Movable-molybdenum-reflector reactivity experiments for control studies of compact space power reactor concepts

NASA Technical Reports Server (NTRS)

Fox, T. A.

1973-01-01

An experimental reflector reactivity study was made with a compact cylindrical reactor using a uranyl fluoride - water fuel solution. The reactor was axially unreflected and radially reflected with segments of molybdenum. The reflector segments were displaced incrementally in both the axial and radial dimensions, and the shutdown of each configuration was measured by using the pulsed-neutron source technique. The reactivity effects for axial and radial displacement of reflector segments are tabulated separately and compared. The experiments provide data for control-system studies of compact-space-power-reactor concepts.

Steady RANS methodology for calculating pressure drop in an in-line molten salt compact crossflow heat exchanger

DOE PAGES

Carasik, Lane B.; Shaver, Dillon R.; Haefner, Jonah B.; ...

2017-08-21

We report the development of molten salt cooled reactors (MSR) and fluoride-salt cooled high temperature reactors (FHR) requires the use of advanced design tools for the primary heat exchanger design. Due to geometric and flow characteristics, compact (pitch to diameter ratios equal to or less than 1.25) heat exchangers with a crossflow flow arrangement can become desirable for these reactors. Unfortunately, the available experimental data is limited for compact tube bundles or banks in crossflow. Computational Fluid Dynamics can be used to alleviate the lack of experimental data in these tube banks. Previous computational efforts have been primarily focused onmore » large S/D ratios (larger than 1.4) using unsteady Reynolds averaged Navier-Stokes and Large Eddy Simulation frameworks. These approaches are useful, but have large computational requirements that make comprehensive design studies impractical. A CFD study was conducted with steady RANS in an effort to provide a starting point for future design work. The study was performed for an in-line tube bank geometry with FLiBe (LiF-BeF2), a frequently selected molten salt, as the working fluid. Based on the estimated pressure drops, the pressure and velocity distributions in the domain, an appropriate meshing strategy was determined and presented. Periodic boundaries in the spanwise direction transverse flow were determined to be an appropriate boundary condition for reduced computational domains. The domain size was investigated and a minimum of 2-flow channels for a domain is recommended to ensure the behavior is accounted for. Finally, the standard low Re κ-ε (Lien) turbulence model was determined to be the most appropriate for steady RANS of this case at the time of writing.« less

Steady RANS methodology for calculating pressure drop in an in-line molten salt compact crossflow heat exchanger

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

Carasik, Lane B.; Shaver, Dillon R.; Haefner, Jonah B.

We report the development of molten salt cooled reactors (MSR) and fluoride-salt cooled high temperature reactors (FHR) requires the use of advanced design tools for the primary heat exchanger design. Due to geometric and flow characteristics, compact (pitch to diameter ratios equal to or less than 1.25) heat exchangers with a crossflow flow arrangement can become desirable for these reactors. Unfortunately, the available experimental data is limited for compact tube bundles or banks in crossflow. Computational Fluid Dynamics can be used to alleviate the lack of experimental data in these tube banks. Previous computational efforts have been primarily focused onmore » large S/D ratios (larger than 1.4) using unsteady Reynolds averaged Navier-Stokes and Large Eddy Simulation frameworks. These approaches are useful, but have large computational requirements that make comprehensive design studies impractical. A CFD study was conducted with steady RANS in an effort to provide a starting point for future design work. The study was performed for an in-line tube bank geometry with FLiBe (LiF-BeF2), a frequently selected molten salt, as the working fluid. Based on the estimated pressure drops, the pressure and velocity distributions in the domain, an appropriate meshing strategy was determined and presented. Periodic boundaries in the spanwise direction transverse flow were determined to be an appropriate boundary condition for reduced computational domains. The domain size was investigated and a minimum of 2-flow channels for a domain is recommended to ensure the behavior is accounted for. Finally, the standard low Re κ-ε (Lien) turbulence model was determined to be the most appropriate for steady RANS of this case at the time of writing.« less

Japan Report No. 173.

DTIC Science & Technology

1983-06-10

nuclear ship Mutsu . We will in parallel pursue the development of an advanced ma- rine reactor of more compact and ef- ficient design. In the field...velopment include such subjects as the High Temperature Gas Reactor - envisaged for uses other than power generation - and nuclear ship pro- pulsion. We...950’centigrade. In the field of nuclear ship pro- pulsion, we will proceed on our plans for experimental voyages to be undertaken by our first

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

NASA Astrophysics Data System (ADS)

Nishimura, Shun; Ebitani, Kohki

2018-01-01

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

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

Batchelor, D.B.; Carreras, B.A.; Hirshman, S.P.

Significant progress has been made in the development of new modest-size compact stellarator devices that could test optimization principles for the design of a more attractive reactor. These are 3 and 4 field period low-aspect-ratio quasi-omnigenous (QO) stellarators based on an optimization method that targets improved confinement, stability, ease of coil design, low-aspect-ratio, and low bootstrap current.

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

NASA Astrophysics Data System (ADS)

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

2013-02-01

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

Neutral Beam Development for the Lockheed Martin Compact Fusion Reactor

NASA Astrophysics Data System (ADS)

Ebersohn, Frans; Sullivan, Regina

2017-10-01

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

Lunar Resource Utilization: Development of a Reactor for Volatile Extraction from Regolith

NASA Technical Reports Server (NTRS)

Kleinhenz, Julie E.; Sacksteder, Kurt R.; Nayagam, Vedha

2007-01-01

The extraction and processing of planetary resources into useful products, known as In- Situ Resource Utilization (ISRU), will have a profound impact on the future of planetary exploration. One such effort is the RESOLVE (Regolith and Environment Science, Oxygen and Lunar Volatiles Extraction) Project, which aims to extract and quantify these resources. As part of the first Engineering Breadboard Unit, the Regolith Volatiles Characterization (RVC) reactor was designed and built at the NASA Glenn Research Center. By heating and agitating the lunar regolith, loosely bound volatiles, such as hydrogen and water, are released and stored in the reactor for later analysis and collection. Intended for operation on a robotic rover, the reactor features a lightweight, compact design, easy loading and unloading of the regolith, and uniform heating of the regolith by means of vibrofluidization. The reactor performance was demonstrated using regolith simulant, JSC1, with favorable results.

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

NASA Technical Reports Server (NTRS)

Mayo, W. T.

1972-01-01

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

ASME code considerations for the compact heat exchanger

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

Nestell, James; Sham, Sam

2015-08-31

The mission of the U.S. Department of Energy (DOE), Office of Nuclear Energy is to advance nuclear power in order to meet the nation's energy, environmental, and energy security needs. Advanced high temperature reactor systems such as sodium fast reactors and high and very high temperature gas-cooled reactors are being considered for the next generation of nuclear reactor plant designs. The coolants for these high temperature reactor systems include liquid sodium and helium gas. Supercritical carbon dioxide (sCO₂), a fluid at a temperature and pressure above the supercritical point of CO₂, is currently being investigated by DOE as a workingmore » fluid for a nuclear or fossil-heated recompression closed Brayton cycle energy conversion system that operates at 550°C (1022°F) at 200 bar (2900 psi). Higher operating temperatures are envisioned in future developments. All of these design concepts require a highly effective heat exchanger that transfers heat from the nuclear or chemical reactor to the chemical process fluid or the to the power cycle. In the nuclear designs described above, heat is transferred from the primary to the secondary loop via an intermediate heat exchanger (IHX) and then from the intermediate loop to either a working process or a power cycle via a secondary heat exchanger (SHX). The IHX is a component in the primary coolant loop which will be classified as "safety related." The intermediate loop will likely be classified as "not safety related but important to safety." These safety classifications have a direct bearing on heat exchanger design approaches for the IHX and SHX. The very high temperatures being considered for the VHTR will require the use of very high temperature alloys for the IHX and SHX. Material cost considerations alone will dictate that the IHX and SHX be highly effective; that is, provide high heat transfer area in a small volume. This feature must be accompanied by low pressure drop and mechanical reliability and robustness. Classic shell and tube designs will be large and costly, and may only be appropriate in steam generator service in the SHX where boiling inside the tubes occurs. For other energy conversion systems, all of these features can be met in a compact heat exchanger design. This report will examine some of the ASME Code issues that will need to be addressed to allow use of a Code-qualified compact heat exchanger in IHX or SHX nuclear service. Most effort will focus on the IHX, since the safety-related (Class A) design rules are more extensive than those for important-to-safety (Class B) or commercial rules that are relevant to the SHX.« less

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

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

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

Dual-phase reactor plant with partitioned isolation condenser

DOEpatents

Hui, Marvin M.

1992-01-01

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

SlimCS—compact low aspect ratio DEMO reactor with reduced-size central solenoid

NASA Astrophysics Data System (ADS)

Tobita, K.; Nishio, S.; Sato, M.; Sakurai, S.; Hayashi, T.; Shibama, Y. K.; Isono, T.; Enoeda, M.; Nakamura, H.; Sato, S.; Ezato, K.; Hayashi, T.; Hirose, T.; Ide, S.; Inoue, T.; Kamada, Y.; Kawamura, Y.; Kawashima, H.; Koizumi, N.; Kurita, G.; Nakamura, Y.; Mouri, K.; Nishitani, T.; Ohmori, J.; Oyama, N.; Sakamoto, K.; Suzuki, S.; Suzuki, T.; Tanigawa, H.; Tsuchiya, K.; Tsuru, D.

2007-08-01

The concept for a compact DEMO reactor named 'SlimCS' is presented. Distinctive features of the concept are low aspect ratio (A = 2.6) and use of a reduced-size centre solenoid (CS) which has the function of plasma shaping rather than poloidal flux supply. The reduced-size CS enables us to introduce a thin toroidal field coil system which contributes to reducing the weight and perhaps lessening the construction cost. Low-A has merits of vertical stability for high elongation (κ) and high normalized beta (βN), which leads to a high power density with reasonable physics requirements. This is because high κ facilitates high nGW (because of an increase in Ip), which allows efficient use of the capacity of high βN. From an engineering aspect, low-A may ensure ease in designing blanket modules robust to electromagnetic forces acting on disruptions. Thus, a superconducting low-A tokamak reactor such as SlimCS can be a promising DEMO concept with physics and engineering advantages.

Progress Toward Attractive Stellarators

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

Neilson, G H; Brown, T G; Gates, D A

The quasi-axisymmetric stellarator (QAS) concept offers a promising path to a more compact stellarator reactor, closer in linear dimensions to tokamak reactors than previous stellarator designs. Concept improvements are needed, however, to make it more maintainable and more compatible with high plant availability. Using the ARIES-CS design as a starting point, compact stellarator designs with improved maintenance characteristics have been developed. While the ARIES-CS features a through-the-port maintenance scheme, we have investigated configuration changes to enable a sector-maintenance approach, as envisioned for example in ARIES AT. Three approaches are reported. The first is to make tradeoffs within the QAS designmore » space, giving greater emphasis to maintainability criteria. The second approach is to improve the optimization tools to more accurately and efficiently target the physics properties of importance. The third is to employ a hybrid coil topology, so that the plasma shaping functions of the main coils are shared more optimally, either with passive conductors made of high-temperature superconductor or with local compensation coils, allowing the main coils to become simpler. Optimization tools are being improved to test these approaches.« less

Safety Testing of AGR-2 UCO Compacts 6-4-2 and 2-3-1

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

Hunn, John D.; Morris, Robert N.; Baldwin, Charles A.

2017-08-01

Post-irradiation examination (PIE) and elevated-temperature safety testing are being performed on tristructural-isotropic (TRISO) coated-particle fuel compacts from the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program second irradiation experiment (AGR-2). Details on this irradiation experiment have been previously reported [Collin 2014]. The AGR-2 PIE effort builds upon the understanding acquired throughout the AGR-1 PIE campaign [Demkowicz et al. 2015] and is establishing a database for the different AGR-2 fuel designs.

Stacked waveguide reactors with gradient embedded scatterers for high-capacity water cleaning

DOE PAGES

Ahsan, Syed Saad; Gumus, Abdurrahman; Erickson, David

2015-11-04

We present a compact water-cleaning reactor with stacked layers of waveguides containing gradient patterns of optical scatterers that enable uniform light distribution and augmented water-cleaning rates. Previous photocatalytic reactors using immersion, external, or distributive lamps suffer from poor light distribution that impedes scalability. Here, we use an external UV-source to direct photons into stacked waveguide reactors where we scatter the photons uniformly over the length of the waveguide to thin films of TiO 2-catalysts. In conclusion, we also show 4.5 times improvement in activity over uniform scatterer designs, demonstrate a degradation of 67% of the organic dye, and characterize themore » degradation rate constant.« less

Stacked waveguide reactors with gradient embedded scatterers for high-capacity water cleaning

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

Ahsan, Syed Saad; Gumus, Abdurrahman; Erickson, David

We present a compact water-cleaning reactor with stacked layers of waveguides containing gradient patterns of optical scatterers that enable uniform light distribution and augmented water-cleaning rates. Previous photocatalytic reactors using immersion, external, or distributive lamps suffer from poor light distribution that impedes scalability. Here, we use an external UV-source to direct photons into stacked waveguide reactors where we scatter the photons uniformly over the length of the waveguide to thin films of TiO 2-catalysts. In conclusion, we also show 4.5 times improvement in activity over uniform scatterer designs, demonstrate a degradation of 67% of the organic dye, and characterize themore » degradation rate constant.« less

An overview of optical diagnostics developed for the Lockheed Martin compact fusion reactor

NASA Astrophysics Data System (ADS)

Sommers, Bradley; Raymond, Anthony; Gucker, Sarah; Lockheed Martin Compact Fusion Reactor Team

2017-10-01

The T4B experiment is a linear, encapsulated ring cusp confinement device, designed to develop a physics and technology basis for a follow-on high beta machine as part of the compact fusion reactor program. Toward this end, a collection of non-invasive optical diagnostics have been developed to investigate confinement, neutral beam heating, and source behavior on the T4B device. These diagnostics include: (1) a multipoint Thomson scattering system employing a 532 nm Nd:YAG laser and high throughput spectrometer to measure 1D profiles of electron density and temperature, (2) a dispersion interferometer utilizing a continuous-wave CO2 laser (10.6 μm) to measure time resolved, line-integrated electron density, and (3) a bolometer suite utilizing four AXUV photodiodes with 64 lines of sight to generate 2D reconstructions of total radiative power and soft x-ray emission (via beryllium filters). An overview of design methods, including laser systems, detection schemes, and data analysis techniques is presented as well as results to date.

High Efficiency Heat Exchanger for High Temperature and High Pressure Applications

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

Sienicki, James J.; Lv, Qiuping; Moisseytsev, Anton

CompRex, LLC (CompRex) specializes in the design and manufacture of compact heat exchangers and heat exchange reactors for high temperature and high pressure applications. CompRex’s proprietary compact technology not only increases heat exchange efficiency by at least 25 % but also reduces footprint by at least a factor of ten compared to traditional shell-and-tube solutions of the same capacity and by 15 to 20 % compared to other currently available Printed Circuit Heat Exchanger (PCHE) solutions. As a result, CompRex’s solution is especially suitable for Brayton cycle supercritical carbon dioxide (sCO2) systems given its high efficiency and significantly lower capitalmore » and operating expenses. CompRex has already successfully demonstrated its technology and ability to deliver with a pilot-scale compact heat exchanger that was under contract by the Naval Nuclear Laboratory for sCO2 power cycle development. The performance tested unit met or exceeded the thermal and hydraulic specifications with measured heat transfer between 95 to 98 % of maximum heat transfer and temperature and pressure drop values all consistent with the modeled values. CompRex’s vision is to commercialize its compact technology and become the leading provider for compact heat exchangers and heat exchange reactors for various applications including Brayton cycle sCO2 systems. One of the limitations of the sCO2 Brayton power cycle is the design and manufacturing of efficient heat exchangers at extreme operating conditions. Current diffusion-bonded heat exchangers have limitations on the channel size through which the fluid travels, resulting in excessive solid material per heat exchanger volume. CompRex’s design allows for more open area and shorter fluid proximity for increased heat transfer efficiency while sustaining the structural integrity needed for the application. CompRex is developing a novel improvement to its current heat exchanger design where fluids are directed to alternating channels so that each fluid is fully surrounded by the opposing fluid. As compared to similar existing compact heat exchangers, the new design converts most secondary surface area to primary surface area, eliminating fin inefficiencies. CompRex requests that all technical information about the heat exchanger designs be protected as proprietary information. To honor that request, only non-proprietay summaries are included in this report.« less

Nuclear design of a vapor core reactor for space nuclear propulsion

NASA Astrophysics Data System (ADS)

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

1993-01-01

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

ADAPTATION OF CRACK GROWTH DETECTION TECHNIQUES TO US MATERIAL TEST REACTORS

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

A. Joseph Palmer; Sebastien P. Teysseyre; Kurt L. Davis

2015-04-01

A key component in evaluating the ability of Light Water Reactors to operate beyond 60 years is characterizing the degradation of materials exposed to radiation and various water chemistries. Of particular concern is the response of reactor materials to Irradiation Assisted Stress Corrosion Cracking (IASCC). Some test reactors outside the United States, such as the Halden Boiling Water Reactor (HBWR), have developed techniques to measure crack growth propagation during irradiation. The basic approach is to use a custom-designed compact loading mechanism to stress the specimen during irradiation, while the crack in the specimen is monitored in-situ using the Direct Currentmore » Potential Drop (DCPD) method. In 2012 the US Department of Energy commissioned the Idaho National Laboratory and the MIT Nuclear Reactor Laboratory (MIT NRL) to take the basic concepts developed at the HBWR and adapt them to a test rig capable of conducting in-pile IASCC tests in US Material Test Reactors. The first two and half years of the project consisted of designing and testing the loader mechanism, testing individual components of the in-pile rig and electronic support equipment, and autoclave testing of the rig design prior to insertion in the MIT Reactor. The load was applied to the specimen by means of a scissor like mechanism, actuated by a miniature metal bellows driven by pneumatic pressure and sized to fit within the small in-core irradiation volume. In addition to the loader design, technical challenges included developing robust connections to the specimen for the applied current and voltage measurements, appropriate ceramic insulating materials that can endure the LWR environment, dealing with the high electromagnetic noise environment of a reactor core at full power, and accommodating material property changes in the specimen, due primarily to fast neutron damage, which change the specimen resistance without additional crack growth. The project culminated with an in-pile demonstration at the MIT Reactor. The test rig and associated support equipment were used to apply loads to a representative Compact Tensile specimen during one MITR operating cycle, while measuring crack growth using the DCPD method. Although the test period was short (approximately 70 days), and the accumulated neutron dose relatively small, successful operation of the test rig was demonstrated. The specimen was cycled more than 8000 times (more than would be typical for a long term IASCC test), which was sufficient to propagate a crack of over 2 mm.« less

Radioactive Waste Streams: Waste Classification for Disposal

DTIC Science & Technology

2006-12-13

INL; and Fort St. Vrain, Colorado .10 In contrast to commercial reactors, naval reactors can operate without refueling for up to 20 years. 11 As of 2003...originally of the states of Arizona, Colorado , Nevada, New Mexico, Utah, and Wyoming.61 Arizona, Utah, and Wyoming later withdrew from the Compact, leaving... Colorado , Nevada, and New Mexico as remaining Compact members.62 The Rocky Mountain Compact defines low-level waste as specifically excluding

Evaluating quantitative 3-D image analysis as a design tool for low enriched uranium fuel compacts for the transient reactor test facility: A preliminary study

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

Kane, J. J.; van Rooyen, I. J.; Craft, A. E.

In this study, 3-D image analysis when combined with a non-destructive examination technique such as X-ray computed tomography (CT) provides a highly quantitative tool for the investigation of a material’s structure. In this investigation 3-D image analysis and X-ray CT were combined to analyze the microstructure of a preliminary subsized fuel compact for the Transient Reactor Test Facility’s low enriched uranium conversion program to assess the feasibility of the combined techniques for use in the optimization of the fuel compact fabrication process. The quantitative image analysis focused on determining the size and spatial distribution of the surrogate fuel particles andmore » the size, shape, and orientation of voids within the compact. Additionally, the maximum effect of microstructural features on heat transfer through the carbonaceous matrix of the preliminary compact was estimated. The surrogate fuel particles occupied 0.8% of the compact by volume with a log-normal distribution of particle sizes with a mean diameter of 39 μm and a standard deviation of 16 μm. Roughly 39% of the particles had a diameter greater than the specified maximum particle size of 44 μm suggesting that the particles agglomerate during fabrication. The local volume fraction of particles also varies significantly within the compact although uniformities appear to be evenly dispersed throughout the analysed volume. The voids produced during fabrication were on average plate-like in nature with their major axis oriented perpendicular to the compaction direction of the compact. Finally, the microstructure, mainly the large preferentially oriented voids, may cause a small degree of anisotropy in the thermal diffusivity within the compact. α∥/α⊥, the ratio of thermal diffusivities parallel to and perpendicular to the compaction direction are expected to be no less than 0.95 with an upper bound of 1.« less

Evaluating quantitative 3-D image analysis as a design tool for low enriched uranium fuel compacts for the transient reactor test facility: A preliminary study

DOE PAGES

Kane, J. J.; van Rooyen, I. J.; Craft, A. E.; ...

2016-02-05

In this study, 3-D image analysis when combined with a non-destructive examination technique such as X-ray computed tomography (CT) provides a highly quantitative tool for the investigation of a material’s structure. In this investigation 3-D image analysis and X-ray CT were combined to analyze the microstructure of a preliminary subsized fuel compact for the Transient Reactor Test Facility’s low enriched uranium conversion program to assess the feasibility of the combined techniques for use in the optimization of the fuel compact fabrication process. The quantitative image analysis focused on determining the size and spatial distribution of the surrogate fuel particles andmore » the size, shape, and orientation of voids within the compact. Additionally, the maximum effect of microstructural features on heat transfer through the carbonaceous matrix of the preliminary compact was estimated. The surrogate fuel particles occupied 0.8% of the compact by volume with a log-normal distribution of particle sizes with a mean diameter of 39 μm and a standard deviation of 16 μm. Roughly 39% of the particles had a diameter greater than the specified maximum particle size of 44 μm suggesting that the particles agglomerate during fabrication. The local volume fraction of particles also varies significantly within the compact although uniformities appear to be evenly dispersed throughout the analysed volume. The voids produced during fabrication were on average plate-like in nature with their major axis oriented perpendicular to the compaction direction of the compact. Finally, the microstructure, mainly the large preferentially oriented voids, may cause a small degree of anisotropy in the thermal diffusivity within the compact. α∥/α⊥, the ratio of thermal diffusivities parallel to and perpendicular to the compaction direction are expected to be no less than 0.95 with an upper bound of 1.« less

Compact power reactor

DOEpatents

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

1978-01-01

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

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.

METHOD OF PREPARING A FUEL ELEMENT FOR A NUCLEAR REACTOR

DOEpatents

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

1962-12-01

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

Non-Nuclear Testing of Compact Reactor Technologies at NASA MSFC

NASA Technical Reports Server (NTRS)

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

2011-01-01

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

Materials interactions between the thermoelectric converter and the 5kwe reactor system

NASA Technical Reports Server (NTRS)

Ferry, P. B.

1973-01-01

The integration of a compact thermoelectric converter with a 5-kwe reactor system is described. Material interaction uncertainties study is also presented. This includes degradation of the required austenitic - refractory metal transition joint during operation at high temperatures; loss of corrosion resistance; embrittlement by the presence of hydrogen; and loss of design margin by transport of interstitial elements. Analysis and limited experimental evidence indicate that these potential materials interactions can be adequately controlled. Group 5-2 refractory metals can be utilized without unacceptable adverse effect on system reliability.

Electronics for the STEREO experiment

NASA Astrophysics Data System (ADS)

HÉLAINE, Victor; STEREO Collaboration

2017-09-01

The STEREO experiment, aiming to probe short baseline neutrino oscillations by precisely measuring reactor anti-neutrino spectrum, is currently under installation. It is located at short distance from the compact research reactor core of the Institut Laue-Langevin, Grenoble, France. Dedicated electronics, hosted in a single µTCA crate, were designed for this experiment. In this article, the electronics requirements, architecture and the performances achieved are described. It is shown how intrinsic Pulse Shape Discrimination properties of the liquid scintillator are preserved and how custom adaptable logic is used to improve the muon veto efficiency.

SoLid: Search for Oscillations with Lithium-6 Detector at the SCK-CEN BR2 reactor

NASA Astrophysics Data System (ADS)

Ban, G.; Beaumont, W.; Buhour, J. M.; Coupé, B.; Cucoanes, A. S.; D'Hondt, J.; Durand, D.; Fallot, M.; Fresneau, S.; Giot, L.; Guillon, B.; Guilloux, G.; Janssen, X.; Kalcheva, S.; Koonen, E.; Labare, M.; Moortgat, C.; Pronost, G.; Raes, L.; Ryckbosch, D.; Ryder, N.; Shitov, Y.; Vacheret, A.; Van Mulders, P.; Van Remortel, N.; Weber, A.; Yermia, F.

2016-04-01

Sterile neutrinos have been considered as a possible explanation for the recent reactor and Gallium anomalies arising from reanalysis of reactor flux and calibration data of previous neutrino experiments. A way to test this hypothesis is to look for distortions of the anti-neutrino energy caused by oscillation from active to sterile neutrino at close stand-off (˜ 6- 8m) of a compact reactor core. Due to the low rate of anti-neutrino interactions the main challenge in such measurement is to control the high level of gamma rays and neutron background. The SoLid experiment is a proposal to search for active-to-sterile anti-neutrino oscillation at very short baseline of the SCK•CEN BR2 research reactor. This experiment uses a novel approach to detect anti-neutrino with a highly segmented detector based on Lithium-6. With the combination of high granularity, high neutron-gamma discrimination using 6LiF:ZnS(Ag) and precise localization of the Inverse Beta Decay products, a better experimental sensitivity can be achieved compared to other state-of-the-art technology. This compact system requires minimum passive shielding allowing for very close stand off to the reactor. The experimental set up of the SoLid experiment and the BR2 reactor will be presented. The new principle of neutrino detection and the detector design with expected performance will be described. The expected sensitivity to new oscillations of the SoLid detector as well as the first measurements made with the 8 kg prototype detector deployed at the BR2 reactor in 2013-2014 will be reported.

Tory II-A: a nuclear ramjet test reactor

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

Hadley, J.W.

Declassified 28 Nov 1973. The first test reactor in the Pluto program, leading to development of a nuclear ramjet engine, is called Tory II-A. While it is not an actual prototype engine, this reactor embodies a core design which is considered feasible for an engine, and operation of the reactor will provide a test of that core type as well as more generalized values in reactor design and testing. The design of Tory II-A and construction of the reactor and of its test facility are described. Operation of the Tory II-A core at a total power of 160 megawatts, withmore » 800 pounds of air per second passing through the core and emerging at a temperature of 2000 deg F, is the central objective of the test program. All other reactor and facility components exist to support operation of the core, and preliminary steps in the test program itself will be directed primarily toward ensuring attalnment of full-power operation and collection of meaningful data on core behavior during that operation. The core, 3 feet in diameter and 41/2 feet long, will be composed of bundled ceramic tubes whose central holes will provide continuous air passages from end to end of the reactor. These tubes are to be composed of a homogeneous mixture of UO/sub 2/ fuel and BeO moderator, compacted and sintered to achieve high strength and density. (30 references) (auth)« less

Effect of Catalytic Cylinders on Autothermal Reforming of Methane for Hydrogen Production in a Microchamber Reactor

PubMed Central

Yan, Yunfei; Guo, Hongliang; Zhang, Li; Zhu, Junchen; Yang, Zhongqing; Tang, Qiang; Ji, Xin

2014-01-01

A new multicylinder microchamber reactor is designed on autothermal reforming of methane for hydrogen production, and its performance and thermal behavior, that is, based on the reaction mechanism, is numerically investigated by varying the cylinder radius, cylinder spacing, and cylinder layout. The results show that larger cylinder radius can promote reforming reaction; the mass fraction of methane decreased from 26% to 21% with cylinder radius from 0.25 mm to 0.75 mm; compact cylinder spacing corresponds to more catalytic surface and the time to steady state is decreased from 40 s to 20 s; alteration of staggered and aligned cylinder layout at constant inlet flow rates does not result in significant difference in reactor performance and it can be neglected. The results provide an indication and optimize performance of reactor; it achieves higher conversion compared with other reforming reactors. PMID:25097877

Progress in space nuclear reactor power systems technology development - The SP-100 program

NASA Technical Reports Server (NTRS)

Davis, H. S.

1984-01-01

Activities related to the development of high-temperature compact nuclear reactors for space applications had reached a comparatively high level in the U.S. during the mid-1950s and 1960s, although only one U.S. nuclear reactor-powered spacecraft was actually launched. After 1973, very little effort was devoted to space nuclear reactor and propulsion systems. In February 1983, significant activities toward the development of the technology for space nuclear reactor power systems were resumed with the SP-100 Program. Specific SP-100 Program objectives are partly related to the determination of the potential performance limits for space nuclear power systems in 100-kWe and 1- to 100-MW electrical classes. Attention is given to potential missions and applications, regimes of possible space power applicability, safety considerations, conceptual system designs, the establishment of technical feasibility, nuclear technology, materials technology, and prospects for the future.

Radiological Shielding Design for the Neutron High-Resolution Backscattering Spectrometer EMU at the OPAL Reactor

NASA Astrophysics Data System (ADS)

Ersez, Tunay; Esposto, Fernando; Souza, Nicolas R. de

2017-09-01

The shielding for the neutron high-resolution backscattering spectrometer (EMU) located at the OPAL reactor (ANSTO) was designed using the Monte Carlo code MCNP 5-1.60. The proposed shielding design has produced compact shielding assemblies, such as the neutron pre-monochromator bunker with sliding cylindrical block shields to accommodate a range of neutron take-off angles, and in the experimental area - shielding of neutron focusing guides, choppers, flight tube, backscattering monochromator, and additional shielding elements inside the Scattering Tank. These shielding assemblies meet safety and engineering requirements and cost constraints. The neutron dose rates around the EMU instrument were reduced to < 0.5 µSv/h and the gamma dose rates to a safe working level of ≤ 3 µSv/h.

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

Heitzenroeder, P.; Dudek, Lawrence E.; Brooks, Arthur W.

The National Compact Stellarator Experiment, NCSX, is being constructed at the Princeton Plasma Physics Laboratory (PPPL) in partnership with the Oak Ridge national Laboratory. The goal of NCSX is to provide the understanding necessary to develop an attractive, disruption free, steady state compact stellaratorbased reactor design. This paper describes the recently revised designs of the critical interfaces between the modular coils, the construction solutions developed to meet assembly tolerances, and the recently revised trim coil system that provides the required compensation to correct for the “as built” conditions and to allow flexibility in the disposition of as-built conditions. In May,more » 2008, the sponsor decided to terminate the NCSX project due to growth in the project’s cost and schedule estimates. However significant technical challenges in design and construction were overcome, greatly reducing the risk in the remaining work to complete the project.« less

Multi-physics design and analyses of long life reactors for lunar outposts

NASA Astrophysics Data System (ADS)

Schriener, Timothy M.

Future human exploration of the solar system is likely to include establishing permanent outposts on the surface of the Moon. These outposts will require reliable sources of electrical power in the range of 10's to 100's of kWe to support exploration and resource utilization activities. This need is best met using nuclear reactor power systems which can operate steadily throughout the long ˜27.3 day lunar rotational period, irrespective of location. Nuclear power systems can potentially open up the entire lunar surface for future exploration and development. Desirable features of nuclear power systems for the lunar surface include passive operation, the avoidance of single point failures in reactor cooling and the integrated power system, moderate operating temperatures to enable the use of conventional materials with proven irradiation experience, utilization of the lunar regolith for radiation shielding and as a supplemental neutron reflector, and safe post-operation decay heat removal and storage for potential retrieval. In addition, it is desirable for the reactor to have a long operational life. Only a limited number of space nuclear reactor concepts have previously been developed for the lunar environment, and these designs possess only a few of these desirable design and operation features. The objective of this research is therefore to perform design and analyses of long operational life lunar reactors and power systems which incorporate the desirable features listed above. A long reactor operational life could be achieved either by increasing the amount of highly enriched uranium (HEU) fuel in the core or by improving the neutron economy in the reactor through reducing neutron leakage and parasitic absorption. The amount of fuel in surface power reactors is constrained by the launch safety requirements. These include ensuring that the bare reactor core remains safely subcritical when submerged in water or wet sand and flooded with seawater in the unlikely event of a launch abort accident. Increasing the amount of fuel in the reactor core, and hence its operational life, would be possible by launching the reactor unfueled and fueling it on the Moon. Such a reactor would, thus, not be subject to launch criticality safety requirements. However, loading the reactor with fuel on the Moon presents a challenge, requiring special designs of the core and the fuel elements, which lend themselves to fueling on the lunar surface. This research investigates examples of both a solid core reactor that would be fueled at launch as well as an advanced concept which could be fueled on the Moon. Increasing the operational life of a reactor fueled at launch is exercised for the NaK-78 cooled Sectored Compact Reactor (SCoRe). A multi-physics design and analyses methodology is developed which iteratively couples together detailed Monte Carlo neutronics simulations with 3-D Computational Fluid Dynamics (CFD) and thermal-hydraulics analyses. Using this methodology the operational life of this compact, fast spectrum reactor is increased by reconfiguring the core geometry to reduce neutron leakage and parasitic absorption, for the same amount of HEU in the core, and meeting launch safety requirements. The multi-physics analyses determine the impacts of the various design changes on the reactor's neutronics and thermal-hydraulics performance. The option of increasing the operational life of a reactor by loading it on the Moon is exercised for the Pellet Bed Reactor (PeBR). The PeBR uses spherical fuel pellets and is cooled by He-Xe gas, allowing the reactor core to be loaded with fuel pellets and charged with working fluid on the lunar surface. The performed neutronics analyses ensure the PeBR design achieves a long operational life, and develops safe launch canister designs to transport the spherical fuel pellets to the lunar surface. The research also investigates loading the PeBR core with fuel pellets on the Moon using a transient Discrete Element Method (DEM) analysis in lunar gravity. In addition, this research addresses the post-operation storage of the SCoRe and PeBR concepts, below the lunar surface, to determine the time required for the radioactivity in the used fuel to decrease to a low level to allow for its safe recovery. The SCoRe and PeBR concepts are designed to operate at coolant temperatures ≤ 900 K and use conventional stainless steels and superalloys for the structure in the reactor core and power system. They are emplaced below grade on the Moon to take advantage of the regolith as a supplemental neutron reflector and as shielding of the lunar outpost from the reactors' neutron and gamma radiation.

