoprene sealing surface against a pipe or rod where it passes through a wall is presented. A ring having a handle or wrench attached is placed eccentric to and between the two banks of shoes. Head bolts from the shoes fit into slots in this ring, which are so arranged that when the ring is rotated a quarter turn in one direction the shoes are thrust inwardly to cramp the neopnrene about the pipe, malting a tight seal. Moving the ring in the reverse direction moves the shoes outwardly and frees the pipe which then may be readily removed from the stuffing box. This device has particular application as a closure for the end of a coolant tube of a neutronic reactor.

Lightweight Exhaust Manifold and Exhaust Pipe Ducting for Internal Combustion Engines

NASA Technical Reports Server (NTRS)

Northam, G. Burton (Inventor); Ransone, Philip O. (Inventor); Rivers, H. Kevin (Inventor)

1999-01-01

An improved exhaust system for an internal combustion gasoline-and/or diesel-fueled engine includes an engine exhaust manifold which has been fabricated from carbon- carbon composite materials in operative association with an exhaust pipe ducting which has been fabricated from carbon-carbon composite materials. When compared to conventional steel. cast iron. or ceramic-lined iron paris. the use of carbon-carbon composite exhaust-gas manifolds and exhaust pipe ducting reduces the overall weight of the engine. which allows for improved acceleration and fuel efficiency: permits operation at higher temperatures without a loss of strength: reduces the "through-the wall" heat loss, which increases engine cycle and turbocharger efficiency and ensures faster "light-off" of catalytic converters: and, with an optional thermal reactor, reduces emission of major pollutants, i.e. hydrocarbons and carbon monoxide.

Residual stress determination in an overlay dissimilar welded pipe by neutron diffraction

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

Woo, Wan Chuck; Em, Vyacheslav; Hubbard, Camden R

2011-01-01

Residual stresses were determined through the thickness of a dissimilar weld overlay pipe using neutron diffraction. The specimen has a complex joining structure consisting of a ferritic steel (SA508), austenitic steel (F316L), Ni-based consumable (Alloy 182), and overlay of Ni-base superalloy (Alloy 52M). It simulates pressurized nozzle components, which have been a critical issue under the severe crack condition of nuclear power reactors. Two neutron diffractometers with different spatial resolutions have been utilized on the identical specimen for comparison. The macroscopic 'stress-free' lattice spacing (d{sub o}) was also obtained from both using a 2-mm width comb-like coupon. The results showmore » significant changes in residual stresses from tension (300-400 MPa) to compression (-600 MPa) through the thickness of the dissimilar weld overlay pipe specimen.« less

Theory of thermionic emission from a two-dimensional conductor and its application to a graphene-semiconductor Schottky junction

NASA Astrophysics Data System (ADS)

Trushin, Maxim

2018-04-01

The standard theory of thermionic emission developed for three-dimensional semiconductors does not apply to two-dimensional materials even for making qualitative predictions because of the vanishing out-of-plane quasiparticle velocity. This study reveals the fundamental origin of the out-of-plane charge carrier motion in a two-dimensional conductor due to the finite quasiparticle lifetime and huge uncertainty of the out-of-plane momentum. The theory is applied to a Schottky junction between graphene and a bulk semiconductor to derive a thermionic constant, which, in contrast to the conventional Richardson constant, is determined by the Schottky barrier height and Fermi level in graphene.

Work function determination of promising electrode materials for thermionic converters

NASA Technical Reports Server (NTRS)

Jacobson, D.

1977-01-01

Work performed on this contract was primarily for the evaluation of selected electrode materials for thermionic energy converters. The original objective was to characterize selected nickel based superalloys up to temperatures of 1400 K. It was found that an early selection, Inconel 800 produced a high vapor pressure which interfered with the vacuum emission measurements. The program then shifted to two other areas. The first area was to obtain emission from the superalloys in a cesiated atmosphere. The cesium plasma helps to suppress the vaporization interference. The second area involved characterization of the Lanthanum-Boron series as thermionic emitters. These final two areas resulted in three journal publications which are attached to this report.

Visualization of Flow in Pressurizer Spray Line Piping and Estimation of Thermal Stress Fluctuation Caused by Swaying of Water Surface

NASA Astrophysics Data System (ADS)

Oumaya, Toru; Nakamura, Akira; Onojima, Daisuke; Takenaka, Nobuyuki

The pressurizer spray line of PWR plants cools reactor coolant by injecting water into pressurizer. Since the continuous spray flow rate during commercial operation of the plant is considered insufficient to fill the pipe completely, there is a concern that a water surface exists in the pipe and may periodically sway. In order to identify the flow regimes in spray line piping and assess their impact on pipe structure, a flow visualization experiment was conducted. In the experiment, air was used substituted for steam to simulate the gas phase of the pressurizer, and the flow instability causing swaying without condensation was investigated. With a full-scale mock-up made of acrylic, flow under room temperature and atmospheric pressure conditions was visualized, and possible flow regimes were identified based on the results of the experiment. Three representative patterns of swaying of water surface were assumed, and the range of thermal stress fluctuation, when the surface swayed instantaneously, was calculated. With the three patterns of swaying assumed based on the visualization experiment, it was confirmed that the thermal stress amplitude would not exceed the fatigue endurance limit prescribed in the Japanese Design and Construction Code.

Modeling, Fabrication, and Electrical Testing of Metal-Insulator-Metal Diode

DTIC Science & Technology

2011-12-01

1 2. MIM Model 1 2.1 Potential Energy and Image Potential . . . . . . . . . . . . . . . . . . . . . . 1 2.2 Thermionic Emission -limited Current ...4 4 Thermionic emission -limited current through the symmetric MIM diode in figure 1...7 7 Absolute value of tunnel-limited, thermal emission -limited, and total currents vs. applied bias for the

Studies of thermionic materials for space power applications

NASA Technical Reports Server (NTRS)

1971-01-01

Service life tests of LC-8 and LC-9 carbide-fueled thermionic converters are discussed. Post operational tests of the converters to show emitter diametric change, microstructures of cladding and fuel, and analysis of fuel composition are described. The fabrication and performance of high temperature thermocouples used in the test procedures are included.

Chemical regeneration of emitter surface increases thermionic diode life

NASA Technical Reports Server (NTRS)

Breiteieser, R.

1966-01-01

Chemical regeneration of sublimated emitter electrode increases the operating efficiency and life of thermionic diodes. A gas which forms chemical compounds with the sublimated emitter material is introduced into the space between the emitter and the collector. The compounds migrate to the emitter where they decompose and redeposit the emitter material.

Energy conversion research and development with diminiodes

NASA Technical Reports Server (NTRS)

Morris, J. F.

1974-01-01

Diminiodes are variable-gap cesium diodes with plane miniature guarded electrodes. These converters allow thermionic evaluations of tiny pieces of rare solids. In addition to smallness, diminiode advantages comprise simplicity, precision, fabrication ease, parts interchangeability, cleanliness, full instrumentation, direct calibration, ruggedness, and economy. Diminiodes with computerized thermionic performance mapping make electrode screening programs practical.

The diminiode: A research and development tool for nuclear thermionics

NASA Technical Reports Server (NTRS)

Morris, J. F.

1972-01-01

Diminiodes are fixed-or variable-gap cesium diodes with plane miniature emitters and guarded collectors. In addition to smallness, their relative advantages are simplicity, precision, ease of fabrication, interchangeability of parts, cleanliness, full instrumentation, ruggedness, and economy. With diminiodes and computers used in thermionic performance mapping, a thorough electrode screening program becomes practical.

ENVIRONMENTAL TECHNOLOGY VERIFICATION REPORT, EXPLOSIVES DETECTION TECHNOLOGY, SRI INSTRUMENTS, MODEL 8610C, GAS CHROMATOGRAPH/THERMIONIC IONIZATION DETECTION

EPA Science Inventory

The SRI Model 86 1 OC gas chromatograph (GC) is a transportable instrument that can provide on-site analysis of soils for explosives. Coupling this transportable gas chromatograph with a thermionic ionization detector (TID) allows for the determination of explosives in soil matri...

Thermionic Energy Conversion Based on Graphene van der Waals Heterostructures

PubMed Central

Liang, Shi-Jun; Liu, Bo; Hu, Wei; Zhou, Kun; Ang, L. K.

2017-01-01

Seeking for thermoelectric (TE) materials with high figure of merit (or ZT), which can directly converts low-grade wasted heat (400 to 500 K) into electricity, has been a big challenge. Inspired by the concept of multilayer thermionic devices, we propose and design a solid-state thermionic devices (as a power generator or a refrigerator) in using van der Waals (vdW) heterostructure sandwiched between two graphene electrodes, to achieve high energy conversion efficiency in the temperature range of 400 to 500 K. The vdW heterostructure is composed of suitable multiple layers of transition metal dichalcogenides (TMDs), such as MoS2, MoSe2, WS2 and WSe2. From our calculations, WSe2 and MoSe2 are identified as two ideal TMDs (using the reported experimental material’s properties), which can harvest waste heat at 400 K with efficiencies about 7% to 8%. To our best knowledge, this design is the first in combining the advantages of graphene electrodes and TMDs to function as a thermionic-based device. PMID:28387363

Thermionic combustor application to combined gas and steam turbine power plants

NASA Astrophysics Data System (ADS)

Miskolczy, G.; Wang, C. C.; Lieb, D. P.; Margulies, A. E.; Fusegni, L. J.; Lovell, B. J.

A design for the insertion of thermionic converters into the wall of a conventional combustor to produce electricity in a topping cycle is described, and a study for applications in gas and steam generators of 70 and 30 MW is evaluated for engineering and economic feasibility. Waste heat from the thermionic elements is used to preheat the combustor air; the heat absorbed by the elements plus further quenching of the exhaust gases with ammonia is projected to reduce NO(x) emissions to acceptable levels. Schematics, flow diagrams, and components of a computer model for cost projections are provided. It was found that temperatures around the emitters must be maintained above 1,600 K, with maximum efficiency and allowable temperature at 1,800 K, while collectors generate maximally at 950 K, with a corresponding work function of 1.5 eV. Cost sensitive studies indicate an installed price of $475/kW for the topping cycle, with improvements in thermionic converter characteristics bringing the cost to $375/kW at a busbar figure of 500 mills/kWh.

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

Faidy, C.

Practical applications of the leak-before break concept are presently limited in French Pressurized Water Reactors (PWR) compared to Fast Breeder Reactors. Neithertheless, different fracture mechanic demonstrations have been done on different primary, auxiliary and secondary PWR piping systems based on similar requirements that the American NUREG 1061 specifications. The consequences of the success in different demonstrations are still in discussion to be included in the global safety assessment of the plants, such as the consequences on in-service inspections, leak detection systems, support optimization,.... A large research and development program, realized in different co-operative agreements, completes the general approach.

Heat transfer system

DOEpatents

Not Available

1980-03-07

A heat transfer system for a nuclear reactor is described. Heat transfer is accomplished within a sealed vapor chamber which is substantially evacuated prior to use. A heat transfer medium, which is liquid at the design operating temperatures, transfers heat from tubes interposed in the reactor primary loop to spaced tubes connected to a steam line for power generation purposes. Heat transfer is accomplished by a two-phase liquid-vapor-liquid process as used in heat pipes. Condensible gases are removed from the vapor chamber through a vertical extension in open communication with the chamber interior.

PROCESS WATER BUILDING, TRA605, INTERIOR. FIRST FLOOR. CAMERA IS IN ...

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

PROCESS WATER BUILDING, TRA-605, INTERIOR. FIRST FLOOR. CAMERA IS IN SOUTHEAST CORNER AND FACES NORTHWEST. CONTROL ROOM AT RIGHT. CRANE MONORAIL IS OVER FLOOR HATCHES AND FLOOR OPENINGS. SIX VALVE HANDWHEELS ALONG FAR WALL IN LEFT CENTER VIEW. SEAL TANK IS ON OTHER SIDE OF WALL; PROCESS WATER PIPES ARE BELOW VALVE WHEELS. NOTE CURBS AROUND FLOOR OPENINGS. INL NEGATIVE NO. HD46-26-3. Mike Crane, Photographer, 2/2005 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Heat transfer system

DOEpatents

McGuire, Joseph C.

1982-01-01

A heat transfer system for a nuclear reactor. Heat transfer is accomplished within a sealed vapor chamber which is substantially evacuated prior to use. A heat transfer medium, which is liquid at the design operating temperatures, transfers heat from tubes interposed in the reactor primary loop to spaced tubes connected to a steam line for power generation purposes. Heat transfer is accomplished by a two-phase liquid-vapor-liquid process as used in heat pipes. Condensible gases are removed from the vapor chamber through a vertical extension in open communication with the chamber interior.

Comparison of Microbial Communities in a Simulated Chloraminated Drinking Water Distribution System Subjected to Episodes of Nitrification (poster)

EPA Science Inventory

Bacterial populations were examined in a simulated chloraminated drinking water distribution system (i.e. PVC pipe loop). After six months of continuous operation, coupons were incubated in CDC reactors receiving water from the simulated system to study biofilm development. The s...

An Overview of Long Duration Sodium Heat Pipe Tests

NASA Astrophysics Data System (ADS)

Rosenfeld, John H.; Ernst, Donald M.; Lindemuth, James E.; Sanzi, James L.; Geng, Steven M.; Zuo, Jon

2004-02-01

High temperature heat pipes are being evaluated for use in energy conversion applications such as fuel cells, gas turbine re-combustors, and Stirling cycle heat sources; with the resurgence of space nuclear power, additional applications include reactor heat removal elements and radiator elements. Long operating life and reliable performance are critical requirements for these applications. Accordingly long-term materials compatibility is being evaluated through the use of high temperature life test heat pipes. Thermacore, Inc. has carried out several sodium heat pipe life tests to establish long term operating reliability. Four sodium heat pipes have recently demonstrated favorable materials compatibility and heat transport characteristics at high operating temperatures in air over long time periods. A 316L stainless steel heat pipe with a sintered porous nickel wick structure and an integral brazed cartridge heater has successfully operated at 650C to 700C for over 115,000 hours without signs of failure. A second 316L stainless steel heat pipe with a specially-designed Inconel 601 rupture disk and a sintered nickel powder wick has demonstrated over 83,000 hours at 600C to 650C with similar success. A representative one-tenth segment Stirling Space Power Converter heat pipe with an Inconel 718 envelope and a stainless steel screen wick has operated for over 41,000 hours at nearly 700C. A hybrid (i.e. gas-fired and solar) heat pipe with a Haynes 230 envelope and a sintered porous nickel wick structure was operated for about 20,000 hours at nearly 700C without signs of degradation. These life test results collectively have demonstrated the potential for high temperature heat pipes to serve as reliable energy conversion system components for power applications that require long operating lifetime with high reliability. Detailed design specifications, operating history, and test results are described for each of these sodium heat pipes. Lessons learned and future life test plans are also discussed.

An Overview of Long Duration Sodium Heat Pipe Tests

NASA Technical Reports Server (NTRS)

Rosenfeld, John H.; Ernst, Donald M.; Lindemuth, James E.; Sanzi, James L.; Geng, Steven M.; Zuo, Jon

2004-01-01

High temperature heat pipes are being evaluated for use in energy conversion applications such as fuel cells, gas turbine re-combustors, and Stirling cycle heat sources; with the resurgence of space nuclear power, additional applications include reactor heat removal elements and radiator elements. Long operating life and reliable performance are critical requirements for these applications. Accordingly long-term materials compatibility is being evaluated through the use of high temperature life test heat pipes. Thermacore International, Inc., has carried out several sodium heat pipe life tests to establish long term operating reliability. Four sodium heat pipes have recently demonstrated favorable materials compatibility and heat transport characteristics at high operating temperatures in air over long time periods. A 3l6L stainless steel heat pipe with a sintered porous nickel wick structure and an integral brazed cartridge heater has successfully operated at 650 to 700 C for over 115,000 hours without signs of failure. A second 3l6L stainless steel heat pipe with a specially-designed Inconel 60 I rupture disk and a sintered nickel powder wick has demonstrated over 83,000 hours at 600 to 650 C with similar success. A representative one-tenth segment Stirling Space Power Converter heat pipe with an Inconel 718 envelope and a stainless steel screen wick has operated for over 41 ,000 hours at nearly 700 0c. A hybrid (i.e. gas-fired and solar) heat pipe with a Haynes 230 envelope and a sintered porous nickel wick structure was operated for about 20,000 hours at nearly 700 C without signs of degradation. These life test results collectively have demonstrated the potential for high temperature heat pipes to serve as reliable energy conversion system components for power applications that require long operating lifetime with high reliability, Detailed design specifications, operating hi story, and test results are described for each of these sodium heat pipes. Lessons learned and future life test plans are also discussed.

The cesiator - A device for cesium vapor control and impurity purge

NASA Astrophysics Data System (ADS)

Rasor, N. S.; Desplat, J.-L.

A new type of liquid cesium reservoir that maintains a temperature-independent cesium pressure, continuously recirculates cesium vapor through the TFE (thermionic fuel element), and purges it of impurities is discussed. This device, the cesiator, is based on well-established gas-buffered heat pipe principles. The cesiator offers new TFE design options for fission product/impurity handling that eliminate the need for an intercell insulator seal and associated failure modes. Cesiator performance requirements are estimated based on data for expected release of fission products and their effect on TFE performance. The effect of design parameters on cesiator performance is described. Experimentation with an ethanol-metal mock-up revealed an unexpected but desirable mode of operation that autoregulates the pressure drop and flow of vapor in the external circuit and that has been incorporated in the reference design for phase II development. Experimental techniques for measuring the local temperature, pressure, and composition in a condensing vapor were successfully developed. A reference design for a TFE cesiator was defined for prototype design, development, and test.

Distributed electrical leads for thermionic converter

DOEpatents

Fitzpatrick, Gary O.; Britt, Edward J.

1979-01-01

In a thermionic converter, means are provided for coupling an electrical lead to at least one of the electrodes thereof. The means include a bus bar and a plurality of distributed leads coupled to the bus bar each of which penetrates through one electrode and are then coupled to the other electrode of the converter in spaced apart relation.

Thermionic energy converters

DOEpatents

Monroe, Jr., James E.

1977-08-09

A thermionic device for converting nuclear energy into electrical energy comprising a tubular anode spaced from and surrounding a cylindrical cathode, the cathode having an outer emitting surface of ruthenium, and nuclear fuel on the inner cylindrical surface. The nuclear fuel is a ceramic composition of fissionable material in a metal matrix. An axial void is provided to collect and contain fission product gases.

Thermionic Emission of Single-Wall Carbon Nanotubes Measured

NASA Technical Reports Server (NTRS)

Landis, Geoffrey A.; Krainsky, Isay L.; Bailey, Sheila G.; Elich, Jeffrey M.; Landi, Brian J.; Gennett, Thomas; Raffaelle, Ryne P.

2004-01-01

Researchers at the NASA Glenn Research Center, in collaboration with the Rochester Institute of Technology, have investigated the thermionic properties of high-purity, single-wall carbon nanotubes (SWNTs) for use as electron-emitting electrodes. Carbon nanotubes are a recently discovered material made from carbon atoms bonded into nanometer-scale hollow tubes. Such nanotubes have remarkable properties. An extremely high aspect ratio, as well as unique mechanical and electronic properties, make single-wall nanotubes ideal for use in a vast array of applications. Carbon nanotubes typically have diameters on the order of 1 to 2 nm. As a result, the ends have a small radius of curvature. It is these characteristics, therefore, that indicate they might be excellent potential candidates for both thermionic and field emission.

Low work function silicon collector for thermionic converters

NASA Technical Reports Server (NTRS)

Chang, K. H.; Shimada, K.

1976-01-01

To improve the efficiency of present thermionic converters, single crystal silicon was investigated as a low work function collector material. The experiments were conducted in a test vehicle which resembled an actual thermionic converter. Work function as low as 1.0eV was obtained with an n-type silicon. The stabilities of the activated surfaces at elevated temperatures were tested by raising the collector temperature up to 829 K. By increasing the Cs arrival rate, it was possible to restore the originally activated low work function of the surface at elevated surface temperatures. These results, plotted in the form of Rasor-Warner curve, show a behavior similar to that of metal electrode except that the minimum work function was much lower with silicon than with metals.

Conceptual design of ACB-CP for ITER cryogenic system

NASA Astrophysics Data System (ADS)

Jiang, Yongcheng; Xiong, Lianyou; Peng, Nan; Tang, Jiancheng; Liu, Liqiang; Zhang, Liang

2012-06-01

ACB-CP (Auxiliary Cold Box for Cryopumps) is used to supply the cryopumps system with necessary cryogen in ITER (International Thermonuclear Experimental Reactor) cryogenic distribution system. The conceptual design of ACB-CP contains thermo-hydraulic analysis, 3D structure design and strength checking. Through the thermohydraulic analysis, the main specifications of process valves, pressure safety valves, pipes, heat exchangers can be decided. During the 3D structure design process, vacuum requirement, adiabatic requirement, assembly constraints and maintenance requirement have been considered to arrange the pipes, valves and other components. The strength checking has been performed to crosscheck if the 3D design meets the strength requirements for the ACB-CP.

Strengthening the fission reactor nuclear science and engineering program at UCLA. Final technical report

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

Okrent, D.