Core Physics and Kinetics Calculations for the Fissioning Plasma Core Reactor

NASA Technical Reports Server (NTRS)

Butler, C.; Albright, D.

2007-01-01

Highly efficient, compact nuclear reactors would provide high specific impulse spacecraft propulsion. This analysis and numerical simulation effort has focused on the technical feasibility issues related to the nuclear design characteristics of a novel reactor design. The Fissioning Plasma Core Reactor (FPCR) is a shockwave-driven gaseous-core nuclear reactor, which uses Magneto Hydrodynamic effects to generate electric power to be used for propulsion. The nuclear design of the system depends on two major calculations: core physics calculations and kinetics calculations. Presently, core physics calculations have concentrated on the use of the MCNP4C code. However, initial results from other codes such as COMBINE/VENTURE and SCALE4a. are also shown. Several significant modifications were made to the ISR-developed QCALC1 kinetics analysis code. These modifications include testing the state of the core materials, an improvement to the calculation of the material properties of the core, the addition of an adiabatic core temperature model and improvement of the first order reactivity correction model. The accuracy of these modifications has been verified, and the accuracy of the point-core kinetics model used by the QCALC1 code has also been validated. Previously calculated kinetics results for the FPCR were described in the ISR report, "QCALC1: A code for FPCR Kinetics Model Feasibility Analysis" dated June 1, 2002.

Preliminary LOCA analysis of the westinghouse small modular reactor using the WCOBRA/TRAC-TF2 thermal-hydraulics code

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

Liao, J.; Kucukboyaci, V. N.; Nguyen, L.

2012-07-01

The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (> 225 MWe) integral pressurized water reactor (iPWR) with all primary components, including the steam generator and the pressurizer located inside the reactor vessel. The reactor core is based on a partial-height 17x17 fuel assembly design used in the AP1000{sup R} reactor core. The Westinghouse SMR utilizes passive safety systems and proven components from the AP1000 plant design with a compact containment that houses the integral reactor vessel and the passive safety systems. A preliminary loss of coolant accident (LOCA) analysis of the Westinghouse SMR has been performed using themore » WCOBRA/TRAC-TF2 code, simulating a transient caused by a double ended guillotine (DEG) break in the direct vessel injection (DVI) line. WCOBRA/TRAC-TF2 is a new generation Westinghouse LOCA thermal-hydraulics code evolving from the US NRC licensed WCOBRA/TRAC code. It is designed to simulate PWR LOCA events from the smallest break size to the largest break size (DEG cold leg). A significant number of fluid dynamics models and heat transfer models were developed or improved in WCOBRA/TRAC-TF2. A large number of separate effects and integral effects tests were performed for a rigorous code assessment and validation. WCOBRA/TRAC-TF2 was introduced into the Westinghouse SMR design phase to assist a quick and robust passive cooling system design and to identify thermal-hydraulic phenomena for the development of the SMR Phenomena Identification Ranking Table (PIRT). The LOCA analysis of the Westinghouse SMR demonstrates that the DEG DVI break LOCA is mitigated by the injection and venting from the Westinghouse SMR passive safety systems without core heat up, achieving long term core cooling. (authors)« less

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

NASA Astrophysics Data System (ADS)

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

2014-10-01

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

Apparatus, components and operating methods for circulating fluidized bed transport gasifiers and reactors

DOEpatents

Vimalchand, Pannalal; Liu, Guohai; Peng, Wan Wang

2015-02-24

The improvements proposed in this invention provide a reliable apparatus and method to gasify low rank coals in a class of pressurized circulating fluidized bed reactors termed "transport gasifier." The embodiments overcome a number of operability and reliability problems with existing gasifiers. The systems and methods address issues related to distribution of gasification agent without the use of internals, management of heat release to avoid any agglomeration and clinker formation, specific design of bends to withstand the highly erosive environment due to high solid particles circulation rates, design of a standpipe cyclone to withstand high temperature gasification environment, compact design of seal-leg that can handle high mass solids flux, design of nozzles that eliminate plugging, uniform aeration of large diameter Standpipe, oxidant injection at the cyclone exits to effectively modulate gasifier exit temperature and reduction in overall height of the gasifier with a modified non-mechanical valve.

Plasma nitriding monitoring reactor: A model reactor for studying plasma nitriding processes using an active screen

NASA Astrophysics Data System (ADS)

Hamann, S.; Börner, K.; Burlacov, I.; Spies, H.-J.; Strämke, M.; Strämke, S.; Röpcke, J.

2015-12-01

A laboratory scale plasma nitriding monitoring reactor (PLANIMOR) has been designed to study the basics of active screen plasma nitriding (ASPN) processes. PLANIMOR consists of a tube reactor vessel, made of borosilicate glass, enabling optical emission spectroscopy (OES) and infrared absorption spectroscopy. The linear setup of the electrode system of the reactor has the advantages to apply the diagnostic approaches on each part of the plasma process, separately. Furthermore, possible changes of the electrical field and of the heat generation, as they could appear in down-scaled cylindrical ASPN reactors, are avoided. PLANIMOR has been used for the nitriding of steel samples, achieving similar results as in an industrial scale ASPN reactor. A compact spectrometer using an external cavity quantum cascade laser combined with an optical multi-pass cell has been applied for the detection of molecular reaction products. This allowed the determination of the concentrations of four stable molecular species (CH4, C2H2, HCN, and NH3). With the help of OES, the rotational temperature of the screen plasma could be determined.

Exclusion Area Radiation Release during the MIT Reactor Design Basis Accident.

DTIC Science & Technology

1983-05-06

Concrete Wall 116 6.2 Concrete Albedo Dose 121 6.3 Steel Door Scattering Dose 124 7.1 Total Dose Results 133 A.1 Values of N /NO for Neutron -Capture...plate fuel elements arranged in x a compact hexagonal core. This core design maximizes the neutron flux in the DO2 reflector region where numerous...sec) V = Volume of the fuel (cm 3 f Ef = Macroscopic fission cross section (cm ) = Thermal neutron flux ( neutrons /cm2 - sec) = Core-averaged value Yi

Advanced Instrumentation for Transient Reactor Testing

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

Corradini, Michael L.; Anderson, Mark; Imel, George

Transient testing involves placing fuel or material into the core of specialized materials test reactors that are capable of simulating a range of design basis accidents, including reactivity insertion accidents, that require the reactor produce short bursts of intense highpower neutron flux and gamma radiation. Testing fuel behavior in a prototypic neutron environment under high-power, accident-simulation conditions is a key step in licensing nuclear fuels for use in existing and future nuclear power plants. Transient testing of nuclear fuels is needed to develop and prove the safety basis for advanced reactors and fuels. In addition, modern fuel development and designmore » increasingly relies on modeling and simulation efforts that must be informed and validated using specially designed material performance separate effects studies. These studies will require experimental facilities that are able to support variable scale, highly instrumented tests providing data that have appropriate spatial and temporal resolution. Finally, there are efforts now underway to develop advanced light water reactor (LWR) fuels with enhanced performance and accident tolerance. These advanced reactor designs will also require new fuel types. These new fuels need to be tested in a controlled environment in order to learn how they respond to accident conditions. For these applications, transient reactor testing is needed to help design fuels with improved performance. In order to maximize the value of transient testing, there is a need for in-situ transient realtime imaging technology (e.g., the neutron detection and imaging system like the hodoscope) to see fuel motion during rapid transient excursions with a higher degree of spatial and temporal resolution and accuracy. There also exists a need for new small, compact local sensors and instrumentation that are capable of collecting data during transients (e.g., local displacements, temperatures, thermal conductivity, neutron flux, etc.).« less

Development of a Radial Deconsolidation Method

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

Helmreich, Grant W.; Montgomery, Fred C.; Hunn, John D.

2015-12-01

A series of experiments have been initiated to determine the retention or mobility of fission products* in AGR fuel compacts [Petti, et al. 2010]. This information is needed to refine fission product transport models. The AGR-3/4 irradiation test involved half-inch-long compacts that each contained twenty designed-to-fail (DTF) particles, with 20-μm thick carbon-coated kernels whose coatings were deliberately fabricated such that they would crack under irradiation, providing a known source of post-irradiation isotopes. The DTF particles in these compacts were axially distributed along the compact centerline so that the diffusion of fission products released from the DTF kernels would be radiallymore » symmetric [Hunn, et al. 2012; Hunn et al. 2011; Kercher, et al. 2011; Hunn, et al. 2007]. Compacts containing DTF particles were irradiated at Idaho National Laboratory (INL) at the Advanced Test Reactor (ATR) [Collin, 2015]. Analysis of the diffusion of these various post-irradiation isotopes through the compact requires a method to radially deconsolidate the compacts so that nested-annular volumes may be analyzed for post-irradiation isotope inventory in the compact matrix, TRISO outer pyrolytic carbon (OPyC), and DTF kernels. An effective radial deconsolidation method and apparatus appropriate to this application has been developed and parametrically characterized.« less

Various methods to improve heat transfer in exchangers

NASA Astrophysics Data System (ADS)

Pavel, Zitek; Vaclav, Valenta

2015-05-01

The University of West Bohemia in Pilsen (Department of Power System Engineering) is working on the selection of effective heat exchangers. Conventional shell and tube heat exchangers use simple segmental baffles. It can be replaced by helical baffles, which increase the heat transfer efficiency and reduce pressure losses. Their usage is demonstrated in the primary circuit of IV. generation MSR (Molten Salt Reactors). For high-temperature reactors we consider the use of compact desk heat exchangers, which are small, which allows the integral configuration of reactor. We design them from graphite composites, which allow up to 1000°C and are usable as exchangers: salt-salt or salt-acid (e.g. for the hydrogen production). In the paper there are shown thermo-physical properties of salts, material properties and principles of calculations.

HRB-22 preirradiation thermal analysis

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

Acharya, R.; Sawa, K.

1995-05-01

This report describes the preirradiation thermal analysis of the HRB-22 capsule designed for irradiation in the removable beryllium (RB) position of the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL). CACA-2 a heavy isotope and fission product concentration calculational code for experimental irradiation capsules was used to determine time dependent fission power for the fuel compacts. The Heat Engineering and Transfer in Nine Geometries (HEATING) computer code, version 7.2, was used to solve the steady-state heat conduction problem. The diameters of the graphite fuel body that contains the compacts and the primary pressure vessel were selected suchmore » that the requirements of running the compacts at an average temperature of < 1,250 C and not exceeding a maximum fuel temperature of 1,350 C was met throughout the four cycles of irradiation.« less

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

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

Not Available

1952-03-29

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

Nuclear powerplants for mobile applications.

NASA Technical Reports Server (NTRS)

Anderson, J. L.

1972-01-01

Mobile nuclear powerplants for applications other than large ships and submarines will require compact, lightweight reactors with especially stringent impact-safety design. This paper examines the technical and economic feasibility that the broadening role of civilian nuclear power, in general, (land-based nuclear electric generating plants and nuclear ships) can extend to lightweight, safe mobile nuclear powerplants. The paper discusses technical experience, identifies potential sources of technology for advanced concepts, cites the results of economic studies of mobile nuclear powerplants, and surveys future technical capabilities needed by examining the current use and projected needs for vehicles, machines, and habitats that could effectively use mobile nuclear reactor powerplants.

Nuclear power plants for mobile applications

NASA Technical Reports Server (NTRS)

Anderson, J. L.

1972-01-01

Mobile nuclear powerplants for applications other than large ships and submarines will require compact, lightweight reactors with especially stringent impact-safety design. The technical and economic feasibility that the broadening role of civilian nuclear power, in general, (land-based nuclear electric generating plants and nuclear ships) can extend to lightweight, safe mobile nuclear powerplants are examined. The paper discusses technical experience, identifies potential sources of technology for advanced concepts, cites the results of economic studies of mobile nuclear powerplants, and surveys future technical capabilities needed by examining the current use and projected needs for vehicles, machines, and habitats that could effectively use mobile nuclear reactor powerplants.

Mass tracking and material accounting in the integral fast reactor (IFR)

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

Orechwa, Y.; Adams, C.H.; White, A.M.

1991-01-01

This paper reports on the Integral Fast Reactor (IFR) which is a generic advanced liquid metal cooled reactor concept being developed at Argonne National Laboratory. There are a number of technical features of the IFR which contribute to its potential as a next-generation reactor. These are associated with large safety margins with regard to off-normal events involving the heat transport system, and the use of metallic fuel which makes possible the utilization of innovative fuel cycle processes. The latter feature permits fuel cycle closure with compact, low-cost reprocessing facilities, collocated with the reactor plant. These primary features are being demonstratedmore » in the facilities at ANL-West, utilizing Experimental Breeder Reactor II and the associated Fuel Cycle Facility (FCF) as an IFR prototype. The demonstration of this IFR prototype includes the design and implementation of the Mass-tracking System (MTG). In this system, data from the operations of the FCF, including weights and batch-process parameters, are collected and maintained by the MTG running on distributed workstations.« less

The Potential of the LFR and the ELSY Project

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

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

2007-03-12

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

Plasma nitriding monitoring reactor: A model reactor for studying plasma nitriding processes using an active screen

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

Hamann, S., E-mail: hamann@inp-greifswald.de; Röpcke, J.; Börner, K.

2015-12-15

A laboratory scale plasma nitriding monitoring reactor (PLANIMOR) has been designed to study the basics of active screen plasma nitriding (ASPN) processes. PLANIMOR consists of a tube reactor vessel, made of borosilicate glass, enabling optical emission spectroscopy (OES) and infrared absorption spectroscopy. The linear setup of the electrode system of the reactor has the advantages to apply the diagnostic approaches on each part of the plasma process, separately. Furthermore, possible changes of the electrical field and of the heat generation, as they could appear in down-scaled cylindrical ASPN reactors, are avoided. PLANIMOR has been used for the nitriding of steelmore » samples, achieving similar results as in an industrial scale ASPN reactor. A compact spectrometer using an external cavity quantum cascade laser combined with an optical multi-pass cell has been applied for the detection of molecular reaction products. This allowed the determination of the concentrations of four stable molecular species (CH{sub 4}, C{sub 2}H{sub 2}, HCN, and NH{sub 3}). With the help of OES, the rotational temperature of the screen plasma could be determined.« less

Fabrication of capsule assemblies, phase 3

NASA Technical Reports Server (NTRS)

Keeton, A. R.; Stemann, L. G.

1973-01-01

Thirteen capsule assemblies were fabricated for evaluation of fuel pin design concepts for a fast spectrum lithium cooled compact space power reactor. These instrumented assemblies were designed for real time test of prototype fuel pins. Uranium mononitride fuel pins were encased in AISI 304L stainless steel capsules. Fabrication procedures were fully qualified by process development and assembly qualification tests. Instrumentation reliability was achieved utilizing specially processed and closely controlled thermocouple hot zone fabrication and by thermal screening tests. Overall capsule reliability was achieved with an all electron beam welded assembly.

Proceedings of a Symposium on Advanced Compact Reactor Systems

NASA Technical Reports Server (NTRS)

1983-01-01

Reactor system technologies suitable for a variety of aerospace and terrestrial applications are considered. Technologies, safety and regulatory considerations, potential applications, and research and development opportunities are covered.

Design of a heatpipe-cooled Mars-surface fission reactor

NASA Astrophysics Data System (ADS)

Poston, David I.; Kapernick, Richard J.; Guffee, Ray M.; Reid, Robert S.; Lipinski, Ronald J.; Wright, Steven A.; Talandis, Regina A.

2002-01-01

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 Heatpipe 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 space 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 Heatpipe-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 heatpipes. Fission energy is conducted from the fuel pins to the heatpipes, which then carry the heat to the Stirling engine. This paper describes the attributes, specifications, and performance of a 15-kWt HOMER reactor. .

SP-100 ground engineering system test site description and progress update

NASA Astrophysics Data System (ADS)

Baxter, William F.; Burchell, Gail P.; Fitzgibbon, Davis G.; Swita, Walter R.

1991-01-01

The SP-100 Ground Engineering System Test Site will provide the facilities for the testing of an SP-100 reactor, which is technically prototypic of the generic design for producing 100 kilowatts of electricity. This effort is part of the program to develop a compact, space-based power system capable of producing several hundred kilowatts of electrical power. The test site is located on the U.S. Department of Energy's Hanford Site near Richland, Washington. The site is minimizing capital equipment costs by utilizing existing facilities and equipment to the maximum extent possible. The test cell is located in a decommissioned reactor containment building, and the secondary sodium cooling loop will use equipment from the Fast Flux Test Facility plant which has never been put into service. Modifications to the facility and special equipment are needed to accommodate the testing of the SP-100 reactor. Definitive design of the Ground Engineering System Test Site facility modifications and systems is in progress. The design of the test facility and the testing equipment will comply with the regulations and specifications of the U.S. Department of Energy and the State of Washington.

PIE on Safety-Tested Loose Particles from Irradiated Compact 4-4-2

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

Hunn, John D.; Gerczak, Tyler J.; Morris, Robert Noel

2016-04-01

Post-irradiation examination (PIE) is being performed in support of tristructural isotropic (TRISO) coated particle fuel development and qualification for High Temperature Gas-cooled Reactors (HTGRs). This work is sponsored by the Department of Energy Office of Nuclear Energy (DOE-NE) through the Advanced Reactor Technologies (ART) Office under the Advanced Gas Reactor Fuel Development and Qualification (AGR) Program. The AGR-1 experiment was the first in a series of TRISO fuel irradiation tests initiated in 2006. The AGR-1 TRISO particles and fuel compacts were fabricated at Oak Ridge National Laboratory (ORNL) in 2006 using laboratory-scale equipment and irradiated for 3 years in themore » Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) to demonstrate and evaluate fuel performance under HTGR irradiation conditions. Post-irradiation examination was performed at INL and ORNL to study how the fuel behaved during irradiation, and to test fuel performance during exposure to elevated temperatures at or above temperatures that could occur during a depressurized conduction cooldown event. This report summarizes safety testing and post-safety testing PIE conducted at ORNL on loose particles extracted from irradiated AGR-1 Compact 4-4-2.« less

Fuel burnup analysis for IRIS reactor using MCNPX and WIMS-D5 codes

NASA Astrophysics Data System (ADS)

Amin, E. A.; Bashter, I. I.; Hassan, Nabil M.; Mustafa, S. S.

2017-02-01

International Reactor Innovative and Secure (IRIS) reactor is a compact power reactor designed with especial features. It contains Integral Fuel Burnable Absorber (IFBA). The core is heterogeneous both axially and radially. This work provides the full core burn up analysis for IRIS reactor using MCNPX and WIMDS-D5 codes. Criticality calculations, radial and axial power distributions and nuclear peaking factor at the different stages of burnup were studied. Effective multiplication factor values for the core were estimated by coupling MCNPX code with WIMS-D5 code and compared with SAS2H/KENO-V code values at different stages of burnup. The two calculation codes show good agreement and correlation. The values of radial and axial powers for the full core were also compared with published results given by SAS2H/KENO-V code (at the beginning and end of reactor operation). The behavior of both radial and axial power distribution is quiet similar to the other data published by SAS2H/KENO-V code. The peaking factor values estimated in the present work are close to its values calculated by SAS2H/KENO-V code.

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

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

Hatayama, Ariyoshi; Ogasawara, Masatada; Yamauchi, Michinori

1994-08-01

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

Design principles of a simple and safe 200-MW(thermal) nuclear district heating plant

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

Goetzmann, C.; Bittermann, D.; Gobel, A.

Kraftwerk Union AG has almost completed the development of a dedicated 200-MW(thermal) nuclear district heating plant to provide environmentally clean energy at a predictably low cost. The concept can easily be adapted to meet power requirements within the 100- to 500-MW(thermal) range. This technology is the product of the experience gained with large pressurized water reactor and boiling water reactor power plants, with respect to both plant and fuel performance. The major development task is that of achieving sufficiently low capital cost by tailoring components and systems designed for large plants to the specific requirements of district heating. These requirementsmore » are small absolute power, low temperatures and pressures, and modest load following, all of which result in the characteristics that are summarized. A fully integrated primary system with natural circulation permits a very compact reactor building containing all safety-related systems and components. Plant safety is essentially guaranteed by inherent features. The reactor containment is tightly fitted around the reactor pressure vessel in such a way that, in the event of any postulated coolant leak, the core cannot become uncovered, even temporarily. Shutdown is assured by gravity drop of the control rods mounted above the core. Decay heat is removed from the core by means of natural circulation via dedicated intermediate circuits of external aircoolers.« less

Comparative analysis of thorium and uranium fuel for transuranic recycle in a sodium cooled Fast Reactor

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

C. Fiorina; N. E. Stauff; F. Franceschini

2013-12-01

The present paper compares the reactor physics and transmutation performance of sodium-cooled Fast Reactors (FRs) for TRansUranic (TRU) burning with thorium (Th) or uranium (U) as fertile materials. The 1000 MWt Toshiba-Westinghouse Advanced Recycling Reactor (ARR) conceptual core has been used as benchmark for the comparison. Both burner and breakeven configurations sustained or started with a TRU supply, and assuming full actinide homogeneous recycle strategy, have been developed. State-of-the-art core physics tools have been employed to establish fuel inventory and reactor physics performances for equilibrium and transition cycles. Results show that Th fosters large improvements in the reactivity coefficients associatedmore » with coolant expansion and voiding, which enhances safety margins and, for a burner design, can be traded for maximizing the TRU burning rate. A trade-off of Th compared to U is the significantly larger fuel inventory required to achieve a breakeven design, which entails additional blankets at the detriment of core compactness as well as fuel manufacturing and separation requirements. The gamma field generated by the progeny of U-232 in the U bred from Th challenges fuel handling and manufacturing, but in case of full recycle, the high contents of Am and Cm in the transmutation fuel impose remote fuel operations regardless of the presence of U-232.« less

A miniature fuel reformer system for portable power sources

NASA Astrophysics Data System (ADS)

Dolanc, Gregor; Belavič, Darko; Hrovat, Marko; Hočevar, Stanko; Pohar, Andrej; Petrovčič, Janko; Musizza, Bojan

2014-12-01

A miniature methanol reformer system has been designed and built to technology readiness level exceeding a laboratory prototype. It is intended to feed fuel cells with electric power up to 100 W and contains a complete setup of the technological elements: catalytic reforming and PROX reactors, a combustor, evaporators, actuation and sensing elements, and a control unit. The system is engineered not only for performance and quality of the reformate, but also for its lightweight and compact design, seamless integration of elements, low internal electric consumption, and safety. In the paper, the design of the system is presented by focussing on its miniaturisation, integration, and process control.

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

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

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

1991-01-01

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

Staged Catalytic Partial Oxidation (SCPO) System - The State of Art Integrated Short Contact Time Hydrogen Generator

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

Ke Liu; Jin Ki Hong; Wei Wei

Research and development on hydrogen and syngas production have great potential in addressing the following challenges in energy arena: (1) produce more clean fuels to meet the increasing demands for clean liquid and gaseous fuels for transportation and electricity generation, (2) increase the efficiency of energy utilization for fuels and electricity production, and (3) eliminate the pollutants and decouple the link between energy utilization and greenhouse gas emissions in end-use systems [Song, 2006, Liu, Song & Subramani 2009]. In this project, GE Global Research (GEGR) collaborated with Argonne National Laboratory (ANL) and the University of Minnesota (UoMn), developed and demonstratedmore » a low cost, compact staged catalytic partial oxidation (SCPO) technology for distributed hydrogen generation. GEGR analyzed different reforming system designs, and developed the SCPO reforming system which is a unique technology staging and integrating 3 different short contact time catalysts in a single, compact reactor: catalytic partial oxidation (CPO), steam methane reforming (SMR) and water-gas shift (WGS). This integration is demonstrated via the fabrication of a prototype scale unit of each key technology. Approaches for key technical challenges of the program includes: · Analyzed different system designs · Designed the SCPO hydrogen production system · Developed highly active and sulfur tolerant CPO catalysts · Designed and built different pilot-scale reactors to demonstrate each key technology · Evaluated different operating conditions · Quantified the efficiency and cost of the system · Developed process design package (PDP) for 1500 kg H2/day distributed H2 production unit. SCPO met the Department of Energy (DOE) and GE’s cost and efficiency targets for distributed hydrogen production.« less

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

Compaction in optical fibres and fibre Bragg gratings under nuclear reactor high neutron and gamma fluence

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

Remy, L.; Cheymol, G.; Gusarov, A.

2015-07-01

In the framework of the development by CEA and SCK.CEN of a Fabry Perot Sensor (FPS) able to measure dimensional changes in Material Testing Reactor (MTR), the first goal of the SAKE 1 (Smirnof extention - Additional Key-tests on Elongation of glass fibres) irradiation was to measure the linear compaction of single mode fibres under high fast neutron fluence. Indeed, the compaction of the fibre which forms one side of the Fabry Perot cavity, may in particular cause a noticeable measurement error. An accurate quantification of this effect is then required to predict the radiation-induced drift and optimize the sensormore » design. To achieve this, an innovative approach was used. Approximately seventy uncoated fibre tips (length: 30 to 50 mm) have been prepared from several different fibre samples and were installed in the SCK.CEN BR2 reactor (Mol Belgium). After 22 days of irradiation a total fast (E > 1 MeV) fluence of 3 to 5x10{sup 19} n{sub fast}/cm{sup 2}, depending on the sample location, was accumulated. The temperature during irradiation was 291 deg. C, which is not far from the condition of the intended FPS use. A precise measurement of each fibre tip length was made before the irradiation and compared to the post irradiation measurement highlighting a decrease of the fibres' length corresponding to about 0.25% of linear compaction. The amplitude of the changes is independent of the capsule, which could mean that the compaction effect saturates even at the lowest considered fluence. In the prospect of performing distributed temperature measurement in MTR, several fibre Bragg gratings written using a femtosecond laser have been also irradiated. All the gratings were written in radiation hardened fibres, and underwent an additional treatment with a procedure enhancing their resistance to ionizing radiations. A special mounting made it possible to test the reflection and the transmission of the gratings on fibre samples cut down to 30 to 50 mm. The comparison of measurements made before and after the irradiation, at the same temperature, allowed us to measure the loss in reflectivity as well as the Bragg wavelength drift. The results are quite promising for some of the investigated gratings. (authors)« 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.

STATUS OF TRISO FUEL IRRADIATIONS IN THE ADVANCED TEST REACTOR SUPPORTING HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGNS

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

Davenport, Michael; Petti, D. A.; Palmer, Joe

2016-11-01

The United States Department of Energy’s Advanced Reactor Technologies (ART) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program is irradiating up to seven low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The experiments will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the experimentsmore » are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of several independent capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and was completed in November 2009. The second experiment (AGR-2) started irradiation in June 2010 and completed in October 2013. The third and fourth experiments have been combined into a single experiment designated (AGR-3/4), which started its irradiation in December 2011 and completed in April 2014. Since the purpose of this experiment was to provide data on fission product migration and retention in the NGNP reactor, the design of this experiment was significantly different from the first two experiments, though the control and monitoring systems are very similar. The final experiment, AGR-5/6/7, is scheduled to begin irradiation in early summer 2017.« 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

Westinghouse Small Modular Reactor nuclear steam supply system design

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

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

2012-07-01

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

ARC: A compact, high-field, disassemblable fusion nuclear science facility and demonstration power plant

NASA Astrophysics Data System (ADS)

Sorbom, Brandon; Ball, Justin; Palmer, Timothy; Mangiarotti, Franco; Sierchio, Jennifer; Bonoli, Paul; Kasten, Cale; Sutherland, Derek; Barnard, Harold; Haakonsen, Christian; Goh, Jon; Sung, Choongki; Whyte, Dennis

2014-10-01

The Affordable, Robust, Compact (ARC) reactor conceptual design aims to reduce the size, cost, and complexity of a combined Fusion Nuclear Science Facility (FNSF) and demonstration fusion pilot power plant. ARC is a 270 MWe tokamak reactor with a major radius of 3.3 m, a minor radius of 1.1 m, and an on-axis magnetic field of 9.2 T. ARC has Rare Earth Barium Copper Oxide (REBCO) superconducting toroidal field coils with joints to allow disassembly, allowing for removal and replacement of the vacuum vessel as a single component. Inboard-launched current drive of 25 MW LHRF power and 13.6 MW ICRF power is used to provide a robust, steady state core plasma far from disruptive limits. ARC uses an all-liquid blanket, consisting of low pressure, slowly flowing Fluorine Lithium Beryllium (FLiBe) molten salt. The liquid blanket acts as a working fluid, coolant, and tritium breeder, and minimizes the solid material that can become activated. The large temperature range over which FLiBe is liquid permits blanket operation at 800-900 K with single phase fluid cooling and allows use of a high-efficiency Brayton cycle for electricity production in the secondary coolant loop.

METHOD AND APPARATUS FOR REACTOR SAFETY CONTROL

DOEpatents

Huston, N.E.

1961-06-01

A self-contained nuclear reactor fuse controlled device tron absorbing material, normally in a compact form but which can be expanded into an extended form presenting a large surface for neutron absorption when triggered by an increase in neutron flux, is described.

PIE on Safety-Tested AGR-1 Compact 5-1-1

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

Hunn, John D.; Morris, Robert Noel; Baldwin, Charles A.

Post-irradiation examination (PIE) is being performed in support of tristructural isotropic (TRISO) coated particle fuel development and qualification for High-Temperature Gas-cooled Reactors (HTGRs). AGR-1 was the first in a series of TRISO fuel irradiation experiments initiated in 2006 under the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program; this work continues to be funded by the Department of Energy's Office of Nuclear Energy as part of the Advanced Reactor Technologies (ART) initiative. AGR-1 fuel compacts were fabricated at Oak Ridge National Laboratory (ORNL) in 2006 and irradiated for three years in the Idaho National Laboratory (INL) Advanced Test Reactormore » (ATR) to demonstrate and evaluate fuel performance under HTGR irradiation conditions. PIE is being performed at INL and ORNL to study how the fuel behaved during irradiation, and to examine fuel performance during exposure to elevated temperatures at or above temperatures that could occur during a depressurized conduction cooldown event. This report summarizes safety testing of irradiated AGR-1 Compact 5-1-1 in the ORNL Core Conduction Cooldown Test Facility (CCCTF) and post-safety testing PIE.« less

Fabrication of fuel pin assemblies, phase 3

NASA Technical Reports Server (NTRS)

Keeton, A. R.; Stemann, L. G.

1972-01-01

Five full size and eight reduced length fuel pins were fabricated for irradiation testing to evaluate design concepts for a fast spectrum lithium cooled compact space power reactor. These assemblies consisted of uranium mononitride fuel pellets encased in a T-111 (Ta-8W-2Hf) clad with a tungsten barrier separating fuel and clad. Fabrication procedures were fully qualified by process development and assembly qualification tests. Detailed specifications and procedures were written for the fabrication and assembly of prototype fuel pins.

Estimate of radiation release from MIT reactor with un-finned LEU core during Maximum Hypothetical Accident

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

Sun, Kaichao; Hu, Lin-wen; Newton, Thomas

2017-05-01

The Massachusetts Institute of Technology Reactor (MITR-II) is a research reactor in Cambridge, Massachusetts designed primarily for experiments using neutron beam and in-core irradiation facilities. At 6 MW, it delivers neutron flux and energy spectrum comparable to light water reactor (LWR) power reactors in a compact core using highly enriched uranium (HEU) fuel. In the framework of nonproliferation policy, the international community aims to minimize the use of HEU in civilian facilities. Within this context, research and test reactors have started a program to convert HEU fuel to low enriched uranium (LEU) fuel. A new type of LEU fuel basedmore » on a high density alloy of uranium and molybdenum (U-10Mo) is expected to allow the conversion of U.S. domestic high performance reactors like MITR. The current study focuses on the impacts of MITR Maximum Hypothetical Accident (MHA), which is also the Design Basis Accident (DBA), with LEU fuel. The MHA for the MITR is postulated to be a coolant flow blockage in the fuel element that contains the hottest fuel plate. It is assumed that the entire active portion of five fuel plates melts. The analysis shows that, within a 2-h period and by considering all the possible radiation sources and dose pathways, the overall off-site dose is 302.1 mrem (1 rem ¼ 0.01 Sv) Total Effective Dose Equivalent (TEDE) at 8 m exclusion area boundary (EAB) and a higher dose of 392.8 mrem TEDE is found at 21 m EAB. In all cases the dose remains below the 500 mrem total TEDE limit goal based on NUREG-1537 guidelines.« less

Identification and Quantification of Carbon Phases in Conversion Fuel for the Transient Reactor Test Facility

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

Steele, Robert; Mata, Angelica; Dunzik-Gougar, Mary Lou

2016-06-01

As part of an overall effort to convert US research reactors to low-enriched uranium (LEU) fuel use, a LEU conversion fuel is being designed for the Transient Reactor Test Facility (TREAT) at the Idaho National Laboratory. TREAT fuel compacts are comprised of UO2 fuel particles in a graphitic matrix material. In order to refine heat transfer modeling, as well as determine other physical and nuclear characteristics of the fuel, the amount and type of graphite and non-graphite phases within the fuel matrix must be known. In this study, we performed a series of complementary analyses, designed to allow detailed characterizationmore » of the graphite and phenolic resin based fuel matrix. Methods included Scanning Electron and Transmission Electron Microscopies, Raman spectroscopy, X-ray Diffraction, and Dual-Beam Focused Ion Beam Tomography. Our results indicate that no single characterization technique will yield all of the desired information; however, through the use of statistical and empirical data analysis, such as curve fitting, partial least squares regression, volume extrapolation and spectra peak ratios, a degree of certainty for the quantity of each phase can be obtained.« less

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

NASA Astrophysics Data System (ADS)

Polzin, Kurt A.; Godfroy, Thomas J.

2008-01-01

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

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

NASA Technical Reports Server (NTRS)

Polzin, Kurt A.; Godfroy, Thomas J.

2008-01-01

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

An analysis of nuclear fuel burnup in the AGR-1 TRISO fuel experiment using gamma spectrometry, mass spectrometry, and computational simulation techniques

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

Harp, Jason M.; Demkowicz, Paul A.; Winston, Philip L.

AGR 1 was the first in a series of experiments designed to test US TRISO fuel under high temperature gas-cooled reactor irradiation conditions. This experiment was irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) and is currently undergoing post irradiation examination (PIE) at INL and Oak Ridge National Laboratory. One component of the AGR 1 PIE is the experimental evaluation of the burnup of the fuel by two separate techniques. Gamma spectrometry was used to non destructively evaluate the burnup of all 72 of the TRISO fuel compacts that comprised the AGR 1 experiment. Two methodsmore » for evaluating burnup by gamma spectrometry were developed, one based on the Cs 137 activity and the other based on the ratio of Cs 134 and Cs 137 activities. Burnup values determined from both methods compared well with the values predicted from simulations. The highest measured burnup was 20.1% FIMA for the direct method and 20.0% FIMA for the ratio method (compared to 19.56% FIMA from simulations). An advantage of the ratio method is that the burnup of the cylindrical fuel compacts can determined in small (2.5 mm) axial increments and an axial burnup profile can be produced. Destructive chemical analysis by inductively coupled mass spectrometry (ICP MS) was then performed on selected compacts that were representative of the expected range of fuel burnups in the experiment to compare with the burnup values determined by gamma spectrometry. The compacts analyzed by mass spectrometry had a burnup range of 19.3% FIMA to 10.7% FIMA. The mass spectrometry evaluation of burnup for the four compacts agreed well with the gamma spectrometry burnup evaluations and the expected burnup from simulation. For all four compacts analyzed by mass spectrometry, the maximum range in the three experimentally determined values and the predicted value was 6% or less. Furthermore, the results confirm the accuracy of the nondestructive burnup evaluation from gamma spectrometry for TRISO fuel compacts across a burnup range of approximately 10 to 20% FIMA and also validate the approach used in the physics simulation of the AGR 1 experiment.« less

An analysis of nuclear fuel burnup in the AGR-1 TRISO fuel experiment using gamma spectrometry, mass spectrometry, and computational simulation techniques

DOE PAGES

Harp, Jason M.; Demkowicz, Paul A.; Winston, Philip L.; ...

2014-09-03

AGR 1 was the first in a series of experiments designed to test US TRISO fuel under high temperature gas-cooled reactor irradiation conditions. This experiment was irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) and is currently undergoing post irradiation examination (PIE) at INL and Oak Ridge National Laboratory. One component of the AGR 1 PIE is the experimental evaluation of the burnup of the fuel by two separate techniques. Gamma spectrometry was used to non destructively evaluate the burnup of all 72 of the TRISO fuel compacts that comprised the AGR 1 experiment. Two methodsmore » for evaluating burnup by gamma spectrometry were developed, one based on the Cs 137 activity and the other based on the ratio of Cs 134 and Cs 137 activities. Burnup values determined from both methods compared well with the values predicted from simulations. The highest measured burnup was 20.1% FIMA for the direct method and 20.0% FIMA for the ratio method (compared to 19.56% FIMA from simulations). An advantage of the ratio method is that the burnup of the cylindrical fuel compacts can determined in small (2.5 mm) axial increments and an axial burnup profile can be produced. Destructive chemical analysis by inductively coupled mass spectrometry (ICP MS) was then performed on selected compacts that were representative of the expected range of fuel burnups in the experiment to compare with the burnup values determined by gamma spectrometry. The compacts analyzed by mass spectrometry had a burnup range of 19.3% FIMA to 10.7% FIMA. The mass spectrometry evaluation of burnup for the four compacts agreed well with the gamma spectrometry burnup evaluations and the expected burnup from simulation. For all four compacts analyzed by mass spectrometry, the maximum range in the three experimentally determined values and the predicted value was 6% or less. Furthermore, the results confirm the accuracy of the nondestructive burnup evaluation from gamma spectrometry for TRISO fuel compacts across a burnup range of approximately 10 to 20% FIMA and also validate the approach used in the physics simulation of the AGR 1 experiment.« 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.

Design of an integrated fuel processor for residential PEMFCs applications

NASA Astrophysics Data System (ADS)

Seo, Yu Taek; Seo, Dong Joo; Jeong, Jin Hyeok; Yoon, Wang Lai

KIER has been developing a novel fuel processing system to provide hydrogen rich gas to residential PEMFCs system. For the effective design of a compact hydrogen production system, each unit process for steam reforming and water gas shift, has a steam generator and internal heat exchangers which are thermally and physically integrated into a single packaged hardware system. The newly designed fuel processor (prototype II) showed a thermal efficiency of 78% as a HHV basis with methane conversion of 89%. The preferential oxidation unit with two staged cascade reactors, reduces, the CO concentration to below 10 ppm without complicated temperature control hardware, which is the prerequisite CO limit for the PEMFC stack. After we achieve the initial performance of the fuel processor, partial load operation was carried out to test the performance and reliability of the fuel processor at various loads. The stability of the fuel processor was also demonstrated for three successive days with a stable composition of product gas and thermal efficiency. The CO concentration remained below 10 ppm during the test period and confirmed the stable performance of the two-stage PrOx reactors.

Retarding field analyzer for ion energy distribution measurements at a radio-frequency biased electrode

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

Gahan, D.; Hopkins, M. B.; Dolinaj, B.

2008-03-15

A retarding field energy analyzer designed to measure ion energy distributions impacting a radio-frequency biased electrode in a plasma discharge is examined. The analyzer is compact so that the need for differential pumping is avoided. The analyzer is designed to sit on the electrode surface, in place of the substrate, and the signal cables are fed out through the reactor side port. This prevents the need for modifications to the rf electrode--as is normally the case for analyzers built into such electrodes. The capabilities of the analyzer are demonstrated through experiments with various electrode bias conditions in an inductively coupledmore » plasma reactor. The electrode is initially grounded and the measured distributions are validated with the Langmuir probe measurements of the plasma potential. Ion energy distributions are then given for various rf bias voltage levels, discharge pressures, rf bias frequencies - 500 kHz to 30 MHz, and rf bias waveforms - sinusoidal, square, and dual frequency.« less

Simultaneous C and N removal from saline salmon effluents in filter reactors comprising anoxic-anaerobic-aerobic processes: effect of recycle ratio.

PubMed

Giustinianovich, Elisa A; Aspé, Estrella R; Huiliñir, César E; Roeckel, Marlene D

2014-01-01

Salmon processing generates saline effluents with high protein load. To treat these effluents, three compact tubular filter reactors were installed and an integrated anoxic/anaerobic/aerobic process was developed with recycling flow from the reactor's exit to the inlet stream in order to save organic matter (OM) for denitrification. The reactors were aerated in the upper section with recycle ratios (RR) of 0, 2, and 10, respectively, at 30°C. A tubular reactor behave as a plug flow reactor when RR = 0, and as a mixed flow reactor when recycle increases, thus, different RR values were used to evaluate how it affects the product distribution and the global performance. Diluted salmon process effluent was prepared as substrate. Using loads of 1.0 kg COD m(-3)d(-1) and 0.15 kg total Kjeldahl nitrogen (TKN) m(-3)d(-1) at HRT of 2 d, 100% removal efficiencies for nitrite and nitrate were achieved in the anoxic-denitrifying section without effect of the dissolved oxygen in the recycled flow on denitrification. Removals >98% for total organic carbon (TOC) was achieved in the three reactors. The RR had no effect on the TOC removal; nevertheless a higher efficiency in total nitrogen removal in the reactor with the highest recycle ratio was observed: 94.3% for RR = 10 and 46.6% for RR = 2. Results showed that the proposed layout with an alternative distribution in a compact reactor can efficiently treat high organic carbon and nitrogen concentrations from a saline fish effluent with OM savings in denitrification.

Nuclear design analysis of square-lattice honeycomb space nuclear rocket engine

NASA Astrophysics Data System (ADS)

Widargo, Reza; Anghaie, Samim

1999-01-01

The square-lattice honeycomb reactor is designed based on a cylindrical core that is determined to have critical diameter and length of 0.50 m and 0.50 c, respectively. A 0.10-cm thick radial graphite reflector, in addition to a 0.20-m thick axial graphite reflector are used to reduce neutron leakage from the reactor. The core is fueled with solid solution of 93% enriched (U, Zr, Nb)C, which is one of several ternary uranium carbides that are considered for this concept. The fuel is to be fabricated as 2 mm grooved (U, Zr, Nb)C wafers. The fuel wafers are used to form square-lattice honeycomb fuel assemblies, 0.10 m in length with 30% cross-sectional flow area. Five fuel assemblies are stacked up axially to form the reactor core. Based on the 30% void fraction, the width of the square flow channel is about 1.3 mm. The hydrogen propellant is passed through these flow channels and removes the heat from the reactor core. To perform nuclear design analysis, a series of neutron transport and diffusion codes are used. The preliminary results are obtained using a simple four-group cross-section model. To optimize the nuclear design, the fuel densities are varied for each assembly. Tantalum, hafnium and tungsten are considered and used as a replacement for niobium in fuel material to provide water submersion sub-criticality for the reactor. Axial and radial neutron flux and power density distributions are calculated for the core. Results of the neutronic analysis indicate that the core has a relatively fast spectrum. From the results of the thermal hydraulic analyses, eight axial temperature zones are chosen for the calculation of group average cross-sections. An iterative process is conducted to couple the neutronic calculations with the thermal hydraulics calculations. Results of the nuclear design analysis indicate that a compact core can be designed based on ternary uranium carbide square-lattice honeycomb fuel. This design provides a relatively high thrust to weight ratio.

Further Development of Crack Growth Detection Techniques for US Test and Research Reactors

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

Kohse, Gordon; Carpenter, David M.; Ostrovsky, Yakov

One of the key issues facing Light Water Reactors (LWRs) in extending lifetimes beyond 60 years is characterizing the combined effect of irradiation and water chemistry on material degradation and failure. Irradiation Assisted Stress Corrosion Cracking (IASCC), in which a crack propagates in a susceptible material under stress in an aggressive environment, is a mechanism of particular concern. Full understanding of IASCC depends on real time crack growth data acquired under relevant irradiation conditions. Techniques to measure crack growth in actively loaded samples under irradiation have been developed outside the US - at the Halden Boiling Water Reactor, for example.more » Several types of IASCC tests have also been deployed at the MITR, including passively loaded crack growth measurements and actively loaded slow strain rate tests. However, there is not currently a facility available in the US to measure crack growth on actively loaded, pre-cracked specimens in LWR irradiation environments. A joint program between the Idaho National Laboratory (INL) and the Massachusetts Institute of Technology (MIT) Nuclear Reactor Laboratory (NRL) is currently underway to develop and demonstrate such a capability for US test and research reactors. Based on the Halden design, the samples will be loaded using miniature high pressure bellows and a compact loading mechanism, with crack length measured in real time using the switched Direct Current Potential Drop (DCPD) method. The basic design and initial mechanical testing of the load system and implementation of the DCPD method have been previously reported. This paper presents the results of initial autoclave testing at INL and the adaptation of the design for use in the high pressure, high temperature water loop at the MITR 6 MW research reactor, where an initial demonstration is planned in mid-2015. Materials considerations for the high pressure bellows are addressed. Design modifications to the loading mechanism required by the size constraints of the MITR water loop are described. The safety case for operation of the high pressure gas-driven bellows mechanism is also presented. Key issues are the design and response of systems to limit gas flow in the event of a high pressure gas leak in the in-core autoclave. Integrity of the autoclave must be maintained and reactivity effects due to voiding of the loop coolant must be shown to be within the reactor technical specifications. The technical development of the crack growth monitor for application in the INL Advanced Test Reactor or the MITR can act as a template for adaptation of this technology in other reactors. (authors)« less

FABRICATION OF TUBE TYPE FUEL ELEMENT FOR NUCLEAR REACTORS

DOEpatents

Loeb, E.; Nicklas, J.H.

1959-02-01

A method of fabricating a nuclear reactor fuel element is given. It consists essentially of fixing two tubes in concentric relationship with respect to one another to provide an annulus therebetween, filling the annulus with a fissionablematerial-containing powder, compacting the powder material within the annulus and closing the ends thereof. The powder material is further compacted by swaging the inner surface of the inner tube to increase its diameter while maintaining the original size of the outer tube. This process results in reduced fabrication costs of powdered fissionable material type fuel elements and a substantial reduction in the peak core temperatures while materially enhancing the heat removal characteristics.

Comparison between instrumented precracked Charpy and compact specimen tests of carbon steels

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

Nanstad, R.K.

1980-01-01

The General Atomic Company High Temperature Gas-Cooled Reactor (HTGR) is housed within a prestressed concrete reactor vessel (PCRV). Various carbon steel structural members serve as closures at penetrations in the vessel. A program of testing and evaluation is underway to determine the need for reference fracture toughness (K/sub IR/) and indexing procedures for these materials as described in Appendix G to Section III, ASME Code for light water reactor steels. The materials of interest are carbon steel forgings (SA508, Class 1) and plates (SA537, Classes 1 and 2) as well as weldments of these steels. The fracture toughness behavior ismore » characterized with instrumented precracked Charpy V-votch specimens (PCVN) - slow-bend and dynamic - and compact specimens (10-mm and 25-mm thicknesses) using both linear elastic (ASTM E399) and elastic-plastic (equivalent Energy and J-Integral) analytical procedures. For the dynamic PCVN tests, force-time traces are analyzed according to the procedures of the Pressure Vessel Research Council (PVRC)/Metal Properties Council (MPC). Testing and analytical procedures are discussed and PCVN results are compared to those obtained with compact specimens.« less

REACTOR

DOEpatents

Christy, R.F.

1961-07-25

A means is described for co-relating the essential physical requirements of a fission chain reaction in order that practical, compact, and easily controllable reactors can be built. These objects are obtained by employing a composition of fissionsble isotope and moderator in fluid form in which the amount of fissionsble isotcpe present governs the reaction. The size of the reactor is no longer a critical factor, the new criterion being the concentration of the fissionable isotope.

A Compact Torus Fusion Reactor Utilizing a Continuously Generated Strings of CT's. The CT String Reactor, CTSR.

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

Hartman, C W; Reisman, D B; McLean, H S

2007-05-30

A fusion reactor is described in which a moving string of mutually repelling compact toruses (alternating helicity, unidirectional Btheta) is generated by repetitive injection using a magnetized coaxial gun driven by continuous gun current with alternating poloidal field. An injected CT relaxes to a minimum magnetic energy equilibrium, moves into a compression cone, and enters a conducting cylinder where the plasma is heated to fusion-producing temperature. The CT then passes into a blanketed region where fusion energy is produced and, on emergence from the fusion region, the CT undergoes controlled expansion in an exit cone where an alternating poloidal fieldmore » opens the flux surfaces to directly recover the CT magnetic energy as current which is returned to the formation gun. The CT String Reactor (CTSTR) reactor satisfies all the necessary MHD stability requirements and is based on extrapolation of experimentally achieved formation, stability, and plasma confinement. It is supported by extensive 2D, MHD calculations. CTSTR employs minimal external fields supplied by normal conductors, and can produce high fusion power density with uniform wall loading. The geometric simplicity of CTSTR acts to minimize initial and maintenance costs, including periodic replacement of the reactor first wall.« less

Low-cost, compact, cooled photomultiplier assembly for use in magnetic fields up to 1400 Gauss

NASA Technical Reports Server (NTRS)

Patch, R. W.; Tashjian, R. A.; Jentner, T. A.

1975-01-01

Use of vortex tube for cooling and concentric shielding have produced smaller and more compact unit than was previously available. Future uses of device could include installation in gas chromatographs and mass spectrometers. Additional uses would include measurements and controls in magnetohydrodynamic power generators and fusion reactors.

Supercell Depletion Studies for Prismatic High Temperature Reactors

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

J. Ortensi

2012-10-01

The traditional two-step method of analysis is not accurate enough to represent the neutronic effects present in the prismatic high temperature reactor concept. The long range coupling of the various regions in high temperature reactors poses a set of challenges that are not seen in either LWRs or fast reactors. Unlike LWRs, which exhibit large, localized effects, the dominant effects in PMRs are, for the most part, distributed over larger regions, but with lower magnitude. The 1-D in-line treatment currently used in pebble bed reactor analysis is not sufficient because of the 2-D nature of the prismatic blocks. Considerable challengesmore » exist in the modeling of blocks in the vicinity of reflectors, which, for current small modular reactor designs with thin annular cores, include the majority of the blocks. Additional challenges involve the treatment of burnable poisons, operational and shutdown control rods. The use of a large domain for cross section preparation provides a better representation of the neutron spectrum, enables the proper modeling of BPs and CRs, allows the calculation of generalized equivalence theory parameters, and generates a relative power distribution that can be used in compact power reconstruction. The purpose of this paper is to quantify the effects of the reflector, burnable poison, and operational control rods on an LEU design and to delineate an analysis approach for the Idaho National Laboratory. This work concludes that the use of supercells should capture these long-range effects in the preparation of cross sections and along with a set of triangular meshes to treat BPs, and CRs a high fidelity neutronics computation is attainable.« less

AGR-1 Compact 1-3-1 Post-Irradiation Examination Results

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

Demkowicz, Paul Andrew

The Advanced Gas Reactor (AGR) Fuel Development and Qualification Program was established to perform the requisite research and development on tristructural isotropic (TRISO) coated particle fuel to support deployment of a high-temperature gas-cooled reactor (HTGR). The work continues as part of the Advanced Reactor Technologies (ART) TRISO Fuel program. The overarching program goal is to provide a baseline fuel qualification data set to support licensing and operation of an HTGR. To achieve these goals, the program includes the elements of fuel fabrication, irradiation, post-irradiation examination (PIE) and safety testing, fuel performance modeling, and fission product transport (INL 2015). A seriesmore » of fuel irradiation experiments is being planned and conducted in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). These experiments will provide data on fuel performance under irradiation, support fuel process development, qualify the fuel for normal operating conditions, provide irradiated fuel for safety testing, and support the development of fuel performance and fission product transport models. The first of these irradiation tests, designated AGR-1, began in the ATR in December 2006 and ended in November 2009. This experiment was conducted primarily to act as a shakedown test of the multicapsule test train design and provide early data on fuel performance for use in fuel fabrication process development. It also provided samples for post-irradiation safety testing, where fission product retention of the fuel at high temperatures will be experimentally measured. The capsule design and details of the AGR-1 experiment have been presented previously (Grover, Petti, and Maki 2010, Maki 2009).« less

AGR-1 Compact 5-3-1 Post-Irradiation Examination Results

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

Demkowicz, Paul; Harp, Jason; Winston, Phil

The Advanced Gas Reactor (AGR) Fuel Development and Qualification Program was established to perform the requisite research and development on tristructural isotropic (TRISO) coated particle fuel to support deployment of a high-temperature gas-cooled reactor (HTGR). The work continues as part of the Advanced Reactor Technologies (ART) TRISO Fuel program. The overarching program goal is to provide a baseline fuel qualification data set to support licensing and operation of an HTGR. To achieve these goals, the program includes the elements of fuel fabrication, irradiation, post-irradiation examination (PIE) and safety testing, fuel performance, and fission product transport (INL 2015). A series ofmore » fuel irradiation experiments is being planned and conducted in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). These experiments will provide data on fuel performance under irradiation, support fuel process development, qualify the fuel for normal operating conditions, provide irradiated fuel for safety testing, and support the development of fuel performance and fission product transport models. The first of these irradiation tests, designated AGR-1, began in the ATR in December 2006 and ended in November 2009. This experiment was conducted primarily to act as a shakedown test of the multicapsule test train design and provide early data on fuel performance for use in fuel fabrication process development. It also provided samples for post-irradiation safety testing, where fission product retention of the fuel at high temperatures will be experimentally measured. The capsule design and details of the AGR-1 experiment have been presented previously.« less

DANSS: Detector of the reactor AntiNeutrino based on Solid Scintillator

NASA Astrophysics Data System (ADS)

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

2016-11-01

The DANSS project is aimed at creating a relatively compact neutrino spectrometer which does not contain any flammable or other dangerous liquids and may therefore be located very close to the core of an industrial power reactor. As a result, it is expected that high neutrino flux would provide about 15,000 IBD interactions per day in the detector with a sensitive volume of 1 m3. High segmentation of the plastic scintillator will allow to suppress a background down to a ~1% level. Numerous tests performed with a simplified pilot prototype DANSSino under a 3 GWth reactor of the Kalinin NPP have demonstrated operability of the chosen design. The DANSS detector surrounded with a composite shield is movable by means of a special lifting gear, varying the distance to the reactor core in a range from 10 m to 12 m. Due to this feature, it could be used not only for the reactor monitoring, but also for fundamental research including short-range neutrino oscillations to the sterile state. Supposing one-year measurement, the sensitivity to the oscillation parameters is expected to reach a level of sin2(2θnew) ~ 5 × 10-3 with Δ m2 ⊂ (0.02-5.0) eV2.

The High Field Ultra Low Aspect Ratio Tokamak (HF-ULART)

NASA Astrophysics Data System (ADS)

Ribeiro, Celso

2017-10-01

Recently, a medium-size HF-ULART has been proposed. The major objective is to explore the high beta and pressure under the high toroidal field, using present day technology. This might be one of pathway scenarios for a potential ultra-compact pulsed neutron source (UCP-NS) based on the spherical tokamak (ST) concept, which may lead to more steady-state NS or even to a fusion reactor, via realistic design scaling. The HF-ULART pulsed mode operation is created by quasi-simultaneous adiabatic compression (AC) in both minor and major radius of a very high beta plasma, possibly with further help of passive-wall stabilization, as envisaged in the RULART concept. This may help the revival of the studies of the AC technique in tokamaks, alongside the less compact and more complex ST-40 device, currently under construction. In addition, by similarities, studies in HF-ULART as a UCP-NS may also help to test the feasibility of the compact NS via the spheromak concept, which also uses the AC technique. Simulations of AC in HF-ULART plasmas will be presented.

One-dimensional MHD simulations of MTF systems with compact toroid targets and spherical liners

NASA Astrophysics Data System (ADS)

Khalzov, Ivan; Zindler, Ryan; Barsky, Sandra; Delage, Michael; Laberge, Michel

2017-10-01

One-dimensional (1D) MHD code is developed in General Fusion (GF) for coupled plasma-liner simulations in magnetized target fusion (MTF) systems. The main goal of these simulations is to search for optimal parameters of MTF reactor, in which spherical liquid metal liner compresses compact toroid plasma. The code uses Lagrangian description for both liner and plasma. The liner is represented as a set of spherical shells with fixed masses while plasma is discretized as a set of nested tori with circular cross sections and fixed number of particles between them. All physical fields are 1D functions of either spherical (liner) or small toroidal (plasma) radius. Motion of liner and plasma shells is calculated self-consistently based on applied forces and equations of state. Magnetic field is determined by 1D profiles of poloidal and toroidal fluxes - they are advected with shells and diffuse according to local resistivity, this also accounts for flux leakage into the liner. Different plasma transport models are implemented, this allows for comparison with ongoing GF experiments. Fusion power calculation is included into the code. We performed a series of parameter scans in order to establish the underlying dependencies of the MTF system and find the optimal reactor design point.

Readiness Review of BWXT for Fabrication of AGR 5/6/7 Compacts

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

Marshall, Douglas William; Sharp, Michelle Tracy

In support of preparations for fabricating compacts for the Advanced Gas Reactor (AGR) fuel qualification irradiation experiments (AGR-5/6/7), Idaho National Laboratory (INL) conducted a readiness review of the BWX Technology (BWXT) procedures, processes, and equipment associated with compact fabrication activities at the BWXT Nuclear Operations Group (BWXT-NOG) facility outside Lynchburg, VirginiaVA. The readiness review used quality assurance requirements taken from the American Society of Mechanical Engineers (ASME) Nuclear Quality Assurance Standard (NQA-1-2008/1a-2009) as a basis to assess readiness to start compact fabrication.

NEET Micro-Pocket Fission Detector. Final Project report

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

Unruh, T.; Rempe, Joy; McGregor, Douglas

2014-09-01

A collaboration between the Idaho National Laboratory (INL), the Kansas State University (KSU), and the French Alternative Energies and Atomic Energy Commission, Commissariat à l'Énergie Atomique et aux Energies Alternatives, (CEA), is funded by the Nuclear Energy Enabling Technologies (NEET) program to develop and test Micro-Pocket Fission Detectors (MPFDs), which are compact fission chambers capable of simultaneously measuring thermal neutron flux, fast neutron flux and temperature within a single package. When deployed, these sensors will significantly advance flux detection capabilities for irradiation tests in US Material Test Reactors (MTRs). Ultimately, evaluations may lead to a more compact, more accurate, andmore » longer lifetime flux sensor for critical mock-ups, and high performance reactors, allowing several Department of Energy Office of Nuclear Energy (DOE-NE) programs to obtain higher accuracy/higher resolution data from irradiation tests of candidate new fuels and materials. Specifically, deployment of MPFDs will address several challenges faced in irradiations performed at MTRs: Current fission chamber technologies do not offer the ability to measure fast flux, thermal flux and temperature within a single compact probe; MPFDs offer this option. MPFD construction is very different than current fission chamber construction; the use of high temperature materials allow MPFDs to be specifically tailored to survive harsh conditions encountered in-core of high performance MTRs. The higher accuracy, high fidelity data available from the compact MPFD will significantly enhance efforts to validate new high-fidelity reactor physics codes and new multi-scale, multi-physics codes. MPFDs can be built with variable sensitivities to survive the lifetime of an experiment or fuel assembly in some MTRs, allowing for more efficient and cost effective power monitoring. The small size of the MPFDs allows multiple sensors to be deployed, offering the potential to accurately measure the flux and temperature profiles in the reactor. This report summarizes the status at the end of year two of this three year project. As documented in this report, all planned accomplishments for developing this unique new, compact, multipurpose sensor have been completed.« less

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

NASA Astrophysics Data System (ADS)

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

2017-01-01

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

PROCESS OF FORMING POWDERED MATERIAL

DOEpatents

Glatter, J.; Schaner, B.E.

1961-07-14

A process of forming high-density compacts of a powdered ceramic material is described by agglomerating the powdered ceramic material with a heat- decompossble binder, adding a heat-decompossble lubricant to the agglomerated material, placing a quantity of the material into a die cavity, pressing the material to form a compact, pretreating the compacts in a nonoxidizing atmosphere to remove the binder and lubricant, and sintering the compacts. When this process is used for making nuclear reactor fuel elements, the ceramic material is an oxide powder of a fissionsble material and after forming, the compacts are placed in a cladding tube which is closed at its ends by vapor tight end caps, so that the sintered compacts are held in close contact with each other and with the interior wall of the cladding tube.

Fission Product Inventory and Burnup Evaluation of the AGR-2 Irradiation by Gamma Spectrometry

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

Harp, Jason Michael; Stempien, John Dennis; Demkowicz, Paul Andrew

Gamma spectrometry has been used to evaluate the burnup and fission product inventory of different components from the US Advanced Gas Reactor Fuel Development and Qualification Program's second TRISO-coated particle fuel irradiation test (AGR-2). TRISO fuel in this irradiation included both uranium carbide / uranium oxide (UCO) kernels and uranium oxide (UO 2) kernels. Four of the 6 capsules contained fuel from the US Advanced Gas Reactor program, and only those capsules will be discussed in this work. The inventories of gamma-emitting fission products from the fuel compacts, graphite compact holders, graphite spacers and test capsule shell were evaluated. Thesemore » data were used to measure the fractional release of fission products such as Cs-137, Cs-134, Eu-154, Ce-144, and Ag-110m from the compacts. The fraction of Ag-110m retained in the compacts ranged from 1.8% to full retention. Additionally, the activities of the radioactive cesium isotopes (Cs-134 and Cs-137) have been used to evaluate the burnup of all US TRISO fuel compacts in the irradiation. The experimental burnup evaluations compare favorably with burnups predicted from physics simulations. Predicted burnups for UCO compacts range from 7.26 to 13.15 % fission per initial metal atom (FIMA) and 9.01 to 10.69 % FIMA for UO 2 compacts. Measured burnup ranged from 7.3 to 13.1 % FIMA for UCO compacts and 8.5 to 10.6 % FIMA for UO 2 compacts. Results from gamma emission computed tomography performed on compacts and graphite holders that reveal the distribution of different fission products in a component will also be discussed. Gamma tomography of graphite holders was also used to locate the position of TRISO fuel particles suspected of having silicon carbide layer failures that lead to in-pile cesium release.« less

Fission Product Inventory and Burnup Evaluation of the AGR-2 Irradiation by Gamma Spectrometry

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

Harp, Jason M.; Demkowicz, Paul A.; Stempien, John D.

Gamma spectrometry has been used to evaluate the burnup and fission product inventory of different components from the US Advanced Gas Reactor Fuel Development and Qualification Program's second TRISO-coated particle fuel irradiation test (AGR-2). TRISO fuel in this irradiation included both uranium carbide / uranium oxide (UCO) kernels and uranium oxide (UO2) kernels. Four of the 6 capsules contained fuel from the US Advanced Gas Reactor program, and only those capsules will be discussed in this work. The inventories of gamma-emitting fission products from the fuel compacts, graphite compact holders, graphite spacers and test capsule shell were evaluated. These datamore » were used to measure the fractional release of fission products such as Cs-137, Cs-134, Eu-154, Ce-144, and Ag-110m from the compacts. The fraction of Ag-110m retained in the compacts ranged from 1.8% to full retention. Additionally, the activities of the radioactive cesium isotopes (Cs-134 and Cs-137) have been used to evaluate the burnup of all US TRISO fuel compacts in the irradiation. The experimental burnup evaluations compare favorably with burnups predicted from physics simulations. Predicted burnups for UCO compacts range from 7.26 to 13.15 % fission per initial metal atom (FIMA) and 9.01 to 10.69 % FIMA for UO2 compacts. Measured burnup ranged from 7.3 to 13.1 % FIMA for UCO compacts and 8.5 to 10.6 % FIMA for UO2 compacts. Results from gamma emission computed tomography performed on compacts and graphite holders that reveal the distribution of different fission products in a component will also be discussed. Gamma tomography of graphite holders was also used to locate the position of TRISO fuel particles suspected of having silicon carbide layer failures that lead to in-pile cesium release.« less

Diffusion-Welded Microchannel Heat Exchanger for Industrial Processes

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

Piyush Sabharwall; Denis E. Clark; Michael V. Glazoff

The goal of next generation reactors is to increase energy ef?ciency in the production of electricity and provide high-temperature heat for industrial processes. The ef?cient transfer of energy for industrial applications depends on the ability to incorporate effective heat exchangers between the nuclear heat transport system and the industrial process. The need for ef?ciency, compactness, and safety challenge the boundaries of existing heat exchanger technology. Various studies have been performed in attempts to update the secondary heat exchanger that is downstream of the primary heat exchanger, mostly because its performance is strongly tied to the ability to employ more ef?cientmore » industrial processes. Modern compact heat exchangers can provide high compactness, a measure of the ratio of surface area-to-volume of a heat exchange. The microchannel heat exchanger studied here is a plate-type, robust heat exchanger that combines compactness, low pressure drop, high effectiveness, and the ability to operate with a very large pressure differential between hot and cold sides. The plates are etched and thereafter joined by diffusion welding, resulting in extremely strong all-metal heat exchanger cores. After bonding, any number of core blocks can be welded together to provide the required ?ow capacity. This study explores the microchannel heat exchanger and draws conclusions about diffusion welding/bonding for joining heat exchanger plates, with both experimental and computational modeling, along with existing challenges and gaps. Also, presented is a thermal design method for determining overall design speci?cations for a microchannel printed circuit heat exchanger for both supercritical (24 MPa) and subcritical (17 MPa) Rankine power cycles.« less

Development and Deployment Assessment of a Melt-Down Proof Modular Micro Reactor (MDP-MMR)

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

Hawari, Ayman I.; Venneri, Francesco

The objective of this project is to perform feasibility assessment and technology gap analysis and establish a development roadmap for an innovative and highly compact Micro Modular Reactor (MMR) concept that integrates power production, power conversion and electricity generation in a single unit. The MMR is envisioned to use fully ceramic micro-encapsulated (FCM) fuel, a particularly robust form of TRISO fuel, and to be gas-cooled (e.g., He or CO 2) and capable of generating power in the range of 10 to 40 MW-thermal. It is designed to be absolutely melt-down proof (MDP) under all circumstances including complete loss of coolantmore » scenarios with no possible release of radioactive material, to be factory produced, to have a cycle length of greater than 20 years, and to be highly proliferation resistant. In addition, it will be transportable, retrievable and suitable for use in remote areas. As such, the MDP-MMR will represent a versatile reactor concept that is suitable for use in various applications including electricity generation, process heat utilization and propulsion.« less

Formation and sustainment of field reversed configuration (FRC) plasmas by spheromak merging and neutral beam injection

NASA Astrophysics Data System (ADS)

Yamada, Masaaki

2016-03-01

This paper briefly reviews a compact toroid reactor concept that addresses critical issues for forming, stabilizing and sustaining a field reversed configuration (FRC) with the use of plasma merging, plasma shaping, conducting shells, neutral beam injection (NBI). In this concept, an FRC plasma is generated by the merging of counter-helicity spheromaks produced by inductive discharges and sustained by the use of neutral beam injection (NBI). Plasma shaping, conducting shells, and the NBI would provide stabilization to global MHD modes. Although a specific FRC reactor design is outside the scope of the present paper, an example of a promising FRC reactor program is summarized based on the previously developed SPIRIT (Self-organized Plasmas by Induction, Reconnection and Injection Techniques) concept in order to connect this concept to the recently achieved the High Performance FRC plasmas obtained by Tri Alpha Energy [Binderbauer et al, Phys. Plasmas 22,056110, (2015)]. This paper includes a brief summary of the previous concept paper by M. Yamada et al, Plasma Fusion Res. 2, 004 (2007) and the recent experimental results from MRX.