1997-06-23

This is the final report on DOE Award No. DE-FG03-92ER75838 A000, a three year matching grant program with Pacific Gas and Electric Company (PG and E) to support strengthening of the fission reactor nuclear science and engineering program at UCLA. The program began on September 30, 1992. The program has enabled UCLA to use its strong existing background to train students in technological problems which simultaneously are of interest to the industry and of specific interest to PG and E. The program included undergraduate scholarships, graduate traineeships and distinguished lecturers. Four topics were selected for research the first year, withmore » the benefit of active collaboration with personnel from PG and E. These topics remained the same during the second year of this program. During the third year, two topics ended with the departure o the students involved (reflux cooling in a PWR during a shutdown and erosion/corrosion of carbon steel piping). Two new topics (long-term risk and fuel relocation within the reactor vessel) were added; hence, the topics during the third year award were the following: reflux condensation and the effect of non-condensable gases; erosion/corrosion of carbon steel piping; use of artificial intelligence in severe accident diagnosis for PWRs (diagnosis of plant status during a PWR station blackout scenario); the influence on risk of organization and management quality; considerations of long term risk from the disposal of hazardous wastes; and a probabilistic treatment of fuel motion and fuel relocation within the reactor vessel during a severe core damage accident.« less

Experiments and models of general corrosion and flow-assisted corrosion of materials in nuclear reactor environments

NASA Astrophysics Data System (ADS)

Cook, William Gordon

Corrosion and material degradation issues are of concern to all industries. However, the nuclear power industry must conform to more stringent construction, fabrication and operational guidelines due to the perceived additional risk of operating with radioactive components. Thus corrosion and material integrity are of considerable concern for the operators of nuclear power plants and the bodies that govern their operations. In order to keep corrosion low and maintain adequate material integrity, knowledge of the processes that govern the material's breakdown and failure in a given environment are essential. The work presented here details the current understanding of the general corrosion of stainless steel and carbon steel in nuclear reactor primary heat transport systems (PHTS) and examines the mechanisms and possible mitigation techniques for flow-assisted corrosion (FAC) in CANDU outlet feeder pipes. Mechanistic models have been developed based on first principles and a 'solution-pores' mechanism of metal corrosion. The models predict corrosion rates and material transport in the PHTS of a pressurized water reactor (PWR) and the influence of electrochemistry on the corrosion and flow-assisted corrosion of carbon steel in the CANDU outlet feeders. In-situ probes, based on an electrical resistance technique, were developed to measure the real-time corrosion rate of reactor materials in high-temperature water. The probes were used to evaluate the effects of coolant pH and flow on FAC of carbon steel as well as demonstrate of the use of titanium dioxide as a coolant additive to mitigated FAC in CANDU outlet feeder pipes.

[Effect of the change in sulphate and dissolved oxygen mass concentration on metal release in old cast iron distribution pipes].

PubMed

Wu, Yong-li; Shi, Bao-you; Sun, Hui-fang; Zhang, Zhi-huan; Gu, Jun-nong; Wang, Dong-sheng

2013-09-01

To understand the processes of corrosion by-product release and the consequent "red water" problems caused by the variation of water chemical composition in drinking water distribution system, the effect of sulphate and dissolved oxygen (DO) concentration on total iron release in corroded old iron pipe sections historically transporting groundwater was investigated in laboratory using small-scale pipe section reactors. The release behaviors of some low-level metals, such as Mn, As, Cr, Cu, Zn and Ni, in the process of iron release were also monitored. The results showed that the total iron and Mn release increased significantly with the increase of sulphate concentration, and apparent red water occurred when sulphate concentration was above 400 mg x L(-1). With the increase of sulfate concentration, the effluent concentrations of As, Cr, Cu, Zn and Ni also increased obviously, however, the effluent concentrations of these metals were lower than the influent concentrations under most circumstances, which indicated that adsorption of these metals by pipe corrosion scales occurred. Increasing DO within a certain range could significantly inhibit the iron release.

Study of a heat rejection system for the Nuclear Electric Propulsion (NEP) spacecraft

NASA Technical Reports Server (NTRS)

Ernest, D. M.

1982-01-01

Two different heat pipe radiator elements, one intended for use with the power conversion subsystem of the NASA funded nuclear electric propulsion (NEP) spacecraft, and one intended for use with the DOE funded space power advanced reactor (SPAR) system were tested and evaluated. The NEP stainless steel/sodium heat pipe was 4.42 meters long and had a 1 cm diameter. Thermal performance testing at 920 K showed a non-limited power level of 3560 watts, well in excess of the design power of 2600 watts. This test verified the applicability of screen arteries for use in long radiator heat pipes. The SPAR titanium/potassium heat pipe was 5.5 meters long and had a semicircular crossection with a 4 cm diameter. Thermal performance testing at 775 K showed a maximum power level of 1.86 kW, somewhat short of the desired 2.6 kW beginning of life design requirement. The reduced performance was shown to be the result of the inability of the evaporator wall wick (shot blasted evaporator wall) to handle the required liquid flow.

Electron Thermionic Emission from Graphene and a Thermionic Energy Converter

NASA Astrophysics Data System (ADS)

Liang, Shi-Jun; Ang, L. K.

2015-01-01

In this paper, we propose a model to investigate the electron thermionic emission from single-layer graphene (ignoring the effects of the substrate) and to explore its application as the emitter of a thermionic energy converter (TIC). An analytical formula is derived, which is a function of the temperature, work function, and Fermi energy level. The formula is significantly different from the traditional Richardson-Dushman (RD) law for which it is independent of mass to account for the supply function of the electrons in the graphene behaving like massless fermion quasiparticles. By comparing with a recent experiment [K. Jiang et al., Nano Res. 7, 553 (2014)] measuring electron thermionic emission from suspended single-layer graphene, our model predicts that the intrinsic work function of single-layer graphene is about 4.514 eV with a Fermi energy level of 0.083 eV. For a given work function, a scaling of T3 is predicted, which is different from the traditional RD scaling of T2. If the work function of the graphene is lowered to 2.5-3 eV and the Fermi energy level is increased to 0.8-0.9 eV, it is possible to design a graphene-cathode-based TIC operating at around 900 K or lower, as compared with the metal-based cathode TIC (operating at about 1500 K). With a graphene-based cathode (work function=4.514 eV ) at 900 K and a metallic-based anode (work function=2.5 eV ) like LaB6 at 425 K, the efficiency of our proposed TIC is about 45%.

Electrical heating tests of uranium dioxide external fuel configuration at emitter temperature of 1900 K

NASA Technical Reports Server (NTRS)

Diianni, D. C.; Mayer, J. T.

1974-01-01

Testing of two fuel clad specimens for thermionic reactor application is described. The annular UO2 fuel was clad on both sides with tungsten; heat rejection was radially inward. The tests were intended to study inner clad stability, fuel redistribution, and fuel melting problems. The specimens were tested in a vacuum chamber using electron bombardment heating. Fuel structural changes were studied using periodic gammagraphs and posttest metallography. The first specimen test was terminated at 50 hours because of a braze failure. The second specimen was tested for 240 hours when an outer clad leak developed due to a tungsten-water reaction. The fuel developed numerous cracks on cooldown but the inner clad remained dimensionally stable. The fuel cover gas did not impede the rate of fuel redistribution. Posttest examination showed the fuel had not melted during operation.

Nuclear fuels for very high temperature applications

NASA Astrophysics Data System (ADS)

Lundberg, L. B.; Hobbins, R. R.

The success of the development of nuclear thermal propulsion devices and thermionic space nuclear power generation systems depends on the successful utilization of nuclear fuel materials at temperatures in the range 2000 to 3500 K. Problems associated with the utilization of uranium bearing fuel materials at these very high temperatures while maintaining them in the solid state for the required operating times are addressed. The critical issues addressed include evaporation, melting, reactor neutron spectrum, high temperature chemical stability, fabrication, fission induced swelling, fission product release, high temperature creep, thermal shock resistance, and fuel density, both mass and fissile atom. Candidate fuel materials for this temperature range are based on UO2 or uranium carbides. Evaporation suppression, such as a sealed cladding, is required for either fuel base. Nuclear performance data needed for design are sparse for all candidate fuel forms in this temperature range, especially at the higher temperatures.

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

Federal Register 2010, 2011, 2012, 2013, 2014

2011-09-19

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

A Review of Fatigue Crack Growth of Pressure Vessel and Piping Steels in High-Temperature, Pressurized Reactor-Grade Water.

DTIC Science & Technology

1980-09-19

ratio, irradiation, materials Germany Kraftwerk Union Materials, temperature, Erlangen, Germany load ratio, frequency, electro- chemical potential Italy...H. Munster, "Frequenzeinfluss auf das Risswachstumsverhalten des Stahles 22NiMoCr37", KWU/R 413/8/80, Kraftwerk Union, Erlangen (1980). 76. D. F

Water inventory management in condenser pool of boiling water reactor

DOEpatents

Gluntz, Douglas M.

1996-01-01

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

Water inventory management in condenser pool of boiling water reactor

DOEpatents

Gluntz, D.M.

1996-03-12

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

Evaluation of corrosion inhibitor simulating conditions of operation

NASA Astrophysics Data System (ADS)

Gómez, O.; Aponte, H.; Vera, E.; Pineda, Y.

2017-12-01

Operating conditions at the head of oil wells are critical, in addition to injecting water to increase the pressure while maintaining production cause deterioration in the metallic structures that transport fluids. One way to maintain integrity is the injection of inhibitors which plays an important role in protecting the pipes. In this study a molecule N-PHENYL NITRONE was obtained, which was evaluated by electrochemical tests (LPR) and Tafel Polarization Curves in a reactor controlling environments containing different dosages, pressure values, temperatures and flow velocity using working electrodes tubing API N 80, the reactor was connected to a potentiostat to determine corrosion rates, allowing the analysis of the influence of each variable on the protective behaviour of the inhibitor, and its efficiency against the decrease of the deterioration of the pipes. Corrosion products are analysed by X-Ray Diffraction (XRD). Photographic records of the surface verify the formation of iron carbonate (FeCO3). In addition, a mathematical analysis of the independent variables is performed to evaluate the effect that it has on the efficiency of corrosion inhibition.

Continuous AE crack monitoring of a dissimilar metal weldment at Limerick Unit 1

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

Hutton, P.H.; Friesel, M.A.; Dawson, J.F.

1993-12-01

Acoustic emission (AE) technology for continuous surveillance of a reactor component(s) to detect crack initiation and/or crack growth has been developed at Pacific Northwest Laboratory (PNL). The technology was validated off-reactor in several major tests, but it had not been validated by monitoring crack growth on an operating reactor system. A flaw indication was identified during normal inservice inspection of piping at Philadelphia Electric Company (PECO) Limerick Unit 1 reactor during the 1989 refueling outage. Evaluation of the flaw indication showed that it could remain in place during the subsequent fuel cycle without compromising safety. The existence of this flawmore » indication offered a long sought opportunity to validate AE surveillance to detect and evaluate crack growth during reactor operation. AE instrumentation was installed by PNL and PECO to monitor the flaw indication during two complete fuel cycles. This report discusses the results obtained from the AE monitoring over the period May 1989 to March 1992 (two fuel cycles).« less

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

Tanaka, T.; Kameyama, M.; Urabe, Y.

At present, cast duplex stainless steel has been used for the primary coolant piping of PWRs in Japan and joints of dissimilar material have been applied for welding to reactor vessels and steam generators. For the primary coolant piping of the next APWR plants, application of low alloy steel that results in designing main loops with the same material is being studied. It means that there is no need to weld low alloy steel with stainless steel and that makes it possible to reduce the welding length. Attenuation of Ultra Sonic Wave Intensity is lower for low alloy steel thanmore » for stainless steel and they have advantageous inspection characteristics. In addition to that, the thermal expansion rate is smaller for low alloy steel than for stainless steel. In consideration of the above features of low alloy steel, the overall reliability of primary coolant piping is expected to be improved. Therefore, for the evaluation of crack stability of low alloy steel piping to be applied for primary loops, elastic-plastic future mechanics analysis was performed by means of a three-dimensioned FEM. The evaluation results for the low alloy steel pipings show that cracks will not grow into unstable fractures under maximum design load conditions, even when such a circumferential crack is assumed to be 6 times the size of the wall thickness.« less

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

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

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

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

Collecting and recirculating condensate in a nuclear reactor containment

DOEpatents

Schultz, Terry L.

1993-01-01

An arrangement passively cools a nuclear reactor in the event of an emergency, condensing and recycling vaporized cooling water. The reactor is surrounded by a containment structure and has a storage tank for cooling liquid, such as water, vented to the containment structure by a port. The storage tank preferably is located inside the containment structure and is thermally coupleable to the reactor, e.g. by a heat exchanger, such that water in the storage tank is boiled off to carry away heat energy. The water is released as a vapor (steam) and condenses on the cooler interior surfaces of the containment structure. The condensed water flows downwardly due to gravity and is collected and routed back to the storage tank. One or more gutters are disposed along the interior wall of the containment structure for collecting the condensate from the wall. Piping is provided for communicating the condensate from the gutters to the storage tank.

Collecting and recirculating condensate in a nuclear reactor containment

DOEpatents

Schultz, T.L.

1993-10-19

An arrangement passively cools a nuclear reactor in the event of an emergency, condensing and recycling vaporized cooling water. The reactor is surrounded by a containment structure and has a storage tank for cooling liquid, such as water, vented to the containment structure by a port. The storage tank preferably is located inside the containment structure and is thermally coupleable to the reactor, e.g. by a heat exchanger, such that water in the storage tank is boiled off to carry away heat energy. The water is released as a vapor (steam) and condenses on the cooler interior surfaces of the containment structure. The condensed water flows downwardly due to gravity and is collected and routed back to the storage tank. One or more gutters are disposed along the interior wall of the containment structure for collecting the condensate from the wall. Piping is provided for communicating the condensate from the gutters to the storage tank. 3 figures.

Thermionic refrigeration at CNT-CNT junctions

NASA Astrophysics Data System (ADS)

Li, C.; Pipe, K. P.

2016-10-01

Monte Carlo (MC) simulation is used to study carrier energy relaxation following thermionic emission at the junction of two van der Waals bonded single-walled carbon nanotubes (SWCNTs). An energy-dependent transmission probability gives rise to energy filtering at the junction, which is predicted to increase the average electron transport energy by as much as 0.115 eV, leading to an effective Seebeck coefficient of 386 μV/K. MC results predict a long energy relaxation length (˜8 μm) for hot electrons crossing the junction into the barrier SWCNT. For SWCNTs of optimal length, an analytical transport model is used to show that thermionic cooling can outweigh parasitic heat conduction due to high SWCNT thermal conductivity, leading to a significant cooling capacity (2.4 × 106 W/cm2).

Power control of SAFE reactor using fuzzy logic

NASA Astrophysics Data System (ADS)

Irvine, Claude

2002-01-01

Controlling the 100 kW SAFE (Safe Affordable Fission Engine) reactor consists of design and implementation of a fuzzy logic process control system to regulate dynamic variables related to nuclear system power. The first phase of development concentrates primarily on system power startup and regulation, maintaining core temperature equilibrium, and power profile matching. This paper discusses the experimental work performed in those areas. Nuclear core power from the fuel elements is simulated using resistive heating elements while heat rejection is processed by a series of heat pipes. Both axial and radial nuclear power distributions are determined from neuronic modeling codes. The axial temperature profile of the simulated core is matched to the nuclear power profile by varying the resistance of the heating elements. The SAFE model establishes radial temperature profile equivalence by establishing 32 control zones as the nodal coordinates. Control features also allow for slow warm up, since complete shutoff can occur in the heat pipes if heat-source temperatures drop/rise below a certain minimum value, depending on the specific fluid and gas combination in the heat pipe. The entire system is expected to be self-adaptive, i.e., capable of responding to long-range changes in the space environment. Particular attention in the development of the fuzzy logic algorithm shall ensure that the system process remains at set point, virtually eliminating overshoot on start-up and during in-process disturbances. The controller design will withstand harsh environments and applications where it might come in contact with water, corrosive chemicals, radiation fields, etc. .

Assessment of SFR Wire Wrap Simulation Uncertainties

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

Delchini, Marc-Olivier G.; Popov, Emilian L.; Pointer, William David

Predictive modeling and simulation of nuclear reactor performance and fuel are challenging due to the large number of coupled physical phenomena that must be addressed. Models that will be used for design or operational decisions must be analyzed for uncertainty to ascertain impacts to safety or performance. Rigorous, structured uncertainty analyses are performed by characterizing the model’s input uncertainties and then propagating the uncertainties through the model to estimate output uncertainty. This project is part of the ongoing effort to assess modeling uncertainty in Nek5000 simulations of flow configurations relevant to the advanced reactor applications of the Nuclear Energy Advancedmore » Modeling and Simulation (NEAMS) program. Three geometries are under investigation in these preliminary assessments: a 3-D pipe, a 3-D 7-pin bundle, and a single pin from the Thermal-Hydraulic Out-of-Reactor Safety (THORS) facility. Initial efforts have focused on gaining an understanding of Nek5000 modeling options and integrating Nek5000 with Dakota. These tasks are being accomplished by demonstrating the use of Dakota to assess parametric uncertainties in a simple pipe flow problem. This problem is used to optimize performance of the uncertainty quantification strategy and to estimate computational requirements for assessments of complex geometries. A sensitivity analysis to three turbulent models was conducted for a turbulent flow in a single wire wrapped pin (THOR) geometry. Section 2 briefly describes the software tools used in this study and provides appropriate references. Section 3 presents the coupling interface between Dakota and a computational fluid dynamic (CFD) code (Nek5000 or STARCCM+), with details on the workflow, the scripts used for setting up the run, and the scripts used for post-processing the output files. In Section 4, the meshing methods used to generate the THORS and 7-pin bundle meshes are explained. Sections 5, 6 and 7 present numerical results for the 3-D pipe, the single pin THORS mesh, and the 7-pin bundle mesh, respectively.« less

Two modelling approaches to water-quality simulation in a flooded iron-ore mine (Saizerais, Lorraine, France): a semi-distributed chemical reactor model and a physically based distributed reactive transport pipe network model.

PubMed

Hamm, V; Collon-Drouaillet, P; Fabriol, R

2008-02-19

The flooding of abandoned mines in the Lorraine Iron Basin (LIB) over the past 25 years has degraded the quality of the groundwater tapped for drinking water. High concentrations of dissolved sulphate have made the water unsuitable for human consumption. This problematic issue has led to the development of numerical tools to support water-resource management in mining contexts. Here we examine two modelling approaches using different numerical tools that we tested on the Saizerais flooded iron-ore mine (Lorraine, France). A first approach considers the Saizerais Mine as a network of two chemical reactors (NCR). The second approach is based on a physically distributed pipe network model (PNM) built with EPANET 2 software. This approach considers the mine as a network of pipes defined by their geometric and chemical parameters. Each reactor in the NCR model includes a detailed chemical model built to simulate quality evolution in the flooded mine water. However, in order to obtain a robust PNM, we simplified the detailed chemical model into a specific sulphate dissolution-precipitation model that is included as sulphate source/sink in both a NCR model and a pipe network model. Both the NCR model and the PNM, based on different numerical techniques, give good post-calibration agreement between the simulated and measured sulphate concentrations in the drinking-water well and overflow drift. The NCR model incorporating the detailed chemical model is useful when a detailed chemical behaviour at the overflow is needed. The PNM incorporating the simplified sulphate dissolution-precipitation model provides better information of the physics controlling the effect of flow and low flow zones, and the time of solid sulphate removal whereas the NCR model will underestimate clean-up time due to the complete mixing assumption. In conclusion, the detailed NCR model will give a first assessment of chemical processes at overflow, and in a second time, the PNM model will provide more detailed information on flow and chemical behaviour (dissolved sulphate concentrations, remaining mass of solid sulphate) in the network. Nevertheless, both modelling methods require hydrological and chemical parameters (recharge flow rate, outflows, volume of mine voids, mass of solids, kinetic constants of the dissolution-precipitation reactions), which are commonly not available for a mine and therefore call for calibration data.

Investigation and Feasibility Assessment of TOPAZ-2 Derivations for Space Power Applications

NASA Technical Reports Server (NTRS)

Parlos, Alexander G.; Peddicord, Kenneth L.

1998-01-01

The ability to provide continuous power at significant levels is of utmost importance for many space missions, from simple satellite operations to manned Mars missions. One of the main problems faced in delivering solar or chemical space power in the tens of kW range, is the increasingly massive nature of the power source and the costs associated with its launch, operation and maintenance. A national program had been initiated to study the feasibility of using certain advanced technologies in developing an efficient lightweight space power source. The starting point for these studies has been the Russian TOPAZ-2 space reactor system, with the ultimate goal to aid in the development of a TOPAZ-2 derivative which will be ready for flight by the year 2000. The main objective of this project has been to perform feasibility assessment and trade studies which would allow the development of an advanced space nuclear power system based on the in-core thermionic fuel element technology currently used in the Russian TOPAZ-2 reactor. Two of the important considerations in developing the concept are: (1) compliance of the current TOPAZ-2 and of any advanced designs with U.S. nuclear safety expectations, and (2) compliance of the design with the seven years lifetime requirement. The project was composed of two major phases. The initial phase of the project has concentrated on understanding the TOPAZ-2 thermionic reactor in sufficient detail to allow several follow-on tasks. The primary interest during this first phase has been given on identifying the potential of the TOPAZ-2 design for further improvements. The second phase of the project has focused on the feasibility of a TOPAZ-2 system capable of delivering 30-50 kWe. Towards the elimination of single-point failures in the load voltage regulation system an active voltage regulator has been designed to be used in conjunction with the available shunt load voltage regulator. The possible use of a dual-loop, model-based adaptive control system for load-following in the TOPAZ-2 has also been investigated. The objective of this fault-tolerant, autonomous control system is to deliver the demanded electric power at the desired voltage level, by appropriately manipulating the neutron power through the control drums. As a result, sufficient thermal power is produced to meet the required demand in the presence of dynamically changing system operating conditions and potential sensor failures. The designed controller is proposed for use in combination with the currently available shunt regulators, or as a back-up controller when other means of power system control, including some of the sensors, fail.

ADAPTION OF NONSTANDARD PIPING COMPONENTS INTO PRESENT DAY SEISMIC CODES

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

D. T. Clark; M. J. Russell; R. E. Spears

2009-07-01

With spiraling energy demand and flat energy supply, there is a need to extend the life of older nuclear reactors. This sometimes requires that existing systems be evaluated to present day seismic codes. Older reactors built in the 1960s and early 1970s often used fabricated piping components that were code compliant during their initial construction time period, but are outside the standard parameters of present-day piping codes. There are several approaches available to the analyst in evaluating these non-standard components to modern codes. The simplest approach is to use the flexibility factors and stress indices for similar standard components withmore » the assumption that the non-standard component’s flexibility factors and stress indices will be very similar. This approach can require significant engineering judgment. A more rational approach available in Section III of the ASME Boiler and Pressure Vessel Code, which is the subject of this paper, involves calculation of flexibility factors using finite element analysis of the non-standard component. Such analysis allows modeling of geometric and material nonlinearities. Flexibility factors based on these analyses are sensitive to the load magnitudes used in their calculation, load magnitudes that need to be consistent with those produced by the linear system analyses where the flexibility factors are applied. This can lead to iteration, since the magnitude of the loads produced by the linear system analysis depend on the magnitude of the flexibility factors. After the loading applied to the nonstandard component finite element model has been matched to loads produced by the associated linear system model, the component finite element model can then be used to evaluate the performance of the component under the loads with the nonlinear analysis provisions of the Code, should the load levels lead to calculated stresses in excess of Allowable stresses. This paper details the application of component-level finite element modeling to account for geometric and material nonlinear component behavior in a linear elastic piping system model. Note that this technique can be applied to the analysis of B31 piping systems.« less

DOE/JPL advanced thermionic technology program

NASA Technical Reports Server (NTRS)

1979-01-01

Progress made in different tasks of the advanced thermionic technology program is described. The tasks include surface and plasma investigations (surface characterization, spectroscopic plasma experiments, and converter theory); low temperature converter development (tungsten emitter, tungsten oxide collector and tungsten emitter, nickel collector); component hardware development (hot shell development); flame-fired silicon carbide converters; high temperature and advanced converter studies; postoperational diagnostics; and correlation of design interfaces.