Development of OTM Syngas Process and Testing of Syngas Derived Ultra-clean Fuels in Diesel Engines and Fuel Cells

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

E.T. Robinson; John Sirman; Prasad Apte

2005-05-01

This final report summarizes work accomplished in the Program from January 1, 2001 through December 31, 2004. Most of the key technical objectives for this program were achieved. A breakthrough material system has lead to the development of an OTM (oxygen transport membrane) compact planar reactor design capable of producing either syngas or hydrogen. The planar reactor shows significant advantages in thermal efficiency and a step change reduction in costs compared to either autothermal reforming or steam methane reforming with CO{sub 2} recovery. Syngas derived ultra-clean transportation fuels were tested in the Nuvera fuel cell modular pressurized reactor and inmore » International Truck and Engine single cylinder test engines. The studies compared emission and engine performance of conventional base fuels to various formulations of ultra-clean gasoline or diesel fuels. A proprietary BP oxygenate showed significant advantage in both applications for reducing emissions with minimal impact on performance. In addition, a study to evaluate new fuel formulations for an HCCI engine was completed.« less

Heat transfer in laminar flow along circular rods in infinite square arrays

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

Kim, J.H.; Li, W.H.

1988-02-01

The need to understand heat transfer characteristics over rods or tube bundles often arises in the design of compact heat exchangers and safety analysis of nuclear reactors. In particular, the fuel bundles of typical light water nuclear reactors are composed of a large number of circular rods arranged in square array pattern. The purpose of the present study is to analyze heat transfer characteristics of flow in such a multirod geometric configuration. The analysis given here will follow as closely as possible the method of Sparrow et al. who analyzed a similar problem for circular cylinders arranged in an equilateralmore » triangular array. The following major assumptions are made in the present analysis: (1) Flow is fully developed laminar flow paralleled to the axis of rods. (2) The axial profile of the surface heat flux to the fluid is uniform.(3) Thermodynamic properties are assumed constant.« less

Experimental physics characteristics of a heavy-metal-reflected fast-spectrum critical assembly

NASA Technical Reports Server (NTRS)

Heneveld, W. H.; Paschall, R. K.; Springer, T. H.; Swanson, V. A.; Thiele, A. W.; Tuttle, R. J.

1972-01-01

A zero-power critical assembly was designed, constructed, and operated for the purpose of conducting a series of benchmark experiments dealing with the physics characteristics of a UN-fueled, Li-cooled, Mo-reflected, drum-controlled compact fast reactor for use with a space-power electric conversion system. The range of the previous experimental investigations has been expanded to include the reactivity effects of:(1) surrounding the reactor with 15.24 cm (6 in.) of polyethylene, (2) reducing the heights of a portion of the upper and lower axial reflectors by factors of 2 and 4, (3) adding 45 kg of W to the core uniformly in two steps, (4) adding 9.54 kg of Ta to the core uniformly, and (5) inserting 2.3 kg of polyethylene into the core proper and determining the effect of a Ta addition on the polyethylene worth.

Fracture toughness testing of Linde 1092 reactor vessel welds in the transition range using Charpy-sized specimens

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

Pavinich, W.A.; Yoon, K.K.; Hour, K.Y.

1999-10-01

The present reference toughness method for predicting the change in fracture toughness can provide over estimates of these values because of uncertainties in initial RT{sub NDT} and shift correlations. It would be preferable to directly measure fracture toughness. However, until recently, no standard method was available to characterize fracture toughness in the transition range. ASTM E08 has developed a draft standard that shows promise for providing lower bound transition range fracture toughness using the master curve approach. This method has been successfully implemented using 1T compact fracture specimens. Combustion Engineering reactor vessel surveillance programs do not have compact fracture specimens.more » Therefore, the CE Owners Group developed a program to validate the master curve method for Charpy-sized and reconstituted Charpy-sized specimens for future application on irradiated specimens. This method was validated for Linde 1092 welds using unirradiated Charpy-sized and reconstituted Charpy-sized specimens by comparison of results with those from compact fracture specimens.« less

Square lattice honeycomb reactor for space power and propulsion

NASA Astrophysics Data System (ADS)

Gouw, Reza; Anghaie, Samim

2000-01-01

The most recent nuclear design study at the Innovative Nuclear Space Power and Propulsion Institute (INSPI) is the Moderated Square-Lattice Honeycomb (M-SLHC) reactor design utilizing the solid solution of ternary carbide fuels. The reactor is fueled with solid solution of 93% enriched (U,Zr,Nb)C. The square-lattice honeycomb design provides high strength and is amenable to the processing complexities of these ultrahigh temperature fuels. The optimum core configuration requires a balance between high specific impulse and thrust level performance, and maintaining the temperature and strength limits of the fuel. The M-SLHC design is based on a cylindrical core that has critical radius and length of 37 cm and 50 cm, respectively. This design utilized zirconium hydrate to act as moderator. The fuel sub-assemblies are designed as cylindrical tubes with 12 cm in diameter and 10 cm in length. Five fuel subassemblies are stacked up axially to form one complete fuel assembly. These fuel assemblies are then arranged in the circular arrangement to form two fuel regions. The first fuel region consists of six fuel assemblies, and 18 fuel assemblies for the second fuel region. A 10-cm radial beryllium reflector in addition to 10-cm top axial beryllium reflector is used to reduce neutron leakage from the system. To perform nuclear design analysis of the M-SLHC design, a series of neutron transport and diffusion codes are used. To optimize the system design, five axial regions are specified. In each axial region, temperature and fuel density are varied. The axial and radial power distributions for the system are calculated, as well as the axial and radial flux distributions. Temperature coefficients of the system are also calculated. A water submersion accident scenario is also analyzed for these systems. Results of the nuclear design analysis indicate that a compact core can be designed based on ternary uranium carbide square-lattice honeycomb fuel, which provides a relatively high thrust to weight ratio. .

Biodiesel production process intensification using a rotor-stator type generator of hydrodynamic cavitation.

PubMed

Crudo, Daniele; Bosco, Valentina; Cavaglià, Giuliano; Grillo, Giorgio; Mantegna, Stefano; Cravotto, Giancarlo

2016-11-01

Triglyceride transesterification for biodiesel production is a model reaction which is used to compare the conversion efficiency, yield, reaction time, energy consumption, scalability and cost estimation of different reactor technology and energy source. This work describes an efficient, fast and cost-effective procedure for biodiesel preparation using a rotating generator of hydrodynamic cavitation (HC). The base-catalyzed transesterification (methanol/sodium hydroxide) has been carried out using refined and bleached palm oil and waste vegetable cooking oil. The novel HC unit is a continuous rotor-stator type reactor in which reagents are directly fed into the controlled cavitation chamber. The high-speed rotation of the reactor creates micron-sized droplets of the immiscible reacting mixture leading to outstanding mass and heat transfer and enhancing the kinetics of the transesterification reaction which completes much more quickly than traditional methods. All the biodiesel samples obtained respect the ASTM standard and present fatty acid methyl ester contents of >99% m/m in both feedstocks. The electrical energy consumption of the HC reactor is 0.030kWh per L of produced crude biodiesel, making this innovative technology really quite competitive. The reactor can be easily scaled-up, from producing a few hundred to thousands of liters of biodiesel per hour while avoiding the risk of orifices clogging with oil impurities, which may occur in conventional HC reactors. Furthermore it requires minimal installation space due to its compact design, which enhances overall security. Copyright © 2016 Elsevier B.V. All rights reserved.

Mass tracking and material accounting in the Integral Fast Reactor (IFR)

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

Orechwa, Y.; Adams, C.H.; White, A.M.

1991-01-01

The Integral Fast Reactor (IFR) is a generic advanced liquid metal cooled reactor concept being developed at Argonne National Laboratory (ANL). There are a number of technical features of the IFR which contribute to its potential as a next-generation reactor. These are associated with large safety margins with regard to off-normal events involving the heat transport system, and the use of metallic fuel which makes possible the utilization of innovative fuel cycle processes. The latter feature permits fuel cycle closure the compact, low-cost reprocessing facilities, collocated with the reactor plant. These primary features are being demonstrated in the facilities atmore » ANL-West, utilizing Experimental Breeder Reactor 2 and the associated Fuel Cycle Facility (FCF) as an IFR prototype. The demonstration of this IFR prototype includes the design and implementation of the Mass-Tracking System (MTG). In this system, data from the operations of the FCF, including weights and batch-process parameters, are collected and maintained by the MTG running on distributed workstations. The components of the MTG System include: (1) an Oracle database manager with a Fortran interface, (2) a set of MTG Tasks'' which collect, manipulate and report data, (3) a set of MTG Terminal Sessions'' which provide some interactive control of the Tasks, and (4) a set of servers which manage the Tasks and which provide the communications link between the MTG System and Operator Control Stations, which control process equipment and monitoring devices within the FCF.« less

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

Tatli, Emre; Ferroni, Paolo; Mazzoccoli, Jason

The possible use of compact heat exchangers (HXs) in sodium-cooled fast reactors (SFR) employing a Brayton cycle is promising due to their high power density and resulting small volume in comparison with conventional shell-and-tube HXs. However, the small diameter of their channels makes them more susceptible to plugging due to Na2O deposition during accident conditions. Although cold traps are designed to reduce oxygen impurity levels in the sodium coolant, their failure, in conjunction with accidental air ingress into the sodium boundary, could result in coolant oxygen levels that are above the saturation limit in the cooler parts of the HXmore » channels. This can result in Na2O crystallization and the formation of solid deposits on cooled channel surfaces, limiting or even blocking coolant flow. The development of analysis tools capable of modeling the formation of these deposits in the presence of sodium flow will allow designers of SFRs to properly size the HX channels so that, in the scenario mentioned above, the reactor operator has sufficient time to detect and react to the affected HX. Until now, analytical methodologies to predict the formation of these deposits have been developed, but never implemented in a high-fidelity computational tool suited to modern reactor design techniques. This paper summarizes the challenges and the current status in the development of a Computational Fluid Dynamics (CFD) methodology to predict deposit formation, with particular emphasis on sensitivity studies on some parameters affecting deposition.« less

Online monitoring of the Osiris reactor with the Nucifer neutrino detector

NASA Astrophysics Data System (ADS)

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

2016-06-01

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

New Water Disinfection Technology for Earth and Space Applications as Part of the NPP Fellowship Research

NASA Technical Reports Server (NTRS)

SilvestryRodriquez, Nadia

2010-01-01

There is the need for a safe, low energy consuming and compact water disinfection technology to maintain water quality for human consumption. The design of the reactor should present no overheating and a constant temperature, with good electrical and optical performance for a UV water treatment system. The study assessed the use of UVA-LEDs to disinfectant water for MS2 Bacteriophage. The log reduction was sufficient to meet US EPA standards as a secondary disinfectant for maintaining water quality control. The study also explored possible inactivation of Pseudomonas aeruginosa and E. coli.

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

NASA Astrophysics Data System (ADS)

Powell, James; Maise, George; Paniagua, John

2006-01-01

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

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

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

Powell, James; Maise, George; Paniagua, John

2006-01-20

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

High-intensity power-resolved radiation imaging of an operational nuclear reactor.

PubMed

Beaumont, Jonathan S; Mellor, Matthew P; Villa, Mario; Joyce, Malcolm J

2015-10-09

Knowledge of the neutron distribution in a nuclear reactor is necessary to ensure the safe and efficient burnup of reactor fuel. Currently these measurements are performed by in-core systems in what are extremely hostile environments and in most reactor accident scenarios it is likely that these systems would be damaged. Here we present a compact and portable radiation imaging system with the ability to image high-intensity fast-neutron and gamma-ray fields simultaneously. This system has been deployed to image radiation fields emitted during the operation of a TRIGA test reactor allowing a spatial visualization of the internal reactor conditions to be obtained. The imaged flux in each case is found to scale linearly with reactor power indicating that this method may be used for power-resolved reactor monitoring and for the assay of ongoing nuclear criticalities in damaged nuclear reactors.

High-intensity power-resolved radiation imaging of an operational nuclear reactor

PubMed Central

Beaumont, Jonathan S.; Mellor, Matthew P.; Villa, Mario; Joyce, Malcolm J.

2015-01-01

Knowledge of the neutron distribution in a nuclear reactor is necessary to ensure the safe and efficient burnup of reactor fuel. Currently these measurements are performed by in-core systems in what are extremely hostile environments and in most reactor accident scenarios it is likely that these systems would be damaged. Here we present a compact and portable radiation imaging system with the ability to image high-intensity fast-neutron and gamma-ray fields simultaneously. This system has been deployed to image radiation fields emitted during the operation of a TRIGA test reactor allowing a spatial visualization of the internal reactor conditions to be obtained. The imaged flux in each case is found to scale linearly with reactor power indicating that this method may be used for power-resolved reactor monitoring and for the assay of ongoing nuclear criticalities in damaged nuclear reactors. PMID:26450669

Design and evaluation of a compact photocatalytic reactor for water treatment.

PubMed

Kete, Marko; Pliekhova, Olena; Matoh, Lev; Štangar, Urška Lavrenčič

2017-08-15

A compact reactor for photocatalytic oxidation and photocatalytic ozonation water treatment was developed and evaluated by using four model pollutants. Additionally, combinations of pollutants were evaluated. Specially produced Al 2 O 3 porous reticulated monolith foams served as TiO 2 carriers, offering a high surface area support. UV lamps were placed in the interior to achieve reduced dimensions of the reactor (12 cm in diameter × 20 cm in height). Despite its small size, the overall photocatalytic cleaning capacity was substantial. It was evaluated by measuring the degradation of LAS + PBIS and RB19 as representatives of surfactants and textile dyes, respectively. These contaminants are commonly found in household grey wastewater with phenol as a trace contaminant. Three different commercial photocatalysts and one mixture of photocatalysts (P25, P90, PC500 and P25 + PC500) were introduced in the sol-gel processing and immobilized on foamed Al 2 O 3 monoliths. RB19 and phenol were easily degradable, while LAS and PBIS were more resistant. The experiments were conducted at neutral-acidic pH because alkaline pH negatively influences both photocatalyic ozonation (PCOZ) and photocatalysis. The synergistic effect of PCOZ was generally much more expressed in mineralization reactions. Total organic carbon TOC half lives were in the range of between 13 and 43 min in the case of individual pollutants in double-deionized water. However, for the mixed pollutants in tap water, the TOC half-life only increased to 53 min with the most efficient catalyst (P90). In comparison to photocatalysis, the PCOZ process is more suitable for treating wastewater with a high loading of organic pollutants due to its higher cleaning capacity. Therefore, PCOZ may prove more effective in industrial applications.

THERMAL FISSION REACTOR COMPOSITIONS AND METHOD OF FABRICATING SAME

DOEpatents

Blainey, A.

1959-10-01

A body is presented for use in a thermal fission reactor comprising a sintered compressed mass of a substance of the group consisting of uranium, thorium, and oxides and carbides of uranium and thorium, enclosed in an envelope of a sintered, compacted, heat-conductive material of the group consisting of beryllium, zirconium, and oxides and carbides of beryllium and zirconium.

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.

Compact light-emitting diode optical fiber immobilized TiO2 reactor for photocatalytic water treatment.

PubMed

O'Neal Tugaoen, Heather; Garcia-Segura, Sergi; Hristovski, Kiril; Westerhoff, Paul

2018-02-01

A key barrier to implementing photocatalysis is delivering light to photocatalysts that are in contact with aqueous pollutants. Slurry photocatalyst systems suffer from poor light penetration and require post-treatment to separate the catalyst. The alternative is to deposit photocatalysts on fixed films and deliver light onto the surface or the backside of the attached catalysts. In this study, TiO 2 -coated quartz optical fibers were coupled to light emitting diodes (OF/LED) to improve in situ light delivery. Design factors and mechanisms studied for OF/LEDs in a flow-through reactor included: (i) the influence of number of LED sources coupled to fibers and (ii) the use of multiple optical fibers bundled to a single LED. The light delivery mechanism from the optical fibers into the TiO 2 coatings is thoroughly discussed. To demonstrate influence of design variables, experiments were conducted in the reactor using the chlorinated pollutant para-chlorobenzoic acid (pCBA). From the degradation kinetics of pCBA, the quantum efficiencies (Φ) of oxidation and electrical energies per order (E EO ) were determined. The use of TiO 2 coated optical fiber bundles reduced the energy requirements to deliver photons and increased available surface area, which improved Φ and enhanced oxidative pollutant removal performance (E EO ). Copyright © 2017 Elsevier B.V. All rights reserved.

Innovative self-powered submersible microbial electrolysis cell (SMEC) for biohydrogen production from anaerobic reactors.

PubMed

Zhang, Yifeng; Angelidaki, Irini

2012-05-15

A self-powered submersible microbial electrolysis cell (SMEC), in which a specially designed anode chamber and external electricity supply were not needed, was developed for in situ biohydrogen production from anaerobic reactors. In batch experiments, the hydrogen production rate reached 17.8 mL/L/d at the initial acetate concentration of 410 mg/L (5 mM), while the cathodic hydrogen recovery ( [Formula: see text] ) and overall systemic coulombic efficiency (CE(os)) were 93% and 28%, respectively, and the systemic hydrogen yield ( [Formula: see text] ) peaked at 1.27 mol-H(2)/mol-acetate. The hydrogen production increased along with acetate and buffer concentration. The highest hydrogen production rate of 32.2 mL/L/d and [Formula: see text] of 1.43 mol-H(2)/mol-acetate were achieved at 1640 mg/L (20 mM) acetate and 100 mM phosphate buffer. Further evaluation of the reactor under single electricity-generating or hydrogen-producing mode indicated that further improvement of voltage output and reduction of electron losses were essential for efficient hydrogen generation. In addition, alternate exchanging the electricity-assisting and hydrogen-producing function between the two cell units of the SMEC was found to be an effective approach to inhibit methanogens. Furthermore, 16S rRNA genes analysis showed that this special operation strategy resulted same microbial community structures in the anodic biofilms of the two cell units. The simple, compact and in situ applicable SMEC offers new opportunities for reactor design for a microbial electricity-assisted biohydrogen production system. Copyright © 2012 Elsevier Ltd. All rights reserved.

A mini-cavity probe reactor.

NASA Technical Reports Server (NTRS)

Hyland, R. E.

1971-01-01

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

Compact Heat Exchanger Design and Testing for Advanced Reactors and Advanced Power Cycles

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

Sun, Xiaodong; Zhang, Xiaoqin; Christensen, Richard

The goal of the proposed research is to demonstrate the thermal hydraulic performance of innovative surface geometries in compact heat exchangers used as intermediate heat exchangers (IHXs) and recuperators for the supercritical carbon dioxide (s-CO 2) Brayton cycle. Printed-circuit heat exchangers (PCHEs) are the primary compact heat exchangers of interest. The overall objectives are: To develop optimized PCHE designs for different working fluid combinations including helium to s-CO 2, liquid salt to s-CO 2, sodium to s-CO 2, and liquid salt to helium; To experimentally and numerically investigate thermal performance, thermal stress and failure mechanism of PCHEs under various transients;more » and To study diffusion bonding techniques for elevated-temperature alloys and examine post-test material integrity of the PCHEs. The project objectives were accomplished by defining and executing five different tasks corresponding to these specific objectives. The first task involved a thorough literature review and a selection of IHX candidates with different surface geometries as well as a summary of prototypic operational conditions. The second task involved optimization of PCHE design with numerical analyses of thermal-hydraulic performances and mechanical integrity. The subsequent task dealt with the development of testing facilities and engineering design of PCHE to be tested in s-CO 2 fluid conditions. The next task involved experimental investigation and validation of the thermal-hydraulic performances and thermal stress distribution of prototype PCHEs manufactured with particular surface geometries. The last task involved an investigation of diffusion bonding process and posttest destructive testing to validate mechanical design methods adopted in the design process. The experimental work utilized the two test facilities at The Ohio State University (OSU) including one existing High-Temperature Helium Test Facility (HTHF) and the newly developed s-CO 2 test loop (STL) facility and s-CO 2 test facility at University of Wisconsin – Madison (UW).« less

FALCON reactor-pumped laser description and program overview

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

None

1989-12-01

The FALCON (Fission Activated Laser CONcept) reactor-pumped laser program at Sandia National Laboratories is examining the feasibility of high-power systems pumped directly by the energy from a nuclear reactor. In this concept we use the highly energetic fission fragments from neutron induced fission to excite a large volume laser medium. This technology has the potential to scale to extremely large optical power outputs in a primarily self-powered device. A laser system of this type could also be relatively compact and capable of long run times without refueling.

A compact self-flowing lithium system for use in an industrial neutron source

NASA Astrophysics Data System (ADS)

Kalathiparambil, Kishor Kumar; Szott, Matthew; Jurczyk, Brian; Ahn, Chisung; Ruzic, David

2016-10-01

A compact trench module to flow liquid lithium in closed loops for handling high heat and particle flux have been fabricated and tested at UIUC. The module was designed to demonstrate the proof of concept in utilizing liquid metals for two principal objectives: i) as self-healing low Z plasma facing components, which is expected to solve the issues facing the current high Z components and ii) using flowing lithium as an MeV-level neutron source. A continuously flowing lithium loop ensures a fresh lithium interface and also accommodate a higher concentration of D, enabling advanced D-Li reactions without using any radioactive tritium. Such a system is expected to have a base yield of 10e7 n/s. For both the applications, the key success factor of the module is attaining the necessary high flow velocity of the lithium especially over the impact area, which will be the disruptive plasma events in fusion reactors and the incident ion beam for the neutron beam source. This was achieved by the efficient shaping of the trenches to exploit the nozzle effect in liquid flow. The compactness of the module, which can also be scaled as desired, was fulfilled by the use of high Tc permanent magnets and air cooled channels attained the necessary temperature gradient for driving the lithium. The design considerations and parameters, experimental arrangements involving lithium filling and attaining flow, data and results obtained will be elaborated. DOE SBIR project DE-SC0013861.

Fabrication methods and anisotropic properties of graphite matrix compacts for use in HTGR

NASA Astrophysics Data System (ADS)

Yeo, Sunghwan; Yun, Jihae; Kim, Sungok; Cho, Moon Sung; Lee, Young-Woo

2018-02-01

This study investigated the anisotropic microstructural, mechanical, and thermal properties of fabricated graphite matrix prismatic compacts for High Temperature Gas Cooled Reactor (HTGR) fuel. When the observed alignment of graphite grains and the coke derived from phenolic resin is in the transverse direction, the result is severely anisotropic thermal properties. Compacts with such orientation in the transverse direction exhibited increases of thermal expansion and conductivity up to 5.8 times and 4.82 times, respectively, more than those in the axial direction. The formation of pores due to the pyrolysis of phenolic resin was observed predominantly on upper region of the fabricated compacts. This anisotropic pore formation created anisotropic Vickers hardness on the planes with different directions.

Generating Breathable Air Through Dissociation of N2O

NASA Technical Reports Server (NTRS)

Zubrin, Robert; Frankie, Brian

2006-01-01

A nitrous oxide-based oxygen-supply system (NOBOSS) is an apparatus in which a breathable mixture comprising 2/3 volume parts of N2 and 1/3 volume part of O2 is generated through dissociation of N2O. The NOBOSS concept can be adapted to a variety of applications in which there are requirements for relatively compact, lightweight systems to supply breathable air. These could include air-supply systems for firefighters, divers, astronauts, and workers who must be protected against biological and chemical hazards. A NOBOSS stands in contrast to compressed-gas and cryogenic air-supply systems. Compressed-gas systems necessarily include massive tanks that can hold only relatively small amounts of gases. Alternatively, gases can be stored compactly in greater quantities and at low pressures when they are liquefied, but then cryogenic equipment is needed to maintain them in liquid form. Overcoming the disadvantages of both compressed-gas and cryogenic systems, the NOBOSS exploits the fact that N2O can be stored in liquid form at room temperature and moderate pressure. The mass of N2O that can be stored in a tank of a given mass is about 20 times the mass of compressed air that can be stored in a tank of equal mass. In a NOBOSS, N2O is exothermically dissociated to N2 and O2 in a main catalytic reactor. In order to ensure the dissociation of N2O to the maximum possible extent, the temperature of the reactor must be kept above 400 C. At the same time, to minimize concentrations of nitrogen oxides (which are toxic), it is necessary to keep the reactor temperature at or below 540 C. To keep the temperature within the required range throughout the reactor and, in particular, to prevent the formation of hot spots that would be generated by local concentrations of the exothermic dissociation reaction, the N2O is introduced into the reactor through an injector tube that features carefully spaced holes to distribute the input flow of N2O widely throughout the reactor. A NOBOSS includes one or more "destroyer" subsystems for removing any nitrogen oxides that remain downstream of the main N2O-dissociation reactor. A destroyer includes a carbon bed in series with a catalytic reactor, and is in thermal contact with the main N2O-dissociation reactor. The gas mixture that leaves the main reactor first goes through a carbon bed, which adsorbs all of the trace NO and most of the trace NO2. The gas mixture then goes through the destroyer catalytic reactor, wherein most or all of the remaining NO2 is dissociated. A NOBOSS can be designed to regulate its reactor temperature across a range of flow rates. One such system includes three destroyer loops; these loops act, in combination with a heat sink, to remove heat from the main N2O-dissociation reactor. In this system, the N2O and product gases play an additional role as coolants; thus, as needed, the coolant flow increases in proportion to the rate of generation of heat, helping to keep the main-reactor temperature below 540 C.

Performance and Mechanisms of Ultrafiltration Membrane Fouling Mitigation by Coupling Coagulation and Applied Electric Field in a Novel Electrocoagulation Membrane Reactor.

PubMed

Sun, Jingqiu; Hu, Chengzhi; Tong, Tiezheng; Zhao, Kai; Qu, Jiuhui; Liu, Huijuan; Elimelech, Menachem

2017-08-01

A novel electrocoagulation membrane reactor (ECMR) was developed, in which ultrafiltration (UF) membrane modules are placed between electrodes to improve effluent water quality and reduce membrane fouling. Experiments with feedwater containing clays (kaolinite) and natural organic matter (humic acid) revealed that the combined effect of coagulation and electric field mitigated membrane fouling in the ECMR, resulting in higher water flux than the conventional combination of electrocoagulation and UF in separate units (EC-UF). Higher current densities and weakly acidic pH in the EMCR favored faster generation of large flocs and effectively reduced membrane pore blocking. The hydraulic resistance of the formed cake layers on the membrane surface in ECMR was reduced due to an increase in cake layer porosity and polarity, induced by both coagulation and the applied electric field. The formation of a polarized cake layer was controlled by the applied current density and voltage, with cake layers formed under higher electric field strengths showing higher porosity and hydrophilicity. Compared to EC-UF, ECMR has a smaller footprint and could achieve significant energy savings due to improved fouling resistance and a more compact reactor design.

Development and Prototyping of the PROSPECT Antineutrino Detector

NASA Astrophysics Data System (ADS)

Commeford, Kelley; Prospect Collaboration

2017-01-01

The PROSPECT experiment will make the most precise measurement of the 235U reactor antineutrino spectrum as well as search for sterile neutrinos using a segmented Li-loaded liquid scintillator neutrino detector. Several prototype detectors of increasing size, complexity, and fidelity have been constructed and tested as part of the PROSPECT detector development program. The challenges to overcome include the efficient rejection of cosmogenic background and collection of optical photons in a compact volume. Design choices regarding segment structure and layout, calibration source deployment, and optical collection methods are discussed. Results from the most recent multi-segment prototype, PROSPECT-50, will also be shown.

Westinghouse Small Modular Reactor passive safety system response to postulated events

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

Smith, M. C.; Wright, R. F.

2012-07-01

The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (>225 MWe) integral pressurized water reactor. This paper is part of a series of four describing the design and safety features of the Westinghouse SMR. This paper focuses in particular upon the passive safety features and the safety system response of the Westinghouse SMR. The Westinghouse SMR design incorporates many features to minimize the effects of, and in some cases eliminates the possibility of postulated accidents. The small size of the reactor and the low power density limits the potential consequences of an accident relative to a large plant. Themore » integral design eliminates large loop piping, which significantly reduces the flow area of postulated loss of coolant accidents (LOCAs). The Westinghouse SMR containment is a high-pressure, compact design that normally operates at a partial vacuum. This facilitates heat removal from the containment during LOCA events. The containment is submerged in water which also aides the heat removal and provides an additional radionuclide filter. The Westinghouse SMR safety system design is passive, is based largely on the passive safety systems used in the AP1000{sup R} reactor, and provides mitigation of all design basis accidents without the need for AC electrical power for a period of seven days. Frequent faults, such as reactivity insertion events and loss of power events, are protected by first shutting down the nuclear reaction by inserting control rods, then providing cold, borated water through a passive, buoyancy-driven flow. Decay heat removal is provided using a layered approach that includes the passive removal of heat by the steam drum and independent passive heat removal system that transfers heat from the primary system to the environment. Less frequent faults such as loss of coolant accidents are mitigated by passive injection of a large quantity of water that is readily available inside containment. An automatic depressurization system is used to reduce the reactor pressure in a controlled manner to facilitate the passive injection. Long-term decay heat removal is accomplished using the passive heat removal systems augmented by heat transfer through the containment vessel to the environment. The passive injection systems are designed so that the fuel remains covered and effectively cooled throughout the event. Like during the frequent faults, the passive systems provide effective cooling without the need for ac power for seven days following the accident. Connections are available to add additional water to indefinitely cool the plant. The response of the safety systems of the Westinghouse SMR to various initiating faults has been examined. Among them, two accidents; an extended station blackout event, and a LOCA event have been evaluated to demonstrate how the plant will remain safe in the unlikely event that either should occur. (authors)« less

DESIGN CHARACTERISTICS OF THE IDAHO NATIONAL LABORATORY HIGH-[TEMPERATURE GAS-COOLED TEST REACTOR

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

Sterbentz, James; Bayless, Paul; Strydom, Gerhard

A point design for a graphite-moderated, high-temperature, gas-cooled test reactor (HTG TR) has been developed by Idaho National Laboratory (INL) as part of a United States (U.S.) Department of Energy (DOE) initiative to explore and potentially expand the existing U.S. test reactor capabilities. This paper provides a summary of the design and its main attributes. The 200 MW HTG TR is a thermal-neutron spectrum reactor composed of hexagonal prismatic fuel and graphite reflector blocks. Twelve fuel columns (96 fuel blocks total and 6.34 m active core height) are arranged in two hexagonal rings to form a relatively compact, high-power density,more » annular core sandwiched between inner, outer, top, and bottom graphite reflectors. The HTG-TR is designed to operate at 7 MPa with a coolant inlet/outlet temperature of 325°C/650°C, and utilizes TRISO particle fuel from the DOE AGR Program with 425 ?m uranium oxycarbide (UCO) kernels and an enrichment of 15.5 wt% 235U. The primary mission of the HTG TR is material irradiation and therefore the core has been specifically designed and optimized to provide the highest possible thermal and fast neutron fluxes. The highest thermal neutron flux (3.90E+14 n/cm2s) occurs in the outer reflector, and the maximum fast flux levels (1.17E+14 n/cm2s) are produced in the central reflector column where most of the graphite has been removed. Due to high core temperatures under accident conditions, all the irradiation test facilities have been located in the inner and outer reflectors where fast flux levels decline. The core features a large number of irradiation positions with large test volumes and long test lengths, ideal for thermal neutron irradiation of large test articles. The total available test volume is more than 1100 liters. Up to four test loop facilities can be accommodated with pressure tube boundaries to isolate test articles and test fluids (e.g., liquid metal, liquid salt, light water) from the helium primary coolant system.« less

Nucifer: A small electron-antineutrino detector for fundamental and safeguard studies

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

Letourneau, A.; Bui, V. M.; Cribier, M.

The Nucifer detector will be deployed in the next few months at the Osiris research reactor in France. Nucifer is a 1-ton Gd-doped liquid scintillator detector devoted to reactor antineutrino studies. It will be installed 7 m away from the compact core of the Osiris reactor. The design of such small volume detector has been focused on high detection efficiency and good background rejection. Over the last decades, our understanding of the neutrino properties has been improved and allows today the possibility to apply the detection of antineutrinos to automatic and to non intrusively survey nuclear power plant. This hasmore » triggered the interest of the International Atomic Energy Agency (IAEA), which is interested by developing new safeguard techniques for next generation reactors. The sensitivity of such technique has to be proved and demonstrated. On the other hand there is still some issues in our understanding of the neutrino properties as the observed deficit in the antineutrino rate at short distances (< 100 m) that can not be explained by oscillations in the 3-flavors neutrino model. If a global systematic error is rejected, such anomaly opens the door to new physic that can be assessed with small detectors placed close to the core. Here we review the Nucifer detector in this context and the tests we are performing. (authors)« less

Flow tests of a single fuel element coolant channel for a compact fast reactor for space power

NASA Technical Reports Server (NTRS)

Springborn, R. H.