EFFECT OF THERMIONIC EMISSION AT ROOM TEMPERATURES IN PHOTOSENSITIVE GEIGERMULLER TUBES

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

Grotowski, K.; Hrynkiewicz, A.Z.; Niewodniczanski, H.

1953-01-01

The temperature-dependence of the background of Geiger-Mueller counting tubes was compared for nonsensitized and sensitized (treated by electric discharge) tubes. A strong increase of background with increasing temperature was observed for phatosensitive counters, while no change was observed in nonsensetized counters. It is shown that the increase is due to thermionic emission of the brass cathode. (T.R.H.)

Thermionic Energy

NASA Technical Reports Server (NTRS)

1985-01-01

Thermionic energy conversion is the production of energy from a nuclear source. It is a technology advanced by SNSO, a joint research and development organization formed by NASA and the AEC. SNSO contracted with Thermo Electron Corporation to develop high temperature applications, i.e., metals with high melting points. Thermo Electron Corporation's expertise resulted in contracts for products made from exotic metals such as bone implants, artificial hips, and heart pacemakers.

Diminiode thermionic energy conversion with lanthanum-hexaboride electrodes

NASA Technical Reports Server (NTRS)

Kroeger, E. W.; Bair, V. L.; Morris, J. F.

1978-01-01

Thermionic conversion data obtained from a variable gap cesium diminiode with a hot pressed, sintered lanthanum hexaboride emitter and an arc melted lanthanum hexaboride collector are presented. Performance curves cover a range of temperatures: emitter 1500 to 1700 K, collector 750 to 1000 K, and cesium reservoir 370 to 510 K. Calculated values of emitter and collector work functions and barrier index are also given.

High aspect ratio catalytic reactor and catalyst inserts therefor

DOEpatents

Lin, Jiefeng; Kelly, Sean M.

2018-04-10

The present invention relates to high efficient tubular catalytic steam reforming reactor configured from about 0.2 inch to about 2 inch inside diameter high temperature metal alloy tube or pipe and loaded with a plurality of rolled catalyst inserts comprising metallic monoliths. The catalyst insert substrate is formed from a single metal foil without a central supporting structure in the form of a spiral monolith. The single metal foil is treated to have 3-dimensional surface features that provide mechanical support and establish open gas channels between each of the rolled layers. This unique geometry accelerates gas mixing and heat transfer and provides a high catalytic active surface area. The small diameter, high aspect ratio tubular catalytic steam reforming reactors loaded with rolled catalyst inserts can be arranged in a multi-pass non-vertical parallel configuration thermally coupled with a heat source to carry out steam reforming of hydrocarbon-containing feeds. The rolled catalyst inserts are self-supported on the reactor wall and enable efficient heat transfer from the reactor wall to the reactor interior, and lower pressure drop than known particulate catalysts. The heat source can be oxygen transport membrane reactors.

An overview of zinc addition for BWR dose rate control

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

Marble, W.J.

1995-03-01

This paper presents an overview of the BWRs employing feedwater zinc addition to reduce primary system dose rates. It identifies which BWRs are using zinc addition and reviews the mechanical injection and passive addition hardware currently being employed. The impact that zinc has on plant chemistry, including the factor of two to four reduction in reactor water Co-60 concentrations, is discussed. Dose rate results, showing the benefits of implementing zinc on either fresh piping surfaces or on pipes with existing films are reviewed. The advantages of using zinc that is isotopically enhanced by the depletion of the Zn-64 precursor tomore » Zn-65 are identified.« less

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

NASA Technical Reports Server (NTRS)

Dochat, George R.; Dhar, Manmohan

1991-01-01

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

Environmental Effect on Evolutionary Cyclic Plasticity Material Parameters of 316 Stainless Steel: An Experimental & Material Modeling Approach

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

Mohanty, Subhasish; Soppet, William K.; Majumdar, Saurin

2014-09-20

This report provides an update on an earlier assessment of environmentally assisted fatigue for light water reactor (LWR) materials under extended service conditions. This report is a deliverable under the work package for environmentally assisted fatigue in the Light Water Reactor Sustainability (LWRS) program. The overall objective of this LWRS project is to assess the degradation by environmentally assisted cracking/fatigue of LWR materials such as various alloy base metals and their welds used in reactor coolant system piping. This effort is to support the Department of Energy LWRS program for developing tools to understand the aging/failure mechanism and to predictmore » the remaining life of LWR components for anticipated 60-80 year operation.« less

Reducing Energy Degradation Due to Back-bombardment Effect with Modulated RF Input in S-band Thermionic RF Gun

NASA Astrophysics Data System (ADS)

Kii, Toshiteru; Nakai, Yoko; Fukui, Toshio; Zen, Heishun; Kusukame, Kohichi; Okawachi, Norihito; Nakano, Masatsugu; Masuda, Kai; Ohgaki, Hideaki; Yoshikawa, Kiyoshi; Yamazaki, Tetsuo

2007-01-01

Energy degradation due to back-bombardment effect is quite serious to produce high-brightness electron beam with long macro-pulse with thermionic rf gun. To avoid the back-bombardment problem, a laser photo cathode is used at many FEL facilities, but usually it costs high and not easy to operate. Thus we have studied long pulse operation of the rf gun with thermionic cathode, which is inexpensive and easy to operate compared to the photocathode rf gun. In this work, to reduce the energy degradation, we controlled input rf power amplitude by controlling pulse forming network of the power modulator for klystron. We have successfully increased the pulse duration up to 4 μs by increasing the rf power from 7.8 MW to 8.5 MW during the macro pulse.

Work function determination of promising electrode materials for thermionic energy converters

NASA Technical Reports Server (NTRS)

Jacobson, D.; Storms, E.; Skaggs, B.; Kouts, T.; Jaskie, J.; Manda, M.

1976-01-01

The work function determinations of candidate materials for low temperature (1400 K) thermionics through vacuum emission tests are discussed. Two systems, a vacuum emission test vehicle and a thermionic emission microscope are used for emission measurements. Some nickel and cobalt based super alloys were preliminarily examined. High temperature physical properties and corrosion behavior of some super alloy candidates are presented. The corrosion behavior of sodium is of particular interest since topping cycles might use sodium heat transfer loops. A Marchuk tube was designed for plasma discharge studies with the carbide and possibly some super alloy samples. A series of metal carbides and other alloys were fabricated and tested in a special high temperature mass spectrometer. This information coupled with work function determinations was evaluated in an attempt to learn how electron bonding occurs in transition alloys.

PBF Cooling Tower (PER720) and its Auxiliary Building (PER625). Camera ...

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

PBF Cooling Tower (PER-720) and its Auxiliary Building (PER-625). Camera facing west shows east facades. Center pipe carried secondary coolant water from reactor. Building to distributor basin. Date: August 2003. INEEL negative no. HD-35-10-1 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

Heat Recuperator Engineering for an ARL Liquid-Fueled Thermophotovoltaic Power Source Demonstrator

DTIC Science & Technology

2014-09-01

using logistics and multiple other fuels. Some potential technologies include thermoelectric , thermophotovoltaic (TPV), and thermionic. For these... thermoelectric , thermophotovoltaic (TPV), and thermionic. For these technologies, thermal efficiency is critical to achieve high energy density and thermal-to... thermoelectric and TPV. The exhaust gas will be above this temperature, but more than 50% of the thermal power of the combustor can be lost to the exhaust

User interface and operational issues with thermionic space power systems

NASA Technical Reports Server (NTRS)

Dahlberg, R. C.; Fisher, C. R.

1987-01-01

Thermionic space power systems have unique features which facilitate predeployment operations, provide operational flexibility and simplify the interface with the user. These were studied in some detail during the SP-100 program from 1983 to 1985. Three examples are reviewed in this paper: (1) system readiness verification in the prelaunch phase; (2) startup, shutdown, and dormancy in the operations phase; (3) part-load operation in the operations phase.

Flow accelerated corrosion of carbon steel feeder pipes from pressurized heavy water reactors

NASA Astrophysics Data System (ADS)

Singh, J. L.; Kumar, Umesh; Kumawat, N.; Kumar, Sunil; Kain, Vivekanand; Anantharaman, S.; Sinha, A. K.

2012-10-01

Detailed investigation of a number of feeder pipes received from Rajasthan Atomic Power Station Unit 2 (RAPS#2) after en-masse feeder pipe replacement after 15.67 Effective Full Power Years (EFPYs) was carried out. Investigations included ultrasonic thickness measurement by ultrasonic testing, optical microscopy, scanning electron microscopy, chemical analysis and X-ray Diffraction (XRD). Results showed that maximum thickness reduction of the feeder had occurred downstream and close to the weld in 32 NB (1.25″/32.75 mm ID) elbows. Rate of Flow Accelerated Corrosion (FAC) was measured to be higher in the lower diameter feeder pipes due to high flow velocity and turbulence. Weld regions had thinned to a lower extent than the parent material due to higher chromium content in the weld. A weld protrusion has been shown to add to the thinning due to FAC and lead to faster thinning rate at localized regions. Surface morphology of inner surface of feeder had shown different size scallop pattern over the weld and parent material. Inter-granular cracks were also observed along the weld fusion line and in the parent material in 32 NB outlet feeder elbow.

Fatigue evaluation of socket welded piping in nuclear power plant

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

Vecchio, R.S.

1996-12-01

Fatigue failures in piping systems occur, almost without exception, at the welded connections. In nuclear power plant systems, such failures occur predominantly at the socket welds of small diameter piping ad fillet attachment welds under high-cycle vibratory conditions. Nearly all socket weld fatigue failures are identified by leaks which, though not high in volume, generally are costly due to attendant radiological contamination. Such fatigue cracking was recently identified in the 3/4 in. diameter recirculation and relief piping socket welds from the reactor coolant system (RCS) charging pumps at a nuclear power plant. Consequently, a fatigue evaluation was performed to determinemore » the cause of cracking and provide an acceptable repair. Socket weld fatigue life was evaluated using S-N type fatigue life curves for welded structures developed by AASHTO and the assessment of an effective cyclic stress range adjacent to each socket weld. Based on the calculated effective tress ranges and assignment of the socket weld details to the appropriate AASHTO S-N curves, the socket weld fatigue lives were calculated and found to be in excellent agreement with the accumulated cyclic life to-date.« less

500(deg)C electronics for harsh environments

NASA Technical Reports Server (NTRS)

Sadwick, Laurence P.; Hwu, R. Jennifer; Chern, J. H. Howard; Lin, Ching-Hsu; Castillo, Linda Del; Johnson, Travis

2005-01-01

Solid state vacuum devices (SSVDs) are a relatively new class of electronic devices. Innosys is a leading producer of high frequency SSVDs for a number of applications, including RF communications. SSVDs combine features inherent to both solid state and vacuum transistors. Electron transport can be by solid state or vacuum or both. The focus of this talk is on thermionic SSVDs, in which the primary vacuum transport is by thermionically liberated electron emission.

Thermionic modules

DOEpatents

King, Donald B.; Sadwick, Laurence P.; Wernsman, Bernard R.

2002-06-18

Modules of assembled microminiature thermionic converters (MTCs) having high energy-conversion efficiencies and variable operating temperatures manufactured using MEMS manufacturing techniques including chemical vapor deposition. The MTCs incorporate cathode to anode spacing of about 1 micron or less and use cathode and anode materials having work functions ranging from about 1 eV to about 3 eV. The MTCs also exhibit maximum efficiencies of just under 30%, and thousands of the devices and modules can be fabricated at modest costs.

Thermionic converter output as a function of collector temperature

NASA Technical Reports Server (NTRS)

Stark, G.; Saunders, M.; Lieb, D.

1980-01-01

Surprisingly few data are available on the variation of thermionic converter output with collector temperature. In this study the output power density has been measured as a function of collector temperature (at a fixed emitter temperature of 1650 K) for six converters with different electrode combinations. Collector temperatures ranged from 750 to 1100 K. For collector temperatures below 900 K, converters built with sublimed molybdenum oxide collectors gave the best performance.

Space station systems analysis study. Part 3: Documentation. Volume 7: SCB alternate EPS evaluation, task 10

NASA Technical Reports Server (NTRS)

1977-01-01

Power levels up to 100 kWe average were baselined for the electrical power system of the space construction base, a long-duration manned facility capable of supporting manufacturing and large scale construction projects in space. Alternatives to the solar array battery systems discussed include: (1) solar concentrator/brayton; (2) solar concentrator/thermionic; (3) isotope/brayton; (4) nuclear/brayton; (5) nuclear thermoelectric; and (6) nuclear thermionic.

Emission properties and back-bombardment for CeB{sub 6} compared to LaB{sub 6}

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

Bakr, Mahmoud, E-mail: m-a-bakr@iae.kyoto-u.ac.jp; Kawai, M.; Kii, T.

The emission properties of CeB{sub 6} compared to LaB{sub 6} thermionic cathodes have been measured using an electrostatic DC gun. Obtaining knowledge of the emission properties is the first step in understanding the back-bombardment effect that limits wide usage of thermionic radio-frequency electron guns. The effect of back-bombardment electrons on CeB{sub 6} compared to LaB{sub 6} was studied using a numerical simulation model. The results show that for 6 μs pulse duration with input radio-frequency power of 8 MW, CeB{sub 6} should experience 14% lower temperature increase and 21% lower current density rise compared to LaB{sub 6}. We conclude that CeB{submore » 6} has the potential to become the future replacement for LaB{sub 6} thermionic cathodes in radio-frequency electron guns.« less

Particle-In-Cell Simulations of a Thermionic Converter

NASA Astrophysics Data System (ADS)

Clark, S. E.

2017-12-01

Simulations of thermionic converters are presented where cesium is used as a work function reducing agent in a nano-fabricated triode configuration. The cathode and anode are spaced on the order of 100 μm, and the grid structure has features on the micron scale near the anode. The hot side is operated near 1600 K, the cold side near 600 K, and the converter has the potential to convert heat to DC electrical current upwards of 20% efficiency. Affordable and robust thermionic converters have the potential to displace century old mechanical engines and turbines as a primary means of electrical power generation in the near future. High efficiency converters that operate at a small scale could be used to generate power locally and alleviate the need for large scale power transmission systems. Electron and negative cesium ion back emission from the anode are considered, as well as device longevity and fabrication feasibility.

Non-equilibrium thermionic electron emission for metals at high temperatures

NASA Astrophysics Data System (ADS)

Domenech-Garret, J. L.; Tierno, S. P.; Conde, L.

2015-08-01

Stationary thermionic electron emission currents from heated metals are compared against an analytical expression derived using a non-equilibrium quantum kappa energy distribution for the electrons. The latter depends on the temperature decreasing parameter κ ( T ) , which decreases with increasing temperature and can be estimated from raw experimental data and characterizes the departure of the electron energy spectrum from equilibrium Fermi-Dirac statistics. The calculations accurately predict the measured thermionic emission currents for both high and moderate temperature ranges. The Richardson-Dushman law governs electron emission for large values of kappa or equivalently, moderate metal temperatures. The high energy tail in the electron energy distribution function that develops at higher temperatures or lower kappa values increases the emission currents well over the predictions of the classical expression. This also permits the quantitative estimation of the departure of the metal electrons from the equilibrium Fermi-Dirac statistics.

Particle-In-Cell Simulations of a Thermionic Converter

NASA Astrophysics Data System (ADS)

Clark, Stephen

2017-10-01

Simulations of thermionic converters are presented where cesium is used as a work function reducing agent in a nano-fabricated triode configuration. The cathode and anode are spaced on the order of 100 μm, and the grid structure has features on the micron scale near the anode. The hot side is operated near 1600 K, the cold side near 600 K, and the converter has the potential to convert heat to DC electrical current upwards of 20% efficiency. Affordable and robust thermionic converters have the potential to displace century old mechanical engines and turbines as a primary means of electrical power generation in the near future. High efficiency converters that operate at a small scale could be used to generate power locally and alleviate the need for large scale power transmission systems. Electron and negative cesium ion back emission from the anode are considered, as well as device longevity and fabrication feasibility.

Enhancement of thermionic emission by light

NASA Astrophysics Data System (ADS)

Sodha, Mahendra Singh; Srivastava, Sweta; Mishra, Rashmi

2017-03-01

In this paper the rate of electron emission from an illuminated hot metallic plate has been evaluated on the basis of the free electron theory of metals and Fowler's theory of photoelectric electron emission. The modification of the electron energy distribution (or enhancement of electron temperature) in the plate by energetic electrons (which get their normal energy enhanced on the surface by incident photons of frequency below threshold and are not emitted) has been taken into account. The thermionic current as modified by the electron temperature so enhanced by irradiation has been evaluated. The results may be applicable to thermionic convertors, as proposed to be operated by Schwede et al. [J.W. Schwede, I. Bargatin, D.C. Riley, B.E. Hardin, S.J. Rosenthal, Y. Sun, F. Schmitt, P. Pianette, R.T. Howe, Z. Shen, N.A. Melosh, Nat. Mater. 9, 762 (2010)]. Numerical results have been presented and discussed.

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

PubMed

Seyfi, Somayye; Gharib, Morteza

2015-10-01

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

Results from the decontamination of and the shielding arrangements in the reactor pressure vessel in Oskarshamn 1-1994

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

Lowendahl, B.

1995-03-01

In September 1992 Oskarshamn 1 was shut down in order to carry out measures to correct discovered deficiencies in the emergency cooling systems. Due to the results of a comprehensive non destructive test programme it was decided to perform a major replacement of pipes in the primary systems including a full system decontamination using the Siemens CORD process. The paper briefly presents the satisfying result of the decontamination performed in May-June 1993. When in late June 1993 cracks also were detected in the feed-water pipes situated inside the reactor pressure vessel (RPV) the plans were reconsidered and a large projectmore » was formed with the aim, in a first phase, to verify the integrity of the RPV. In order to make it possible to perform work manually inside the RPV special radiation protection measures had to be carried out. In January 1994 the lower region of the RPV was decontaminated, again using the CORD-process, followed by the installation of a special shielding construction in the RPV. The surprisingly good results of these efforts are also briefly described in the paper.« less

J-R fracture characteristics of ferritic steels for RPVs and RCS piping of nuclear power plants

NASA Astrophysics Data System (ADS)

Yoon, Ji-Hyun; Lee, Bong-Sang; Hong, Jun-Hwa

2001-10-01

J-R fracture resistance tests have been performed on 3 heats of SA508-Gr.3 nuclear reactor pressure vessel (RPV) steel as well as 2 heats of SA516-Gr.70 and a heat of SA508-Gr.1a steels for nuclear reactor coolant system (RCS) piping. For the latter two steels, dynamic in addition to static J-R fracture resistances were investigated. From the test results of the SA508-Gr.3 steels, the J-R fracture resistance was superior in the following order: Si-killing steel, modified VCD steel and VCD steel. Microstructural analyses were carried out to correlate J-R fracture resistances with microstructural characteristics. According to the test results for SA508-Gr.1a and SA516-Gr.70 steels, all of the tested steels showed steep drops in fracture resistance at certain temperature and loading rate combinations. One heat of SA516-Gr.70 steel was very sensitive to dynamic strain aging and its fracture resistance was significantly low. It was concluded that microstructural and chemical factors affect the J-R fracture and DSA characteristics of SA516-Gr.70 steels.

Study of radioisotope safety devices for electric propulsion system, Volume 1: Summary report

NASA Technical Reports Server (NTRS)

Bradshaw, G. B.; Homeyer, W. G.; Postula, F. D.; Steeger, E. J.

1972-01-01

A new reference design was prepared for the 5 kW(e) thermionic power supply. The safety equipment in this design is a passive containment system which does not rely on the operation of any mechanisms such as a launch escape rocket or deployment of parachutes. It includes: (1) a blast shield to protect against the explosion of the launch vehicle; (2) a combination of refractory thermal insulation and heat storage material to protect against a sustained launch pad fire; (3) a reentry body with a spherical nose and a large conical flare at the aft end to stabilize the reentry attitude and lower the terminal velocity in air; (4) composite graphite thermal protection to sustain the reentry heat pulse; (5) crushable honeycomb behind the nose to limit the deceleration of the radioisotope source due to impact on land at terminal velocity; (6) a double-walled secondary containment vessel surrounding the isotopic capsules; (7) neutron shielding to reduce external dose rates; (8) an auxiliary cooling system employing redundant heat pipes to remove the radioactive decay heat from the heat source and reject it to the surroundings or to a forced convection loop.

Design, Operation, and Modeling of a Vertical APCVD Reactor for Silicon Carbide Film Growth

NASA Technical Reports Server (NTRS)

DeAnna, Russell G.; Fleischman, Aaron J.; Zorman, Christian A.; Mehregany, Mehran

1998-01-01

An atmospheric pressure chemical vapor deposition (APCVD) reactor utilizing a unique vertical geometry which enables 3C-SiC films to be grown on two, 4-inch diameter Si wafers has been constructed. Contrary to expectations, 3C-SiC films grown in this reactor are thickest at the downstream end of the substrates. To better understand the reason for the thickness distribution on the wafers, an axisymmetric finite-element model of the gas flow in the reactor was constructed. The model uses the ANSYS53 Flowtran package and includes compressible and temperature-dependent fluid properties in laminar or turbulent flow. It does not include reaction chemistry or unsteady flow. The ANSYS53 results predict that the cool, inlet fluid falls through the inlet pipe and the warm, diffuser region like a jet. This jet impinges on top of the susceptor and gets diverted to the reactor side walls, where it flows to the bottom of the reactor, turns, and slowly rises along the face of the susceptor. This may explain why the SiC films are thickest at the downstream side of the wafers, as gas containing fresh reactants first passes over this region. Modeling results are presented for both one atmosphere and one half atmosphere reactor pressure.

Microfabricated thermionic detector

DOEpatents

Lewis, Patrick R; Manginell, Ronald P; Wheeler, David R; Trudell, Daniel E

2012-10-30

A microfabricated TID comprises a microhotplate and a thermionic source disposed on the microhotplate. The microfabricated TID can provide high sensitivity and selectivity to nitrogen- and phosphorous-containing compounds and other compounds containing electronegative function groups. The microfabricated TID can be microfabricated with semiconductor-based materials. The microfabricated TID can be combined with a microfabricated separation column and used in microanalytical system for the rapid on-site detection of pesticides, chemical warfare agents, explosives, pharmaceuticals, and other organic compounds that contain nitrogen or phosphorus.

PBF Cooling Tower under construction. Cold water basin is five ...

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

PBF Cooling Tower under construction. Cold water basin is five feet deep. Foundation and basin walls are reinforced concrete. Camera facing west. Pipe openings through wall in front are outlets for return flow of cool water to reactor building. Photographer: John Capek. Date: September 4, 1968. INEEL negative no. 68-3473 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

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.

Numerical study of heat and mass transfer in inertial suspensions in pipes.

NASA Astrophysics Data System (ADS)

Niazi Ardekani, Mehdi; Brandt, Luca

2017-11-01

Controlling heat and mass transfer in particulate suspensions has many important applications such as packed and fluidized bed reactors and industrial dryers. In this work, we study the heat and mass transfer within a suspension of spherical particles in a laminar pipe flow, using the immersed boundary method (IBM) to account for the solid fluid interactions and a volume of fluid (VoF) method to resolve temperature equation both inside and outside of the particles. Tracers that follow the fluid streamlines are considered to investigate mass transfer within the suspension. Different particle volume fractions 5, 15, 30 and 40% are simulated for different pipe to particle diameter ratios: 5, 10 and 15. The preliminary results quantify the heat and mass transfer enhancement with respect to a single-phase laminar pipe flow. We show in particular that the heat transfer from the wall saturates for volume fractions more than 30%, however at high particle Reynolds numbers (small diameter ratios) the heat transfer continues to increase. Regarding the dispersion of tracer particles we show that the diffusivity of tracers increases with volume fraction in radial and stream-wise directions however it goes through a peak at 15% in the azimuthal direction. European Research Council, Grant No. ERC-2013-CoG- 616186, TRITOS; SNIC (the Swedish National Infrastructure for Computing).

Heat Rejection Concepts for Lunar Fission Surface Power Applications

NASA Technical Reports Server (NTRS)

Siamidis, John

2006-01-01

This paper describes potential heat rejection design concepts for lunar surface Brayton power conversion systems. Brayton conversion systems are currently under study by NASA for surface power applications. Surface reactors may be used for the moon to power human outposts enabling extended stays and closed loop life support. The Brayton Heat Rejection System (HRS) must dissipate waste heat generated by the power conversion system due to inefficiencies in the thermal-to-electric conversion process. Space Brayton conversion system designs tend to optimize at efficiencies of about 20 to 25 percent with radiator temperatures in the 400 K to 600 K range. A notional HRS was developed for a 100 kWe-class Brayton power system that uses a pumped water heat transport loop coupled to a water heat pipe radiator. The radiator panels employ a tube and fin construction consisting of regularly-spaced circular heat pipes contained within two composite facesheets. The water heat pipes interface to the coolant through curved sections partially contained within the cooling loop. The paper evaluates various design parameters including radiator panel orientation, coolant flow path, and facesheet thickness. Parameters were varied to compare design options on the basis of H2O pump pressure rise and required power, heat pipe unit power and radial flux, radiator area, radiator panel areal mass, and overall HRS mass.

Modeling of the steam hydrolysis in a two-step process for hydrogen production by solar concentrated energy

NASA Astrophysics Data System (ADS)

Valle-Hernández, Julio; Romero-Paredes, Hernando; Pacheco-Reyes, Alejandro

2017-06-01

In this paper the simulation of the steam hydrolysis for hydrogen production through the decomposition of cerium oxide is presented. The thermochemical cycle for hydrogen production consists of the endothermic reduction of CeO2 to lower-valence cerium oxide, at high temperature, where concentrated solar energy is used as a source of heat; and of the subsequent steam hydrolysis of the resulting cerium oxide to produce hydrogen. The modeling of endothermic reduction step was presented at the Solar Paces 2015. This work shows the modeling of the exothermic step; the hydrolysis of the cerium oxide (III) to form H2 and the corresponding initial cerium oxide made at lower temperature inside the solar reactor. For this model, three sections of the pipe where the reaction occurs were considered; the steam water inlet, the porous medium and the hydrogen outlet produced. The mathematical model describes the fluid mechanics; mass and energy transfer occurring therein inside the tungsten pipe. Thermochemical process model was simulated in CFD. The results show a temperature distribution in the solar reaction pipe and allow obtaining the fluid dynamics and the heat transfer within the pipe. This work is part of the project "Solar Fuels and Industrial Processes" from the Mexican Center for Innovation in Solar Energy (CEMIE-Sol).

The Effect of Flow Swirling on the Safety and Reliability of Nuclear Power Installations of New Generation

NASA Astrophysics Data System (ADS)

Mitrofanova, O. V.; Ivlev, O. A.; Urtenov, D. S.

2018-03-01

Hydrodynamics and heat exchange in the elements of thermal hydraulic tracts of ship nuclear reactors of the new generation were numerically simulated in this work. Parts of the coolant circuit in the collector and piping systems with geometries that may lead to generation of stable large-scale vortexes, causing a wide range of acoustic oscillations of the coolant, were selected as modeling objects. The purpose of the research is to develop principles of physical and mathematical modeling for scientific substantiation of optimal layout solutions that ensure enhanced operational life of icebreaker’s nuclear power installations of new generation with reactors of integral type.

21. View from the work area of the front face ...

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

21. View from the work area of the front face of the pile in the 105 building, in this case at the F Reactor in February 1945. The 2,004 pigtails and process tube nozzles are neatly aligned in rows and columns across the face of the pile. The cooling water risers stand at the left and right of the pile and the distribution crossheaders run across its face. The pipes running vertically at the bottom of the pile carry cooling water to the thermal shield. The low railing along the floor in front of the face prevented workers from accidentally falling into the charging elevator pit. D-8326 - B Reactor, Richland, Benton County, WA

RELAP5 Analysis of the Hybrid Loop-Pool Design for Sodium Cooled Fast Reactors

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

Hongbin Zhang; Haihua Zhao; Cliff Davis

2008-06-01

An innovative hybrid loop-pool design for sodium cooled fast reactors (SFR-Hybrid) has been recently proposed. This design takes advantage of the inherent safety of a pool design and the compactness of a loop design to improve economics and safety of SFRs. In the hybrid loop-pool design, primary loops are formed by connecting the reactor outlet plenum (hot pool), intermediate heat exchangers (IHX), primary pumps and the reactor inlet plenum with pipes. The primary loops are immersed in the cold pool (buffer pool). Passive safety systems -- modular Pool Reactor Auxiliary Cooling Systems (PRACS) – are added to transfer decay heatmore » from the primary system to the buffer pool during loss of forced circulation (LOFC) transients. The primary systems and the buffer pool are thermally coupled by the PRACS, which is composed of PRACS heat exchangers (PHX), fluidic diodes and connecting pipes. Fluidic diodes are simple, passive devices that provide large flow resistance in one direction and small flow resistance in reverse direction. Direct reactor auxiliary cooling system (DRACS) heat exchangers (DHX) are immersed in the cold pool to transfer decay heat to the environment by natural circulation. To prove the design concepts, especially how the passive safety systems behave during transients such as LOFC with scram, a RELAP5-3D model for the hybrid loop-pool design was developed. The simulations were done for both steady-state and transient conditions. This paper presents the details of RELAP5-3D analysis as well as the calculated thermal response during LOFC with scram. The 250 MW thermal power conventional pool type design of GNEP’s Advanced Burner Test Reactor (ABTR) developed by Argonne National Laboratory was used as the reference reactor core and primary loop design. The reactor inlet temperature is 355 °C and the outlet temperature is 510 °C. The core design is the same as that for ABTR. The steady state buffer pool temperature is the same as the reactor inlet temperature. The peak cladding, hot pool, cold pool and reactor inlet temperatures were calculated during LOFC. The results indicate that there are two phases during LOFC transient – the initial thermal equilibration phase and the long term decay heat removal phase. The initial thermal equilibration phase occurs over a few hundred seconds, as the system adjusts from forced circulation to natural circulation flow. Subsequently, during long-term heat removal phase all temperatures evolve very slowly due to the large thermal inertia of the primary and buffer pool systems. The results clearly show that passive safety PRACS can effectively transfer decay heat from the primary system to the buffer pool by natural circulation. The DRACS system in turn can effectively transfer the decay heat to the environment.« less

An overview on the reactors to study drinking water biofilms.

PubMed

Gomes, I B; Simões, M; Simões, L C

2014-10-01

The development of biofilms in drinking water distribution systems (DWDS) can cause pipe degradation, changes in the water organoleptic properties but the main problem is related to the public health. Biofilms are the main responsible for the microbial presence in drinking water (DW) and can be reservoirs for pathogens. Therefore, the understanding of the mechanisms underlying biofilm formation and behavior is of utmost importance in order to create effective control strategies. As the study of biofilms in real DWDS is difficult, several devices have been developed. These devices allow biofilm formation under controlled conditions of physical (flow velocity, shear stress, temperature, type of pipe material, etc), chemical (type and amount of nutrients, type of disinfectant and residuals, organic and inorganic particles, ions, etc) and biological (composition of microbial community - type of microorganism and characteristics) parameters, ensuring that the operational conditions are similar as possible to the DWDS conditions in order to achieve results that can be applied to the real scenarios. The devices used in DW biofilm studies can be divided essentially in two groups, those usually applied in situ and the bench top laboratorial reactors. The selection of a device should be obviously in accordance with the aim of the study and its advantages and limitations should be evaluated to obtain reproducible results that can be transposed into the reality of the DWDS. The aim of this review is to provide an overview on the main reactors used in DW biofilm studies, describing their characteristics and applications, taking into account their main advantages and limitations. Copyright © 2014 Elsevier Ltd. All rights reserved.

JSUS solar thermal thruster and its integration with thermionic power converter

NASA Astrophysics Data System (ADS)

Shimizu, Morio; Eguchi, Kunihisa; Itoh, Katsuya; Sato, Hitoshi; Fujii, Tadayuki; Okamoto, Ken-Ichi; Igarashi, Tadashi

1998-01-01

This paper describes solar heating test results of a single crystal Mo thruster of solar thermal propulsion (STP) with super high-temperature brazing of Mo/Ru for hydrogen-gas sealing, using the paraboloidal concentrator of 1.6 m diameter newly installed in NAL in the Japan Solar Upper Stage (JSUS) research program. The designed thruster has a target Isp about 800 sec for 2,250 K or higher temperatures of hydrogen propellant. Additionally, tungsten CVD-coating was applied to a outer surface of the thruster in order to prevent vaporization of the wall material and Mo/Ru under the condition of high temperature over 2,500K and high vacuum. Also addressed in our paper is solar thermionic power module design for the integration with the STP receiver. The thermionic converter (TIC) module is of a planar type in a Knudsen-mode operation and provides a high conversion efficiency of 23% at the TIC emitter temperature of nearly 1,850 K for a heat input flux of 24 W/cm2.

High-efficiency AlxGa1-xAs/GaAs cathode for photon-enhanced thermionic emission solar energy converters

NASA Astrophysics Data System (ADS)

Feng, Cheng; Zhang, Yijun; Qian, Yunsheng; Wang, Ziheng; Liu, Jian; Chang, Benkang; Shi, Feng; Jiao, Gangcheng

2018-04-01

A theoretical emission model for AlxGa1-xAs/GaAs cathode with complex structure based on photon-enhanced thermionic emission is developed by utilizing one-dimensional steady-state continuity equations. The cathode structure comprises a graded-composition AlxGa1-xAs window layer and an exponential-doping GaAs absorber layer. In the deduced model, the physical properties changing with the Al composition are taken into consideration. Simulated current-voltage characteristics are presented and some important factors affecting the conversion efficiency are also illustrated. Compared with the graded-composition and uniform-doping cathode structure, and the uniform-composition and uniform-doping cathode structure, the graded-composition and exponential-doping cathode structure can effectively improve the conversion efficiency, which is ascribed to the twofold built-in electric fields. More strikingly, this graded bandgap structure is especially suitable for photon-enhanced thermionic emission devices since a higher conversion efficiency can be achieved at a lower temperature.

Effects of nanoscale vacuum gap on photon-enhanced thermionic emission devices

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

Wang, Yuan; Liao, Tianjun; Zhang, Yanchao

2016-01-28

A new model of the photon-enhanced thermionic emission (PETE) device with a nanoscale vacuum gap is established by introducing the quantum tunneling effect and the image force correction. Analytic expressions for both the thermionic emission and tunneling currents are derived. The electron concentration and the temperature of the cathode are determined by the particle conservation and energy balance equations. The effects of the operating voltage on the maximum potential barrier, cathode temperature, electron concentration and equilibrium electron concentration of the conduction band, and efficiency of the PETE device are discussed in detail for different given values of the vacuum gapmore » length. The influence of the band gap of the cathode and flux concentration on the efficiency is further analyzed. The maximum efficiency of the PETE and the corresponding optimum values of the band gap and the operating voltage are determined. The results obtained here show that the efficiency of the PETE device can be significantly improved by employing a nanoscale vacuum gap.« less

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

Gerard, R.; Malekian, C.; Meessen, O.

The Leak Before Break (LBB) concept allows to eliminate from the design basis the double-ended guillotine break of the primary loop piping, provided it can be demonstrated by a fracture mechanics analysis that a through-wall flaw, of a size giving rise to a leakage still well detectable by the plant leak detection systems, remains stable even under accident conditions (including the Safe Shutdown Earthquake (SSE)). This concept was successfully applied to the primary loop piping of several Belgian Pressurized Water Reactor (PWR) units, operated by the Utility Electrabel. One of the main benefits is to permit justification of supports inmore » the primary loop and justification of the integrity of the reactor pressure vessel and internals in case of a Loss Of Coolant Accident (LOCA) in stretch-out conditions. For two of the Belgian PWR units, the LBB approach also made it possible to reduce the number of large hydraulic snubbers installed on the primary coolant pumps. Last but not least, the LBB concept also facilitates the steam generator replacement operations, by eliminating the need for some pipe whip restraints located close to the steam generator. In addition to the U.S. regulatory requirements, the Belgian safety authorities impose additional requirements which are described in details in a separate paper. An novel aspect of the studies performed in Belgium is the way in which residual loads in the primary loop are taken into account. Such loads may result from displacements imposed to close the primary loop in a steam generator replacement operation, especially when it is performed using the {open_quote}two cuts{close_quotes} technique. The influence of such residual loads on the LBB margins is discussed in details and typical results are presented.« less

[Case study of red water phenomenon in drinking water distribution systems caused by water source switch].

PubMed

Wang, Yang; Zhang, Xiao-jian; Chen, Chao; Pan, An-jun; Xu, Yang; Liao, Ping-an; Zhang, Su-xia; Gu, Jun-nong

2009-12-01

Red water phenomenon occurred in some communities of a city in China after water source switch in recent days. The origin of this red water problem and mechanism of iron release were investigated in the study. Water quality of local and new water sources was tested and tap water quality in suffered area had been monitored for 3 months since red water occurred. Interior corrosion scales on the pipe which was obtained from the suffered area were analyzed by XRD, SEM, and EDS. Corrosion rates of cast iron under the conditions of two source water were obtained by Annular Reactor. The influence of different source water on iron release was studied by pipe section reactor to simulate the distribution systems. The results indicated that large increase of sulfate concentration by water source shift was regarded as the cause of red water problem. The Larson ratio increased from about 0.4 to 1.7-1.9 and the red water problem happened in the taps of some urban communities just several days after the new water source was applied. The mechanism of iron release was concluded that the stable shell of scales in the pipes had been corrupted by this kind of high-sulfate-concentration source water and it was hard to recover soon spontaneously. The effect of sulfate on iron release of the old cast iron was more significant than its effect on enhancing iron corrosion. The rate of iron release increased with increasing Larson ratio, and the correlation of them was nonlinear on the old cast-iron. The problem remained quite a long time even if the water source re-shifted into the blended one with only small ratio of the new source and the Larson ratio reduced to about 0.6.

Optimizing pneumatic conveying of biomass materials

NASA Astrophysics Data System (ADS)

DiCianni, Matthew Edward Michael

2011-12-01

Biomass is a readily available but underutilized energy resource. One of the main challenges is the inability of biomass feed stocks like corn stover or wood chips to flow freely without intermittent jamming. This research integrated an automated pneumatic conveying system to efficiently transport biomass into a biomass reactor. Material was held in a storage container until an end effector attached to a 3-axis controller engaged the material to flow through pneumatic vacuum in the carrier fluid of air. The material was disengaged from the carrier fluid through centripetal forces induced by a cyclone separator. As the air was pulled out of the cyclone, the biomass drops out the bottom due to gravitational forces and fell into a secondary storage hopper. The second storage container was for testing purposes only, where the actual apparatus would use a vertically oriented lock hopper to feed material into the biomass reactor. In the experimental test apparatus, sensors measured the storage hopper weight (mass-flow rate), pressure drop from the blower, and input power consumption of the motor. Parameters that were adjusted during testing include pipe diameter, material type, and motor speed. Testing indicated that decreasing the motor speed below its maximum still allows for conveyance of the material without blockage forming in the piping. The data shows that the power consumption of the system can be reduced based on the size and weight of the material introduced to the conveying pipe. Also, conveying certain materials proved to be problematic with particular duct diameters. Ultimately, an optimal duct diameter that can perform efficiently for a broad range of materials was chosen for the given system. Through these improvements, the energy return on investment will be improved for biomass feed stocks, which is taking a step in the right direction to secure the nation's energy independence.

Work function and surface stability of tungsten-based thermionic electron emission cathodes

NASA Astrophysics Data System (ADS)

Jacobs, Ryan; Morgan, Dane; Booske, John

2017-11-01

Materials that exhibit a low work function and therefore easily emit electrons into vacuum form the basis of electronic devices used in applications ranging from satellite communications to thermionic energy conversion. W-Ba-O is the canonical materials system that functions as the thermionic electron emitter commercially used in a range of high-power electron devices. However, the work functions, surface stability, and kinetic characteristics of a polycrystalline W emitter surface are still not well understood or characterized. In this study, we examined the work function and surface stability of the eight lowest index surfaces of the W-Ba-O system using density functional theory methods. We found that under the typical thermionic cathode operating conditions of high temperature and low oxygen partial pressure, the most stable surface adsorbates are Ba-O species with compositions in the range of Ba0.125O-Ba0.25O per surface W atom, with O passivating all dangling W bonds and Ba creating work function-lowering surface dipoles. Wulff construction analysis reveals that the presence of O and Ba significantly alters the surface energetics and changes the proportions of surface facets present under equilibrium conditions. Analysis of previously published data on W sintering kinetics suggests that fine W particles in the size range of 100-500 nm may be at or near equilibrium during cathode synthesis and thus may exhibit surface orientation fractions well described by the calculated Wulff construction.

Coupling a Supercritical Carbon Dioxide Brayton Cycle to a Helium-Cooled Reactor.

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

Middleton, Bobby; Pasch, James Jay; Kruizenga, Alan Michael

2016-01-01

This report outlines the thermodynamics of a supercritical carbon dioxide (sCO 2) recompression closed Brayton cycle (RCBC) coupled to a Helium-cooled nuclear reactor. The baseline reactor design for the study is the AREVA High Temperature Gas-Cooled Reactor (HTGR). Using the AREVA HTGR nominal operating parameters, an initial thermodynamic study was performed using Sandia's deterministic RCBC analysis program. Utilizing the output of the RCBC thermodynamic analysis, preliminary values of reactor power and of Helium flow rate through the reactor were calculated in Sandia's HelCO 2 code. Some research regarding materials requirements was then conducted to determine aspects of corrosion related tomore » both Helium and to sCO 2 , as well as some mechanical considerations for pressures and temperatures that will be seen by the piping and other components. This analysis resulted in a list of materials-related research items that need to be conducted in the future. A short assessment of dry heat rejection advantages of sCO 2> Brayton cycles was also included. This assessment lists some items that should be investigated in the future to better understand how sCO 2 Brayton cycles and nuclear can maximally contribute to optimizing the water efficiency of carbon free power generation« less

Convective cooling in a pool-type research reactor

NASA Astrophysics Data System (ADS)

Sipaun, Susan; Usman, Shoaib

2016-01-01

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

Heat transfer in turbulent magneto-fluid-mechanic pipe flow

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

Andelman, M.P.

1975-12-01

The ability to predict heat transfer in Magneto-Fluid-Mechanic flow is of importance in light of the development of MHD generators and the proposed development of thermonuclear reactors. In both cases heat transfer from (or to) a conducting fluid in the presence of a magnetic field plays an important part in the overall economics of the system. A semi-empirical analytical method is given for obtaining heat transfer coefficients in turbulent liquid metal pipe flow in the presence of a magnetic field aligned to the flow. The analysis was based on the Lykoudis turbulent transport model with the influence of a longitudinalmore » magnetic field included. The results are shown to be in agreement with available experimental values. Experimental velocity profiles in mercury for pipe flow in a transverse magnetic field were made at a Reynolds number of 315,000; for Hartmann numbers of 0, 92, 184, 369, and 1198; and at orientations of 0 degrees, 45 degrees, and 90 degrees from the magnetic field. These results provide a basis for the determination of the effect of a transverse magnetic field on turbulent diffusivities.« less

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.

Application of Simulated Reactivity Feedback in Nonnuclear Testing of a Direct-Drive Gas-Cooled Reactor

NASA Technical Reports Server (NTRS)

Bragg-Sitton, S. M.; Webster, K. L.

2007-01-01

Nonnuclear testing can be a valuable tool in the development of an in-space nuclear power or propulsion system. In a nonnuclear test facility, electric heaters are used to simulate heat from nuclear fuel. Standard testing allows one to fully assess thermal, heat transfer, and stress related attributes of a given system but fails to demonstrate the dynamic response that would be present in an integrated, fueled reactor system. The integration of thermal hydraulic hardware tests with simulated neutronic response provides a bridge between electrically heated testing and full nuclear testing. By implementing a neutronic response model to simulate the dynamic response that would be expected in a fueled reactor system, one can better understand system integration issues, characterize integrated system response times and response and response characteristics, and assess potential design improvements with a relatively small fiscal investment. Initial system dynamic response testing was demonstrated on the integrated SAFE 100a heat pipe cooled, electrically heated reactor and heat exchanger hardware. This Technical Memorandum discusses the status of the planned dynamic test methodology for implementation in the direct-drive gas-cooled reactor testing and assesses the additional instrumentation needed to implement high-fidelity dynamic testing.