1971-01-01

Water flow tests were conducted on a single-fuel-element cooling channel for a nuclear concept to be used for space power. The tests established a method for measuring coolant flow rate which is applicable to water flow testing of a complete mockup of the reference reactor. The inlet plenum-to-outlet plenum pressure drop, which approximates the overall core pressure drop, was measured and correlated with flow rate. This information can be used for reactor coolant flow and heat transfer calculations. An analytical study of the flow characteristics was also conducted.

Satellite nuclear power station: An engineering analysis

NASA Technical Reports Server (NTRS)

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

1973-01-01

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

Diffusion of cesium and iodine in compressed IG-110 graphite compacts

NASA Astrophysics Data System (ADS)

Carter, L. M.; Brockman, J. D.; Robertson, J. D.; Loyalka, S. K.

2016-08-01

Nuclear graphite grade IG-110 is currently used in the High Temperature Engineering Test Reactor (HTTR) in Japan for certain permanent and replaceable core components, and is a material of interest in general. Therefore, transport parameters for fission products in this material are needed. Measurement of diffusion through pressed compacts of IG-110 graphite is experimentally attractive because they are easy to prepare with homogeneous distributions of fission product surrogates. In this work, we measured diffusion coefficients for Cs and I in pressed compacts made from IG-110 powder in the 1079-1290 K temperature range, and compared them to those obtained in as-received IG-110.

Conceptual design of a fast-ignition laser fusion reactor based on a dry wall chamber

NASA Astrophysics Data System (ADS)

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

2008-05-01

The fast ignition is quite attractive for a compact laser fusion reactor, because a sufficiently high pellet gain is available with a small input energy. We designed an inertial fusion reactor based on Fast-ignition Advanced Laser fusion reactor CONcept, called FALCON-D, where a dry wall is employed for a chamber wall. A simple point model shows that the pellet gain G~100 is available with laser energies of 350kJ for implosion, 50kJ for heating. This results in the fusion yield of 40 MJ in one shot. By increasing the repetition rate up to 30 Hz, the fusion power of 1.2 GWth becomes available. Plant system analysis shows the net electric power to be about 0.4 GWe In the fast ignition it is available to employ a low aspect ratio pellet, which is favorable for the stability during the implosion phase. Here the pellet aspect ratio is reduced to be 2 ~ 4, and the optimization of the pulse shape for the implosion laser are carried out by using the 1-D hydrodynamic simulation code ILESTA-1D. A ferritic steel with a tungsten armour is employed for the chamber wall. The feasibility of this dry wall concept is studied from various engineering aspects such as surface melting, physical and chemical sputtering, blistering and exfoliation by helium retention, and thermo-mechanical fatigue, and it is found that blistering and exfoliation due to the helium retention and fatigue failure due to cyclic thermal load are major concerns. The cost analysis shows that the construction cost is moderate but the cost of electricity is slightly expensive.

Validation of the Physics Analysis used to Characterize the AGR-1 TRISO Fuel Irradiation Test

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

Sterbentz, James W.; Harp, Jason M.; Demkowicz, Paul A.

2015-05-01

The results of a detailed physics depletion calculation used to characterize the AGR-1 TRISO-coated particle fuel test irradiated in the Advanced Test Reactor (ATR) at the Idaho National Laboratory are compared to measured data for the purpose of validation. The particle fuel was irradiated for 13 ATR power cycles over three calendar years. The physics analysis predicts compact burnups ranging from 11.30-19.56% FIMA and cumulative neutron fast fluence from 2.21?4.39E+25 n/m 2 under simulated high-temperature gas-cooled reactor conditions in the ATR. The physics depletion calculation can provide a full characterization of all 72 irradiated TRISO-coated particle compacts during and post-irradiation,more » so validation of this physics calculation was a top priority. The validation of the physics analysis was done through comparisons with available measured experimental data which included: 1) high-resolution gamma scans for compact activity and burnup, 2) mass spectrometry for compact burnup, 3) flux wires for cumulative fast fluence, and 4) mass spectrometry for individual actinide and fission product concentrations. The measured data are generally in very good agreement with the calculated results, and therefore provide an adequate validation of the physics analysis and the results used to characterize the irradiated AGR-1 TRISO fuel.« less

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

A Compact Nuclear Fusion Reactor for Space Flights

NASA Astrophysics Data System (ADS)

Nastoyashchiy, Anatoly F.

2006-05-01

A small-scale nuclear fusion reactor is suggested based on the concepts of plasma confinement (with a high pressure gas) which have been patented by the author. The reactor considered can be used as a power setup in space flights. Among the advantages of this reactor is the use of a D3He fuel mixture which at burning gives main reactor products — charged particles. The energy balance considerably improves, as synchrotron radiation turn out "captured" in the plasma volume, and dangerous, in the case of classical magnetic confinement, instabilities in the direct current magnetic field configuration proposed do not exist. As a result, the reactor sizes are quite suitable (of the order of several meters). A possibility of making reactive thrust due to employment of ejection of multiply charged ions formed at injection of pellets from some adequate substance into the hot plasma center is considered.

Materials, Turbomachinery and Heat Exchangers for Supercritical CO2 Systems

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

Anderson, Mark; Nellis, Greg; Corradini, Michael

2012-10-19

The objective of this project is to produce the necessary data to evaluate the performance of the supercritical carbon dioxide cycle. The activities include a study of materials compatibility of various alloys at high temperatures, the heat transfer and pressure drop in compact heat exchanger units, and turbomachinery issues, primarily leakage rates through dynamic seals. This experimental work will serve as a test bed for model development and design calculations, and will help define further tests necessary to develop high-efficiency power conversion cycles for use on a variety of reactor designs, including the sodium fast reactor (SFR) and very high-temperaturemore » gas reactor (VHTR). The research will be broken into three separate tasks. The first task deals with the analysis of materials related to the high-temperature S-CO{sub 2} Brayton cycle. The most taxing materials issues with regard to the cycle are associated with the high temperatures in the reactor side heat exchanger and in the high-temperature turbine. The system could experience pressures as high as 20MPa and temperatures as high as 650°C. The second task deals with optimization of the heat exchangers required by the S-CO{sub 2} cycle; the S-CO{sub 2} flow passages in these heat exchangers are required whether the cycle is coupled with a VHTR or an SFR. At least three heat exchangers will be required: the pre-cooler before compression, the recuperator, and the heat exchanger that interfaces with the reactor coolant. Each of these heat exchangers is unique and must be optimized separately. The most challenging heat exchanger is likely the pre-cooler, as there is only about a 40°C temperature change but it operates close to the CO{sub 2} critical point, therefore inducing substantial changes in properties. The proposed research will focus on this most challenging component. The third task examines seal leakage through various dynamic seal designs under the conditions expected in the S-CO{sub 2} cycle, including supercritical, choked, and two-phase flow conditions.« less

Compact reactor for onboard hydrogen generation

NASA Technical Reports Server (NTRS)

Brabbs, T. A.

1980-01-01

Hydrogen, chemically stored as methanol, is promising internal-combustion fuel. Methanol is readily obtainable from natural products such as wood, compost, or various organic wastes. Steam reformation of methanol as source for hydrogen is relatively simple operation.

AGR-3/4 Irradiation Test Train Disassembly and Component Metrology First Look Report

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

Stempien, John Dennis; Rice, Francine Joyce; Harp, Jason Michael

2016-03-01

The AGR-3/4 experiment was designed to study fission product transport within graphitic matrix material and nuclear-grade graphite. To this end, this experiment consisted of 12 capsules, each fueled with 4 compacts containing UCO TRISO particles as driver fuel and 20 UCO designed-to-fail (DTF) fuel particles in each compact. The DTF fuel was fabricated with a thin pyrocarbon layer which was intended to fail during irradiation and provide a source of fission products. These fission products could then migrate through the compact and into the surrounding concentric rings of graphitic matrix material and/or nuclear graphite. Through post-irradiation examination (PIE) of themore » rings (including physical sampling and gamma scanning) fission product concentration profiles within the rings can be determined. These data can be used to elucidate fission product transport parameters (e.g. diffusion coefficients within the test materials) which will be used to inform and refine models of fission product transport. After irradiation in the Advanced Test Reactor (ATR) had been completed in April 2014, the AGR-3/4 experiment was shipped to the Hot Fuel Examination Facility (HFEF) at the Materials and Fuels Complex (MFC) for inspection, disassembly, and metrology. The AGR-3/4 test train was received at MFC in two separate shipments between February and April 2015. Visual examinations of the test train exterior did not indicate dimensional distortion, and only two small discolored areas were observed at the bottom of Capsules 8 and 9. No corresponding discoloration was found on the inside of these capsules, however. Prior to disassembly, the two test train sections were subject to analysis via the Precision Gamma Scanner (PGS), which did not indicate that any gross fuel relocation had occurred. A series of specialized tools (including clamps, cutters, and drills) had been designed and fabricated in order to carry out test train disassembly and recovery of capsule components (graphite rings and fuel compacts). This equipment performed well for separating each capsule in the test train and extracting the capsule components. Only a few problems were encountered. In one case, the outermost ring (the sink ring) was cracked during removal of the capsule through tubes. Although the sink ring will be analyzed in order to obtain a mass balance of fission products in the experiment, these cracks do not pose a major concern because the sink ring will not be analyzed in detail to obtain the spatial distribution of fission products. In Capsules 4 and 5, the compacts could not be removed from the inner rings. Strategies for removing the compacts are being evaluated. Sampling the inner rings with the compacts in-place is also an option. Dimensional measurements were made on the compacts, inner rings, outer rings, and sink rings. The diameters of all compacts decreased by 0.5 to 2.0 %. Generally, the extent of diametric shrinkage increased linearly with increasing neutron fluence. Most compact lengths also decreased. Compact lengths decreased with increasing fluence, reaching maximum shrinkage of about 0.9 % at a fast fluence of 4.0x10 25 n/m 2 E > 0.18 MeV. Above this fluence, the extent of length shrinkage appeared to decrease with fluence, and two compacts from Capsule 7 were found to have slightly increased in length (< 0.1 %) after a fluence of 5.2x10 25 n/m 2.« less

AGR-3/4 Irradiation Test Train Disassembly and Component Metrology First Look Report

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

Stempien, John Dennis; Rice, Francine Joyce; Harp, Jason Michael

The AGR-3/4 experiment was designed to study fission product transport within graphitic matrix material and nuclear-grade graphite. To this end, this experiment consisted of 12 capsules, each fueled with 4 compacts containing UCO TRISO particles as driver fuel and 20 UCO designed-to-fail (DTF) fuel particles in each compact. The DTF fuel was fabricated with a thin pyrocarbon layer which was intended to fail during irradiation and provide a source of fission products. These fission products could then migrate through the compact and into the surrounding concentric rings of graphitic matrix material and/or nuclear graphite. Through post-irradiation examination (PIE) of themore » rings (including physical sampling and gamma scanning) fission product concentration profiles within the rings can be determined. These data can be used to elucidate fission product transport parameters (e.g. diffusion coefficients within the test materials) which will be used to inform and refine models of fission product transport. After irradiation in the Advanced Test Reactor (ATR) had been completed in April 2014, the AGR-3/4 experiment was shipped to the Hot Fuel Examination Facility (HFEF) at the Materials and Fuels Complex (MFC) for inspection, disassembly, and metrology. The AGR-3/4 test train was received at MFC in two separate shipments between February and April 2015. Visual examinations of the test train exterior did not indicate dimensional distortion, and only two small discolored areas were observed at the bottom of Capsules 8 and 9. No corresponding discoloration was found on the inside of these capsules, however. Prior to disassembly, the two test train sections were subject to analysis via the Precision Gamma Scanner (PGS), which did not indicate that any gross fuel relocation had occurred. A series of specialized tools (including clamps, cutters, and drills) had been designed and fabricated in order to carry out test train disassembly and recovery of capsule components (graphite rings and fuel compacts). This equipment performed well for separating each capsule in the test train and extracting the capsule components. Only a few problems were encountered. In one case, the outermost ring (the sink ring) was cracked during removal of the capsule through tubes. Although the sink ring will be analyzed in order to obtain a mass balance of fission products in the experiment, these cracks do not pose a major concern because the sink ring will not be analyzed in detail to obtain the spatial distribution of fission products. In Capsules 4 and 5, the compacts could not be removed from the inner rings. Strategies for removing the compacts are being evaluated. Sampling the inner rings with the compacts in-place is also an option. Dimensional measurements were made on the compacts, inner rings, outer rings, and sink rings. The diameters of all compacts decreased by 0.5 to 2.0 %. Generally, the extent of diametric shrinkage increased linearly with increasing neutron fluence. Most compact lengths also decreased. Compact lengths decreased with increasing fluence, reaching maximum shrinkage of about 0.9 % at a fast fluence of 4.0x1025 n/m2 E > 0.18 MeV. Above this fluence, the extent of length shrinkage appeared to decrease with fluence, and two compacts from Capsule 7 were found to have slightly increased in length (< 0.1 %) after a fluence of 5.2x1025 n/m2.« less

Strategic need for a multi-purpose thermal hydraulic loop for support of advanced reactor technologies

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

O'Brien, James E.; Sabharwall, Piyush; Yoon, Su -Jong

2014-09-01

This report presents a conceptual design for a new high-temperature multi fluid, multi loop test facility for the INL to support thermal hydraulic, materials, and thermal energy storage research for nuclear and nuclear-hybrid applications. In its initial configuration, the facility will include a high-temperature helium loop, a liquid salt loop, and a hot water/steam loop. The three loops will be thermally coupled through an intermediate heat exchanger (IHX) and a secondary heat exchanger (SHX). Research topics to be addressed with this facility include the characterization and performance evaluation of candidate compact heat exchangers such as printed circuit heat exchangers (PCHEs)more » at prototypical operating conditions, flow and heat transfer issues related to core thermal hydraulics in advanced helium-cooled and salt-cooled reactors, and evaluation of corrosion behavior of new cladding materials and accident-tolerant fuels for LWRs at prototypical conditions. Based on its relevance to advanced reactor systems, the new facility has been named the Advanced Reactor Technology Integral System Test (ARTIST) facility. Research performed in this facility will advance the state of the art and technology readiness level of high temperature intermediate heat exchangers (IHXs) for nuclear applications while establishing the INL as a center of excellence for the development and certification of this technology. The thermal energy storage capability will support research and demonstration activities related to process heat delivery for a variety of hybrid energy systems and grid stabilization strategies. Experimental results obtained from this research will assist in development of reliable predictive models for thermal hydraulic design and safety codes over the range of expected advanced reactor operating conditions. Proposed/existing IHX heat transfer and friction correlations and criteria will be assessed with information on materials compatibility and instrumentation needs. The experimental database will guide development of appropriate predictive methods and be available for code verification and validation (V&V) related to these systems.« less

THETRIS: A MICRO-SCALE TEMPERATURE AND GAS RELEASE MODEL FOR TRISO FUEL

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

J. Ortensi; A.M. Ougouag

2011-12-01

The dominating mechanism in the passive safety of gas-cooled, graphite-moderated, high-temperature reactors (HTRs) is the Doppler feedback effect. These reactor designs are fueled with sub-millimeter sized kernels formed into TRISO particles that are imbedded in a graphite matrix. The best spatial and temporal representation of the feedback effect is obtained from an accurate approximation of the fuel temperature. Most accident scenarios in HTRs are characterized by large time constants and slow changes in the fuel and moderator temperature fields. In these situations a meso-scale, pebble and compact scale, solution provides a good approximation of the fuel temperature. Micro-scale models aremore » necessary in order to obtain accurate predictions in faster transients or when parameters internal to the TRISO are needed. Since these coated particles constitute one of the fundamental design barriers for the release of fission products, it becomes important to understand the transient behavior inside this containment system. An explicit TRISO fuel temperature model named THETRIS has been developed and incorporated into the CYNOD-THERMIX-KONVEK suite of coupled codes. The code includes gas release models that provide a simple predictive capability of the internal pressure during transients. The new model yields similar results to those obtained with other micro-scale fuel models, but with the added capability to analyze gas release, internal pressure buildup, and effects of a gap in the TRISO. The analyses show the instances when the micro-scale models improve the predictions of the fuel temperature and Doppler feedback. In addition, a sensitivity study of the potential effects on the transient behavior of high-temperature reactors due to the presence of a gap is included. Although the formation of a gap occurs under special conditions, its consequences on the dynamic behavior of the reactor can cause unexpected responses during fast transients. Nevertheless, the strong Doppler feedback forces the reactor to quickly stabilize.« less

Detection and analysis of particles with failed SiC in AGR-1 fuel compacts

DOE PAGES

Hunn, John D.; Baldwin, Charles A.; Gerczak, Tyler J.; ...

2016-04-06

As the primary barrier to release of radioactive isotopes emitted from the fuel kernel, retention performance of the SiC layer in tristructural isotropic (TRISO) coated particles is critical to the overall safety of reactors that utilize this fuel design. Most isotopes are well-retained by intact SiC coatings, so pathways through this layer due to cracking, structural defects, or chemical attack can significantly contribute to radioisotope release. In the US TRISO fuel development effort, release of 134Cs and 137Cs are used to detect SiC failure during fuel compact irradiation and safety testing because the amount of cesium released by a compactmore » containing one particle with failed SiC is typically ten or more times higher than that released by compacts without failed SiC. Compacts with particles that released cesium during irradiation testing or post-irradiation safety testing at 1600–1800 °C were identified, and individual particles with abnormally low cesium retention were sorted out with the Oak Ridge National Laboratory (ORNL) Irradiated Microsphere Gamma Analyzer (IMGA). X-ray tomography was used for three-dimensional imaging of the internal coating structure to locate low-density pathways through the SiC layer and guide subsequent materialography by optical and scanning electron microscopy. In addition, all three cesium-releasing particles recovered from as-irradiated compacts showed a region where the inner pyrocarbon (IPyC) had cracked due to radiation-induced dimensional changes in the shrinking buffer and the exposed SiC had experienced concentrated attack by palladium; SiC failures observed in particles subjected to safety testing were related to either fabrication defects or showed extensive Pd corrosion through the SiC where it had been exposed by similar IPyC cracking.« less

PROSPECT - A precision oscillation and spectrum experiment

NASA Astrophysics Data System (ADS)

Langford, T. J.; PROSPECT Collaboration

2015-08-01

Segmented antineutrino detectors placed near a compact research reactor provide an excellent opportunity to probe short-baseline neutrino oscillations and precisely measure the reactor antineutrino spectrum. Close proximity to a reactor combined with minimal overburden yield a high background environment that must be managed through shielding and detector technology. PROSPECT is a new experimental effort to detect reactor antineutrinos from the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory, managed by UT Battelle for the U.S. Department of Energy. The detector will use novel lithium-loaded liquid scintillator capable of neutron/gamma pulse shape discrimination and neutron capture tagging. These enhancements improve the ability to identify neutrino inverse-beta decays (IBD) and reject background events in analysis. Results from these efforts will be covered along with their implications for an oscillation search and a precision spectrum measurement.

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.

Heat exchanger for reactor core and the like

DOEpatents

Kaufman, Jay S.; Kissinger, John A.

1986-01-01

A compact bayonet tube type heat exchanger which finds particular application as an auxiliary heat exchanger for transfer of heat from a reactor gas coolant to a secondary fluid medium. The heat exchanger is supported within a vertical cavity in a reactor vessel intersected by a reactor coolant passage at its upper end and having a reactor coolant return duct spaced below the inlet passage. The heat exchanger includes a plurality of relatively short length bayonet type heat exchange tube assemblies adapted to pass a secondary fluid medium therethrough and supported by primary and secondary tube sheets which are releasibly supported in a manner to facilitate removal and inspection of the bayonet tube assemblies from an access area below the heat exchanger. Inner and outer shrouds extend circumferentially of the tube assemblies and cause the reactor coolant to flow downwardly internally of the shrouds over the tube bundle and exit through the lower end of the inner shroud for passage to the return duct in the reactor vessel.

Pebble Fuel Handling and Reactivity Control for Salt-Cooled High Temperature Reactors

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

Peterson, Per; Greenspan, Ehud

2015-02-09

This report documents the work completed on the X-PREX facility under NEUP Project 11- 3172. This project seeks to demonstrate the viability of pebble fuel handling and reactivity control for fluoride salt-cooled high-temperature reactors (FHRs). The research results also improve the understanding of pebble motion in helium-cooled reactors, as well as the general, fundamental understanding of low-velocity granular flows. Successful use of pebble fuels in with salt coolants would bring major benefits for high-temperature reactor technology. Pebble fuels enable on-line refueling and operation with low excess reactivity, and thus simpler reactivity control and improved fuel utilization. If fixed fuel designsmore » are used, the power density of salt- cooled reactors is limited to 10 MW/m 3 to obtain adequate duration between refueling, but pebble fuels allow power densities in the range of 20 to 30 MW/m 3. This can be compared to the typical modular helium reactor power density of 5 MW/m3. Pebble fuels also permit radial zoning in annular cores and use of thorium or graphite pebble blankets to reduce neutron fluences to outer radial reflectors and increase total power production. Combined with high power conversion efficiency, compact low-pressure primary and containment systems, and unique safety characteristics including very large thermal margins (>500°C) to fuel damage during transients and accidents, salt-cooled pebble fuel cores offer the potential to meet the major goals of the Advanced Reactor Concepts Development program to provide electricity at lower cost than light water reactors with improved safety and system performance.This report presents the facility description, experimental results, and supporting simulation methods of the new X-Ray Pebble Recirculation Experiment (X-PREX), which is now operational and being used to collect data on the behavior of slow dense granular flows relevant to pebble bed reactor core designs. The X-PREX facility uses novel digital x-ray tomography methods to track both the translational and rotational motion of spherical pebbles, which provides unique experimental results that can be used to validate discrete element method (DEM) simulations of pebble motion. The validation effort supported by the X-PREX facility provides a means to build confidence in analysis of pebble bed configuration and residence time distributions that impact the neutronics, thermal hydraulics, and safety analysis of pebble bed reactor cores. Experimental and DEM simulation results are reported for silo drainage, a classical problem in the granular flow literature, at several hopper angles. These studies include conventional converging and novel diverging geometries that provide additional flexibility in the design of pebble bed reactor cores. Excellent agreement is found between the X-PREX experimental and DEM simulation results. This report also includes results for additional studies relevant to the design and analysis of pebble bed reactor cores including the study of forces on shut down blades inserted directly into a packed bed and pebble flow in a cylindrical hopper that is representative of a small test reactor.« less

Modeling of Flow Blockage in a Liquid Metal-Cooled Reactor Subassembly with a Subchannel Analysis Code

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

Jeong, Hae-Yong; Ha, Kwi-Seok; Chang, Won-Pyo

The local blockage in a subassembly of a liquid metal-cooled reactor (LMR) is of importance to the plant safety because of the compact design and the high power density of the core. To analyze the thermal-hydraulic parameters in a subassembly of a liquid metal-cooled reactor with a flow blockage, the Korea Atomic Energy Research Institute has developed the MATRA-LMR-FB code. This code uses the distributed resistance model to describe the sweeping flow formed by the wire wrap around the fuel rods and to model the recirculation flow after a blockage. The hybrid difference scheme is also adopted for the descriptionmore » of the convective terms in the recirculating wake region of low velocity. Some state-of-the-art turbulent mixing models were implemented in the code, and the models suggested by Rehme and by Zhukov are analyzed and found to be appropriate for the description of the flow blockage in an LMR subassembly. The MATRA-LMR-FB code predicts accurately the experimental data of the Oak Ridge National Laboratory 19-pin bundle with a blockage for both the high-flow and low-flow conditions. The influences of the distributed resistance model, the hybrid difference method, and the turbulent mixing models are evaluated step by step with the experimental data. The appropriateness of the models also has been evaluated through a comparison with the results from the COMMIX code calculation. The flow blockage for the KALIMER design has been analyzed with the MATRA-LMR-FB code and is compared with the SABRE code to guarantee the design safety for the flow blockage.« less

Numerical Study on Crossflow Printed Circuit Heat Exchanger for Advanced Small Modular Reactors

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

Yoon, Su-Jong; Sabharwall, Piyush; Kim, Eung-Soo

2014-03-01

Various fluids such as water, gases (helium), molten salts (FLiNaK, FLiBe) and liquid metal (sodium) are used as a coolant of advanced small modular reactors (SMRs). The printed circuit heat exchanger (PCHE) has been adopted as the intermediate and/or secondary heat exchanger of SMR systems because this heat exchanger is compact and effective. The size and cost of PCHE can be changed by the coolant type of each SMR. In this study, the crossflow PCHE analysis code for advanced small modular reactor has been developed for the thermal design and cost estimation of the heat exchanger. The analytical solution ofmore » single pass, both unmixed fluids crossflow heat exchanger model was employed to calculate a two dimensional temperature profile of a crossflow PCHE. The analytical solution of crossflow heat exchanger was simply implemented by using built in function of the MATLAB program. The effect of fluid property uncertainty on the calculation results was evaluated. In addition, the effect of heat transfer correlations on the calculated temperature profile was analyzed by taking into account possible combinations of primary and secondary coolants in the SMR systems. Size and cost of heat exchanger were evaluated for the given temperature requirement of each SMR.« less

Upgrade of the compact neutron spectrometer for high flux environments

NASA Astrophysics Data System (ADS)

Osipenko, M.; Bellucci, A.; Ceriale, V.; Corsini, D.; Gariano, G.; Gatti, F.; Girolami, M.; Minutoli, S.; Panza, F.; Pillon, M.; Ripani, M.; Trucchi, D. M.

2018-03-01

In this paper new version of the 6Li-based neutron spectrometer for high flux environments is described. The new spectrometer was built with commercial single crystal Chemical Vapour Deposition diamonds of electronic grade. These crystals feature better charge collection as well as higher radiation hardness. New metal contacts approaching ohmic conditions were deposited on the diamonds suppressing build-up of space charge observed in the previous prototypes. New passive preamplification of the signal at detector side was implemented to improve its resolution. This preamplification is based on the RF transformer not sensitive to high neutron flux. The compact mechanical design allowed to reduce detector size to a tube of 1 cm diameter and 13 cm long. The spectrometer was tested in the thermal column of TRIGA reactor and at the DD neutron generator. The test results indicate an energy resolution of 300 keV (FWHM), reduced to 72 keV (RMS) excluding energy loss, and coincidence timing resolution of 160 ps (FWHM). The measured data are in agreement with Geant4 simulations except for larger energy loss tail presumably related to imperfections of metal contacts and glue expansion.

Whole Device Modeling of Compact Tori: Stability and Transport Modeling of C-2W

NASA Astrophysics Data System (ADS)

Dettrick, Sean; Fulton, Daniel; Lau, Calvin; Lin, Zhihong; Ceccherini, Francesco; Galeotti, Laura; Gupta, Sangeeta; Onofri, Marco; Tajima, Toshiki; TAE Team

2017-10-01

Recent experimental evidence from the C-2U FRC experiment shows that the confinement of energy improves with inverse collisionality, similar to other high beta toroidal devices, NSTX and MAST. This motivated the construction of a new FRC experiment, C-2W, to study the energy confinement scaling at higher electron temperature. Tri Alpha Energy is working towards catalysing a community-wide collaboration to develop a Whole Device Model (WDM) of Compact Tori. One application of the WDM is the study of stability and transport properties of C-2W using two particle-in-cell codes, ANC and FPIC. These codes can be used to find new stable operating points, and to make predictions of the turbulent transport at those points. They will be used in collaboration with the C-2W experimental program to validate the codes against C-2W, mitigate experimental risk inherent in the exploration of new parameter regimes, accelerate the optimization of experimental operating scenarios, and to find operating points for future FRC reactor designs.

A New Equivalence Theory Method for Treating Doubly Heterogeneous Fuel - I. Theory

DOE PAGES

Williams, Mark L.; Lee, Deokjung; Choi, Sooyoung

2015-03-04

A new methodology has been developed to treat resonance self-shielding in doubly heterogeneous very high temperature gas-cooled reactor systems in which the fuel compact region of a reactor lattice consists of small fuel grains dispersed in a graphite matrix. This new method first homogenizes the fuel grain and matrix materials using an analytically derived disadvantage factor from a two-region problem with equivalence theory and intermediate resonance method. This disadvantage factor accounts for spatial self-shielding effects inside each grain within the framework of an infinite array of grains. Then the homogenized fuel compact is self-shielded using a Bondarenko method to accountmore » for interactions between the fuel compact regions in the fuel lattice. In the final form of the equations for actual implementations, the double-heterogeneity effects are accounted for by simply using a modified definition of a background cross section, which includes geometry parameters and cross sections for both the grain and fuel compact regions. With the new method, the doubly heterogeneous resonance self-shielding effect can be treated easily even with legacy codes programmed only for a singly heterogeneous system by simple modifications in the background cross section for resonance integral interpolations. This paper presents a detailed derivation of the new method and a sensitivity study of double-heterogeneity parameters introduced during the derivation. The implementation of the method and verification results for various test cases are presented in the companion paper.« less

SP-100 flight qualification testing assessment

NASA Technical Reports Server (NTRS)

Jeanmougin, Nanette M.; Moore, Roger M.; Wait, David L.; Jacox, Michael G.

1988-01-01

The SP-100 is a compact space power system driven by a nuclear reactor that provides 100 kWe to the user at 200 VDC. The thermal energy generated by the nuclear reactor is converted into electrical energy by passive thermoelectric devices. Various options for tailoring the MIL-STD-1540B guidelines to the SP-100 nuclear power system are discussed. This study aids in selecting the appropriate qualification test program based on the cost, schedule, and test effectiveness of the various options.

NEUTRONIC REACTOR CONTROL ROD AND METHOD OF FABRICATION

DOEpatents

Porembka, S.W. Jr.

1961-06-27

A reactor control rod formed from a compacted powder dispersion is patented. The rod consists of titanium sheathed with a cladding alloy. The cladding alloy contains 1.3% to 1.6% by weight of tin, 0.07% to 0.12% by weight of chromium, 0.04% to 0.08% by weight of nickel, 0.09% to 0.16% by weight of iron, carbon not exceeding 0.05%, less than 0.5% by weight of incidental impurities, and the balance zirconium.

Steam reforming of heptane in a fluidized bed membrane reactor

NASA Astrophysics Data System (ADS)

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

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

Development of NASA's Small Fission Power System for Science and Human Exploration

NASA Technical Reports Server (NTRS)

Gibson, Marc A.; Mason, Lee; Bowman, Cheryl; Poston, David I.; McClure, Patrick R.; Creasy, John; Robinson, Chris

2014-01-01

Exploration of our solar system has brought great knowledge to our nation's scientific and engineering community over the past several decades. As we expand our visions to explore new, more challenging destinations, we must also expand our technology base to support these new missions. NASA's Space Technology Mission Directorate is tasked with developing these technologies for future mission infusion and continues to seek answers to many existing technology gaps. One such technology gap is related to compact power systems (greater than 1 kWe) that provide abundant power for several years where solar energy is unavailable or inadequate. Below 1 kWe, Radioisotope Power Systems have been the workhorse for NASA and will continue, assuming its availability, to be used for lower power applications similar to the successful missions of Voyager, Ulysses, New Horizons, Cassini, and Curiosity. Above 1 kWe, fission power systems become an attractive technology offering a scalable modular design of the reactor, shield, power conversion, and heat transport subsystems. Near term emphasis has been placed in the 1-10kWe range that lies outside realistic radioisotope power levels and fills a promising technology gap capable of enabling both science and human exploration missions. History has shown that development of space reactors is technically, politically, and financially challenging and requires a new approach to their design and development. A small team of NASA and DOE experts are providing a solution to these enabling FPS technologies starting with the lowest power and most cost effective reactor series named "Kilopower" that is scalable from approximately 1-10 kWe.

Development of NASA's Small Fission Power System for Science and Human Exploration

NASA Technical Reports Server (NTRS)

Gibson, Marc A.; Mason, Lee S.; Bowman, Cheryl L.; Poston, David I.; McClure, Patrick R.; Creasy, John; Robinson, Chris

2015-01-01

Exploration of our solar system has brought many exciting challenges to our nations scientific and engineering community over the past several decades. As we expand our visions to explore new, more challenging destinations, we must also expand our technology base to support these new missions. NASAs Space Technology Mission Directorate is tasked with developing these technologies for future mission infusion and continues to seek answers to many existing technology gaps. One such technology gap is related to compact power systems (1 kWe) that provide abundant power for several years where solar energy is unavailable or inadequate. Below 1 kWe, Radioisotope Power Systems have been the workhorse for NASA and will continue to be used for lower power applications similar to the successful missions of Voyager, Ulysses, New Horizons, Cassini, and Curiosity. Above 1 kWe, fission power systems become an attractive technology offering a scalable modular design of the reactor, shield, power conversion, and heat transport subsystems. Near term emphasis has been placed in the 1-10kWe range that lies outside realistic radioisotope power levels and fills a promising technology gap capable of enabling both science and human exploration missions. History has shown that development of space reactors is technically, politically, and financially challenging and requires a new approach to their design and development. A small team of NASA and DOE experts are providing a solution to these enabling FPS technologies starting with the lowest power and most cost effective reactor series named Kilopower that is scalable from approximately 1-10 kWe.

Performance of AGR-1 high-temperature reactor fuel during post-irradiation heating tests

DOE PAGES

Morris, Robert N.; Baldwin, Charles A.; Demkowicz, Paul A.; ...