Effect of Light Water Reactor Water Environments on the Fatigue Life of Reactor Materials

DOE PAGES

Chopra, O. K.; Stevens, G. L.; Tregoning, R.; ...

2017-10-06

The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code) provides rules for the design of Class 1 components of nuclear power plants. Figures I-9.1 through I-9.6 of Appendix I to Section III of the Code specify fatigue design curves for applicable structural materials. However, the Code design curves do not explicitly address the effects of light water reactor (LWR) water environments. Existing fatigue strain-vs.-life (ε-N) laboratory data illustrate potentially significant effects of LWR water environments on the fatigue resistance of pressure vessel and piping steels. Extensive studies have been conducted at Argonne National Laboratory and elsewheremore » since 1990 to investigate the effects of LWR environments on the fatigue life of piping and pressure vessel steels. This article summarizes the results of these studies. Existing fatigue ε-N data were evaluated to identify the various material, environmental, and loading conditions that influence fatigue crack initiation; a methodology for estimating fatigue lives as a function of these parameters was developed. The effects were incorporated into the ASME Code Section III fatigue evaluations in terms of an environmental correction factor, F en, which is defined as the ratio of fatigue life in air at room temperature to the fatigue life in the LWR water environment at reactor operating temperatures. Available fatigue data were used to develop fatigue design curves for carbon and low-alloy steels, austenitic stainless steels, and nickel-chromium-iron (NiCr-Fe) alloys and their weld metals in air at room temperature. A review of the Code Section III fatigue adjustment factors of 2 on strain and 20 on life is also presented and the possible conservatism inherent in the choice of these adjustment factors is evaluated. A brief description of potential effects of neutron irradiation on fatigue crack initiation for these structural materials is also presented.« less

Effect of Light Water Reactor Water Environments on the Fatigue Life of Reactor Materials

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

Chopra, O. K.; Stevens, G. L.; Tregoning, R.

The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code) provides rules for the design of Class 1 components of nuclear power plants. Figures I-9.1 through I-9.6 of Appendix I to Section III of the Code specify fatigue design curves for applicable structural materials. However, the Code design curves do not explicitly address the effects of light water reactor (LWR) water environments. Existing fatigue strain-vs.-life (ε-N) laboratory data illustrate potentially significant effects of LWR water environments on the fatigue resistance of pressure vessel and piping steels. Extensive studies have been conducted at Argonne National Laboratory and elsewheremore » since 1990 to investigate the effects of LWR environments on the fatigue life of piping and pressure vessel steels. This article summarizes the results of these studies. Existing fatigue ε-N data were evaluated to identify the various material, environmental, and loading conditions that influence fatigue crack initiation; a methodology for estimating fatigue lives as a function of these parameters was developed. The effects were incorporated into the ASME Code Section III fatigue evaluations in terms of an environmental correction factor, F en, which is defined as the ratio of fatigue life in air at room temperature to the fatigue life in the LWR water environment at reactor operating temperatures. Available fatigue data were used to develop fatigue design curves for carbon and low-alloy steels, austenitic stainless steels, and nickel-chromium-iron (NiCr-Fe) alloys and their weld metals in air at room temperature. A review of the Code Section III fatigue adjustment factors of 2 on strain and 20 on life is also presented and the possible conservatism inherent in the choice of these adjustment factors is evaluated. A brief description of potential effects of neutron irradiation on fatigue crack initiation for these structural materials is also presented.« less

Status of the Development of Low Cost Radiator for Surface Fission Power - II

NASA Technical Reports Server (NTRS)

Tarau, Calin; Maxwell, Taylor; Anderson, William G.; Wagner, Corey; Wrosch, Matthew; Briggs, Maxwell H.

2016-01-01

NASA Glenn Research Center (GRC) is developing fission power system technology for future Lunar and Martian surface power applications. The systems are envisioned in the 10 to 100kWe range and have an anticipated design life of 8 to 15 years with no maintenance. NASA GRC is currently setting up a 55 kWe non-nuclear system ground test in thermal-vacuum to validate technologies required to transfer reactor heat, convert the heat into electricity, reject waste heat, process the electrical output, and demonstrate overall system performance. The paper reports on the development of the heat pipe radiator to reject the waste heat from the Stirling convertors. Reducing the radiator mass, size, and cost is essential to the success of the program. To meet these goals, Advanced Cooling Technologies, Inc. (ACT) and Vanguard Space Technologies, Inc. (VST) are developing a single facesheet radiator with heat pipes directly bonded to the facesheet. The facesheet material is a graphite fiber reinforced composite (GFRC) and the heat pipes are titanium/water Variable Conductance Heat Pipes (VCHPs). By directly bonding a single facesheet to the heat pipes, several heavy and expensive components can be eliminated from the traditional radiator design such as, POCO"TM" foam saddles, aluminum honeycomb, and a second facesheet. As mentioned in previous papers by the authors, the final design of the waste heat radiator is described as being modular with independent GFRC panels for each heat pipe. The present paper reports on test results for a single radiator module as well as a radiator cluster consisting of eight integral modules. These tests were carried out in both ambient and vacuum conditions. While the vacuum testing of the single radiator module was performed in the ACT's vacuum chamber, the vacuum testing of the eight heat pipe radiator cluster took place in NASA GRC's vacuum chamber to accommodate the larger size of the cluster. The results for both articles show good agreement with the predictions and are presented in the paper.

COMPUTATIONAL MODELING OF CIRCULATING FLUIDIZED BED REACTORS

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

Ibrahim, Essam A

2013-01-09

Details of numerical simulations of two-phase gas-solid turbulent flow in the riser section of Circulating Fluidized Bed Reactor (CFBR) using Computational Fluid Dynamics (CFD) technique are reported. Two CFBR riser configurations are considered and modeled. Each of these two riser models consist of inlet, exit, connecting elbows and a main pipe. Both riser configurations are cylindrical and have the same diameter but differ in their inlet lengths and main pipe height to enable investigation of riser geometrical scaling effects. In addition, two types of solid particles are exploited in the solid phase of the two-phase gas-solid riser flow simulations tomore » study the influence of solid loading ratio on flow patterns. The gaseous phase in the two-phase flow is represented by standard atmospheric air. The CFD-based FLUENT software is employed to obtain steady state and transient solutions for flow modulations in the riser. The physical dimensions, types and numbers of computation meshes, and solution methodology utilized in the present work are stated. Flow parameters, such as static and dynamic pressure, species velocity, and volume fractions are monitored and analyzed. The differences in the computational results between the two models, under steady and transient conditions, are compared, contrasted, and discussed.« less

Chemical vapor deposition techniques and related methods for manufacturing microminiature thermionic converters

DOEpatents

King, Donald B.; Sadwick, Laurence P.; Wernsman, Bernard R.

2002-06-25

Methods of manufacturing microminiature thermionic converters (MTCs) having high energy-conversion efficiencies and variable operating temperatures using MEMS manufacturing techniques including chemical vapor deposition. The MTCs made using the methods of the invention incorporate cathode to anode spacing of about 1 micron or less and use cathode and anode materials having work functions ranging from about 1 eV to about 3 eV. The MTCs also exhibit maximum efficiencies of just under 30%, and thousands of the devices can be fabricated at modest costs.

Thermionic gas switch

DOEpatents

Hatch, George L.; Brummond, William A.; Barrus, Donald M.

1986-01-01

A temperature responsive thermionic gas switch having folded electron emitting surfaces. An ionizable gas is located between the emitter and an interior surface of a collector, coaxial with the emitter. In response to the temperature exceeding a predetermined level, sufficient electrons are derived from the emitter to cause the gas in the gap between the emitter and collector to become ionized, whereby a very large increase in current in the gap occurs. Due to the folded emitter surface area of the switch, increasing the "on/off" current ratio and adjusting the "on" current capacity is accomplished.

Thermionic energy converter investigations

NASA Technical Reports Server (NTRS)

Goodale, D. B.; Lee, C.; Lieb, D.; Oettinger, P. E.

1979-01-01

This paper presents evaluation of a variety of thermionic converter configurations to obtain improved efficiency. A variable-spacing diode using an iridium emitter gave emission properties comparable to platinum, but the power output from a sintered LaB6 collector diode was not consistent with its work function. Reflectivities above 0.5 were measured at thermal energies on oxygenated-cesiated surfaces using a field emission retarding potential gun. Performance of converters with structured electrodes and the characteristics of a pulsed triode were studied as a function of emitter, collector, cesium reservoir, interelectrode spacing, xenon pressure, and pulsing parameters.

Experimental investigation of electron guns for THz microwave vacuum amplifiers

NASA Astrophysics Data System (ADS)

Burtsev, A. A.; Grigor'ev, Yu. A.; Navrotsky, I. A.; Rogovin, V. I.; Sakhadzhi, G. V.; Shumikhin, K. V.

2016-05-01

Single-sheet and multiple beam electron emitters based on thermionic minicathodes for terahertz traveling-wave tubes have been studied. Data are presented for impregnated blade thermionic cathode with dimensions 0.1 × 0.7 mm and a maximum current density of 114 A/cm2 in a pulsed mode. A variant of the five-beam electron gun with 0.25-mm-diameter cylindrical minicathodes in cells of a control grid is proposed that provides a current density of 85.5 A/cm2 at a grid potential of 900-1000 V.

Attractive potential around a thermionically emitting microparticle.

PubMed

Delzanno, G L; Lapenta, G; Rosenberg, M

2004-01-23

We present a simulation study of the charging of a dust grain immersed in a plasma, considering the effect of thermionic electron emission from the grain. It is shown that the orbit motion limited theory is no longer reliable when electron emission becomes large: screening can no longer be treated within the Debye-Huckel approach and an attractive potential well can form, leading to the possibility of attractive forces on other grains with the same polarity. We suggest to perform laboratory experiments where emitting dust grains could be used to create nonconventional dust crystals or macromolecules.

Excited-state thermionic emission in III-antimonides: Low emittance ultrafast photocathodes

NASA Astrophysics Data System (ADS)

Berger, Joel A.; Rickman, B. L.; Li, T.; Nicholls, A. W.; Andreas Schroeder, W.

2012-11-01

The normalized rms transverse emittance of an electron source is shown to be proportional to √m* , where m* is the effective mass of the state from which the electron is emitted, by direct observation of the transverse momentum distribution for excited-state thermionic emission from two III-V semiconductor photocathodes, GaSb and InSb, together with a control experiment employing two-photon emission from gold. Simulations of the experiment using an extended analytical Gaussian model of electron pulse propagation are in close agreement with the data.

Gridded thermionic gun and integral superconducting ballistic bunch compression cavity

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

Schultheiss, Thomas

Electron-Ion colliders such as the Medium energy Electron Ion Collider (MEIC) being developed by JLAB require high current electrons with low energy spread for electron cooling of the collider ring. Accelerator techniques for improving bunch charge, average current, emittance, and energy spread are required for Energy Recovery Linacs (ERLs) and Circulator Rings (CR) for next generation colliders for nuclear physics experiments. Example candidates include thermionic-cathode electron guns with RF accelerating structures. Thermionic cathodes are known to produce high currents and have excellent lifetime. The success of the IR and THz Free-Electron Laser (FEL) designed and installed by Advanced Energy Systemsmore » at the Fritz Haber Institute (FHI) of the Max Planck Society in Berlin [1,2] demonstrates that gridded thermionic cathodes and rf systems be considered for next generation collider technology. In Phase 1 Advanced Energy Systems (AES) developed and analyzed a design concept using a superconducting cavity pair and gridded thermionic cathode. Analysis included Beam Dynamics and thermal analysis to show that a design of this type is feasible. The latest design goals for the MEIC electron cooler were for electron bunches of 420 pC at a frequency of 952.6 MHz with a magnetic field on the cathode of 2kG. This field magnetizes the beam imparting angular momentum that provides for helical motion of the electrons in the cooling solenoid. The helical motion increases the interaction time and improves the cooling efficiency. A coil positioned around the cathode providing 2kG field was developed. Beam dynamics simulations were run to develop the particle dynamics near the cathode and grid. Lloyd Young added capability to Tstep to include space charge effects between two plates and include image charge effects from the grid. He also added new pepper-pot geometry capability to account for honeycomb grids. These additions were used to develop the beam dynamics for this gun. The general design is a modified ballistic compression cavity pair with two independently powered cells [3]. The first is a cathode cell that includes the thermionic cathode and grid to provide for beam bunching. The second is a full cell with independent phasing and field levels designed to minimize energy spread. The primary goal for Phase II is to manufacture a superconducting gun with a thermionic cathode and imbedded coil. The system developed here is applicable to many high current electron accelerators. The analysis and design constraints imposed by the magnetized cathode make the cathode system developed here more complicated and limited than one without the magnetized beam constraints. High power ERLs would benefit by a gun with the capabilities shown here, 400 mA or more of current. ERLs hold great promise for electron cooling experiments, advanced light sources and Free Electron Lasers. This high current electron injector is a technological advance that will place the requirements for an ERL capable of providing quality bunches needed for cooling within the MEIC circulator ring within reach. This injector would have application to future ERLs around the world.« less

Magnetohydrodynamic flow and heat transfer around a heated cylinder of arbitrary conductivity

NASA Astrophysics Data System (ADS)

Tassone, A.; Nobili, M.; Caruso, G.

2017-11-01

The interaction of the liquid metal with the plasma confinement magnetic field constitutes a challenge for the design of fusion reactor blankets, due to the arise of MHD effects: increased pressure drops, heat transfer suppression, etc. To overcome these issues, a dielectric fluid can be employed as coolant for the breeding zone. A typical configuration involves pipes transverse to the liquid metal flow direction. This numerical study is conducted to assess the influence of pipe conductivity on the MHD flow and heat transfer. The CFD code ANSYS CFX was employed for this purpose. The fluid is assumed to be bounded by rectangular walls with non-uniform thickness and subject to a skewed magnetic field with the main component aligned with the cylinder axis. The simulations were restricted to Re = (20; 40) and M = (10; 50). Three different scenarios for the obstacle were considered: perfectly insulating, finite conductivity and perfectly conducting. The electrical conductivity was found to affect the channel pressure penalty due to the obstacle insertion only for M = 10 and just for the two limiting cases. A general increment of the heat transfer with M was found due to the tendency of the magnetic field to equalize the flow rate between the sub-channels individuated by the pipe. The best results were obtained with the insulating pipe, due to the reduced electromagnetic drag. The generation of counter-rotating vortices close to the lateral duct walls was observed for M = 50 and perfectly conducting pipe as a result of the modified currents distribution.

PROCESS FOR PRODUCING URANIUM HALIDES

DOEpatents

Murphree, E.V.

1957-10-29

A process amd associated apparatus for producing UF/sub 4/ from U/sub 3/ O/sub 8/ by a fluidized'' technique are reported. The U/sub 3/O/sub 8/ is first reduced to UO/sub 2/ by reaction with hydrogen, and the lower oxide of uranium is then reacted with gaseous HF to produce UF/sub 4/. In each case the reactant gas is used, alone or in combination with inert gases, to fluidize'' the finely divided reactant solid. The complete setup of the plant equipment including bins, reactor and the associated piping and valving, is described. An auxiliary fluorination reactor allows for the direct production of UF/sub 6/ from UF/sub 4/ and fluorine gas, or if desired, UF/sub 4/ may be collected as the product.

DEMINERALIZER BUILDING, TRA608. CAMERA IS ON RAW WATER TOWER AND ...

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

DEMINERALIZER BUILDING, TRA-608. CAMERA IS ON RAW WATER TOWER AND FACES WEST. STEAM PLANT, TRA-609, AT UPPER EDGE OF VIEW. ABSENCE OF ROOF EXPOSES FIVE-BAY STRUCTURE AND INTERIOR DIVISION OF SPACE. CORRIDOR AT WEST END OF BUILDING WILL SEPARATE LABORATORY AND OFFICE SPACE FROM POTABLE WATER TANKS. ALONG NORTH WALL ARE SPACES FOR CATION AND ANION EXCHANGE UNITS. PENTHOUSE WILL ENCLOSE DEGASSIFIER. TANK AT LEFT (SOUTH) OF BUILDING STORES DEMINERALIZED WATER. NOTE BRINE STORAGE PIT, TRA-631, AT RIGHT OF VIEW, ABOVE PAIR OF CAUSTIC STORAGE TANKS. NOTE TRENCHES FOR BURIED WATER PIPES. INL NEGATIVE NO. 2732. Unknown Photographer, 6/29/1951 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Apparatus for detecting leakage of liquid sodium

DOEpatents

Himeno, Yoshiaki

1978-01-01

An apparatus for detecting the leakage of liquid sodium includes a cable-like sensor adapted to be secured to a wall of piping or other equipment having sodium on the opposite side of the wall, and the sensor includes a core wire electrically connected to the wall through a leak current detector and a power source. An accidental leakage of the liquid sodium causes the corrosion of a metallic layer and an insulative layer of the sensor by products resulted from a reaction of sodium with water or oxygen in the atmospheric air so as to decrease the resistance between the core wire and the wall. Thus, the leakage is detected as an increase in the leaking electrical current. The apparatus is especially adapted for use in detecting the leakage of liquid sodium from sodium-conveying pipes or equipment in a fast breeder reactor.

Measurement of Coolant in a Flat Heat Pipe Using Neutron Radiography

NASA Astrophysics Data System (ADS)

Mizuta, Kei; Saito, Yasushi; Goshima, Takashi; Tsutsui, Toshio

A newly developed flat heat pipe FGHPTM (Morex Kiire Co.) was experimentally investigated by using neutron radiography. The test sample of the FGHP heat spreader was 65 × 65 × 2 mm3 composed of several etched copper plates and pure water was used as the coolant. Neutron radiography was performed at the E-2 port of the Kyoto University Research Reactor (KUR). The coolant distributions in the wick area of the FGHP and its heat transfer characteristics were measured at heating conditions. Experimental results show that the coolant distributions depend slightly on its installation posture and that the liquid thickness in the wick region remains constant with increasing heat input to the FGHP. In addition, it is found that the wick surface does not dry out even in the vertical posture at present experimental conditions.

Treatment of iron(II)-rich acid mine water with limestone and oxygen.

PubMed

Mohajane, G B; Maree, J P; Panichev, N

2014-01-01

The main components of acid mine water are free acid, sulphate, and Fe²⁺. Limestone is the most cost-effective alkali that can be used for neutralization. The purpose of this investigation was to identify conditions where Fe²⁺ is removed with limestone and simultaneously oxidized with oxygen to Fe³⁺, in a polyvinyl chloride pipe under pressure. Gypsum scaling is prevented by passing rubber balls through the pipe of the so-called Oxygen-Pipe-Neutralization (OPeN) process pilot plant. Two synthetic waters were treated: (A) acid mine water containing 123 mg L⁻¹ Fe²⁺ representing gold mine water, and (B) acid mine water containing 6,032 mg L⁻¹ Fe²⁺ representing coal mine water. Batch studies were carried out in a pipe reactor and showed that the rate of Fe²⁺ oxidation depended on the Fe²⁺ concentration, oxygen pressure, amount of recycled sludge, limestone dosage and the mixing rate. Continuous studies in an OPeN process pilot plant resulted in 100% removal of total acidity from synthetic coal mine water and a 98% removal from synthetic gold mine water. Fe²⁺ was removed completely as precipitated Fe(OH)₃ from both synthetic coal and gold mine water at around pH 7 at 200 and 100 kPa oxygen pressure, respectively.

Hybrid thermionic-photovoltaic converter

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

Datas, A.

2016-04-04

A conceptual device for the direct conversion of heat into electricity is presented. This concept hybridizes thermionic (TI) and thermophotovoltaic (TPV) energy conversion in a single thermionic-photovoltaic (TIPV) solid-state device. This device transforms into electricity both the electron and photon fluxes emitted by an incandescent surface. This letter presents an idealized analysis of this device in order to determine its theoretical potential. According to this analysis, the key advantage of this converter, with respect to either TPV or TI, is the higher power density in an extended temperature range. For low temperatures, TIPV performs like TPV due to the negligiblemore » electron flux. On the contrary, for high temperatures, TIPV performs like TI due to the great enhancement of the electron flux, which overshadows the photon flux contribution. At the intermediate temperatures, ∼1650 K in the case of this particular study, I show that the power density potential of TIPV converter is twice as great as that of TPV and TI. The greatest impact concerns applications in which the temperature varies in a relatively wide range, for which averaged power density enhancement above 500% is attainable.« less

Thermionic Properties of Carbon Based Nanomaterials Produced by Microhollow Cathode PECVD

NASA Technical Reports Server (NTRS)

Haase, John R.; Wolinksy, Jason J.; Bailey, Paul S.; George, Jeffrey A.; Go, David B.

2015-01-01

Thermionic emission is the process in which materials at sufficiently high temperature spontaneously emit electrons. This process occurs when electrons in a material gain sufficient thermal energy from heating to overcome the material's potential barrier, referred to as the work function. For most bulk materials very high temperatures (greater than 1500 K) are needed to produce appreciable emission. Carbon-based nanomaterials have shown significant promise as emission materials because of their low work functions, nanoscale geometry, and negative electron affinity. One method of producing these materials is through the process known as microhollow cathode PECVD. In a microhollow cathode plasma, high energy electrons oscillate at very high energies through the Pendel effect. These high energy electrons create numerous radical species and the technique has been shown to be an effective method of growing carbon based nanomaterials. In this work, we explore the thermionic emission properties of carbon based nanomaterials produced by microhollow cathode PECVD under a variety of synthesis conditions. Initial studies demonstrate measureable current at low temperatures (approximately 800 K) and work functions (approximately 3.3 eV) for these materials.

From the first nuclear power plant to fourth-generation nuclear power installations [on the 60th anniversary of the World's First nuclear power plant

NASA Astrophysics Data System (ADS)

Rachkov, V. I.; Kalyakin, S. G.; Kukharchuk, O. F.; Orlov, Yu. I.; Sorokin, A. P.