2016-05-18

The fission product retention of irradiated low-enriched uranium oxide/uranium carbide tri-structural isotropic (TRISO) fuel compacts from the Advanced Gas-Cooled Reactor 1 (AGR-1) experiment has been evaluated at temperatures of 1600–1800 °C during post-irradiation safety tests. Fourteen compacts (a total of ~58,000 particles) with a burnup ranging from 13.4% to 19.1% fissions per initial metal atom (FIMA) have been tested using dedicated furnace systems at Idaho National Laboratory and Oak Ridge National Laboratory. The release of fission products 110mAg, 134Cs, 137Cs, 154Eu, 155Eu, 90Sr, and 85Kr was monitored while heating the fuel specimens in flowing helium. The behavior of silver, europium,more » and strontium appears to be dominated by inventory that was originally released through intact SiC coating layers during irradiation, but was retained in the compact at the end of irradiation and subsequently released during the safety tests. However, at a test temperature of 1800 °C, the data suggest that release of these elements through intact coatings may become significant after ~100 h. Cesium was very well retained by intact SiC layers, with a fractional release <5 × 10–6 after 300 h at 1600 °C or 100 h at 1800 °C. However, it was rapidly released from individual particles if the SiC layer failed, and therefore the overall cesium release fraction was dominated by the SiC defect and failure fractions in the fuel compacts. No complete TRISO coating layer failures were observed after 300 h at 1600 or 1700 °C, and 85Kr release was very low during the tests (particles with failed SiC, but intact outer pyrocarbon, retained most of their krypton). Krypton release from TRISO failures was only observed after ~210 h at 1800 °C in one compact. As a result, post-safety-test examination of fuel compacts and particles has focused on identifying specific particles from each compact with notable fission product release and detailed analysis of the coating layers to understand particle behavior.« less

Performance of AGR-1 high-temperature reactor fuel during post-irradiation heating tests

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

Morris, Robert N.; Baldwin, Charles A.; Demkowicz, Paul A.

The fission product retention of irradiated low-enriched uranium oxide/uranium carbide tri-structural isotropic (TRISO) fuel compacts from the Advanced Gas-Cooled Reactor 1 (AGR-1) experiment has been evaluated at temperatures of 1600–1800 °C during post-irradiation safety tests. Fourteen compacts (a total of ~58,000 particles) with a burnup ranging from 13.4% to 19.1% fissions per initial metal atom (FIMA) have been tested using dedicated furnace systems at Idaho National Laboratory and Oak Ridge National Laboratory. The release of fission products 110mAg, 134Cs, 137Cs, 154Eu, 155Eu, 90Sr, and 85Kr was monitored while heating the fuel specimens in flowing helium. The behavior of silver, europium,more » and strontium appears to be dominated by inventory that was originally released through intact SiC coating layers during irradiation, but was retained in the compact at the end of irradiation and subsequently released during the safety tests. However, at a test temperature of 1800 °C, the data suggest that release of these elements through intact coatings may become significant after ~100 h. Cesium was very well retained by intact SiC layers, with a fractional release <5 × 10–6 after 300 h at 1600 °C or 100 h at 1800 °C. However, it was rapidly released from individual particles if the SiC layer failed, and therefore the overall cesium release fraction was dominated by the SiC defect and failure fractions in the fuel compacts. No complete TRISO coating layer failures were observed after 300 h at 1600 or 1700 °C, and 85Kr release was very low during the tests (particles with failed SiC, but intact outer pyrocarbon, retained most of their krypton). Krypton release from TRISO failures was only observed after ~210 h at 1800 °C in one compact. As a result, post-safety-test examination of fuel compacts and particles has focused on identifying specific particles from each compact with notable fission product release and detailed analysis of the coating layers to understand particle behavior.« less

Spherical torus fusion reactor

DOEpatents

Peng, Yueng-Kay M.

1989-04-04

A fusion reactor is provided having a near spherical-shaped plasma with a modest central opening through which straight segments of toroidal field coils extend that carry electrical current for generating a toroidal magnet plasma confinement fields. By retaining only the indispensable components inboard of the plasma torus, principally the cooled toroidal field conductors and in some cases a vacuum containment vessel wall, the fusion reactor features an exceptionally small aspect ratio (typically about 1.5), a naturally elongated plasma cross section without extensive field shaping, requires low strength magnetic containment fields, small size and high beta. These features combine to produce a spherical torus plasma in a unique physics regime which permits compact fusion at low field and modest cost.

Spherical torus fusion reactor

DOEpatents

Peng, Yueng-Kay M.

1989-01-01

A fusion reactor is provided having a near spherical-shaped plasma with a modest central opening through which straight segments of toroidal field coils extend that carry electrical current for generating a toroidal magnet plasma confinement fields. By retaining only the indispensable components inboard of the plasma torus, principally the cooled toroidal field conductors and in some cases a vacuum containment vessel wall, the fusion reactor features an exceptionally small aspect ratio (typically about 1.5), a naturally elongated plasma cross section without extensive field shaping, requires low strength magnetic containment fields, small size and high beta. These features combine to produce a spherical torus plasma in a unique physics regime which permits compact fusion at low field and modest cost.

Formation and sustainment of field reversed configuration (FRC) plasmas by spheromak merging and neutral beam injection

DOE PAGES

Yamada, Masaaki

2016-01-01

This study briefly reviews a compact toroid reactor concept that addresses critical issues for forming, stabilizing and sustaining a field reversed configuration (FRC) with the use of plasma merging, plasma shaping, conducting shells, neutral beam injection (NBI). In this concept, an FRC plasma is generated by the merging of counter-helicity spheromaks produced by inductive discharges and sustained by the use of neutral beam injection (NBI). Plasma shaping, conducting shells, and the NBI would provide stabilization to global MHD modes. Although a specific FRC reactor design is outside the scope of the present paper, an example of a promising FRC reactormore » program is summarized based on the previously developed SPIRIT (Self-organized Plasmas by Induction, Reconnection and Injection Techniques) concept in order to connect this concept to the recently achieved the High Performance FRC plasmas obtained by Tri Alpha Energy [Binderbauer et al, Phys. Plasmas 22,056110, (2015)]. This paper includes a brief summary of the previous concept paper by M. Yamada et al, Plasma Fusion Res. 2, 004 (2007) and the recent experimental results from MRX.« less

Formation and sustainment of field reversed configuration (FRC) plasmas by spheromak merging and neutral beam injection

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

Yamada, Masaaki

2016-03-25

This paper briefly reviews a compact toroid reactor concept that addresses critical issues for forming, stabilizing and sustaining a field reversed configuration (FRC) with the use of plasma merging, plasma shaping, conducting shells, neutral beam injection (NBI). In this concept, an FRC plasma is generated by the merging of counter-helicity spheromaks produced by inductive discharges and sustained by the use of neutral beam injection (NBI). Plasma shaping, conducting shells, and the NBI would provide stabilization to global MHD modes. Although a specific FRC reactor design is outside the scope of the present paper, an example of a promising FRC reactormore » program is summarized based on the previously developed SPIRIT (Self-organized Plasmas by Induction, Reconnection and Injection Techniques) concept in order to connect this concept to the recently achieved the High Performance FRC plasmas obtained by Tri Alpha Energy [Binderbauer et al, Phys. Plasmas 22,056110, (2015)]. This paper includes a brief summary of the previous concept paper by M. Yamada et al, Plasma Fusion Res. 2, 004 (2007) and the recent experimental results from MRX.« less

STUDIES ON CONTAMINANT BIODEGRADATION IN SLURRY, WAFER, AND COMPACTED SOIL TUBE REACTORS

EPA Science Inventory

A systematic experimental approach is presented to quantitatively evaluate biodegradation rates in intact soil systems. Knowledge of bioremediation rates in intact soil systems is important for evaluating the efficacy of in-situ biodegradation and approaches for enhancing degrad...

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

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

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

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

Microbial Population Dynamics and Ecosystem Functions of Anoxic/Aerobic Granular Sludge in Sequencing Batch Reactors Operated at Different Organic Loading Rates

PubMed Central

Szabó, Enikö; Liébana, Raquel; Hermansson, Malte; Modin, Oskar; Persson, Frank; Wilén, Britt-Marie

2017-01-01

The granular sludge process is an effective, low-footprint alternative to conventional activated sludge wastewater treatment. The architecture of the microbial granules allows the co-existence of different functional groups, e.g., nitrifying and denitrifying communities, which permits compact reactor design. However, little is known about the factors influencing community assembly in granular sludge, such as the effects of reactor operation strategies and influent wastewater composition. Here, we analyze the development of the microbiomes in parallel laboratory-scale anoxic/aerobic granular sludge reactors operated at low (0.9 kg m-3d-1), moderate (1.9 kg m-3d-1) and high (3.7 kg m-3d-1) organic loading rates (OLRs) and the same ammonium loading rate (0.2 kg NH4-N m-3d-1) for 84 days. Complete removal of organic carbon and ammonium was achieved in all three reactors after start-up, while the nitrogen removal (denitrification) efficiency increased with the OLR: 0% at low, 38% at moderate, and 66% at high loading rate. The bacterial communities at different loading rates diverged rapidly after start-up and showed less than 50% similarity after 6 days, and below 40% similarity after 84 days. The three reactor microbiomes were dominated by different genera (mainly Meganema, Thauera, Paracoccus, and Zoogloea), but these genera have similar ecosystem functions of EPS production, denitrification and polyhydroxyalkanoate (PHA) storage. Many less abundant but persistent taxa were also detected within these functional groups. The bacterial communities were functionally redundant irrespective of the loading rate applied. At steady-state reactor operation, the identity of the core community members was rather stable, but their relative abundances changed considerably over time. Furthermore, nitrifying bacteria were low in relative abundance and diversity in all reactors, despite their large contribution to nitrogen turnover. The results suggest that the OLR has considerable impact on the composition of the granular sludge communities, but also that the granule communities can be dynamic even at steady-state reactor operation due to high functional redundancy of several key guilds. Knowledge about microbial diversity with specific functional guilds under different operating conditions can be important for engineers to predict the stability of reactor functions during the start-up and continued reactor operation. PMID:28507540

Experimental physics characteristics of a heavy-metal-reflected fast-spectrum critical assembly

NASA Technical Reports Server (NTRS)

Heneveld, W. H.; Paschall, R. K.; Springer, T. H.; Swanson, V. A.; Thiele, A. W.; Tuttle, R. J.

1971-01-01

A zero-power critical assembly was designed, constructed, and operated for the purpose 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 electric conversion system. The experimental program consisted basically of measuring the differential neutron spectra and the changes in critical mass that accompanied the stepwise addition of (Li-7)3N, Hf, Ta, and W to a basic core fueled with U metal in a pin-type Ta honeycomb structure. In addition, experimental results were obtained on power distributions, control characteristics, neutron lifetime, and reactivity worths of numerous absorber, structural, and scattering materials.

Laser-based sensor for a coolant leak detection in a nuclear reactor

NASA Astrophysics Data System (ADS)

Kim, T.-S.; Park, H.; Ko, K.; Lim, G.; Cha, Y.-H.; Han, J.; Jeong, D.-Y.

2010-08-01

Currently, the nuclear industry needs strongly a reliable detection system to continuously monitor a coolant leak during a normal operation of reactors for the ensurance of nuclear safety. In this work, we propose a new device for the coolant leak detection based on tunable diode laser spectroscopy (TDLS) by using a compact diode laser. For the feasibility experiment, we established an experimental setup consisted of a near-IR diode laser with a wavelength of about 1392 nm, a home-made multi-pass cell and a sample injection system. The feasibility test was performed for the detection of the heavy water (D2O) leaks which can happen in a pressurized heavy water reactor (PWHR). As a result, the device based on the TDLS is shown to be operated successfully in detecting a HDO molecule, which is generated from the leaked heavy water by an isotope exchange reaction between D2O and H2O. Additionally, it is suggested that the performance of the new device, such as sensitivity and stability, can be improved by adapting a cavity enhanced absorption spectroscopy and a compact DFB diode laser. We presume that this laser-based leak detector has several advantages over the conventional techniques currently employed in the nuclear power plant, such as radiation monitoring, humidity monitoring and FT-IR spectroscopy.

High-rate wastewater treatment combining a moving bed biofilm reactor and enhanced particle separation.

PubMed

Helness, H; Melin, E; Ulgenes, Y; Järvinen, P; Rasmussen, V; Odegaard, H

2005-01-01

Many cities around the world are looking for compact wastewater treatment alternatives since space for treatment plants is becoming scarce. In this paper development of a new compact, high-rate treatment concept with results from experiments in lab-scale and pilot-scale are presented. The idea behind the treatment concept is that coagulation/floc separation may be used to separate suspended and colloidal matter (resulting in > 70% organic matter removal in normal wastewater) while a high-rate biofilm process (based on Moving Bed biofilm reactors) may be used for removing low molecular weight, easily biodegradable, soluble organic matter. By using flotation for floc/biomass separation, the total residence time for a plant according to this concept will normally be < 1 hour. A cationic polymer combined with iron is used as coagulant at low dosages (i.e. 1-2 mg polymer/l, 5-10 mg Fe/l) resulting in low sludge production (compared to conventional chemical treatment) and sufficient P-removal.

Method of making a catalytic reactor for automobile

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

Vroman, W.R.

1976-09-07

A catalytic reactor is made by providing a generally cylindrical catalytic substrate of oval transverse section and clamping the same between paired housing shells to form a housing of oval section spaced from the substrate by means of a pair of wire mesh ropes seated within a corresponding pair of grooves extending around the periphery of the substrate at axially spaced locations. Each rope is compacted from a matrix of multiple layers of resilient stainless steel knitted wire and is interlocked with the housing by means of a pair of inwardly opening channels of the housing spaced axially by anmore » inwardly projecting rib of the housing. The grooves in the substrate are pressed radially into the latter while the same is in a plastic uncured condition, thereby to compact and reinforce the grooves to withstand the localized compressional force of the ropes seated therein after the substrate is cured and hardened and clamped between the housing shells.« less

Combustion of Na 2B 4O 7 + Mg + C to synthesis B 4C powders

NASA Astrophysics Data System (ADS)

Guojian, Jiang; Jiayue, Xu; Hanrui, Zhuang; Wenlan, Li

2009-09-01

Boron carbide powder was fabricated by combustion synthesis (CS) method directly from mixed powders of borax (Na 2B 4O 7), magnesium (Mg) and carbon. The adiabatic temperature of the combustion reaction of Na 2B 4O 7 + 6 Mg + C was calculated. The control of the reactions was achieved by selecting reactant composition, relative density of powder compact and gas pressure in CS reactor. The effects of these different influential factors on the composition and morphologies of combustion products were investigated. The results show that, it is advantageous for more Mg/Na 2B 4O 7 than stoichiometric ratio in Na 2B 4O 7 + Mg + C system and high atmosphere pressure in the CS reactor to increase the conversion degree of reactants to end product. The final product with the minimal impurities' content could be fabricated at appropriate relative density of powder compact. At last, boron carbide without impurities could be obtained after the acid enrichment and distilled water washing.

Blanket activation and afterheat for the Compact Reversed-Field Pinch Reactor

NASA Astrophysics Data System (ADS)

Davidson, J. W.; Battat, M. E.

A detailed assessment has been made of the activation and afterheat for a Compact Reversed-Field Pinch Reactor (CRFPR) blanket using a two-dimensional model that included the limiter, the vacuum ducts, and the manifolds and headers for cooling the limiter and the first and second walls. Region-averaged, multigroup fluxes and prompt gamma-ray/neutron heating rates were calculated using the two-dimensional, discrete-ordinates code TRISM. Activation and depletion calculations were performed with the code FORIG using one-group cross sections generated with the TRISM region-averaged fluxes. Afterheat calculations were performed for regions near the plasma, i.e., the limiter, first wall, etc. assuming a 10-day irradiation. Decay heats were computed for decay periods up to 100 minutes. For the activation calculations, the irradiation period was taken to be one year and blanket activity inventories were computed for decay times to 4 x 10 years. These activities were also calculated as the toxicity-weighted biological hazard potential (BHP).

Advanced particle-in-cell simulation techniques for modeling the Lockheed Martin Compact Fusion Reactor

NASA Astrophysics Data System (ADS)

Welch, Dale; Font, Gabriel; Mitchell, Robert; Rose, David

2017-10-01

We report on particle-in-cell developments of the study of the Compact Fusion Reactor. Millisecond, two and three-dimensional simulations (cubic meter volume) of confinement and neutral beam heating of the magnetic confinement device requires accurate representation of the complex orbits, near perfect energy conservation, and significant computational power. In order to determine initial plasma fill and neutral beam heating, these simulations include ionization, elastic and charge exchange hydrogen reactions. To this end, we are pursuing fast electromagnetic kinetic modeling algorithms including a two implicit techniques and a hybrid quasi-neutral algorithm with kinetic ions. The kinetic modeling includes use of the Poisson-corrected direct implicit, magnetic implicit, as well as second-order cloud-in-cell techniques. The hybrid algorithm, ignoring electron inertial effects, is two orders of magnitude faster than kinetic but not as accurate with respect to confinement. The advantages and disadvantages of these techniques will be presented. Funded by Lockheed Martin.

Concepts and Tests for the Remote-Controlled Dismantling of the Biological Shield and Form work of the KNK Reactor - 13425

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

Neff, Sylvia; Graf, Anja; Petrick, Holger

The compact sodium-cooled nuclear reactor facility Karlsruhe (KNK), a prototype Fast Breeder, is currently in an advanced stage of dismantling. Complete dismantling is based on 10 partial licensing steps. In the frame of the 9. decommissioning permit, which is currently ongoing, the dismantling of the biological shield is foreseen. The biological shield consists of heavy reinforced concrete with built-in steel fitments, such as form-work of the reactor tank, pipe sleeves, ventilation channels, and measuring devices. Due to the activation of the inner part of the biological shield, dismantling has to be done remote-controlled. During a comprehensive basic design phase amore » practical dismantling strategy was developed. Necessary equipment and tools were defined. Preliminary tests revealed that hot wire plasma cutting is the most favorable cutting technology due to the geometrical boundary conditions, the varying distance between cutter and material, and the heavy concrete behind the steel form-work. The cutting devices will be operated remotely via a carrier system with an industrial manipulator. The carrier system has expandable claws to adjust to the varying diameter of the reactor shaft during dismantling progress. For design approval of this prototype development, interaction between manipulator and hot wire plasma cutting was tested in a real configuration. For the demolition of the concrete structure, an excavator with appropriate tools, such as a hydraulic hammer, was selected. Other mechanical cutting devices, such as a grinder or rope saw, were eliminated because of concrete containing steel spheres added to increase the shielding factor of the heavy concrete. Dismantling of the biological shield will be done in a ring-wise manner due to static reasons. During the demolition process, the excavator is positioned on its tripod in three concrete recesses made prior to the dismantling of the separate concrete rings. The excavator and the manipulator carrier system will be operated alternately. Main boundary condition for all the newly designed equipment is the decommissioning housing of limited space within the reactor building containment. To allow for a continuous removal of the concrete rubble, an additional opening on the lowest level of the reactor shaft will be made. All equipment and the interaction of the tools have to be tested before use in the controlled area. Therefore a full-scale model of the biological shield will be provided in a mock-up. The tests will be performed in early 2014. The dismantling of the biological shield is scheduled for 2015. (authors)« less

Experimental Measurement and Numerical Modeling of the Effective Thermal Conductivity of TRISO Fuel Compacts

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

Folsom, Charles; Xing, Changhu; Jensen, Colby

2015-03-01

Accurate modeling capability of thermal conductivity of tristructural-isotropic (TRISO) fuel compacts is important to fuel performance modeling and safety of Generation IV reactors. To date, the effective thermal conductivity (ETC) of tristructural-isotropic (TRISO) fuel compacts has not been measured directly. The composite fuel is a complicated structure comprised of layered particles in a graphite matrix. In this work, finite element modeling is used to validate an analytic ETC model for application to the composite fuel material for particle-volume fractions up to 40%. The effect of each individual layer of a TRISO particle is analyzed showing that the overall ETC ofmore » the compact is most sensitive to the outer layer constituent. In conjunction with the modeling results, the thermal conductivity of matrix-graphite compacts and the ETC of surrogate TRISO fuel compacts have been successfully measured using a previously developed measurement system. The ETC of the surrogate fuel compacts varies between 50 and 30 W m -1 K -1 over a temperature range of 50-600°C. As a result of the numerical modeling and experimental measurements of the fuel compacts, a new model and approach for analyzing the effect of compact constituent materials on ETC is proposed that can estimate the fuel compact ETC with approximately 15-20% more accuracy than the old method. Using the ETC model with measured thermal conductivity of the graphite matrix-only material indicate that, in the composite form, the matrix material has a much greater thermal conductivity, which is attributed to the high anisotropy of graphite thermal conductivity. Therefore, simpler measurements of individual TRISO compact constituents combined with an analytic ETC model, will not provide accurate predictions of overall ETC of the compacts emphasizing the need for measurements of composite, surrogate compacts.« less

Aircraft and Bases Powered by Compact Nuclear Reactors: Solutions to Projecting Power in Highly Contested Environments and Fossil Fuel Dependence

DTIC Science & Technology

2015-05-01

pushed the depletion date past 2100.21 David Archibald, author of books and papers on climate science and a fellow at the Institute of World...Politics, does not predict explicitly the date of complete exhaustion, but he does note that humans have consumed about half of the world’s supply.22...deuterium, and lithium are plentiful on the earth and in the solar system. As far as fuel for existing and future fission reactors, uranium and

Comparison of silver release predictions using PARFUME with results from the AGR-2 irradiation experiment

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

Collin, Blaise P.; Demkowicz, Paul A.; Baldwin, Charles A.

2016-11-01

The PARFUME (PARticle FUel ModEl) code was used to predict silver release from tristructural isotropic (TRISO) coated fuel particles and compacts during the second irradiation experiment (AGR-2) of the Advanced Gas Reactor Fuel Development and Qualification program. The PARFUME model for the AGR-2 experiment used the fuel compact volume average temperature for each of the 559 days of irradiation to calculate the release of fission product silver from a representative particle for a select number of AGR-2 compacts and individual fuel particles containing either mixed uranium carbide/oxide (UCO) or 100% uranium dioxide (UO2) kernels. Post-irradiation examination (PIE) measurements were performedmore » to provide data on release of silver from these compacts and individual fuel particles. The available experimental fractional releases of silver were compared to their corresponding PARFUME predictions. Preliminary comparisons show that PARFUME under-predicts the PIE results in UCO compacts and is in reasonable agreement with experimental data for UO2 compacts. The accuracy of PARFUME predictions is impacted by the code limitations in the modeling of the temporal and spatial distributions of the temperature across the compacts. Nevertheless, the comparisons on silver release lie within the same order of magnitude.« less

Synthesis gas and olefins from the catalytic autothermal reforming of volatile and non-volatile liquids

NASA Astrophysics Data System (ADS)

Dreyer, Bradon Justin

2007-12-01

The research presented in this thesis develops an understanding of a clean energy process technology, catalytic partial oxidation (CPO). CPO is a process in which a carbon containing fuel, such as a hydrocarbon, is passed over a noble metal catalyst (e.g. rhodium and platinum) to efficiently generate synthesis gas (H2 and CO) and olefins (e.g. ethylene and propylene) in millisecond contact times. Chapter 1 introduces CPO and compares this technology with conventional methods for synthesis gas and olefin production. CPO has several advantages over the traditional synthesis gas and olefin production methods. One advantage includes autothermal operation, requiring no external heat input from furnaces or heat exchangers. Autothermal operation allows these reactors to be built compactly. The short contact-times associated with CPO further enable for high throughput in relatively small reactor systems, and more compact reactors typically translate to faster response times if transient operation is required. Nobel metal based CPO catalysts are also resistant to deactivation, resulting in less catalyst replacement, regeneration, and maintenance, and an increase in operating efficiency. An overview of the many applications of the chemicals produced from CPO is also presented in Chapter 1. The chemicals produced are crucial in generating valuable chemical intermediates that are eventually incorporated in consumer products, medical devices, building structures, and fertilizers. Additionally, H2 can be used as a source of energy in mobile fuel applications. Fuel cells convert H2 and O2 into electricity and water at higher efficiencies than thermal engine generators. Due to the difficulties in H2 storage, these more efficient energy generators are dependent on hydrogen obtained from synthesis gas production in compact, portable fuel reformers, such as CPO reactors. Furthermore, H2 and CO can be used in reducing environmentally harmful emissions. Particularly, the implementation of NOx traps and hydrogen into diesel engines has shown potential in reducing NOx emissions into the environment. Both concepts are dependent on synthesis gas generated from portable, compact fuel reformers, such as CPO reactors. Chapter 1 also reviews previous research in CPO, along with several important experimental parameters, and outlines the remaining research directions in the remaining chapters. In Chapter 2, steam addition to the CPO of higher hydrocarbons was explored over rhodium-coated ceramic foam supports at millisecond contact times. Steam addition to the CPO of n-decane and n-hexadecane in air produced considerably higher H2 and CO2 and lower olefin and CO selectivities than traditional CPO. For steam to carbon feed ratios from 0.0 to 4.0, the reactor operated autothermally, and the H2 to CO product ratio increased from ˜1.0 to ˜4.0, which is essentially the equilibrium product composition near synthesis gas stoichiometry (C/O ˜1) at contact times of ˜7 milliseconds. In fuel-rich feeds exceeding the synthesis gas ratio (C/O > 1), steam addition suppressed olefins, promoted synthesis gas and water-gas shift products, and reduced catalyst surface carbon. Furthermore, steam addition to the CPO of the military fuel JP-8 was performed successfully, also increasing H2 and suppressing olefins. (Abstract shortened by UMI.)

SoLid: An innovative anti-neutrino detector for searching oscillations at the SCK•CEN BR2 reactor

NASA Astrophysics Data System (ADS)

Abreu, Yamiel; SoLid Collaboration

2017-02-01

The SoLid experiment intends to search for active-to-sterile anti-neutrino oscillations at a very short baseline from the SCK•CEN BR2 research reactor (Mol, Belgium). A novel detector approach to measure reactor anti-neutrinos was developed based on an innovative sandwich of composite polyvinyl-toluene and 6LiF:ZnS(Ag) scintillators. The system is highly segmented and read out by a network of wavelength shifting fibers and SiPM. High experimental sensitivity can be achieved compared to other standard technologies thanks to the combination of high granularity, good neutron-gamma discrimination using 6LiF:ZnS(Ag) scintillator and precise localisation of the Inverse Beta Decay products. This technology can be considered as a new generation of an anti-neutrino detector. This compact system requires limited passive shielding and relies on spatial topology to determine the different classes of backgrounds. We will describe the principle of detection and the detector design. Particular focus on the neutron discrimination will be made, as well as on the capability to use cosmic muons for channel equalisation and energy calibration. The performance of the first 288 kg SoLid module (SM1), based on the data taken at BR2 from February to September 2015, will be presented. We will conclude with the next phase, which will start in 2016, and the future plans of the experiment.

From Confrontation to Cooperation: 8th International Seminar on Nuclear War

NASA Astrophysics Data System (ADS)

Zichichi, A.; Dardo, M.

1992-09-01

The Table of Contents for the full book PDF is as follows: * OPENING SESSION * A. Zichichi: Opening Statements * R. Nicolosi: Opening Statements * MESSAGES * CONTRIBUTIONS * "The Contribution of the Erice Seminars in East-West-North-South Scientific Relations" * 1. LASER TECHNOLOGY * "Progress in laser technology" * "Progress in laboratory high gain ICF: prospects for the future" * "Applications of laser in metallurgy" * "Laser tissue interactions in medicine and surgery" * "Laser fusion" * "Compact X-ray lasers in the laboratory" * "Alternative method for inertial confinement" * "Laser technology in China" * 2. NUCLEAR AND CHEMICAL SAFETY * "Reactor safety and reactor design" * "Thereotical analysis and numerical modelling of heat transfer and fuel migration in underlying soils and constructive elements of nuclear plants during an accident release from the core" * "How really to attain reactor safely" * "The problem of chemical weapons" * "Long terms genetic effects of nuclear and chemical accidents" * "Features of the brain which are of importance in understanding the mode of operation of toxic substances and of radiation" * "CO2 and ultra safe reactors" * 3. USE OF MISSILES * "How to convert INF technology for peaceful scientific purposes" * "Beating words into plowshares: a proposal for the peaceful uses of retired nuclear warheads" * "Some thoughts on the peaceful use of retired nuclear warheads" * "Status of the HEFEST project" * 4. OZONE * "Status of the ozone layer problem" * 5. CONVENTIONAL AND NUCLEAR FORCE RESTRUCTURING IN EUROPE * 6. CONFLICT AVOIDANCE MODEL * 7. GENERAL DISCUSSION OF THE WORLD LAB PROJECTS * "East-West-North-South Collaboration in Subnuclear Physics" * "Status of the World Lab in the USSR" * CLOSING SESSION

Determination of Trace Concentration in TMD Detectors using PGAA

NASA Astrophysics Data System (ADS)

Tomandl, I.; Viererbl, L.; Kudějová, P.; Lahodová, Z.; Klupák, V.; Fikrle, M.

2015-05-01

Transmutation detectors could be alternative to the traditional activation detector method for neutron fluence dosimetry at power nuclear reactors. This new method require an isotopically highly-sensitive, non-destructive in sense of compactness as well as isotopic content, precise and standardly used analytical method for trace concentration determination. The capability of Prompt Gamma-ray Activation Analysis (PGAA) for determination of trace concentrations of transmuted stable nuclides in the metallic foils of Ni, Au, Cu and Nb, which were irradiated for 21 days in the reactor core at the LVR-15 research reactor in Řež, is reported. The PGAA measurements of these activation foils were performed at the PGAA facility at Forschungs-Neutronenquelle Heinz Maier-Leibnitz (FRMII) in Garching.

Effect of end-point compaction on superpave hot mix asphalt (HMA) mix designs.

DOT National Transportation Integrated Search

2004-01-09

In the Superpave hot mix asphalt (HMA) mix design system, gyratory specime ns are compacted to varying levels of initial (Ninitial), : design (Ndesign) and maximum (Nmaximum) gyrations. Initially, in the Superpave system, specimens were compacted to ...

PREPARATION OF HIGH-DENSITY, COMPACTIBLE THORIUM OXIDE PARTICLES

DOEpatents

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

1962-05-22

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

Progress towards an Optimization Methodology for Combustion-Driven Portable Thermoelectric Power Generation Systems

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

Krishnan, Shankar; Karri, Naveen K.; Gogna, Pawan K.

2012-03-13

Enormous military and commercial interests exist in developing quiet, lightweight, and compact thermoelectric (TE) power generation systems. This paper investigates design integration and analysis of an advanced TE power generation system implementing JP-8 fueled combustion and thermal recuperation. Design and development of a portable TE power system using a JP-8 combustor as a high temperature heat source and optimal process flows depend on efficient heat generation, transfer, and recovery within the system are explored. Design optimization of the system required considering the combustion system efficiency and TE conversion efficiency simultaneously. The combustor performance and TE sub-system performance were coupled directlymore » through exhaust temperatures, fuel and air mass flow rates, heat exchanger performance, subsequent hot-side temperatures, and cold-side cooling techniques and temperatures. Systematic investigation of this system relied on accurate thermodynamic modeling of complex, high-temperature combustion processes concomitantly with detailed thermoelectric converter thermal/mechanical modeling. To this end, this work reports on design integration of systemlevel process flow simulations using commercial software CHEMCADTM with in-house thermoelectric converter and module optimization, and heat exchanger analyses using COMSOLTM software. High-performance, high-temperature TE materials and segmented TE element designs are incorporated in coupled design analyses to achieve predicted TE subsystem level conversion efficiencies exceeding 10%. These TE advances are integrated with a high performance microtechnology combustion reactor based on recent advances at the Pacific Northwest National Laboratory (PNNL). Predictions from this coupled simulation established a basis for optimal selection of fuel and air flow rates, thermoelectric module design and operating conditions, and microtechnology heat-exchanger design criteria. This paper will discuss this simulation process that leads directly to system efficiency power maps defining potentially available optimal system operating conditions and regimes. This coupled simulation approach enables pathways for integrated use of high-performance combustor components, high performance TE devices, and microtechnologies to produce a compact, lightweight, combustion driven TE power system prototype that operates on common fuels.« less

Investigation of a Possibility of Chromium-51 Accumulation in the SM-3 Reactor to Fabricate a Neutrino Source

NASA Astrophysics Data System (ADS)

Romanov, E. G.; Gavrin, V. N.; Tarasov, V. A.; Malkov, A. P.; Kupriyanov, A. V.; Danshin, S. N.; Veretenkin, E. P.

2017-01-01

Compact high intensity neutrino sources based on 51Cr isotope are demanded for very short baseline neutrino experiments. In particular, a 3 MCi 51Cr neutrino source is needed for the experiment BEST on search for transitions of electron neutrinos to sterile states. The paper presents the results of the analysis of options of the irradiation of highly enriched 50Cr in the existing trap of thermal neutrons of high-flux reactor SM-3, as well as using the most promising variants of the trap after upcoming reconstruction of the reactor. It is shown that it is possible to to obtain the intensity of 51Cr up to 3.85 MCi at the end of irradiation of 50Cr enriched to 97% in the high-flux reactor SM-3 of the JSC “SSC NIIAR”.

Human factors engineering verification and validation for APR1400 computerized control room

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

Shin, Y. C.; Moon, H. K.; Kim, J. H.

2006-07-01

This paper introduces the Advanced Power Reactor 1400 (APR1400) HFE V and V activities the Korea Hydro Nuclear Plant Co. LTD. (KHNP) has performed for the last 10 years and some of the lessons learned through these activities. The features of APR1400 main control room include large display panel, redundant compact workstations, computer-based procedure, and safety console. Several iterations of human factors evaluations have been performed from small scale proof of concept tests to large scale integrated system tests for identifying human engineering deficiencies in the human system interface design. Evaluations in the proof of concept test were focused onmore » checking the presence of any show stopper problems in the design concept. Later evaluations were mostly for finding design problems and for assuring the resolution of human factors issues of advanced control room. The results of design evaluations were useful not only for refining the control room design, but also for licensing the standard design. Several versions of APR1400 mock-ups with dynamic simulation models of currently operating Korea Standard Nuclear Plant (KSNP) have been used for the evaluations with the participation of operators from KSNP plants. (authors)« less

Compact CFB: The next generation CFB boiler

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

Utt, J.

1996-12-31

The next generation of compact circulating fluidized bed (CFB) boilers is described in outline form. The following topics are discussed: compact CFB = pyroflow + compact separator; compact CFB; compact separator is a breakthrough design; advantages of CFB; new design with substantial development history; KUHMO: successful demo unit; KUHMO: good performance over load range with low emissions; KOKKOLA: first commercial unit and emissions; KOKKOLA: first commercial unit and emissions; compact CFB installations; next generation CFB boiler; grid nozzle upgrades; cast segmented vortex finders; vortex finder installation; ceramic anchors; pre-cast vertical bullnose; refractory upgrades; and wet gunning.

Development of compact fuel processor for 2 kW class residential PEMFCs

NASA Astrophysics Data System (ADS)

Seo, Yu Taek; Seo, Dong Joo; Jeong, Jin Hyeok; Yoon, Wang Lai

Korea Institute of Energy Research (KIER) has been developing a novel fuel processing system to provide hydrogen rich gas to residential polymer electrolyte membrane fuel cells (PEMFCs) cogeneration system. For the effective design of a compact hydrogen production system, the unit processes of steam reforming, high and low temperature water gas shift, steam generator and internal heat exchangers are thermally and physically integrated into a packaged hardware system. Several prototypes are under development and the prototype I fuel processor showed thermal efficiency of 73% as a HHV basis with methane conversion of 81%. Recently tested prototype II has been shown the improved performance of thermal efficiency of 76% with methane conversion of 83%. In both prototypes, two-stage PrOx reactors reduce CO concentration less than 10 ppm, which is the prerequisite CO limit condition of product gas for the PEMFCs stack. After confirming the initial performance of prototype I fuel processor, it is coupled with PEMFC single cell to test the durability and demonstrated that the fuel processor is operated for 3 days successfully without any failure of fuel cell voltage. Prototype II fuel processor also showed stable performance during the durability test.

Microstructural evolution of nanochannel CrN films under ion irradiation at elevated temperature and post-irradiation annealing

NASA Astrophysics Data System (ADS)

Tang, Jun; Hong, Mengqing; Wang, Yongqiang; Qin, Wenjing; Ren, Feng; Dong, Lan; Wang, Hui; Hu, Lulu; Cai, Guangxu; Jiang, Changzhong

2018-03-01

High-performance radiation tolerance materials are crucial for the success of future advanced nuclear reactors. In this paper, we present a further investigation that the "vein-like" nanochannel films can enhance radiation tolerance under ion irradiation at high temperature and post-irradiation annealing. The chromium nitride (CrN) nanochannel films with different nanochannel densities and the compact CrN film are chosen as a model system for these studies. Microstructural evolution of these films were investigated using Transmission Electron Microscopy (TEM), Scanning Electron Microscopy (SEM), Elastic Recoil Detection (ERD) and Grazing Incidence X-ray Diffraction (GIXRD). Under the high fluence He+ ion irradiation at 500 °C, small He bubbles with low bubble densities are observed in the irradiated nanochannel CrN films, while the aligned large He bubbles, blistering and texture reconstruction are found in the irradiated compact CrN film. For the heavy Ar2+ ion irradiation at 500 °C, the microstructure of the nanochannel CrN RT film is more stable than that of the compact CrN film due to the effective releasing of defects via the nanochannel structure. Under the He+ ion irradiation and subsequent annealing, compared with the compact film, the nanochannel films have excellent performance for the suppression of He bubble growth and possess the strong microstructural stability. Basing on the analysis on the sizes and number densities of bubbles as well as the concentrations of He retained in the nanochannel CrN films and the compact CrN film under different experimental conditions, potential mechanism for the enhanced radiation tolerance are discussed. Nanochannels play a crucial role on the release of He/defects under ion irradiation. We conclude that the tailored "vein-like" nanochannel structure may be used as advanced radiation tolerance materials for future nuclear reactors.

Testing of an advanced thermochemical conversion reactor system

NASA Astrophysics Data System (ADS)

1990-01-01

This report presents the results of work conducted by MTCI to verify and confirm experimentally the ability of the MTCI gasification process to effectively generate a high-quality, medium-Btu gas from a wider variety of feedstock and waste than that attainable in air-blown, direct gasification systems. The system's overall simplicity, due to the compact nature of the pulse combustor, and the high heat transfer rates attainable within the pulsating flow resonance tubes, provide a decided and near-term potential economic advantage for the MTCI indirect gasification system. The primary objective was the design, construction, and testing of a Process Design Verification System for an indirectly heated, thermochemical fluid-bed reactor and a pulse combustor an an integrated system that can process alternative renewable sources of energy such as biomass, black liquor, municipal solid waste and waste hydrocarbons, including heavy oils into a useful product gas. The test objectives for the biomass portion of this program were to establish definitive performance data on biomass feedstocks covering a wide range of feedstock qualities and characteristics. The test objectives for the black liquor portion of this program were to verify the operation of the indirect gasifier on commercial black liquor containing 65 percent solids at several temperature levels and to characterize the bed carbon content, bed solids particle size and sulfur distribution as a function of gasification conditions.

Application Research of Quality Control Technology of Asphalt Pavement based on GPS Intelligent

NASA Astrophysics Data System (ADS)

Wang, Min; Gao, Bo; Shang, Fei; Wang, Tao

2017-10-01

Due to the difficulty of steel deck pavement asphalt layer compaction caused by the effect of the flexible supporting system (orthotropic steel deck plate), it is usually hard and difficult to control for the site compactness to reach the design goal. The intelligent compaction technology is based on GPS control technology and real-time acquisition of actual compaction tracks, and then forms a cloud maps of compaction times, which guide the roller operator to do the compaction in accordance with the design requirement to ensure the deck compaction technology and compaction quality. From the actual construction situation of actual bridge and checked data, the intelligent compaction technology is significant in guaranteeing the steel deck asphalt pavement compactness and quality stability.

Design and characterization of an irradiation facility with real-time monitoring

NASA Astrophysics Data System (ADS)

Braisted, Jonathan David

Radiation causes performance degradation in electronics by inducing atomic displacements and ionizations. While radiation hardened components are available, non-radiation hardened electronics can be preferable because they are generally more compact, require less power, and less expensive than radiation tolerant equivalents. It is therefore important to characterize the performance of electronics, both hardened and non-hardened, to prevent costly system or mission failures. Radiation effects tests for electronics generally involve a handful of step irradiations, leading to poorly-resolved data. Step irradiations also introduce uncertainties in electrical measurements due to temperature annealing effects. This effect may be intensified if the time between exposure and measurement is significant. Induced activity in test samples also complicates data collection of step irradiated test samples. The University of Texas at Austin operates a 1.1 MW Mark II TRIGA research reactor. An in-core irradiation facility for radiation effects testing with a real-time monitoring capability has been designed for the UT TRIGA reactor. The facility is larger than any currently available non-central location in a TRIGA, supporting testing of larger electronic components as well as other in-core irradiation applications requiring significant volume such as isotope production or neutron transmutation doping of silicon. This dissertation describes the design and testing of the large in-core irradiation facility and the experimental campaign developed to test the real-time monitoring capability. This irradiation campaign was performed to test the real-time monitoring capability at various reactor power levels. The device chosen for characterization was the 4N25 general-purpose optocoupler. The current transfer ratio, which is an important electrical parameter for optocouplers, was calculated as a function of neutron fluence and gamma dose from the real-time voltage measurements. The resultant radiation effects data was seen to be repeatable and exceptionally finely-resolved. Therefore, the capability at UT TRIGA has been proven competitive with world-class effects characterization facilities.

Comparison of Homogeneous and Heterogeneous CFD Fuel Models for Phase I of the IAEA CRP on HTR Uncertainties Benchmark

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

Gerhard Strydom; Su-Jong Yoon

2014-04-01

Computational Fluid Dynamics (CFD) evaluation of homogeneous and heterogeneous fuel models was performed as part of the Phase I calculations of the International Atomic Energy Agency (IAEA) Coordinate Research Program (CRP) on High Temperature Reactor (HTR) Uncertainties in Modeling (UAM). This study was focused on the nominal localized stand-alone fuel thermal response, as defined in Ex. I-3 and I-4 of the HTR UAM. The aim of the stand-alone thermal unit-cell simulation is to isolate the effect of material and boundary input uncertainties on a very simplified problem, before propagation of these uncertainties are performed in subsequent coupled neutronics/thermal fluids phasesmore » on the benchmark. In many of the previous studies for high temperature gas cooled reactors, the volume-averaged homogeneous mixture model of a single fuel compact has been applied. In the homogeneous model, the Tristructural Isotropic (TRISO) fuel particles in the fuel compact were not modeled directly and an effective thermal conductivity was employed for the thermo-physical properties of the fuel compact. On the contrary, in the heterogeneous model, the uranium carbide (UCO), inner and outer pyrolytic carbon (IPyC/OPyC) and silicon carbide (SiC) layers of the TRISO fuel particles are explicitly modeled. The fuel compact is modeled as a heterogeneous mixture of TRISO fuel kernels embedded in H-451 matrix graphite. In this study, a steady-state and transient CFD simulations were performed with both homogeneous and heterogeneous models to compare the thermal characteristics. The nominal values of the input parameters are used for this CFD analysis. In a future study, the effects of input uncertainties in the material properties and boundary parameters will be investigated and reported.« less

PREPARATION OF UO$sub 2$ FOR NUCLEAR REACTOR FUEL PELLETS

DOEpatents

Googin, J.M.

1962-06-01

A method is given for preparing high-density UO/sub 2/ compacts. An aqueous uranyl fluoride solution is contacted with an aqueous ammonium hydroxide solution at an ammonium to-uranium ratio of 25: 1 to 30:1 to form a precipitate. The precipitate is separated from the- mother liquor, dried, and contacted with steam at a uniform temperature within the range of 400 to 650 deg C to produce U/ sub 3/O/sub 8/. The U/sub 3/O/sub 8/ is red uced to UO/sub 2/ with hydrogen at a uniform temperature within the range of 550 to 600 deg C. The UO/sub 2/ is then compressed into compacts and sintered. High-density compacts are fabricated to close tolerances without use of a binder and without machining or grinding. (AEC)

Space reactor power 1986 - A year of choices and transition

NASA Technical Reports Server (NTRS)

Wiley, R. L.; Verga, R. L.; Schnyer, A. D.; Sholtis, J. A., Jr.; Wahlquist, E. J.

1986-01-01

Both the SP-100 and Multimegawatt programs have made significant progress over the last year and that progress is the focus of this paper. In the SP-100 program the thermoelectric energy conversion concept powered by a compact, high-temperature, lithium-cooled, uranium-nitride-fueled fast spectrum reactor was selected for engineering development and ground demonstration testing at an electrical power level of 300 kilowatts. In the Multimegawatt program, activities moved from the planning phase into one of technology development and assessment with attendant preliminary definition and evaluation of power concepts against requirements of the Strategic Defense Initiative.

Simulation of High-Beta Plasma Confinement

NASA Astrophysics Data System (ADS)

Font, Gabriel; Welch, Dale; Mitchell, Robert; McGuire, Thomas

2017-10-01

The Lockheed Martin Compact Fusion Reactor concept utilizes magnetic cusps to confine the plasma. In order to minimize losses through the axial and ring cusps, the plasma is pushed to a high-beta state. Simulations were made of the plasma and magnetic field system in an effort to quantify particle confinement times and plasma behavior characteristics. Computations are carried out with LSP using implicit PIC methods. Simulations of different sub-scale geometries at high-Beta fusion conditions are used to determine particle loss scaling with reactor size, plasma conditions, and gyro radii. ©2017 Lockheed Martin Corporation. All Rights Reserved.

A strategy for intensive production of molybdenum-99 isotopes for nuclear medicine using CANDU reactors.

PubMed

Morreale, A C; Novog, D R; Luxat, J C

2012-01-01

Technetium-99m is an important medical isotope utilized worldwide in nuclear medicine and is produced from the decay of its parent isotope, molybdenum-99. The online fueling capability and compact fuel of the CANDU(®)(1) reactor allows for the potential production of large quantities of (99)Mo. This paper proposes (99)Mo production strategies using modified target fuel bundles loaded into CANDU fuel channels. Using a small group of channels a yield of 89-113% of the weekly world demand for (99)Mo can be obtained. Copyright © 2011 Elsevier Ltd. All rights reserved.

Spherical tokamaks with plasma centre-post

NASA Astrophysics Data System (ADS)

Ribeiro, Celso

2013-10-01

The metal centre-post (MCP) in tokamaks is a structure which carries the total toroidal field current and also houses the Ohmic heating solenoid in conventional or low aspect ratio (Spherical)(ST) tokamaks. The MCP and solenoid are critical components for producing the toroidal field and for the limited Ohmic flux in STs. Constraints for a ST reactor related to these limitations lead to a minimum plasma aspect ratio of 1.4 which reduces the benefit of operation at higher betas in a more compact ST reactor. Replacing the MCP is of great interest for reactor-based ST studies since the device is simplified, compactness increased, and maintenance reduced. An experiment to show the feasibility of using a plasma centre-post (PCP) is being currently under construction and involves a high level of complexity. A preliminary study of a very simple PCP, which is ECR(Electron Cyclotron Resonance)-assisted and which includes an innovative fuelling system based on pellet injection, has recently been reported. This is highly suitable for an ultra-low aspect ratio tokamak (ULART) device. Advances on this PCP ECR-assisted concept within a ULART and the associated fuelling system are presented here, and will include the field topology for the PCP ECR-assisted scheme, pellet ablation modeling, and a possible global equilibrium simulation. VIE-ITCR, IAEA-CRP contr.17592, National Instruments-Costa Rica.

Health physics aspects of advanced reactor licensing reviews

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

Hinson, C.S.

1995-03-01

The last Construction Permit to be issued by the U.S. Nuclear Regulatory Commission (NRC) for a U.S. light water reactor (LWR) was granted in the late 1970s. In 1989 the NRC issued 10 CFR Part 52 which is intended to serve as a framework for the licensing of future reactor designs. The NRC is currently reviewing four different future on {open_quotes}next-generation{close_quotes} reactor designs. Two of these designs are classified as evolutionary designs (modified versions of current generation LWRs) and two are advanced designs (reactors incorporating simplified designs and passive means for accident mitigation). These {open_quotes}next-generation{close_quotes} reactor designs incorporate many innovativemore » design features which are intended to maintain personnel doses ALARA and ensure that the annual average collective dose at these reactors does not exceed 100 person-rems (1 person-sievert) per year. This paper discusses some of the ALARA design features which are incorporated in the four {open_quotes}next-generation{close_quotes} reactor designs incorporate many innovative design features which are intended to maintain personnel doses ALARA and ensure that the annual average collective dose at these reactors does not exceed 100 person-rems (1 person-sievert) per year. This paper discusses some of the ALARA design features which are incorporated in the four {open_quotes}next-generation{close_quotes} reactor designs currently being reviewed by the NRC.« less

CONTROL ROD

DOEpatents

Walker, D.E.; Matras, S.

1963-04-30

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

Gyrotron-driven high current ECR ion source for boron-neutron capture therapy neutron generator

NASA Astrophysics Data System (ADS)

Skalyga, V.; Izotov, I.; Golubev, S.; Razin, S.; Sidorov, A.; Maslennikova, A.; Volovecky, A.; Kalvas, T.; Koivisto, H.; Tarvainen, O.

2014-12-01

Boron-neutron capture therapy (BNCT) is a perspective treatment method for radiation resistant tumors. Unfortunately its development is strongly held back by a several physical and medical problems. Neutron sources for BNCT currently are limited to nuclear reactors and accelerators. For wide spread of BNCT investigations more compact and cheap neutron source would be much more preferable. In present paper an approach for compact D-D neutron generator creation based on a high current ECR ion source is suggested. Results on dense proton beams production are presented. A possibility of ion beams formation with current density up to 600 mA/cm2 is demonstrated. Estimations based on obtained experimental results show that neutron target bombarded by such deuteron beams would theoretically yield a neutron flux density up to 6·1010 cm-2/s. Thus, neutron generator based on a high-current deuteron ECR source with a powerful plasma heating by gyrotron radiation could fulfill the BNCT requirements significantly lower price, smaller size and ease of operation in comparison with existing reactors and accelerators.

Development of a repetitive compact torus injector

NASA Astrophysics Data System (ADS)

Onchi, Takumi; McColl, David; Dreval, Mykola; Rohollahi, Akbar; Xiao, Chijin; Hirose, Akira; Zushi, Hideki

2013-10-01

A system for Repetitive Compact Torus Injection (RCTI) has been developed at the University of Saskatchewan. CTI is a promising fuelling technology to directly fuel the core region of tokamak reactors. In addition to fuelling, CTI has also the potential for (a) optimization of density profile and thus bootstrap current and (b) momentum injection. For steady-state reactor operation, RCTI is necessary. The approach to RCTI is to charge a storage capacitor bank with a large capacitance and quickly charge the CT capacitor bank through a stack of integrated-gate bipolar transistors (IGBTs). When the CT bank is fully charged, the IGBT stack will be turned off to isolate banks, and CT formation/acceleration sequence will start. After formation of each CT, the fast bank will be replenished and a new CT will be formed and accelerated. Circuits for the formation and the acceleration in University of Saskatchewan CT Injector (USCTI) have been modified. Three CT shots at 10 Hz or eight shots at 1.7 Hz have been achieved. This work has been sponsored by the CRC and NSERC, Canada.

Zirconium Hydride Space Power Reactor design.

NASA Technical Reports Server (NTRS)

Asquith, J. G.; Mason, D. G.; Stamp, S.

1972-01-01

The Zirconium Hydride Space Power Reactor being designed and fabricated at Atomics International is intended for a wide range of potential applications. Throughout the program a series of reactor designs have been evaluated to establish the unique requirements imposed by coupling with various power conversion systems and for specific applications. Current design and development emphasis is upon a 100 kilowatt thermal reactor for application in a 5 kwe thermoelectric space power generating system, which is scheduled to be fabricated and ground tested in the mid 70s. The reactor design considerations reviewed in this paper will be discussed in the context of this 100 kwt reactor and a 300 kwt reactor previously designed for larger power demand applications.

HYDRAULIC SERVO CONTROL MECHANISM

DOEpatents

Hussey, R.B.; Gottsche, M.J. Jr.

1963-09-17

A hydraulic servo control mechanism of compact construction and low fluid requirements is described. The mechanism consists of a main hydraulic piston, comprising the drive output, which is connected mechanically for feedback purposes to a servo control piston. A control sleeve having control slots for the system encloses the servo piston, which acts to cover or uncover the slots as a means of controlling the operation of the system. This operation permits only a small amount of fluid to regulate the operation of the mechanism, which, as a result, is compact and relatively light. This mechanism is particuiarly adaptable to the drive and control of control rods in nuclear reactors. (auth)

METHOD OF MAKING FUEL BODIES

DOEpatents

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

1962-04-24

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

Secondary Heat Exchanger Design and Comparison for Advanced High Temperature Reactor

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

Piyush Sabharwall; Ali Siahpush; Michael McKellar

2012-06-01

The goals of next generation nuclear reactors, such as the high temperature gas-cooled reactor and advance high temperature reactor (AHTR), are to increase energy efficiency in the production of electricity and provide high temperature heat for industrial processes. The efficient transfer of energy for industrial applications depends on the ability to incorporate effective heat exchangers between the nuclear heat transport system and the industrial process heat transport system. The need for efficiency, compactness, and safety challenge the boundaries of existing heat exchanger technology, giving rise to the following study. Various studies have been performed in attempts to update the secondarymore » heat exchanger that is downstream of the primary heat exchanger, mostly because its performance is strongly tied to the ability to employ more efficient conversion cycles, such as the Rankine super critical and subcritical cycles. This study considers two different types of heat exchangers—helical coiled heat exchanger and printed circuit heat exchanger—as possible options for the AHTR secondary heat exchangers with the following three different options: (1) A single heat exchanger transfers all the heat (3,400 MW(t)) from the intermediate heat transfer loop to the power conversion system or process plants; (2) Two heat exchangers share heat to transfer total heat of 3,400 MW(t) from the intermediate heat transfer loop to the power conversion system or process plants, each exchanger transfers 1,700 MW(t) with a parallel configuration; and (3) Three heat exchangers share heat to transfer total heat of 3,400 MW(t) from the intermediate heat transfer loop to the power conversion system or process plants. Each heat exchanger transfers 1,130 MW(t) with a parallel configuration. A preliminary cost comparison will be provided for all different cases along with challenges and recommendations.« less

Composting on Mars or the Moon: I. Comparative evaluation of process design alternatives

NASA Technical Reports Server (NTRS)

Finstein, M. S.; Strom, P. F.; Hogan, J. A.; Cowan, R. M.; Janes, H. W. (Principal Investigator)

1999-01-01

As a candidate technology for treating solid wastes and recovering resources in bioregenerative Advanced Life Support, composting potentially offers such advantages as compactness, low mass, near ambient reactor temperatures and pressures, reliability, flexibility, simplicity, and forgiveness of operational error or neglect. Importantly, the interactions among the physical, chemical, and biological factors that govern composting system behavior are well understood. This article comparatively evaluates five Generic Systems that describe the basic alternatives to composting facility design and control. These are: 1) passive aeration; 2) passive aeration abetted by mechanical agitation; 3) forced aeration--O2 feedback control; 4) forced aeration--temperature feedback control; 5) forced aeration--integrated O2 and temperature feedback control. Each of the five has a distinctive pattern of behavior and process performance characteristics. Only Systems 4 and 5 are judged to be viable candidates for ALS on alien worlds, though which is better suited in this application is yet to be determined.

UV Disinfection of Wastewater and Combined Sewer Overflows.

PubMed

Gibson, John; Drake, Jennifer; Karney, Bryan

2017-01-01

Municipal wastewater contains bacteria, viruses, and other pathogens that adversely affect the environment, human health, and economic activity. One way to mitigate these effects is a final disinfection step using ultraviolet light (UVL). The advantages of UVL disinfection, when compared to the more traditional chlorine, include no chlorinated by-products, no chemical residual, and relatively compact size. The design of most UV reactors is complex. It involves lamp selection, power supply design, optics, and hydraulics. In general, medium pressure lamps are more compact, powerful, and emit over a wider range of light than the more traditional low pressure lamps. Low pressure lamps, however, may be electrically more efficient. In UV disinfection, the fraction of surviving organisms (e.g. E. coli) will decrease exponentially with increasing UV dose. However, the level of disinfection that can be achieved is often limited by particle-associated organisms. Efforts to remove or reduce the effects of wastewater particles will often improve UV disinfection effectiveness. Regrowth, photoreactivation, or dark repair after UV exposure are sometimes cited as disadvantages of UV disinfection. Research is continuing in this area, however there is little evidence that human pathogens can photoreactivate in environmental conditions, at doses used in wastewater treatment. The UV disinfection of combined sewer overflows, a form of wet weather pollution, is challenging and remains largely at the research phase. Pre-treatment of combined sewer overflows (CSOs) with a cationic polymer to induce fast settling, and a low dose of alum to increase UV transmittance, has shown promise at the bench scale.

Compact and Lightweight Sabatier Reactor for Carbon Dioxide Reduction

NASA Technical Reports Server (NTRS)

Junaedi, Christian; Hawley, Kyle; Walsh, Dennis; Roychoudhury, Subir; Abney, Morgan B.; Perry, Jay L.

2011-01-01

The utilization of CO2 to produce life support consumables, such as O2 and H2O, via the Sabatier reaction is an important aspect of NASA s cabin Atmosphere Revitalization System and In-Situ Resource Utilization architectures for both low-earth orbit and long-term manned space missions. In the current International Space Station (ISS) and other low orbit missions, metabolically-generated CO2 is removed from the cabin air and vented into space, resulting in a net loss of O2. This requires a continuous resupply of O2 via water electrolysis, and thus highlights the need for large water storage capacity. For long-duration space missions, the amount of life support consumables is limited and resupply options are practically nonexistent, thus atmosphere resource management and recycle becomes crucial to significantly reduce necessary O2 and H2O storage. Additionally, the potential use of the Martian CO2-rich atmosphere and Lunar regolith to generate life support consumables and propellant fuels is of interest to NASA. Precision Combustion, Inc. (PCI) has developed a compact, lightweight Microlith(Registered TradeMark)-based Sabatier (CO2 methanation) reactor which demonstrates the capability of achieving high CO2 conversion and near 100% CH4 selectivity at space velocities of 30,000-60,000 hr-1. The combination of the Microlith(Registered TradeMark) substrates and durable, novel catalyst coating permitted efficient Sabatier reactor operation that favors high reactant conversion, high selectivity, and long-term durability. This paper presents the reactor development and performance results at various operating conditions. Additionally, results from 100-hr durability tests and mechanical vibration tests are discussed.

The UASB reactor as an alternative for the septic tank for on-site sewage treatment.

PubMed

Coelho, A L S S; do Nascimento, M B H; Cavalcanti, P F F; van Haandel, A C

2003-01-01

Although septic tanks are amply used for on site sewage treatment, these units have serious drawbacks: the removal efficiency of organic material and suspended solids is low, the units are costly and occupy a large area and operational cost is high due to the need for periodic desludging. In this paper an innovative variant of the UASB reactor is proposed as an alternative for the septic tank. This alternative has several important advantages in comparison with the conventional septic tank: (1) Although the volume of the UASB reactor was about 4 times smaller than the septic tank, its effluent quality was superior, even though small sludge particles were present, (2) desludging of the UASB reactor is unnecessary and even counterproductive, as the sludge mass guarantees proper performance, (3) the UASB reactor is easily transportable (compact and light) and therefore can be produced in series, strongly reducing construction costs and (4) since the concentration of colloids in the UASB effluent is much smaller than in the ST effluent, it is expected that the infiltration of the effluent will be much less problematic.

Search for eV Sterile Neutrinos - The Stereo Experiment

NASA Astrophysics Data System (ADS)

Haser, J.; Stereo Collaboration

2017-07-01

In the recent years, major milestones in neutrino physics were accomplished at nuclear reactors: the smallest neutrino mixing angle $\\theta_{13}$ was determined with high precision and the emitted antineutrino spectrum was measured at unprecedented resolution. However, two anomalies, the first one related to the absolute flux and the second one to the spectral shape, have yet to be solved. The flux anomaly is known as the Reactor Antineutrino Anomaly and could be caused by the existence of a light sterile neutrino participating in the neutrino oscillation phenomenon. Introducing a sterile state implies the presence of a fourth mass eigenstate, global fits favour oscillation parameters around $\\sin^2({2\\theta}) \\approx 0.09$ and $\\Delta m^2 \\approx 1\\,\\mathrm{eV}^2$. The Stereo experiment was built to finally solve this puzzle. It is one of the first running experiments built to search for eV sterile neutrinos and takes data since end of 2016 at ILL Grenoble (France). At a short baseline of 10 metres, it measures the antineutrino flux and spectrum emitted by a compact research reactor. The segmentation of the detector in six target cells allows for measurements of the neutrino spectrum at multiple baselines. An active-sterile flavour oscillation could be unambiguously detected, as it distorts the spectral shape of each cell's measurement differently. This contribution gives an overview on the Stereo experiment, along with details on the detector design, detection principle and the current status of data analysis.

Plasma catalytic reforming of methane

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

Bromberg, L.; Cohn, D.R.; Rabinovich, A.

1998-08-01

Thermal plasma technology can be efficiently used in the production of hydrogen and hydrogen-rich gases from methane and a variety of fuels. This paper describes progress in plasma reforming experiments and calculations of high temperature conversion of methane using heterogeneous processes. The thermal plasma is a highly energetic state of matter that is characterized by extremely high temperatures (several thousand degrees Celsius) and high degree of dissociation and substantial degree of ionization. The high temperatures accelerate the reactions involved in the reforming process. Hydrogen-rich gas (50% H{sub 2}, 17% CO and 33% N{sub 2}, for partial oxidation/water shifting) can bemore » efficiently made in compact plasma reformers. Experiments have been carried out in a small device (2--3 kW) and without the use of efficient heat regeneration. For partial oxidation/water shifting, it was determined that the specific energy consumption in the plasma reforming processes is 16 MJ/kg H{sub 2} with high conversion efficiencies. Larger plasmatrons, better reactor thermal insulation, efficient heat regeneration and improved plasma catalysis could also play a major role in specific energy consumption reduction and increasing the methane conversion. A system has been demonstrated for hydrogen production with low CO content ({approximately} 1.5%) with power densities of {approximately} 30 kW (H{sub 2} HHV)/liter of reactor, or {approximately} 10 m{sup 3}/hr H{sub 2} per liter of reactor. Power density should further increase with increased power and improved design.« less

Wastewater treatment using a novel bioreactor with submerged packing bed of polyethylene tape.

PubMed

Mijaylova Nacheva, P; Moeller Chávez, G

2010-01-01

The performance of a novel aerobic bioreactor with a specially designed submerged packing bed of high specific surface area density, made of polyethylene tape, was studied for the treatment of domestic wastewater. The reactor has a volume of 0.71 m(3) and the specific area of the packing bed was 1,098 m(2)/m(3). The operation was performed with and without effluent recycling, applying different organic loads in the range of 4.0-17.6 g COD m(-2) d(-1). No back-washings were carried out. Overall BOD(5) removals of 90-95% were obtained with organic loads of 4.0-17.6 g COD m(-2) d(-1) and HRT of 0.2-1.1 h. Overall TN removal of 69-72% was obtained at loads of 0.8-4.6 g TN m(-2) d(-1) when effluent recycling was used. The reactor allowed obtaining high quality water for urban reuse and demonstrated an effective process performance and resistance to load variations. The developed biofilm was completely penetrated by the organic matter, ammonia and oxygen, providing high removal rates. Large biomass quantities, up to 13 g dry VS/m(2), were reached in the reactor and the determined sludge yield coefficient was relatively low, of 0.25 g VSS/g COD. These results allow obtaining compact treatment systems with low sludge production and make the technology a suitable option for small wastewater treatment plants.

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.

Evaluating Georgia DOT's compaction requirements for stone matrix asphalt mixes.

DOT National Transportation Integrated Search

2006-06-01

This study determined a compactive effort for Stone Mastic Asphalt (SMA) mixes with the Superpave gyratory compactor (SGC) that would match a 50-blow Marshall compactive effort using aggregates and mix designs common in Georgia. SMA mix designs were ...

Design and experimental evaluation of compact radial-inflow turbines

NASA Technical Reports Server (NTRS)

Fredmonski, A. J.; Huber, F. W.; Roelke, R. J.; Simonyi, S.

1991-01-01

The application of a multistage 3D Euler solver to the aerodynamic design of two compact radial-inflow turbines is presented, along with experimental results evaluating and validating the designs. The objectives of the program were to design, fabricate, and rig test compact radial-inflow turbines with equal or better efficiency relative to conventional designs, while having 40 percent less rotor length than current traditionally-sized radial turbines. The approach to achieving these objectives was to apply a calibrated 3D multistage Euler code to accurately predict and control the high rotor flow passage velocities and high aerodynamic loadings resulting from the reduction in rotor length. A comparison of the advanced compact designs to current state-of-the-art configurations is presented.

Development of the Packed Bed Reactor ISS Flight Experiment

NASA Technical Reports Server (NTRS)

Patton, Martin O.; Bruzas, Anthony E.; Rame, Enrique; Motil, Brian J.

2012-01-01

Packed bed reactors are compact, require minimum power and maintenance to operate, and are highly reliable. These features make this technology a leading candidate as a potential unit operation in support of long duration human space exploration. On earth, this type of reactor accounts for approximately 80% of all the reactors used in the chemical process industry today. Development of this technology for space exploration is truly crosscutting with many other potential applications (e.g., in-situ chemical processing of planetary materials and transport of nutrients through soil). NASA is developing an ISS experiment to address this technology with particular focus on water reclamation and air revitalization. Earlier research and development efforts funded by NASA have resulted in two hydrodynamic models which require validation with appropriate instrumentation in an extended microgravity environment. The first model developed by Motil et al., (2003) is based on a modified Ergun equation. This model was demonstrated at moderate gas and liquid flow rates, but extension to the lower flow rates expected in many advanced life support systems must be validated. The other model, developed by Guo et al., (2004) is based on Darcy s (1856) law for two-phase flow. This model has been validated for a narrow range of flow parameters indirectly (without full instrumentation) and included test points where the flow was not fully developed. The flight experiment presented will be designed with removable test sections to test the hydrodynamic models. The experiment will provide flexibility to test additional beds with different types of packing in the future. One initial test bed is based on the VRA (Volatile Removal Assembly), a packed bed reactor currently on ISS whose behavior in micro-gravity is not fully understood. Improving the performance of this system through an accurate model will increase our ability to purify water in the space environment.

Reactor antineutrino detector iDREAM.

NASA Astrophysics Data System (ADS)

Gromov, M. B.; Lukyanchenko, G. A.; Novikova, G. J.; Obinyakov, B. A.; Oralbaev, A. Y.; Skorokhvatov, M. D.; Sukhotin, S. V.; Chepurnov, A. S.; Etenko, A. V.

2017-09-01

Industrial Detector for Reactor Antineutrino Monitoring (iDREAM) is a compact (≈ 3.5m 2) industrial electron antineutrino spectrometer. It is dedicated for remote monitoring of PWR reactor operational modes by neutrino method in real-time. Measurements of antineutrino flux from PWR allow to estimate a fuel mixture in active zone and to check the status of the reactor campaign for non-proliferation purposes. LAB-based gadolinium doped scintillator is exploited as a target. Multizone architecture of the detector with gamma-catcher surrounding fiducial volume and plastic muon veto above and below ensure high efficiency of IBD detection and background suppression. DAQ is based on Flash ADC with PSD discrimination algorithms while digital trigger is programmable and flexible due to FPGA. The prototype detector was started up in 2014. Preliminary works on registration Cerenkov radiation produced by cosmic muons were established with distilled water inside the detector in order to test electronic and slow control systems. Also in parallel a long-term measurements with different scintillator samples were conducted.