2014-05-01

Successful commissioning in the 1954 of the World's First nuclear power plant constructed at the Institute for Physics and Power Engineering (IPPE) in Obninsk signaled a turn from military programs to peaceful utilization of atomic energy. Up to the decommissioning of this plant, the AM reactor served as one of the main reactor bases on which neutron-physical investigations and investigations in solid state physics were carried out, fuel rods and electricity generating channels were tested, and isotope products were bred. The plant served as a center for training Soviet and foreign specialists on nuclear power plants, the personnel of the Lenin nuclear-powered icebreaker, and others. The IPPE development history is linked with the names of I.V. Kurchatov, A.I. Leipunskii, D.I. Blokhintsev, A.P. Aleksandrov, and E.P. Slavskii. More than 120 projects of various nuclear power installations were developed under the scientific leadership of the IPPE for submarine, terrestrial, and space applications, including two water-cooled power units at the Beloyarsk NPP in Ural, the Bilibino nuclear cogeneration station in Chukotka, crawler-mounted transportable TES-3 power station, the BN-350 reactor in Kazakhstan, and the BN-600 power unit at the Beloyarsk NPP. Owing to efforts taken on implementing the program for developing fast-neutron reactors, Russia occupied leading positions around the world in this field. All this time, IPPE specialists worked on elaborating the principles of energy supertechnologies of the 21st century. New large experimental installations have been put in operation, including the nuclear-laser setup B, the EGP-15 accelerator, the large physical setup BFS, the high-pressure setup SVD-2; scientific, engineering, and technological schools have been established in the field of high- and intermediate-energy nuclear physics, electrostatic accelerators of multicharge ions, plasma processes in thermionic converters and nuclear-pumped lasers, physics of compact nuclear reactors and radiation protection, thermal physics, physical chemistry and technology of liquid metal coolants, and physics of radiation-induced defects, and radiation materials science. The activity of the institute is aimed at solving matters concerned with technological development of large-scale nuclear power engineering on the basis of a closed nuclear fuel cycle with the use of fast-neutron reactors (referred to henceforth as fast reactors), development of innovative nuclear and conventional technologies, and extension of their application fields.

Technical Information on the Carbonation of the EBR-II Reactor, Summary Report Part 2: Application to EBR-II Primary Sodium System and Related Systems

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

Steven R. Sherman; Collin J. Knight

Residual sodium is defined as sodium metal that remains behind in pipes, vessels, and tanks after the bulk sodium metal has been melted and drained from such components. The residual sodium has the same chemical properties as bulk sodium, and differs from bulk sodium only in the thickness of the sodium deposit. Typically, sodium is considered residual when the thickness of the deposit is less than 5-6 cm. This residual sodium must be removed or deactivated when a pipe, vessel, system, or entire reactor is permanently taken out of service, in order to make the component or system safer and/ormore » to comply with decontamination and decomissioning regulations. As an alternative to the established residual sodium deactivation techniques (steam-and-nitrogen, wet vapor nitrogen, etc.), a technique involving the use of moisture and carbon dioxide has been developed. With this technique, sodium metal is converted into sodium bicarbonate by reacting it with humid carbon dioxide. Hydrogen is emitted as a by-product. This technique was first developed in the laboratory by exposing sodium samples to humidifed carbon dioxide under controlled conditions, and then demonstrated on a larger scale by treating residual sodium within the Experimental Breeder Reactor II (EBR-II) secondary cooling system, followed by the primary cooling system, respectively. The EBR-II facility is located at the Idaho National Laboratory (INL) in southeastern Idaho, USA. This report is Part 2 of a two-part report. This second report provides a supplement to the first report and describes the application of the humdidified carbon dioxide technique ("carbonation") to the EBR-II primary tank, primary cover gas systems, and the intermediate heat exchanger. Future treatment plans are also provided.« less

Convective cooling in a pool-type research reactor

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

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

2016-01-22

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

Conceptual Design for CLIC Gun Pulser

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

Tang, Tao

The Compact Linear Collider (CLIC) is a proposed future electron-positron collider, designed to perform collisions at energies from 0.5 to 5 TeV, with a nominal design optimized for 3 TeV (Dannheim, 2012). The Drive Beam Accelerator consists of a thermionic DC gun, bunching section and an accelerating section. The thermionic gun needs deliver a long (~143us) pulse of current into the buncher. A pulser is needed to drive grid of the gun to generate a stable current output. This report explores the requirements of the gun pulser and potential solutions to regulate grid current.

Temperature dependence of photon-enhanced thermionic emission from GaAs surface with nonequilibrium Cs overlayers

NASA Astrophysics Data System (ADS)

Zhuravlev, A. G.; Alperovich, V. L.

2017-02-01

The temperature influence on the Cs/GaAs surface electronic properties, which determine the photon-enhanced thermionic emission (PETE), is studied. It was found that heating to moderate temperatures of about 100 °С leads to substantial changes in the magnitude and shape of Cs coverage dependences of photoemission current and surface band bending, along with the changes of relaxation kinetics after Cs deposition. A spectral proof of the PETE process is obtained under thermal cycling of the Cs/GaAs surface with 0.45 monolayer (ML) of Cs.

Carbide fuel pin and capsule design for irradiations at thermionic temperatures

NASA Technical Reports Server (NTRS)

Siegel, B. L.; Slaby, J. G.; Mattson, W. F.; Dilanni, D. C.

1973-01-01

The design of a capsule assembly to evaluate tungsten-emitter - carbide-fuel combinations for thermionic fuel elements is presented. An inpile fuel pin evaluation program concerned with clad temperture, neutron spectrum, carbide fuel composition, fuel geometry,fuel density, and clad thickness is discussed. The capsule design was a compromise involving considerations between heat transfer, instrumentation, materials compatibility, and test location. Heat-transfer calculations were instrumental in determining the method of support of the fuel pin to minimize axial temperature variations. The capsule design was easily fabricable and utilized existing state-of-the-art experience from previous programs.

Hollow-Cathode Source Generates Plasma

NASA Technical Reports Server (NTRS)

Deininger, W. D.; Aston, G.; Pless, L. C.

1989-01-01

Device generates argon, krypton, or xenon plasma via thermionic emission and electrical discharge within hollow cathode and ejects plasma into surrounding vacuum. Goes from cold start up to full operation in less than 5 s after initial application of power. Exposed to moist air between operations without significant degradation of starting and running characteristics. Plasma generated by electrical discharge in cathode barrel sustained and aided by thermionic emission from emitter tube. Emitter tube does not depend on rare-earth oxides, making it vulnerable to contamination by exposure to atmosphere. Device modified for use as source of plasma in laboratory experiments or industrial processes.

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

NASA Astrophysics Data System (ADS)

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

2017-03-01

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

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.

Pressure polymerization of polyester

DOEpatents

Maurer, Charles J.; Shaw, Gordon; Smith, Vicky S.; Buelow, Steven J.; Tumas, William; Contreras, Veronica; Martinez, Ronald J.

2000-08-29

A process is disclosed for the preparation of a polyester polymer or polyester copolymer under superatmospheric pressure conditions in a pipe or tubular reaction under turbannular flow conditions. Reaction material having a glycol equivalents to carboxylic acid equivalents mole ratio of from 1.0:1 to 1.2:1, together with a superatmospheric dense gaseous medium are fed co-currently to the reactor. Dicarboxylic acid and/or diol raw materials may be injected into any of the reaction zones in the process during operation to achieve the overall desired mole ratio balance. The process operates at temperatures of from about 220.degree. C. to about 320.degree. C., with turbannular flow achieved before the polymer product and gas exit the reactor process. The pressure in the reaction zones can be in the range from 15 psia to 2500 psia. A polymer product having a DP of a greater than 40, more preferably at least about 70, is achieved by the transfer of water from the reacting material polymer melt to the gaseous medium in the reactor.

Helium heater design for the helium direct cycle component test facility. [for gas-cooled nuclear reactor power plant

NASA Technical Reports Server (NTRS)

Larson, V. R.; Gunn, S. V.; Lee, J. C.

1975-01-01

The paper describes a helium heater to be used to conduct non-nuclear demonstration tests of the complete power conversion loop for a direct-cycle gas-cooled nuclear reactor power plant. Requirements for the heater include: heating the helium to a 1500 F temperature, operating at a 1000 psia helium pressure, providing a thermal response capability and helium volume similar to that of the nuclear reactor, and a total heater system helium pressure drop of not more than 15 psi. The unique compact heater system design proposed consists of 18 heater modules; air preheaters, compressors, and compressor drive systems; an integral control system; piping; and auxiliary equipment. The heater modules incorporate the dual-concentric-tube 'Variflux' heat exchanger design which provides a controlled heat flux along the entire length of the tube element. The heater design as proposed will meet all system requirements. The heater uses pressurized combustion (50 psia) to provide intensive heat transfer, and to minimize furnace volume and heat storage mass.

Solar photochemical process engineering for production of fuels and chemicals

NASA Technical Reports Server (NTRS)

Biddle, J. R.; Peterson, D. B.; Fujita, T.

1984-01-01

The engineering costs and performance of a nominal 25,000 scmd (883,000 scfd) photochemical plant to produce dihydrogen from water were studied. Two systems were considered, one based on flat-plate collector/reactors and the other on linear parabolic troughs. Engineering subsystems were specified including the collector/reactor, support hardware, field transport piping, gas compression equipment, and balance-of-plant (BOP) items. Overall plant efficiencies of 10.3 and 11.6% are estimated for the flat-plate and trough systems, respectively, based on assumed solar photochemical efficiencies of 12.9 and 14.6%. Because of the opposing effects of concentration ratio and operating temperature on efficiency, it was concluded that reactor cooling would be necessary with the trough system. Both active and passive cooling methods were considered. Capital costs and energy costs, for both concentrating and non-concentrating systems, were determined and their sensitivity to efficiency and economic parameters were analyzed. The overall plant efficiency is the single most important factor in determining the cost of the fuel.

Solar photochemical process engineering for production of fuels and chemicals

NASA Technical Reports Server (NTRS)

Biddle, J. R.; Peterson, D. B.; Fujita, T.

1985-01-01

The engineering costs and performance of a nominal 25,000 scmd (883,000 scfd) photochemical plant to produce dihydrogen from water were studied. Two systems were considered, one based on flat-plate collector/reactors and the other on linear parabolic troughs. Engineering subsystems were specified including the collector/reactor, support hardware, field transport piping, gas compression equipment, and balance-of-plant (BOP) items. Overall plant efficiencies of 10.3 and 11.6 percent are estimated for the flat-plate and trough systems, respectively, based on assumed solar photochemical efficiencies of 12.9 and 14.6 percent. Because of the opposing effects of concentration ratio and operating temperature on efficiency, it was concluded that reactor cooling would be necessary with the trough system. Both active and passive cooling methods were considered. Capital costs and energy costs, for both concentrating and non-concentrating systems, were determined and their sensitivity to efficiency and economic parameters were analyzed. The overall plant efficiency is the single most important factor in determining the cost of the fuel.

A novel ultrasonic NDE for shrink fit welded structures using interface waves.

PubMed

Lee, Jaesun; Park, Junpil; Cho, Younho

2016-05-01

Reactor vessel inspection is a critical part of safety maintenance in a nuclear power plant. The inspection of shrink fit welded structures in a reactor nozzle can be a challenging task due to the complicated geometry. Nozzle inspection using pseudo interface waves allows us to inspect the nozzle from outside of the nuclear reactor. In this study, layered concentric pipes were manufactured with perfect shrink fit conditions using stainless steel 316. The displacement distributions were calculated with boundary conditions for a shrink fit welded structure. A multi-transducer guided wave phased array system was employed to monitor the welding quality of the nozzle end at a distance from a fixed position. The complicated geometry of a shrink fit welded structure can be overcome by using the pseudo interface waves in identifying the location and size of defects. The experimental results demonstrate the feasibility of detecting weld delamination and defects. Copyright © 2016 Elsevier B.V. All rights reserved.

A Small Fission Power System with Stirling Power Conversion for NASA Science Missions

NASA Technical Reports Server (NTRS)

Mason, Lee; Carmichael, Chad

2011-01-01

In early 2010, a joint National Aeronautics and Space Administration (NASA) and Department of Energy (DOE) study team developed a concept for a 1 kWe Fission Power System with a 15-year design life that could be available for a 2020 launch to support future NASA science missions. The baseline concept included a solid block uranium-molybdenum reactor core with embedded heat pipes and distributed thermoelectric converters directly coupled to aluminum radiator fins. A short follow-on study was conducted at NASA Glenn Research Center (GRC) to evaluate an alternative power conversion approach. The GRC study considered the use of free-piston Stirling power conversion as a substitution to the thermoelectric converters. The resulting concept enables a power increase to 3 kWe with the same reactor design and scalability to 10 kW without changing the reactor technology. This paper presents the configuration layout, system performance, mass summary, and heat transfer analysis resulting from the study.

Calculation of Local Stress and Fatigue Resistance due to Thermal Stratification on Pressurized Surge Line Pipe

NASA Astrophysics Data System (ADS)

Bandriyana, B.; Utaja

2010-06-01

Thermal stratification introduces thermal shock effect which results in local stress and fatique problems that must be considered in the design of nuclear power plant components. Local stress and fatique calculation were performed on the Pressurize Surge Line piping system of the Pressurize Water Reactor of the Nuclear Power Plant. Analysis was done on the operating temperature between 177 to 343° C and the operating pressure of 16 MPa (160 Bar). The stagnant and transient condition with two kinds of stratification model has been evaluated by the two dimensional finite elements method using the ANSYS program. Evaluation of fatigue resistance is developed based on the maximum local stress using the ASME standard Code formula. Maximum stress of 427 MPa occurred at the upper side of the top half of hot fluid pipe stratification model in the transient case condition. The evaluation of the fatigue resistance is performed on 500 operating cycles in the life time of 40 years and giving the usage value of 0,64 which met to the design requirement for class 1 of nuclear component. The out surge transient were the most significant case in the localized effects due to thermal stratification.

Lightweight Carbon-Carbon High-Temperature Space Radiator

NASA Technical Reports Server (NTRS)

Miller, W.O.; Shih, Wei

2008-01-01

A document summarizes the development of a carbon-carbon composite radiator for dissipating waste heat from a spacecraft nuclear reactor. The radiator is to be bonded to metal heat pipes and to operate in conjunction with them at a temperature approximately between 500 and 1,000 K. A goal of this development is to reduce the average areal mass density of a radiator to about 2 kg/m(exp 2) from the current value of approximately 10 kg/m(exp 2) characteristic of spacecraft radiators made largely of metals. Accomplishments thus far include: (1) bonding of metal tubes to carbon-carbon material by a carbonization process that includes heating to a temperature of 620 C; (2) verification of the thermal and mechanical integrity of the bonds through pressure-cycling, axial-shear, and bending tests; and (3) construction and testing of two prototype heat-pipe/carbon-carbon-radiator units having different radiator areas, numbers of heat pipes, and areal mass densities. On the basis of the results achieved thus far, it is estimated that optimization of design could yield an areal mass density of 2.2 kg/m (exp 2) close to the goal of 2 kg/m(exp 2).

Dynamic Modeling and Control of Nuclear Reactors Coupled to Closed-Loop Brayton Cycle Systems using SIMULINK{sup TM}

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

Wright, Steven A.; Sanchez, Travis

2005-02-06

The operation of space reactors for both in-space and planetary operations will require unprecedented levels of autonomy and control. Development of these autonomous control systems will require dynamic system models, effective control methodologies, and autonomous control logic. This paper briefly describes the results of reactor, power-conversion, and control models that are implemented in SIMULINK{sup TM} (Simulink, 2004). SIMULINK{sup TM} is a development environment packaged with MatLab{sup TM} (MatLab, 2004) that allows the creation of dynamic state flow models. Simulation modules for liquid metal, gas cooled reactors, and electrically heated systems have been developed, as have modules for dynamic power-conversion componentsmore » such as, ducting, heat exchangers, turbines, compressors, permanent magnet alternators, and load resistors. Various control modules for the reactor and the power-conversion shaft speed have also been developed and simulated. The modules are compiled into libraries and can be easily connected in different ways to explore the operational space of a number of potential reactor, power-conversion system configurations, and control approaches. The modularity and variability of these SIMULINK{sup TM} models provides a way to simulate a variety of complete power generation systems. To date, both Liquid Metal Reactors (LMR), Gas Cooled Reactors (GCR), and electric heaters that are coupled to gas-dynamics systems and thermoelectric systems have been simulated and are used to understand the behavior of these systems. Current efforts are focused on improving the fidelity of the existing SIMULINK{sup TM} modules, extending them to include isotopic heaters, heat pipes, Stirling engines, and on developing state flow logic to provide intelligent autonomy. The simulation code is called RPC-SIM (Reactor Power and Control-Simulator)« less

Mineralization of TNT, RDX, and By-Products in an Anaerobic Granular Activated Carbon-Fluidized Bed Reactor

DTIC Science & Technology

2003-04-01

pipe fitters and safety personnel who constructed and installed the system. CERL personnel include Don Cropek, Pat Kemme, Neil Adrian and Clint Arnett...al. (2000), demonstrated the sequential conversion of the nitro-groups to amino-groups in TNT degradation, and Adrian and Sutherland , 1998...dinitrotoluene extended the research to TNT, RDX and pinkwater (Adrian and Sutherland , 1998, VanderLoop et al., 1998 and Hwang et al., 2000). Initial pilot

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

NASA Astrophysics Data System (ADS)

Van Dyke, Melissa; Martin, James

2005-02-01

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

Enhanced Thermionic Emission and Low 1/f Noise in Exfoliated Graphene/GaN Schottky Barrier Diode.

PubMed

Kumar, Ashutosh; Kashid, Ranjit; Ghosh, Arindam; Kumar, Vikram; Singh, Rajendra

2016-03-01

Temperature-dependent electrical transport characteristics of exfoliated graphene/GaN Schottky diodes are investigated and compared with conventional Ni/GaN Schottky diodes. The ideality factor of graphene/GaN and Ni/GaN diodes are measured to be 1.33 and 1.51, respectively, which is suggestive of comparatively higher thermionic emission current in graphene/GaN diode. The barrier height values for graphene/GaN diode obtained using thermionic emission model and Richardson plots are found to be 0.60 and 0.72 eV, respectively, which are higher than predicted barrier height ∼0.40 eV as per the Schottky-Mott model. The higher barrier height is attributed to hole doping of graphene due to graphene-Au interaction which shifts the Fermi level in graphene by ∼0.3 eV. The magnitude of flicker noise of graphene/GaN Schottky diode increases up to 175 K followed by its decrease at higher temperatures. This indicates that diffusion currents and barrier inhomogeneities dominate the electronic transport at lower and higher temperatures, respectively. The exfoliated graphene/GaN diode is found to have lower level of barrier inhomogeneities than conventional Ni/GaN diode, as well as earlier reported graphene/GaN diode fabricated using chemical vapor deposited graphene. The lesser barrier inhomogeneities in graphene/GaN diode results in lower flicker noise by 2 orders of magnitude as compared to Ni/GaN diode. Enhanced thermionic emission current, lower level of inhomogeneities, and reduced flicker noise suggests that graphene-GaN Schottky diodes may have the underlying trend for replacing metal-GaN Schottky diodes.

Comparison of Fission Product Yields and Their Impact

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

S. Harrison

2006-02-01

This memorandum describes the Naval Reactors Prime Contractor Team (NRPCT) Space Nuclear Power Program (SNPP) interest in determining the expected fission product yields from a Prometheus-type reactor and assessing the impact of these species on materials found in the fuel element and balance of plant. Theoretical yield calculations using ORIGEN-S and RACER computer models are included in graphical and tabular form in Attachment, with focus on the desired fast neutron spectrum data. The known fission product interaction concerns are the corrosive attack of iron- and nickel-based alloys by volatile fission products, such as cesium, tellurium, and iodine, and the radiologicalmore » transmutation of krypton-85 in the coolant to rubidium-85, a potentially corrosive agent to the coolant system metal piping.« less

Poisoning of Heat Pipes

NASA Technical Reports Server (NTRS)

Gillies, Donald; Lehoczky, Sandor; Palosz, Witold; Carpenter, Paul; Salvail, Pat

2007-01-01

Thermal management is critical to space exploration efforts. In particular, efficient transfer and control of heat flow is essential when operating high energy sources such as nuclear reactors. Thermal energy must be transferred to various energy conversion devices, and to radiators for safe and efficient rejection of excess thermal energy. Applications for space power demand exceptionally long periods of time with equipment that is accessible for limited maintenance only. Equally critical is the hostile and alien environment which includes high radiation from the reactor and from space (galactic) radiation. In space or lunar applications high vacuum is an issue, while in Martian operations the systems will encounter a CO2 atmosphere. The effect of contact at high temperature with local soil (regolith) in surface operations on the moon or other terrestrial bodies (Mars, asteroids) must be considered.

Effect of water addition in a microwave assisted thermal cracking of biomass tar gasification

NASA Astrophysics Data System (ADS)

Warsita, A.; Surya, I.

2018-02-01

Producer gas from biomass gasification is plagued by the presence of tar which causes pipe blockages. Thermal and catalytic treatments in a microwave reactor have been shown to be effective methods for removing tar from producer gas. A question arises as to the possibility of enhancing the removal mechanism by adding water into the reactor. Thermal treatment with a various amount of water was added at temperatures in the range of 800-1200°C. The tar removal efficiency obtained 96.32% at the optimum temperature of 1200°C at the water to tar ratio (W/T) of 0.3. This study shows that the removal of tar by microwave irradiation with water addition is a significant and effective method in tar cracking.

Nitrile bioconversion by Microbacterium imperiale CBS 498-74 resting cells in batch and ultrafiltration membrane bioreactors.

PubMed

Cantarella, M; Cantarella, L; Gallifuoco, A; Spera, A

2006-03-01

The biohydration of acrylonitrile, propionitrile and benzonitrile catalysed by the NHase activity contained in resting cells of Microbacterium imperiale CBS 498-74 was operated at 5, 10 and 20 degrees C in laboratory-scale batch and membrane bioreactors. The bioreactions were conducted in buffered medium (50 mM Na(2)HPO(4)/NaH(2)PO(4), pH 7.0) in the presence of distilled water or tap-water, to simulate a possible end-pipe biotreatment process. The integral bioreactor performances were studied with a cell loading (dry cell weight; DCW) varying from 0.1 mg(DCW) per reactor to 16 mg(DCW) per reactor, in order to realize near 100% bioconversion of acrylonitrile, propionitrile and benzonitrile without consistent loss of NHase activity.

Design for On-Sun Evaluation of Evaporator Receivers

NASA Technical Reports Server (NTRS)

Jaworske, Donald A.; Colozza, Anthony; Sechkar, Edward A.

2011-01-01

A heat pipe designed for operation as a solar power receiver should be optimized to accept the solar energy flux and transfer this heat into a reactor. Optical properties of the surface, thermal conductance of the receiver wall, contact resistance of the heat pipe wick, and other heat pipe wick properties ultimately define the maximum amount of power that can be extracted from the concentrated sunlight impinging on the evaporator surface. Modeling of solar power receivers utilizing optical and physical properties provides guidance to their design. On-sun testing is another important means of gathering information on performance. A test rig is being designed and built to conduct on-sun testing. The test rig is incorporating a composite strip mirror concentrator developed as part of a Small Business Innovative Research effort and delivered to NASA Glenn Research Center. In the strip concentrator numerous, lightweight composite parabolic strips of simple curvature were combined to form an array 1.5 m x 1.5 m in size. The line focus of each strip is superimposed in a central area simulating a point of focus. A test stand is currently being developed to hold the parabolic strip concentrator, track the sun, and turn the beam downward towards the ground. The hardware is intended to be sufficiently versatile to accommodate on-sun testing of several receiver concepts, including those incorporating heat pipe evaporators. Characterization devices are also being developed to evaluate the effectiveness of the solar concentrator, including a receiver designed to conduct calorimetry. This paper describes the design and the characterization devices of the on-sun test rig, and the prospect of coupling the concentrated sunlight to a heat pipe solar power receiver developed as part of another Small Business Innovative Research effort.

Framework for Structural Online Health Monitoring of Aging and Degradation of Secondary Systems due to some Aspects of Erosion

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

Gribok, Andrei; Patnaik, Sobhan; Williams, Christian

This report describes the current state of research related to critical aspects of erosion and selected aspects of degradation of secondary components in nuclear power plants. The report also proposes a framework for online health monitoring of aging and degradation of secondary components. The framework consists of an integrated multi-sensor modality system which can be used to monitor different piping configurations under different degradation conditions. The report analyses the currently known degradation mechanisms and available predictive models. Based on this analysis, the structural health monitoring framework is proposed. The Light Water Reactor Sustainability Program began to evaluate technologies that couldmore » be used to perform online monitoring of piping and other secondary system structural components in commercial NPPs. These online monitoring systems have the potential to identify when a more detailed inspection is needed using real-time measurements, rather than at a pre-determined inspection interval. This transition to condition-based, risk informed automated maintenance will contribute to a significant reduction of operations and maintenance costs that account for the majority of nuclear power generation costs. There is unanimous agreement between industry experts and academic researchers that identifying and prioritizing inspection locations in secondary piping systems (for example, in raw water piping or diesel piping) would eliminate many excessive in-service inspections. The proposed structural health monitoring framework takes aim at answering this challenge by combining long-range guided wave technologies with other monitoring techniques, which can significantly increase the inspection length and pinpoint the locations that degraded the most. More widely, the report suggests research efforts aimed at developing, validating, and deploying online corrosion monitoring techniques for complex geometries, which are pervasive in NPPs.« less

Low-Cost Radiator for Fission Power Thermal Control

NASA Technical Reports Server (NTRS)

Maxwell, Taylor; Tarau, Calin; Anderson, William; Hartenstine, John; Stern, Theodore; Walmsley, Nicholas; Briggs, Maxwell

2014-01-01

NASA Glenn Research Center (GRC) is developing fission power system technology for future Lunar surface power applications. The systems are envisioned in the 10 to 100kW(sub e) range and have an anticipated design life of 8 to 15 years with no maintenance. NASA GRC is currently setting up a 55 kW(sub e) non-nuclear system ground test in thermal-vacuum to validate technologies required to transfer reactor heat, convert the heat into electricity, reject waste heat, process the electrical output, and demonstrate overall system performance. Reducing the radiator mass, size, and cost is essential to the success of the program. To meet these goals, Advanced Cooling Technologies, Inc. (ACT) and Vanguard Space Technologies, Inc. (VST) are developing a single facesheet radiator with heat pipes directly bonded to the facesheet. The facesheet material is a graphite fiber reinforced composite (GFRC) and the heat pipes are titanium/water. By directly bonding a single facesheet to the heat pipes, several heavy and expensive components can be eliminated from the traditional radiator design such as, POC(TradeMark) foam saddles, aluminum honeycomb, and a second facesheet. A two-heat pipe radiator prototype, based on the single facesheet direct-bond concept, was fabricated and tested to verify the ability of the direct-bond joint to withstand coefficient of thermal expansion (CTE) induced stresses during thermal cycling. The thermal gradients along the bonds were measured before and after thermal cycle tests to determine if the performance degraded. Overall, the results indicated that the initial uniformity of the adhesive was poor along one of the heat pipes. However, both direct bond joints showed no measureable amount of degradation after being thermally cycled at both moderate and aggressive conditions.

Comparison of evolving photovoltaic and nuclear power systems for earth orbital applications

NASA Technical Reports Server (NTRS)

Rockey, D. E.; Jones, R. M.; Schulman, I.

1982-01-01

Photovoltaic and fission reactor orbital power systems are compared in terms of the end-to-end system power-to-mass ratios. Three PV systems are examined, i.e., a solid substrate with a cell array and a NiCd battery, a modified SEP array and an NiH2 battery, and a 62-micron Si cell array and a fuel cell. All arrays were modeled to be 13.5% efficient and to produce 25 kW dc. The SP-100 reactor consists of the heat source, radiation shield, heat pipes to transfer thermal energy from the reactor to thermoelectric elements, and a waste heat radiator. Consideration is given to system applications in orbits ranging from LEO to GEO, and to mission durations of 1, 5, and 10 yr. PV systems are concluded to be flight-proven, useful out of radiation belts, and best for low to moderate power levels. Limitations exist for operations where atmospheric drag may become a factor and due to the size of a large PV power supply. Space nuclear reactors will continue under development and uses at high power levels and in low altitude orbits are foreseen.

Double Retort System for Materials Compatibility Testing

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

V. Munne; EV Carelli

2006-02-23

With Naval Reactors (NR) approval of the Naval Reactors Prime Contractor Team (NRPCT) recommendation to develop a gas cooled reactor directly coupled to a Brayton power conversion system as the Space Nuclear Power Plant (SNPP) for Project Prometheus (References a and b) there was a need to investigate compatibility between the various materials to be used throughout the SNPP. Of particular interest was the transport of interstitial impurities from the nickel-base superalloys, which were leading candidates for most of the piping and turbine components to the refractory metal alloys planned for use in the reactor core. This kind of contaminationmore » has the potential to affect the lifetime of the core materials. This letter provides technical information regarding the assembly and operation of a double retort materials compatibility testing system and initial experimental results. The use of a double retort system to test materials compatibility through the transfer of impurities from a source to a sink material is described here. The system has independent temperature control for both materials and is far less complex than closed loops. The system is described in detail and the results of three experiments are presented.« less

Removal properties of diesel exhaust particles by a dielectric barrier discharge reactor.

PubMed

Suzuki, Ken-ichiro; Takeuchi, Naomi; Madokoro, Kazuhiko; Fushimi, Chihiro; Yao, Shuiliang; Fujioka, Yuichi; Nihei, Yoshimasa

2008-02-01

The removal properties of diesel exhaust particles (DEP) were investigated using an engine exhaust particle size spectrometer (EEPS), field emission-type scanning electron microscopy (FE-SEM) and time-of-flight secondary ion mass spectrometry (TOF-SIMS). DEP were treated using a dielectric barrier discharge (DBD) reactor installed in the tail pipe of a diesel engine, and a model DBD reactor fed with DEP in the mixture of N(2) and O(2). When changing the experimental parameters of both the plasma conditions and the engine load conditions, we obtained characteristic information of DEP treated with plasma discharges from the particle diameter and the composition. In evaluating the model DBD reactor, it became clear that there were two types of plasma processes (reactions with active oxygen species to yield CO(2) and reactions with active nitrogen species to yield nitrogen containing compounds). Moreover, from the result of a TOF-SIMS analysis, the characteristic secondary ions, such as C(2)H(6)N(+), C(4)H(12)N(+), and C(10)H(20)N(2)(+), were strongly detected from the DEP surfaces during the plasma discharges. This indicates that the nitrogen contained hydrocarbons were generated by plasma reactions.

Insoluble coatings for Stirling engine heat pipe condenser surfaces

NASA Astrophysics Data System (ADS)

Dussinger, Peter M.

1993-09-01

The work done by Thermacore, Inc., Lancaster, Pennsylvania, for the Phase 1, 1992 SBIR National Aeronautics and Space Administration Contract, 'Insoluble Coatings for Stirling Engine Heat Pipe Condenser Surfaces' is described. The work was performed between January 1992 and July 1992. Stirling heat engines are being developed for electrical power generation use on manned and unmanned earth orbital and planetary missions. Dish Stirling solar systems and nuclear reactor Stirling systems are two of the most promising applications of the Stirling engine electrical power generation technology. The sources of thermal energy used to drive the Stirling engine typically are non-uniform in temperature and heat flux. Liquid metal heat pipe receivers are used as thermal transformers and isothermalizers to deliver the thermal energy at a uniform high temperature to the heat input section of the Stirling engine. The use of a heat pipe receiver greatly enhances system efficiency and potential life span. One issue that is raised during the design phase of heat pipe receivers is the potential solubility corrosion of the Stirling engine heat input section by the liquid metal working fluid. This Phase 1 effort initiated a program to evaluate and demonstrate coatings, applied to nickel based Stirling engine heater head materials, that are practically 'insoluble' in sodium, potassium, and NaK. This program initiated a study of nickel aluminide as a coating and developed and demonstrated a heat pipe test vehicle that can be used to test candidate materials and coatings. Nickel 200 and nickel aluminide coated Nickel 200 were tested for 1000 hours at 800 C at a condensation heat flux of 25 W/sq cm. Subsequent analyses of the samples showed no visible sign of solubility corrosion of either coated or uncoated samples. The analysis technique, photomicrographs at 200X, has a resolution of better than 2.5 microns (.0001 in). The results indicate that the heat pipe environment is not directly comparable to liquid metal pumped loop data, that nickel aluminide is still a leading candidate for solubility corrosion protection, and that longer duration tests are required to reach a definitive conclusion whether coatings are required at all. Should further testing be required, the test vehicle and analytical tools were developed.

Insoluble coatings for Stirling engine heat pipe condenser surfaces

NASA Technical Reports Server (NTRS)

Dussinger, Peter M.

1993-01-01

The work done by Thermacore, Inc., Lancaster, Pennsylvania, for the Phase 1, 1992 SBIR National Aeronautics and Space Administration Contract, 'Insoluble Coatings for Stirling Engine Heat Pipe Condenser Surfaces' is described. The work was performed between January 1992 and July 1992. Stirling heat engines are being developed for electrical power generation use on manned and unmanned earth orbital and planetary missions. Dish Stirling solar systems and nuclear reactor Stirling systems are two of the most promising applications of the Stirling engine electrical power generation technology. The sources of thermal energy used to drive the Stirling engine typically are non-uniform in temperature and heat flux. Liquid metal heat pipe receivers are used as thermal transformers and isothermalizers to deliver the thermal energy at a uniform high temperature to the heat input section of the Stirling engine. The use of a heat pipe receiver greatly enhances system efficiency and potential life span. One issue that is raised during the design phase of heat pipe receivers is the potential solubility corrosion of the Stirling engine heat input section by the liquid metal working fluid. This Phase 1 effort initiated a program to evaluate and demonstrate coatings, applied to nickel based Stirling engine heater head materials, that are practically 'insoluble' in sodium, potassium, and NaK. This program initiated a study of nickel aluminide as a coating and developed and demonstrated a heat pipe test vehicle that can be used to test candidate materials and coatings. Nickel 200 and nickel aluminide coated Nickel 200 were tested for 1000 hours at 800 C at a condensation heat flux of 25 W/sq cm. Subsequent analyses of the samples showed no visible sign of solubility corrosion of either coated or uncoated samples. The analysis technique, photomicrographs at 200X, has a resolution of better than 2.5 microns (.0001 in). The results indicate that the heat pipe environment is not directly comparable to liquid metal pumped loop data, that nickel aluminide is still a leading candidate for solubility corrosion protection, and that longer duration tests are required to reach a definitive conclusion whether coatings are required at all. Should further testing be required, the test vehicle and analytical tools were developed.

Effect of Combined Loading Due to Bending and Internal Pressure on Pipe Flaw Evaluation Criteria

NASA Astrophysics Data System (ADS)

Miura, Naoki; Sakai, Shinsuke

Considering a rule for the rationalization of maintenance of Light Water Reactor piping, reliable flaw evaluation criteria are essential for determining how a detected flaw will be detrimental to continuous plant operation. Ductile fracture is one of the dominant failure modes that must be considered for carbon steel piping and can be analyzed by elastic-plastic fracture mechanics. Some analytical efforts have provided various flaw evaluation criteria using load correction factors, such as the Z-factors in the JSME codes on fitness-for-service for nuclear power plants and the section XI of the ASME boiler and pressure vessel code. The present Z-factors were conventionally determined, taking conservativity and simplicity into account; however, the effect of internal pressure, which is an important factor under actual plant conditions, was not adequately considered. Recently, a J-estimation scheme, LBB.ENGC for the ductile fracture analysis of circumferentially through-wall-cracked pipes subjected to combined loading was developed for more accurate prediction under more realistic conditions. This method explicitly incorporates the contributions of both bending and tension due to internal pressure by means of a scheme that is compatible with an arbitrary combined-loading history. In this study, the effect of internal pressure on the flaw evaluation criteria was investigated using the new J-estimation scheme. The Z-factor obtained in this study was compared with the presently used Z-factors, and the predictability of the current flaw evaluation criteria was quantitatively evaluated in consideration of the internal pressure.

Thermionic performance of a cesium diminiode with relatively impure 110-tungsten electrodes

NASA Technical Reports Server (NTRS)

Smith, A. L.; Manista, E. J.; Morris, J. F.

1974-01-01

Thermionic performance data from a miniature plane cesium diode (diminiode) with 110-tungsten electrodes are presented. The diminiode has a guard-ringed collector and a spacing of 0.23 mm. The data were obtained by using a computerized acquisition system. The diode was tested at increments between 1700 and 1900 K for the emitter, 694 and 1101 K for the collector, and 519 and 650 K for the reservoir. A maximum power density of 4.5 W/sq cm was obtained at an emitter temperature of 1900 K. This relatively low output probably results from high carbon and sodium impurities in the electrode materials.

ANALYTICAL MODELING OF ELECTRON BACK-BOMBARDMENT INDUCED CURRENT INCREASE IN UN-GATED THERMIONIC CATHODE RF GUNS

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

Edelen, J. P.; Sun, Y.; Harris, J. R.

In this paper we derive analytical expressions for the output current of an un-gated thermionic cathode RF gun in the presence of back-bombardment heating. We provide a brief overview of back-bombardment theory and discuss comparisons between the analytical back-bombardment predictions and simulation models. We then derive an expression for the output current as a function of the RF repetition rate and discuss relationships between back-bombardment, fieldenhancement, and output current. We discuss in detail the relevant approximations and then provide predictions about how the output current should vary as a function of repetition rate for some given system configurations.

The NASA thermionic-conversion (TEC-ART) program

NASA Technical Reports Server (NTRS)

Morris, J. F.

1977-01-01

The current emphasis is on out-of-core thermionic conversion (TEC). The additional degrees of freedom offer new potentialities, but high-temperature material effects determine the level and lifetime of TEC performance: New electrodes not only raise power outputs but also maintain them regardless of emitter-vapor deposition on collectors. In addition, effective electrodes serve compatibly with hot-shell alloys. Space TEC withstands external and internal high-temperature vaporization problems, and terrestrial TEC tolerates hot corrosive atmospheres outside and near-vacuum inside. Finally, reduction of losses between converter electrodes is essential even though rather demanding geometries appear to be required for some modes of enhanced operation.

Device for providing high-intensity ion or electron beam

DOEpatents

McClanahan, Edwin D.; Moss, Ronald W.

1977-01-01

A thin film of a low-thermionic-work-function material is maintained on the cathode of a device for producing a high-current, low-pressure gas discharge by means of sputter deposition from an auxiliary electrode. The auxiliary electrode includes a surface with a low-work-function material, such as thorium, uranium, plutonium or one of the rare earth elements, facing the cathode but at a disposition and electrical potential so as to extract ions from the gas discharge and sputter the low-work-function material onto the cathode. By continuously replenishing the cathode film, high thermionic emissions and ion plasmas can be realized and maintained over extended operating periods.

A program-management plan with critical-path definition for Combustion Augmentation with Thermionic Energy Conversion (CATEC)

NASA Technical Reports Server (NTRS)

Morris, J. F.; Merrill, O. S.; Reddy, H. K.

1981-01-01

Thermionic energy conversion (TEC) is discussed. In recent TEC-topping analyses, overall plant efficiency (OPE) and cost of electricity (COE) improved slightly with current capabilities and substantially with fully matured technologies. Enhanced credibility derives from proven hot-corrosion protection for TEC by silicon-carbide clads in fossil fuel combustion products. Combustion augmentation with TEC (CATEC) affords minimal cost and plant perturbation, but with smaller OPE and COE improvements than more conventional topping applications. Risk minimization as well as comparative simplicity and convenience, favor CATEC for early market penetration. A program-management plan is proposed. Inputs, characteristics, outputs and capabilities are discussed.

High-efficiency, low-temperature cesium diodes with lanthanum-hexaboride electrodes

NASA Technical Reports Server (NTRS)

Morris, J. F.

1974-01-01

Lanthanum hexaboride electrodes in 1700 K cesium diodes may triple power outputs compared with those demonstrated for nuclear thermionic space applications. Still greater relative gains seem possible for emitters below 1700 K. Further improvements in cesium diode performance should result from the lower collector temperatures allowed for earth and low power space duties. Decreased temperatures will lessen thermal transport losses that attend thermionic conversion mechanisms. Such advantages will add to those from collector Carnot and electrode effects. If plasma ignition difficulties impede diode temperature reductions, recycling small fractions of the output power could provide ionization. So high efficiency, low temperature cesium diodes with lanthanum hexaboride electrodes appear feasible.

Fundamental studies on a heat driven lamp

NASA Technical Reports Server (NTRS)

Lawless, J. L.

1985-01-01

A detailed theoretical study of a heat-driven lamp has been performed. This lamp uses a plasma produced in a thermionic diode. The light is produced by the resonance transition of cesium. An important result of this study is that up to 30% of the input heat is predicted to be converted to light in this device. This is a major improvement over ordinary thermionic energy converters in which only approx. 1% is converted to resonance radiation. Efficiencies and optimum inter-electrode spacings have been found as a function of cathode temperature and the radiative escape factor. The theory developed explains the operating limits of the device.

A program-management plan with critical-path definition for Combustion Augmentation with Thermionic Energy Conversion (CATEC)

NASA Astrophysics Data System (ADS)

Morris, J. F.; Merrill, O. S.; Reddy, H. K.

Thermionic energy conversion (TEC) is discussed. In recent TEC-topping analyses, overall plant efficiency (OPE) and cost of electricity (COE) improved slightly with current capabilities and substantially with fully matured technologies. Enhanced credibility derives from proven hot-corrosion protection for TEC by silicon-carbide clads in fossil fuel combustion products. Combustion augmentation with TEC (CATEC) affords minimal cost and plant perturbation, but with smaller OPE and COE improvements than more conventional topping applications. Risk minimization as well as comparative simplicity and convenience, favor CATEC for early market penetration. A program-management plan is proposed. Inputs, characteristics, outputs and capabilities are discussed.

Measurement and Modeling of Blocking Contacts for Cadmium Telluride Gamma Ray Detectors

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

Beck, Patrick R.

2010-01-07

Gamma ray detectors are important in national security applications, medicine, and astronomy. Semiconductor materials with high density and atomic number, such as Cadmium Telluride (CdTe), offer a small device footprint, but their performance is limited by noise at room temperature; however, improved device design can decrease detector noise by reducing leakage current. This thesis characterizes and models two unique Schottky devices: one with an argon ion sputter etch before Schottky contact deposition and one without. Analysis of current versus voltage characteristics shows that thermionic emission alone does not describe these devices. This analysis points to reverse bias generation current ormore » leakage through an inhomogeneous barrier. Modeling the devices in reverse bias with thermionic field emission and a leaky Schottky barrier yields good agreement with measurements. Also numerical modeling with a finite-element physics-based simulator suggests that reverse bias current is a combination of thermionic emission and generation. This thesis proposes further experiments to determine the correct model for reverse bias conduction. Understanding conduction mechanisms in these devices will help develop more reproducible contacts, reduce leakage current, and ultimately improve detector performance.« less

Neutron tomography of axially symmetric objects using 14 MeV neutrons from a portable neutron generator

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

Andersson, P., E-mail: peter.andersson@physics.uu.se; Andersson-Sunden, E.; Sjöstrand, H.

2014-08-01

In nuclear boiling water reactor cores, the distribution of water and steam (void) is essential for both safety and efficiency reasons. In order to enhance predictive capabilities, void distribution assessment is performed in two-phase test-loops under reactor-relevant conditions. This article proposes the novel technique of fast-neutron tomography using a portable deuterium-tritium neutron generator to determine the time-averaged void distribution in these loops. Fast neutrons have the advantage of high transmission through the metallic structures and pipes typically concealing a thermal-hydraulic test loop, while still being fairly sensitive to the water/void content. However, commercially available fast-neutron generators also have the disadvantagemore » of a relatively low yield and fast-neutron detection also suffers from relatively low detection efficiency. Fortunately, some loops are axially symmetric, a property which can be exploited to reduce the amount of data needed for tomographic measurement, thus limiting the interrogation time needed. In this article, three axially symmetric test objects depicting a thermal-hydraulic test loop have been examined; steel pipes with outer diameter 24 mm, thickness 1.5 mm, and with three different distributions of the plastic material POM inside the pipes. Data recorded with the FANTOM fast-neutron tomography instrument have been used to perform tomographic reconstructions to assess their radial material distribution. Here, a dedicated tomographic algorithm that exploits the symmetry of these objects has been applied, which is described in the paper. Results are demonstrated in 20 rixel (radial pixel) reconstructions of the interior constitution and 2D visualization of the pipe interior is demonstrated. The local POM attenuation coefficients in the rixels were measured with errors (RMS) of 0.025, 0.020, and 0.022 cm{sup −1}, solid POM attenuation coefficient. The accuracy and precision is high enough to provide a useful indication on the flow mode, and a visualization of the radial material distribution can be obtained. A benefit of this system is its potential to be mounted at any axial height of a two-phase test section without requirements for pre-fabricated entrances or windows. This could mean a significant increase in flexibility of the void distribution assessment capability at many existing two-phase test loops.« less

Neutron tomography of axially symmetric objects using 14 MeV neutrons from a portable neutron generator.

PubMed

Andersson, P; Andersson-Sunden, E; Sjöstrand, H; Jacobsson-Svärd, S

2014-08-01

In nuclear boiling water reactor cores, the distribution of water and steam (void) is essential for both safety and efficiency reasons. In order to enhance predictive capabilities, void distribution assessment is performed in two-phase test-loops under reactor-relevant conditions. This article proposes the novel technique of fast-neutron tomography using a portable deuterium-tritium neutron generator to determine the time-averaged void distribution in these loops. Fast neutrons have the advantage of high transmission through the metallic structures and pipes typically concealing a thermal-hydraulic test loop, while still being fairly sensitive to the water/void content. However, commercially available fast-neutron generators also have the disadvantage of a relatively low yield and fast-neutron detection also suffers from relatively low detection efficiency. Fortunately, some loops are axially symmetric, a property which can be exploited to reduce the amount of data needed for tomographic measurement, thus limiting the interrogation time needed. In this article, three axially symmetric test objects depicting a thermal-hydraulic test loop have been examined; steel pipes with outer diameter 24 mm, thickness 1.5 mm, and with three different distributions of the plastic material POM inside the pipes. Data recorded with the FANTOM fast-neutron tomography instrument have been used to perform tomographic reconstructions to assess their radial material distribution. Here, a dedicated tomographic algorithm that exploits the symmetry of these objects has been applied, which is described in the paper. Results are demonstrated in 20 rixel (radial pixel) reconstructions of the interior constitution and 2D visualization of the pipe interior is demonstrated. The local POM attenuation coefficients in the rixels were measured with errors (RMS) of 0.025, 0.020, and 0.022 cm(-1), solid POM attenuation coefficient. The accuracy and precision is high enough to provide a useful indication on the flow mode, and a visualization of the radial material distribution can be obtained. A benefit of this system is its potential to be mounted at any axial height of a two-phase test section without requirements for pre-fabricated entrances or windows. This could mean a significant increase in flexibility of the void distribution assessment capability at many existing two-phase test loops.

TEM Pump With External Heat Source And Sink

NASA Technical Reports Server (NTRS)

Nesmith, Bill J.

1991-01-01

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

Proceedings of the AFOSR Special Conference on Prime-Power for High Energy Space Systems, Norfolk, Virginia, 22-25 February 1982. Volume I.

DTIC Science & Technology

1982-02-25

source both liquid and solid fuel combustion devices have been successfully demonstrated during various development programs . Nuclear reactor heat...U02 fuel in the core . Improving the heat pipe model to correlate more closely with the experimental data is a major concern in the development of...ORGANIZATION NAME AND ADDRESS 10. PROGRAM ELEMENT, PROJECT. TASK Research & Development Associates (RDA) AREA &WKNT AE (X Rosslyn, VA 22209 61102F 2301

Hydrogen Assisted Cracking and Corrosion of Some Highly Corrosion Resistant Alloys

DTIC Science & Technology

1990-01-01

Stainless Steel", June 1985, and "On the Roles of Corrosion Products in Local Cell Processes", January 1986. Research on the latter has occurred in the...concern. In closed systems. howevter, such as nuclear reactor cooling pipes. acid container systems, fuel cells, and so on. the production of ti, gas and...mernhra lie is also imiportant. fihe stirf.ice should he flat. m-e1I-polished and free of filims. (Whde or other corrosion product film-. :Are easil% formed

Structural Integrity of Water Reactor Pressure Boundary Components.

DTIC Science & Technology

1981-02-20

environment, and load waveform parameters . A theory of the influence of dissolved oxygen content on the fatigue crack growth results in simulated PWR ...simulated PWR coolant is - (Continues ) DD IJN7 1473 EDITION OF I NOV S ..OSL- -C 2 S/ 0102-LF-014-6601 S1ECURITY CLASSI1FICATION OF THIS PAGE (When...not seem to influence the data, which was produced for a load ratio of 0.2 and a simulated PWR coolant environment. Test results for A106 Grade C piping

Synthesis of Struvite using a Vertical Canted Reactor with Continuous Laminar Flow Process

NASA Astrophysics Data System (ADS)

Sutiyono, S.; Edahwati, L.; Muryanto, S.; Jamari, J.; Bayuseno, A. P.

2018-01-01

Struvite is a white crystalline that is chemically known as magnesium ammonium phosphorus hexahydrate (MgNH4PO4·6H2O). It can easily dissolve in acidic conditions and slightly soluble in neutral and alkaline conditions. In industry, struvite forms as a scale deposit on a pipe with hot flow fluid. However, struvite can be used as fertilizer because of its phosphate content. A vertical canted reactor is a promising technology for recovering phosphate levels in wastewater through struvite crystallization. The study was carried out with the vertical canted reactor by mixing an equimolar stock solution of MgCl2, NH4OH, and H3PO4 in 1: 1: 1 ratio. The crystallization process worked with the flow rate of three stock solution entering the reactor in the range of 16-38 ml/min, the temperature in the reactor is worked on 20°, 30°, and 40°C, while the incoming air rate is kept constant at 0.25 liters/min. Moreover, pH was maintained at a constant value of 9. The struvite crystallization process run until the steady state was reached. Then, the result of crystal precipitates was filtered and dried at standard temperature room for 48 hours. After that, struvite crystals were stored for the subsequent analysis by Scanning Electron Microscope (SEM) and XRD (X-Ray Diffraction) method. The use of canted reactor provided the high pure struvite with a prismatic crystal morphology.

Phenomenology and modeling of particulate corrosion product behavior in Hanford N Reactor primary coolant

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

Bechtold, D.B.

1983-12-31

The levels and composition of filterable corrosion products in the Hanford N Reactor Primary Loop are measurable by filtration. The suspended crud level has ranged from 0.0005 ppM to 6.482 ppM with a median 0.050 ppM. The composition approximates magnetite. The particle size distribution has been found in 31 cases to be uniformly a log normal distribution with a count median ranging from 1.10 to 2.31 microns with a median of 1.81 microns, and the geometric standard deviation ranging from 1.60 to 2.34 with a median of 1.84. An auto-correcting inline turbidimeter was found to respond to linearly to suspendedmore » crud levels over a range 0.05 to at least 6.5 ppM by direct comparison with filter sample weights. Cause of crud bursts in the primary loop were found to be power decreases. The crud transients associated with a reactor power drop, several reactor shutdowns, and several reactor startups could be modeled consistently with each other using a simple stirred-tank, first order exchange model of particulate between makeup, coolant, letdown, and loosely adherent crud on pipe walls. Over 3/10 of the average steady running particulate crud level could be accounted for by magnetically filterable particulate in the makeup feed. A simulation model of particulate transport has been coded in FORTRAN.« less

Re-weldability tests of irradiated 316L(N) stainless steel using laser welding technique

NASA Astrophysics Data System (ADS)

Yamada, Hirokazu; Kawamura, Hiroshi; Tsuchiya, Kunihiko; Kalinin, George; Kohno, Wataru; Morishima, Yasuo

2002-12-01

SS316L(N)-IG is the candidate material for the in-vessel and ex-vessel components of fusion reactors such as ITER (International Thermonuclear Experimental Reactor). This paper describes a study on re-weldability of un-irradiated and/or irradiated SS316L(N)-IG and the effect of helium generation on the mechanical properties of the weld joint. The laser welding process is used for re-welding of the water cooling branch pipeline repairs. It is clarified that re-welding of SS316L(N)-IG irradiated up to about 0.2 dpa (3.3 appm He) can be carried out without a serious deterioration of tensile properties due to helium accumulation. Therefore, repair of the ITER blanket cooling pipes can be performed by the laser welding process.

The nuclear battery

NASA Astrophysics Data System (ADS)

Kozier, K. S.; Rosinger, H. E.

The evolution and present status of an Atomic Energy of Canada Limited program to develop a small, solid-state, passively cooled reactor power supply known as the Nuclear Battery is reviewed. Key technical features of the Nuclear Battery reactor core include a heat-pipe primary heat transport system, graphite neutron moderator, low-enriched uranium TRISO coated-particle fuel and the use of burnable poisons for long-term reactivity control. An external secondary heat transport system extracts useful heat energy, which may be converted into electricity in an organic Rankine cycle engine or used to produce high-pressure steam. The present reference design is capable of producing about 2400 kW(t) (about 600 kW(e) net) for 15 full-power years. Technical and safety features are described along with recent progress in component hardware development programs and market assessment work.

Ab initio investigation of the surface properties of dispenser B-type and scandate thermionic emission cathodes

NASA Astrophysics Data System (ADS)

Vlahos, Vasilios; Lee, Yueh-Lin; Booske, John H.; Morgan, Dane; Turek, Ladislav; Kirshner, Mark; Kowalczyk, Richard; Wilsen, Craig

2009-05-01

Scandate cathodes (BaxScyOz on W) are important thermionic electron emission materials whose emission mechanism remains unclear. Ab initio modeling is used to investigate the surface properties of both scandate and traditional B-type (Ba-O on W) cathodes. We demonstrate that the Ba-O dipole surface structure believed to be present in active B-type cathodes is not thermodynamically stable, suggesting that a nonequilibrium steady state dominates the active cathode's surface structure. We identify a stable, low work function BaxScyOz surface structure, which may be responsible for some scandate cathode properties and demonstrate that multicomponent surface coatings can lower cathode work functions.

Why do aged fluorescent tubes flicker?

NASA Astrophysics Data System (ADS)

Plihon, Nicolas; Ferrand, Jérémy; Guyomar, Tristan; Museur, Flavien; Taberlet, Nicolas

2017-11-01

Our everyday experience of aged and defective fluorescent tubes or bulbs informs us that they may flicker and emit a clicking sound while struggling to light up. In this article, the physical mechanisms controlling the initial illumination of a functioning fluorescent tube are investigated using a simple and affordable experimental setup. Thermionic emission from the electrodes of the tube controls the startup of fluorescent tubes. The origin of the faulty startup of aged fluorescent tubes is discussed and flickering regimes using functional tubes are artificially produced using a dedicated setup that decreases electron emission by the thermionic effect in a controlled manner. The physical parameters controlling the occurrence of flickering light are discussed, and their temporal statistics are reported.

Beam dynamics studies of a 30 MeV RF linac for neutron production

NASA Astrophysics Data System (ADS)

Nayak, B.; Krishnagopal, S.; Acharya, S.

2018-02-01

Design of a 30 MeV, 10 Amp RF linac as neutron source has been carried out by means of ASTRA simulation code. Here we discuss details of design simulations for three different cases i.e Thermionic , DC and RF photocathode guns and compare them as injectors to a 30 MeV RF linac for n-ToF production. A detailed study on choice of input parameters of the beam from point of view of transmission efficiency and beam quality at the output have been described. We found that thermionic gun isn't suitable for this application. Both DC and RF photocathode gun can be used. RF photocathode gun would be of better performance.

Advanced thermionic energy conversion

NASA Technical Reports Server (NTRS)

1979-01-01

Developments towards space and terrestrial applications of thermionic energy conversion are presented. Significant accomplishments for the three month period include: (1) devised a blade-type distributed lead design with many advantages compared to the stud-type distributed lead; (2) completed design of Marchuk tube test apparatus; (3) concluded, based on current understanding, that residual hydrogen should not contribute to a negative space charge barrier at the collector; (4) modified THX design program to include series-coupled designs as well as inductively-coupled designs; (5) initiated work on the heat transfer technology, THX test module, output power transfer system, heat transfer system, and conceptual plant design tasks; and (6) reached 2200 hours of operation in JPL-5 cylindrical converter envelope test.

Single-size thermometric measurements on a size distribution of neutral fullerenes.

PubMed

Cauchy, C; Bakker, J M; Huismans, Y; Rouzée, A; Redlich, B; van der Meer, A F G; Bordas, C; Vrakking, M J J; Lépine, F

2013-05-10

We present measurements of the velocity distribution of electrons emitted from mass-selected neutral fullerenes, performed at the intracavity free electron laser FELICE. We make use of mass-specific vibrational resonances in the infrared domain to selectively heat up one out of a distribution of several fullerene species. Efficient energy redistribution leads to decay via thermionic emission. Time-resolved electron kinetic energy distributions measured give information on the decay rate of the selected fullerene. This method is generally applicable to all neutral species that exhibit thermionic emission and provides a unique tool to study the stability of mass-selected neutral clusters and molecules that are only available as part of a size distribution.

Fabrication, Densification and Thermionic Emission Property of Lanthanum Hexaboride

NASA Astrophysics Data System (ADS)

Yu, Yiping; Wang, Song; Li, Wei; Chen, Hongmei; Chen, Zhaohui

2018-03-01

An effective way to improve sintering densification of LaB6 was proposed and confirmed experimentally. Firstly, LaB6 nanopowders with a cube-like shape of 94.7 nm were fabricated by molten salt synthesis route at 800 °C for 1 h. Then, LaB6 bulk material of 98% density was prepared by hot pressing sintering of as-synthesized LaB6 nanopowders under 1800 °C/50 MPa/30 min. The acquired LaB6 bulk material had a work function of 2.87 eV and exhibited an excellent thermionic emission property. The saturation emission current density at 1500 and 1600 °C reached 37.4 and 44.3 A/cm2, respectively.

Photon enhanced thermionic emission

DOEpatents

Schwede, Jared; Melosh, Nicholas; Shen, Zhixun

2014-10-07

Photon Enhanced Thermionic Emission (PETE) is exploited to provide improved efficiency for radiant energy conversion. A hot (greater than 200.degree. C.) semiconductor cathode is illuminated such that it emits electrons. Because the cathode is hot, significantly more electrons are emitted than would be emitted from a room temperature (or colder) cathode under the same illumination conditions. As a result of this increased electron emission, the energy conversion efficiency can be significantly increased relative to a conventional photovoltaic device. In PETE, the cathode electrons can be (and typically are) thermalized with respect to the cathode. As a result, PETE does not rely on emission of non-thermalized electrons, and is significantly easier to implement than hot-carrier emission approaches.

Investigation of miniaturized radioisotope thermionic power generation for general use

NASA Astrophysics Data System (ADS)

Duzik, Adam J.; Choi, Sang H.

2016-04-01

Radioisotope thermoelectric generators (RTGs) running off the radioisotope Pu238 are the current standard in deep space probe power supplies. While reliable, these generators are very inefficient, operating at only ~7% efficiency. As an alternative, more efficient radioisotope thermionic emission generators (RTIGs) are being explored. Like RTGs, current RTIGs concepts use exotic materials for the emitter, limiting applicability to space and other niche applications. The high demand for long-lasting mobile power sources would be satisfied if RTIGs could be produced inexpensively. This work focuses on exposing several common materials, such as Al, stainless steel, W, Si, and Cu, to elevated temperatures under vacuum to determine the efficiency of each material as inexpensive replacements for thermoelectric materials.

Lunar electric power systems utilizing the SP-100 reactor coupled to dynamic conversion systems

NASA Technical Reports Server (NTRS)

Harty, Richard B.; Durand, Richard E.

1993-01-01

An integration study was performed by Rocketdyne under contract to NASA-LeRC. The study was concerned with coupling an SP-0100 reactor to either a Brayton or Stirling power conversion system. The application was for a surface power system to supply power requirements to a lunar base. A power level of 550 kWe was selected based on the NASA Space Exploration Initiative 90-day study. Reliability studies were initially performed to determine optimum power conversion redundancy. This study resulted in selecting three operating engines and one stand-by unit. Integration design studies indicated that either the Brayton or Stirling power conversion systems could be integrated with the PS-100 reactor. The Stirling system had an integration advantage because of smaller piping size and fewer components. The Stirling engine, however, is more complex and heavier than the Brayton rotating unit, which tends to off-set the Stirling integration advantage. From a performance consideration, the Brayton had a 9 percent mass advantage, and the Stirling had a 50 percent radiator advantage.

Where does the removal of H₂S from biogas occur in microaerobic reactors?

PubMed

Ramos, I; Peña, M; Fdz-Polanco, M

2014-08-01

In order to maximise the efficiency of biogas desulphurisation and reduce the oxygen cost during microaerobic digestion, it is essential to know how the process occurs. For this purpose, a reactor with a total volume of 266 L, treating 10 L/d of sewage sludge, was operated with 25.0 L and without headspace. Under anaerobic conditions, the H2S concentration in the biogas varied between 0.21 and 0.38%v/v. Next, O2 was supplied from the bottom of the reactor. At 0.25-0.30 NLO₂/Lfed, the biogas was entirely desulphurised, and its O₂ content remained below 1.03%v/v, when the digester had 25.0 L of gas space. However, with almost no headspace, the H2S content in the biogas fluctuated from 0.08 to 0.21%v/v, while the average O2 concentration was 1.66%v/v. The removed H2S accumulated in the outlet pipe of the biogas in the form of S(0) due to the insufficient headspace. Copyright © 2014 Elsevier Ltd. All rights reserved.

A power propulsion system based on a second-generation thermionic NPS of the ``Topaz'' type

NASA Astrophysics Data System (ADS)

Gryaznov, Georgi M.; Zhabotinski, Eugene E.; Andreev, Pavel V.; Zaritski, Gennadie a.; Koroteev, Anatoly S.; Martishin, Viktor M.; Akimov, Vladimir N.; Ponomarev-Stepnoi, Nikolai N.; Usov, Veniamin A.; Britt, Edward J.

1992-01-01

The paper considers the concept of power propulsion systems-universal space platforms (USPs) on the basis of second-generation thermionic nuclear power system (NPSs) and stationary plasma electric thrusters (SPETs). The composition and the principles of layout of such a system, based on a thermionic NPS with a continuous power of up to 30 kWe allowing power augmentation by a factor of 2-2.5 as long as during a year, as well as SPETs with a specific impulse of at least 20 km/s and a propulsion efficiency of 0.6-0.7 are discussed. The layouts and the basic parameters are presented for a power propulsion system ensuring cargo transportation from an initial radiation-safe 800 km high orbit into a geostationary one using the ``Zenit'' and ``Proton'' launch systems for injection into an initial orbit. It is shown that the mass of mission-oriented equipment in the geostationary orbit in the cases under consideration ranges from 2500 to 5500 kg on condition that the flight time is not longer than a year. The power propulsion system can be applied to autonomous power supply of various spacecraft including remote power delivery. It can be also used for deep space exploration.

Radiation Shielding Design and Orientation Considerations for a 1 kWe Heat Pipe Cooled Reactor Utilized to Bore Through the Ice Caps of Mars

NASA Astrophysics Data System (ADS)

Fensin, Michael L.; Elliott, John O.; Lipinski, Ronald J.; Poston, David I.

2006-01-01

The goal in designing any space power system is to develop a system able to meet the mission requirements for success while minimizing the overall costs. The mission requirements for the this study was to develop a reactor (with Stirling engine power conversion) and shielding configuration able to fit, along with all the other necessary science equipment, in a Cryobot 3 m high with ~0.5 m diameter hull, produce 1 kWe for 5yrs, and not adversely affect the mission science by keeping the total integrated dose to the science equipment below 150 krad. Since in most space power missions the overall system mass dictates the mission cost, the shielding designs in this study incorporated Martian water extracted at the startup site in order to minimize the tungsten and LiH mass loading at launch. Different reliability and mass minimization concerns led to three design configuration evolutions. With the help of implementing Martian water and configuring the reactor as far from the science equipment as possible, the needed tungsten and LiH shield mass was minimized. This study further characterizes the startup dose and the necessary mission requirements in order to ensure integrity of the surface equipment during reactor startup phase.

Correlating Fast Fluence to dpa in Atypical Locations

NASA Astrophysics Data System (ADS)

Drury, Thomas H.

2016-02-01

Damage to a nuclear reactor's materials by high-energy neutrons causes changes in the ductility and fracture toughness of the materials. The reactor vessel and its associated piping's ability to withstand stress without brittle fracture are paramount to safety. Theoretically, the material damage is directly related to the displacements per atom (dpa) via the residual defects from induced displacements. However in practice, the material damage is based on a correlation to the high-energy (E > 1.0 MeV) neutron fluence. While the correlated approach is applicable when the material in question has experienced the same neutron spectrum as test specimens which were the basis of the correlation, this approach is not generically acceptable. Using Monte Carlo and discrete ordinates transport codes, the energy dependent neutron flux is determined throughout the reactor structures and the reactor vessel. Results from the models provide the dpa response in addition to the high-energy neutron flux. Ratios of dpa to fast fluence are calculated throughout the models. The comparisons show a constant ratio in the areas of historical concern and thus the validity of the correlated approach to these areas. In regions above and below the fuel however, the flux spectrum has changed significantly. The correlated relationship of material damage to fluence is not valid in these regions without adjustment. An adjustment mechanism is proposed.

Report on FY15 alloy 617 code rules development

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

Sham, Sam; Jetter, Robert I; Hollinger, Greg

2015-09-01

Due to its strength at very high temperatures, up to 950°C (1742°F), Alloy 617 is the reference construction material for structural components that operate at or near the outlet temperature of the very high temperature gas-cooled reactors. However, the current rules in the ASME Section III, Division 5 Subsection HB, Subpart B for the evaluation of strain limits and creep-fatigue damage using simplified methods based on elastic analysis have been deemed inappropriate for Alloy 617 at temperatures above 650°C (1200°F) (Corum and Brass, Proceedings of ASME 1991 Pressure Vessels and Piping Conference, PVP-Vol. 215, p.147, ASME, NY, 1991). The rationalemore » for this exclusion is that at higher temperatures it is not feasible to decouple plasticity and creep, which is the basis for the current simplified rules. This temperature, 650°C (1200°F), is well below the temperature range of interest for this material for the high temperature gas-cooled reactors and the very high temperature gas-cooled reactors. The only current alternative is, thus, a full inelastic analysis requiring sophisticated material models that have not yet been formulated and verified. To address these issues, proposed code rules have been developed which are based on the use of elastic-perfectly plastic (EPP) analysis methods applicable to very high temperatures. The proposed rules for strain limits and creep-fatigue evaluation were initially documented in the technical literature (Carter, Jetter and Sham, Proceedings of ASME 2012 Pressure Vessels and Piping Conference, papers PVP 2012 28082 and PVP 2012 28083, ASME, NY, 2012), and have been recently revised to incorporate comments and simplify their application. Background documents have been developed for these two code cases to support the ASME Code committee approval process. These background documents for the EPP strain limits and creep-fatigue code cases are documented in this report.« less