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

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

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

1992-12-31

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

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

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

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

1992-01-01

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

Addition of acetate improves stability of power generation using microbial fuel cells treating domestic wastewater.

PubMed

Stager, Jennifer L; Zhang, Xiaoyuan; Logan, Bruce E

2017-12-01

Power generation using microbial fuel cells (MFCs) must provide stable, continuous conversion of organic matter in wastewaters into electricity. However, when relatively small diameter (0.8cm) graphite fiber brush anodes were placed close to the cathodes in MFCs, power generation was unstable during treatment of low strength domestic wastewater. One reactor produced 149mW/m 2 before power generation failed, while the other reactor produced 257mW/m 2 , with both reactors exhibiting severe power overshoot in polarization tests. Using separators or activated carbon cathodes did not result in stable operation as the reactors continued to exhibit power overshoot based on polarization tests. However, adding acetate (1g/L) to the wastewater produced stable performance during fed batch and continuous flow operation, and there was no power overshoot in polarization tests. These results highlight the importance of wastewater strength and brush anode size for producing stable and continuous power in compact MFCs. Copyright © 2017 Elsevier B.V. All rights reserved.

An MFC-Based Online Monitoring and Alert System for Activated Sludge Process

PubMed Central

Xu, Gui-Hua; Wang, Yun-Kun; Sheng, Guo-Ping; Mu, Yang; Yu, Han-Qing

2014-01-01

In this study, based on a simple, compact and submersible microbial fuel cell (MFC), a novel online monitoring and alert system with self-diagnosis function was established for the activated sludge (AS) process. Such a submersible MFC utilized organic substrates and oxygen in the AS reactor as the electron donor and acceptor respectively, and could provide an evaluation on the status of the AS reactor and thus give a reliable early warning of potential risks. In order to evaluate the reliability and sensitivity of this online monitoring and alert system, a series of tests were conducted to examine the response of this system to various shocks imposed on the AS reactor. The results indicate that this online monitoring and alert system was highly sensitive to the performance variations of the AS reactor. The stability, sensitivity and repeatability of this online system provide feasibility of being incorporated into current control systems of wastewater treatment plants to real-time monitor, diagnose, alert and control the AS process. PMID:25345502

AGR-2: The first irradiation of French HTR fuel in Advanced Test Reactor

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

T. Lambert; B. Grover; P. Guillermier

AGR-2, the second irradiation of the US program for qualification of the NGNP fuel, is open to international participation within the scope of the Generation IV International Forum. In this frame, it includes in its multi-capsule irradiation rig an irradiation of French HTR fuel manufactured in the CAPRI line (GAIA facility at CEA/Cadarache and AREVA/CERCA compacting line at Romans). The AGR-2 irradiation is designed to place our first fabrications of HTR particles under operating conditions that are representative of ANTARES project while keeping close to the test range of the German fuel as much as possible, which is the referencemore » in terms of irradiation behavior. A few batches of particles and 12 fuel compacts were produced and characterized in 2009 by CEA and CERCA. The fuel main characteristics are in conformity with our specifications and in compliance with INL requirements. The AGR-2 experiment is based on the design and devices used in the first experiment of the AGR program. The design makes it possible to monitor the irradiation conditions and in particular, the temperature, the power and the fission products released from fuel particles. The in pile equipment consists of a multi-capsule device designed to simultaneously irradiate six independent capsules with temperature control. The out-of-core part consists of the equipment for actively controlling temperature and measuring the fission products release on-line. The target conditions for the irradiation experiment were defined with the aim of comparing the results obtained under irradiation with German particles along with the objectives of reaching burn-up and fluence targets to validate the behavior of our fuel in a significant range (15% FIMA – 5 × 1025 n/m2 at 600 EFPD with centerline fuel temperature about 1100 degrees C). These conditions have to be representative of ANTARES project characteristics. These target conditions were compared with final results from neutron and thermal design studies performed by INL team, and preliminary thermal mechanical ATLAS calculations were carried out by CEA from this pre-design. Despite the mean burn-up achieved in approximately 600 EFPD being a little high (16.3% FIMA max. associated with a low fluence up to 2.85 × 1025 n/m2), this irradiation will nevertheless encompass the range of irradiation effects covered in our experimental objectives (maximum stress peak at start of irradiation then sign inversion of the stress in the SiC layer). In addition, the fluence and burn-up acceleration factors are very similar to those of the German reference experiments. This experimental irradiation began in July 2010 in the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) and first results have been acquired.« less

A novel plant protection strategy for transient reactors

NASA Astrophysics Data System (ADS)

Bhattacharyya, Samit K.; Lipinski, Walter C.; Hanan, Nelson A.

A novel plant protection system designed for use in the TREAT Upgrade (TU) reactor is described. The TU reactor is designed for controlled transient operation in the testing of reactor fuel behavior under simulated reactor accident conditions. Safe operation of the reactor is of paramount importance and the Plant Protection System (PPS) had to be designed to exacting requirements. Researchers believe that the strategy developed for the TU has potential application to the multimegawatt space reactors and represents the state of the art in terrestrial transient reactor protection systems.

Design and manufacture of a D-shape coil-based toroid-type HTS DC reactor using 2nd generation HTS wire

NASA Astrophysics Data System (ADS)

Kim, Kwangmin; Go, Byeong-Soo; Sung, Hae-Jin; Park, Hea-chul; Kim, Seokho; Lee, Sangjin; Jin, Yoon-Su; Oh, Yunsang; Park, Minwon; Yu, In-Keun

2014-09-01

This paper describes the design specifications and performance of a real toroid-type high temperature superconducting (HTS) DC reactor. The HTS DC reactor was designed using 2G HTS wires. The HTS coils of the toroid-type DC reactor magnet were made in the form of a D-shape. The target inductance of the HTS DC reactor was 400 mH. The expected operating temperature was under 20 K. The electromagnetic performance of the toroid-type HTS DC reactor magnet was analyzed using the finite element method program. A conduction cooling method was adopted for reactor magnet cooling. Performances of the toroid-type HTS DC reactor were analyzed through experiments conducted under the steady-state and charge conditions. The fundamental design specifications and the data obtained from this research will be applied to the design of a commercial-type HTS DC reactor.

Fast start-up reactor for partial oxidation of methane with electrically-heated metallic monolith catalyst

NASA Astrophysics Data System (ADS)

Jung, Heon; Yoon, Wang Lai; Lee, Hotae; Park, Jong Soo; Shin, Jang Sik; La, Howon; Lee, Jong Dae

A palladium-washcoated metallic monolith catalyst is applied to the partial oxidation of methane to syngas. This catalyst is highly active at a gas hourly space velocity (GHSV) of 100,000 h -1. The compact partial oxidation (POX) reactor equipped with both 96 cc of the metallic monolith catalyst and an electrically-heated catalyst (EHC) has a start-up time of less than 1.5 min and a syngas generation capacity of 9.5 Nm 3 h -1. The POX reaction is sustained without the need for an external heater. With the stand-alone POX reactor, the methane conversion can be increased either by preheating the reactant mixture heat-exchanged with the product gas, or by supplying a larger amount of oxygen than is necessary for the reaction stoichiometry.

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.

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

PubMed

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

2007-10-01

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

Detection of on-surface objects with an underground radiography detector system using cosmic-ray muons

NASA Astrophysics Data System (ADS)

Fujii, Hirofumi; Hara, Kazuhiko; Hayashi, Kohei; Kakuno, Hidekazu; Kodama, Hideyo; Nagamine, Kanetada; Sato, Kazuyuki; Sato, Kotaro; Kim, Shin-Hong; Suzuki, Atsuto; Takahashi, Kazuki; Takasaki, Fumihiko

2017-05-01

We have developed a compact muon radiography detector to investigate the status of the nuclear debris in the Fukushima Daiichi Reactors. Our previous observation showed that a large portion of the Unit-1 Reactor fuel had fallen to floor level. The detector must be located underground to further investigate the status of the fallen debris. To investigate the performance of muon radiography in such a situation, we observed 2 m cubic iron blocks located on the surface of the ground through different lengths of ground soil. The iron blocks were imaged and their corresponding iron density was derived successfully.

Post Irradiation Evaluation of Thermal Control Coatings and Solid Lubricants to Support Fission Surface Power Systems

NASA Astrophysics Data System (ADS)

Bowman, Cheryl L.; Jaworske, Donald A.; Stanford, Malcolm K.; Persinger, Justin A.; Khorsandi, Behrooz; Blue, Thomas E.

2007-01-01

The development of a nuclear power system for space missions, such as the Jupiter Icy Moons Orbiter or a lunar outpost, requires substantially more compact reactor design than conventional terrestrial systems. In order to minimize shielding requirements and hence system weight, the radiation tolerance of component materials within the power conversion and heat rejection systems must be defined. Two classes of coatings, thermal control paints and solid lubricants, were identified as material systems for which limited radiation hardness information was available. Screening studies were designed to explore candidate coatings under a predominately fast neutron spectrum. The Ohio State Research Reactor Facility staff performed irradiation in a well characterized, mixed energy spectrum and performed post irradiation analysis of representative coatings for thermal control and solid lubricant applications. Thermal control paints were evaluated for 1 MeV equivalent fluences from 1013 to 1015 n/cm2. No optical degradation was noted although some adhesive degradation was found at higher fluence levels. Solid lubricant coatings were evaluated for 1 MeV equivalent fluences from 1015 to 1016 n/cm2 with coating adhesion and flexibility used for post irradiation evaluation screening. The exposures studied did not lead to obvious property degradation indicating the coatings would have survived the radiation environment for the previously proposed Jupiter mission. The results are also applicable to space power development programs such as fission surface power for future lunar and Mars missions.

Post Irradiation Evaluation of Thermal Control Coatings and Solid Lubricants to Support Fission Surface Power Systems

NASA Technical Reports Server (NTRS)

Bowman, Cheryl L.; Jaworske, Donald A.; Stanford, Malcolm K.; Persinger, Justin A.; Khorsandi, Behrooz; Blue, Thomas E.

2007-01-01

The development of a nuclear power system for space missions, such as the Jupiter Icy Moons Orbiter or a lunar outpost, requires substantially more compact reactor design than conventional terrestrial systems. In order to minimize shielding requirements and hence system weight, the radiation tolerance of component materials within the power conversion and heat rejection systems must be defined. Two classes of coatings, thermal control paints and solid lubricants, were identified as material systems for which limited radiation hardness information was available. Screening studies were designed to explore candidate coatings under a predominately fast neutron spectrum. The Ohio State Research Reactor Facility staff performed irradiation in a well characterized, mixed energy spectrum and performed post irradiation analysis of representative coatings for thermal control and solid lubricant applications. Thermal control paints were evaluated for 1 MeV equivalent fluences from 10(exp 13) to 10(exp 15) n per square centimeters. No optical degradation was noted although some adhesive degradation was found at higher fluence levels. Solid lubricant coatings were evaluated for 1 MeV equivalent fluences from 10(exp 15) to 10(exp 16) n per square centimeters with coating adhesion and flexibility used for post irradiation evaluation screening. The exposures studied did not lead to obvious property degradation indicating the coatings would have survived the radiation environment for the previously proposed Jupiter mission. The results are also applicable to space power development programs such as fission surface power for future lunar and Mars missions.

Variability aware compact model characterization for statistical circuit design optimization

NASA Astrophysics Data System (ADS)

Qiao, Ying; Qian, Kun; Spanos, Costas J.

2012-03-01

Variability modeling at the compact transistor model level can enable statistically optimized designs in view of limitations imposed by the fabrication technology. In this work we propose an efficient variabilityaware compact model characterization methodology based on the linear propagation of variance. Hierarchical spatial variability patterns of selected compact model parameters are directly calculated from transistor array test structures. This methodology has been implemented and tested using transistor I-V measurements and the EKV-EPFL compact model. Calculation results compare well to full-wafer direct model parameter extractions. Further studies are done on the proper selection of both compact model parameters and electrical measurement metrics used in the method.

Nuclear code case development of printed-circuit heat exchangers with thermal and mechanical performance testing

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

Aakre, Shaun R.; Jentz, Ian W.; Anderson, Mark H.

The U.S. Department of Energy has agreed to fund a three-year integrated research project to close technical gaps involved with compact heat exchangers to be used in nuclear applications. This paper introduces the goals of the project, the research institutions, and industrial partners working in collaboration to develop a draft Boiler and Pressure Vessel Code Case for this technology. Heat exchanger testing, as well as non-destructive and destructive evaluation, will be performed by researchers across the country to understand the performance of compact heat exchangers. Testing will be performed using coolants and conditions proposed for Gen IV Reactor designs. Preliminarymore » observations of the mechanical failure mechanisms of the heat exchangers using destructive and non-destructive methods is presented. Unit-cell finite element models assembled to help predict the mechanical behavior of these high-temperature components are discussed as well. Performance testing methodology is laid out in this paper along with preliminary modeling results, an introduction to x-ray and neutron inspection techniques, and results from a recent pressurization test of a printed-circuit heat exchanger. The operational and quality assurance knowledge gained from these models and validation tests will be useful to developers of supercritical CO 2 systems, which commonly employ printed-circuit heat exchangers.« less

Coupling of anaerobic waste treatment to produce protein- and lipid-rich bacterial biomass.

PubMed

Steinberg, Lisa M; Kronyak, Rachel E; House, Christopher H

2017-11-01

Future long-term manned space missions will require effective recycling of water and nutrients as part of a life support system. Biological waste treatment is less energy intensive than physicochemical treatment methods, yet anaerobic methanogenic waste treatment has been largely avoided due to slow treatment rates and safety issues concerning methane production. However, methane is generated during atmosphere regeneration on the ISS. Here we propose waste treatment via anaerobic digestion followed by methanotrophic growth of Methylococcus capsulatus to produce a protein- and lipid-rich biomass that can be directly consumed, or used to produce other high-protein food sources such as fish. To achieve more rapid methanogenic waste treatment, we built and tested a fixed-film, flow-through, anaerobic reactor to treat an ersatz wastewater. During steady-state operation, the reactor achieved a 97% chemical oxygen demand (COD) removal rate with an organic loading rate of 1740 g d -1  m -3 and a hydraulic retention time of 12.25 d. The reactor was also tested on three occasions by feeding ca. 500 g COD in less than 12 h, representing 50x the daily feeding rate, with COD removal rates ranging from 56-70%, demonstrating the ability of the reactor to respond to overfeeding events. While investigating the storage of treated reactor effluent at a pH of 12, we isolated a strain of Halomonas desiderata capable of acetate degradation under high pH conditions. We then tested the nutritional content of the alkaliphilic Halomonas desiderata strain, as well as the thermophile Thermus aquaticus, as supplemental protein and lipid sources that grow in conditions that should preclude pathogens. The M. capsulatus biomass consisted of 52% protein and 36% lipids, the H. desiderata biomass consisted of 15% protein and 7% lipids, and the Thermus aquaticus biomass consisted of 61% protein and 16% lipids. This work demonstrates the feasibility of rapid waste treatment in a compact reactor design, and proposes recycling of nutrients back into foodstuffs via heterotrophic (including methanotrophic, acetotrophic, and thermophilic) microbial growth. Copyright © 2017. Published by Elsevier Ltd.

Coupling of anaerobic waste treatment to produce protein- and lipid-rich bacterial biomass

NASA Astrophysics Data System (ADS)

Steinberg, Lisa M.; Kronyak, Rachel E.; House, Christopher H.

2017-11-01

Future long-term manned space missions will require effective recycling of water and nutrients as part of a life support system. Biological waste treatment is less energy intensive than physicochemical treatment methods, yet anaerobic methanogenic waste treatment has been largely avoided due to slow treatment rates and safety issues concerning methane production. However, methane is generated during atmosphere regeneration on the ISS. Here we propose waste treatment via anaerobic digestion followed by methanotrophic growth of Methylococcus capsulatus to produce a protein- and lipid-rich biomass that can be directly consumed, or used to produce other high-protein food sources such as fish. To achieve more rapid methanogenic waste treatment, we built and tested a fixed-film, flow-through, anaerobic reactor to treat an ersatz wastewater. During steady-state operation, the reactor achieved a 97% chemical oxygen demand (COD) removal rate with an organic loading rate of 1740 g d-1 m-3 and a hydraulic retention time of 12.25 d. The reactor was also tested on three occasions by feeding ca. 500 g COD in less than 12 h, representing 50x the daily feeding rate, with COD removal rates ranging from 56-70%, demonstrating the ability of the reactor to respond to overfeeding events. While investigating the storage of treated reactor effluent at a pH of 12, we isolated a strain of Halomonas desiderata capable of acetate degradation under high pH conditions. We then tested the nutritional content of the alkaliphilic Halomonas desiderata strain, as well as the thermophile Thermus aquaticus, as supplemental protein and lipid sources that grow in conditions that should preclude pathogens. The M. capsulatus biomass consisted of 52% protein and 36% lipids, the H. desiderata biomass consisted of 15% protein and 7% lipids, and the Thermus aquaticus biomass consisted of 61% protein and 16% lipids. This work demonstrates the feasibility of rapid waste treatment in a compact reactor design, and proposes recycling of nutrients back into foodstuffs via heterotrophic (including methanotrophic, acetotrophic, and thermophilic) microbial growth.

Design of a compact microfludic device for controllable cell distribution.

PubMed

Li, Jing-Liang; Day, Daniel; Gu, Min

2010-11-21

A compact microfluidic device with 96 microchambers allocated within four circular units was designed and examined for cell distribution. In each unit, cells were distributed to the surrounding chambers radially from the center. The circular arrangement of the chambers makes the design simple and compact. A controllable and quantitative cell distribution is achievable in this device. This design is significant to the microfluidic applications where controllable distribution of cells in multipule microchambers is demanded.

PREPARATION OF HIGH-DENSITY THORIUM OXIDE SPHERES

DOEpatents

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

1963-12-31

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

Eddy Current Flow Measurements in the FFTF

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

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

2017-02-02

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

Space station prototype Sabatier reactor design verification testing

NASA Technical Reports Server (NTRS)

Cusick, R. J.

1974-01-01

A six-man, flight prototype carbon dioxide reduction subsystem for the SSP ETC/LSS (Space Station Prototype Environmental/Thermal Control and Life Support System) was developed and fabricated for the NASA-Johnson Space Center between February 1971 and October 1973. Component design verification testing was conducted on the Sabatier reactor covering design and off-design conditions as part of this development program. The reactor was designed to convert a minimum of 98 per cent hydrogen to water and methane for both six-man and two-man reactant flow conditions. Important design features of the reactor and test conditions are described. Reactor test results are presented that show design goals were achieved and off-design performance was stable.

2D stepping drive for hyperspectral systems

NASA Astrophysics Data System (ADS)

Endrödy, Csaba; Mehner, Hannes; Grewe, Adrian; Sinzinger, Stefan; Hoffmann, Martin

2015-07-01

We present the design, fabrication and characterization of a compact 2D stepping microdrive for pinhole array positioning. The miniaturized solution enables a highly integrated compact hyperspectral imaging system. Based on the geometry of the pinhole array, an inch-worm drive with electrostatic actuators was designed resulting in a compact (1 cm2) positioning system featuring a step size of about 15 µm in a 170 µm displacement range. The high payload (20 mg) as required for the pinhole array and the compact system design exceed the known electrostatic inch-worm-based microdrives.

Ethanol dehydration to ethylene in a stratified autothermal millisecond reactor.

PubMed

Skinner, Michael J; Michor, Edward L; Fan, Wei; Tsapatsis, Michael; Bhan, Aditya; Schmidt, Lanny D

2011-08-22

The concurrent decomposition and deoxygenation of ethanol was accomplished in a stratified reactor with 50-80 ms contact times. The stratified reactor comprised an upstream oxidation zone that contained Pt-coated Al(2)O(3) beads and a downstream dehydration zone consisting of H-ZSM-5 zeolite films deposited on Al(2)O(3) monoliths. Ethanol conversion, product selectivity, and reactor temperature profiles were measured for a range of fuel:oxygen ratios for two autothermal reactor configurations using two different sacrificial fuel mixtures: a parallel hydrogen-ethanol feed system and a series methane-ethanol feed system. Increasing the amount of oxygen relative to the fuel resulted in a monotonic increase in ethanol conversion in both reaction zones. The majority of the converted carbon was in the form of ethylene, where the ethanol carbon-carbon bonds stayed intact while the oxygen was removed. Over 90% yield of ethylene was achieved by using methane as a sacrificial fuel. These results demonstrate that noble metals can be successfully paired with zeolites to create a stratified autothermal reactor capable of removing oxygen from biomass model compounds in a compact, continuous flow system that can be configured to have multiple feed inputs, depending on process restrictions. Copyright © 2011 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

Status of Fuel Development and Manufacturing for Space Nuclear Reactors at BWX Technologies

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

Carmack, W.J.; Husser, D.L.; Mohr, T.C.

2004-02-04

New advanced nuclear space propulsion systems will soon seek a high temperature, stable fuel form. BWX Technologies Inc (BWXT) has a long history of fuel manufacturing. UO2, UCO, and UCx have been fabricated at BWXT for various US and international programs. Recent efforts at BWXT have focused on establishing the manufacturing techniques and analysis capabilities needed to provide a high quality, high power, compact nuclear reactor for use in space nuclear powered missions. To support the production of a space nuclear reactor, uranium nitride has recently been manufactured by BWXT. In addition, analytical chemistry and analysis techniques have been developedmore » to provide verification and qualification of the uranium nitride production process. The fabrication of a space nuclear reactor will require the ability to place an unclad fuel form into a clad structure for assembly into a reactor core configuration. To this end, BWX Technologies has reestablished its capability for machining, GTA welding, and EB welding of refractory metals. Specifically, BWX Technologies has demonstrated GTA welding of niobium flat plate and EB welding of niobium and Nb-1Zr tubing. In performing these demonstration activities, BWX Technologies has established the necessary infrastructure to manufacture UO2, UCx, or UNx fuel, components, and complete reactor assemblies in support of space nuclear programs.« less

Laminated grid and web magnetic cores

DOEpatents

Sefko, John; Pavlik, Norman M.

1984-01-01

A laminated magnetic core characterized by an electromagnetic core having core legs which comprise elongated apertures and edge notches disposed transversely to the longitudinal axis of the legs, such as high reluctance cores with linear magnetization characteristics for high voltage shunt reactors. In one embodiment the apertures include compact bodies of microlaminations for more flexibility and control in adjusting permeability and/or core reluctance.

Continuous-Flow Synthesis of N-Succinimidyl 4-[18F]fluorobenzoate Using a Single Microfluidic Chip

PubMed Central

Kimura, Hiroyuki; Tomatsu, Kenji; Saiki, Hidekazu; Arimitsu, Kenji; Ono, Masahiro; Kawashima, Hidekazu; Iwata, Ren; Nakanishi, Hiroaki; Ozeki, Eiichi; Kuge, Yuji; Saji, Hideo

2016-01-01

In the field of positron emission tomography (PET) radiochemistry, compact microreactors provide reliable and reproducible synthesis methods that reduce the use of expensive precursors for radiolabeling and make effective use of the limited space in a hot cell. To develop more compact microreactors for radiosynthesis of 18F-labeled compounds required for the multistep procedure, we attempted radiosynthesis of N-succinimidyl 4-[18F]fluorobenzoate ([18F]SFB) via a three-step procedure using a microreactor. We examined individual steps for [18F]SFB using a batch reactor and microreactor and developed a new continuous-flow synthetic method with a single microfluidic chip to achieve rapid and efficient radiosynthesis of [18F]SFB. In the synthesis of [18F]SFB using this continuous-flow method, the three-step reaction was successfully completed within 6.5 min and the radiochemical yield was 64 ± 2% (n = 5). In addition, it was shown that the quality of [18F]SFB synthesized on this method was equal to that synthesized by conventional methods using a batch reactor in the radiolabeling of bovine serum albumin with [18F]SFB. PMID:27410684

Investigation of the characteristics of a compact steam reformer integrated with a water-gas shift reactor

NASA Astrophysics Data System (ADS)

Seo, Yong-Seog; Seo, Dong-Joo; Seo, Yu-Taek; Yoon, Wang-Lai

The objective of this study is to investigate numerically a compact steam methane reforming (SMR) system integrated with a water-gas shift (WGS) reactor. Separate numerical models are established for the combustion part, SMR and WGS reaction bed. The concentration of species at the exits of the SMR and WGS bed, and the temperatures in the WGS bed are in good agreement with the measured data. Heat transfer to the catalyst beds and the catalytic reactions in the SMR and WGS catalyst bed are investigated as a function of the operation parameters. The conversion of methane at the exit of the SMR catalyst bed is calculated to be 87%, and the carbon monoxide concentration at the outlet of the WGS bed is estimated to be 0.45%. The effects of the cooling heat flux at the outside wall of the system and steam-to-carbon (S/C) ratio are also examined. As the cooling heat flux increases, both the methane conversion and carbon monoxide content are reduced in the SMR bed, and the carbon monoxide conversion is improved in the WGS bed. Both methane conversion and carbon dioxide reduction increase with increasing steam-to-carbon ratio.

Animal Guts as Ideal Reactors: An Open-Ended Project for a Course in Kinetics and Reactor Design.

ERIC Educational Resources Information Center

Carlson, Eric D.; Gast, Alice P.

1998-01-01

Presents an open-ended project tailored for a senior kinetics and reactor design course in which basic reactor design equations are used to model the digestive systems of several animals. Describes the assignment as well as the results. (DDR)

Characteristics of aerobic granules grown on glucose and acetate in sequential aerobic sludge blanket reactors.

PubMed

Tay, J H; Liu, Q S; Liu, Y

2002-08-01

Aerobic granules were cultivated in two column-type sequential aerobic sludge blanket reactors fed with glucose and acetate, respectively. The characteristics of aerobic granules were investigated. Results indicated that the glucose- and acetate-fed granules have comparable characteristics in terms of settling velocity, size, shape, biomass density, hydrophobicity, physical strength, microbial activity and storage stability. Substrate component does not seem to be a key factor on the formation of aerobic granules. However, microbial diversity of the granules is closely associated with the carbon sources supplied to the reactors. Compared with the conventional activated sludge flocs, aerobic granules exhibit excellent physical characteristics that would be essential for industrial application. This research provides a complete set of characteristics data of aerobic granules grown on glucose and acetate, which would be useful for further development of aerobic granules-based compact bioreactor for handling high strength organic wastewater.

Potential civil mission applications for space nuclear power systems

NASA Technical Reports Server (NTRS)

Ambrus, J. H.; Beatty, R. G. G.

1985-01-01

It is pointed out that the energy needs of spacecraft over the last 25 years have been met by photovoltaic arrays with batteries, primary fuel cells, and radioisotope thermoelectric generators (RTG). However, it might be difficult to satisfy energy requirements for the next generation of space missions with the currently used energy sources. Applications studies have emphasized the need for a lighter, cheaper, and more compact high-energy source than the scaling up of current technologies would permit. These requirements could be satisfied by a nuclear reactor power system. The joint NASA/DOD/DOE SP-100 program is to explore and evaluate this option. Critical elements of the technology are also to be developed, taking into account space reactor systems of the 100 kW class. The present paper is concerned with some of the civil mission categories and concepts which are enabled or significantly enhanced by the performance characteristics of a nuclear reactor energy system.

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

NASA Technical Reports Server (NTRS)

Stanojev, B. J.; Houts, M.

2004-01-01

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

Applications of power beaming from space-based nuclear power stations. [Laser beaming to airplanes; microwave beaming to ground

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

Powell, J.R.; Botts, T.E.; Hertzberg, A.

1981-01-01

Power beaming from space-based reactor systems is examined using an advanced compact, lightweight Rotating Bed Reactor (RBR). Closed Brayton power conversion efficiencies in the range of 30 to 40% can be achieved with turbines, with reactor exit temperatures on the order of 2000/sup 0/K and a liquid drop radiator to reject heat at temperatures of approx. 500/sup 0/K. Higher RBR coolant temperatures (up to approx. 3000/sup 0/K) are possible, but gains in power conversion efficiency are minimal, due to lower expander efficiency (e.g., a MHD generator). Two power beaming applications are examined - laser beaming to airplanes and microwave beamingmore » to fixed ground receivers. Use of the RBR greatly reduces system weight and cost, as compared to solar power sources. Payback times are a few years at present prices for power and airplane fuel.« less

Experimental Criticality Benchmarks for SNAP 10A/2 Reactor Cores

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

Krass, A.W.

2005-12-19

This report describes computational benchmark models for nuclear criticality derived from descriptions of the Systems for Nuclear Auxiliary Power (SNAP) Critical Assembly (SCA)-4B experimental criticality program conducted by Atomics International during the early 1960's. The selected experimental configurations consist of fueled SNAP 10A/2-type reactor cores subject to varied conditions of water immersion and reflection under experimental control to measure neutron multiplication. SNAP 10A/2-type reactor cores are compact volumes fueled and moderated with the hydride of highly enriched uranium-zirconium alloy. Specifications for the materials and geometry needed to describe a given experimental configuration for a model using MCNP5 are provided. Themore » material and geometry specifications are adequate to permit user development of input for alternative nuclear safety codes, such as KENO. A total of 73 distinct experimental configurations are described.« less

High Temperature Gas-Cooled Test Reactor Point Design: Summary Report

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

Sterbentz, James William; Bayless, Paul David; Nelson, Lee Orville

2016-01-01

A point design has been developed for a 200-MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched uranium oxycarbide fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technology readiness level, licensing approach, and costs of the test reactor point design.

High Temperature Gas-Cooled Test Reactor Point Design: Summary Report

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

Sterbentz, James William; Bayless, Paul David; Nelson, Lee Orville

2016-03-01

A point design has been developed for a 200-MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched uranium oxycarbide fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technology readiness level, licensing approach, and costs of the test reactor point design.

AGR-3/4 Irradiation Test Predictions using PARFUME

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

Skerjanc, William Frances; Collin, Blaise Paul

2016-03-01

PARFUME, a fuel performance modeling code used for high temperature gas reactors, was used to model the AGR-3/4 irradiation test using as-run physics and thermal hydraulics data. The AGR-3/4 test is the combined third and fourth planned irradiations of the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. The AGR-3/4 test train consists of twelve separate and independently controlled and monitored capsules. Each capsule contains four compacts filled with both uranium oxycarbide (UCO) unaltered “driver” fuel particles and UCO designed-to-fail (DTF) fuel particles. The DTF fraction was specified to be 1×10-2. This report documents the calculations performed to predictmore » failure probability of TRISO-coated fuel particles during the AGR-3/4 experiment. In addition, this report documents the calculated source term from both the driver fuel and DTF particles. The calculations include the modeling of the AGR-3/4 irradiation that occurred from December 2011 to April 2014 in the Advanced Test Reactor (ATR) over a total of ten ATR cycles including seven normal cycles, one low power cycle, one unplanned outage cycle, and one Power Axial Locator Mechanism cycle. Results show that failure probabilities are predicted to be low, resulting in zero fuel particle failures per capsule. The primary fuel particle failure mechanism occurred as a result of localized stresses induced by the calculated IPyC cracking. Assuming 1,872 driver fuel particles per compact, failure probability calculated by PARFUME leads to no predicted particle failure in the AGR-3/4 driver fuel. In addition, the release fraction of fission products Ag, Cs, and Sr were calculated to vary depending on capsule location and irradiation temperature. The maximum release fraction of Ag occurs in Capsule 7 reaching up to 56% for the driver fuel and 100% for the DTF fuel. The release fraction of the other two fission products, Cs and Sr, are much smaller and in most cases less than 1% for the driver fuel. The notable exception occurs in Capsule 7 where the release fraction for Cs and Sr reach up to 0.73% and 2.4%, respectively, for the driver fuel. For the DTF fuel in Capsule 7, the release fraction for Cs and Sr are estimated to be 100% and 5%, respectively.« less

Design of energy-storage reactors for single-winding constant-frequency dc-to-dc converters operating in the discontinuous-reactor-current mode

NASA Technical Reports Server (NTRS)

Chen, D. Y.; Owen, H. A., Jr.; Wilson, T. G.

1980-01-01

This paper presents an algorithm and equations for designing the energy-storage reactor for dc-to-dc converters which are constrained to operate in the discontinuous-reactor-current mode. This design procedure applied to the three widely used single-winding configurations: the voltage step-up, the current step-up, and the voltage-or-current step-up converters. A numerical design example is given to illustrate the use of the design algorithm and design equations.

Trench fast reactor design using the microcomputer

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

Rohach, A.F.; Sankoorikal, J.T.; Schmidt, R.R.

1987-01-01

This project is a study of alternative liquid-metal-cooled fast power reactor system concepts. Specifically, an unconventional primary system is being conceptually designed and evaluated. The project design is based primarily on microcomputer analysis through the use of computational modules. The reactor system concept is a long, narrow pool with a long, narrow reactor called a trench-type pool reactor in it. The reactor consists of five core-blanket modules in a line. Specific power is to be modest, permitting long fuel residence time. Two fuel cycles are currently being considered. The reactor design philosophy is that of the inherently safe concept. Thismore » requires transient analysis dependent on reactivity coefficients: prompt fuel, including Doppler and expansion, fuel expansion, sodium temperature and void, and core expansion. Conceptual reactor design is done on a microcomputer. A part of the trench reactor project is to develop a microcomputer-based system that can be used by the user for scoping studies and design. Current development includes the neutronics and fuel management aspects of the design. Thermal-hydraulic analysis and economics are currently being incorporated into the microcomputer system. The system is menu-driven including preparation of program input data and of output data for displays in graphics form.« less

Performance of Transuranic-Loaded Fully Ceramic Micro-Encapsulated Fuel in LWRs Final Report, Including Void Reactivity Evaluation

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

Michael A. Pope; R. Sonat Sen; Brian Boer

2011-09-01

The current focus of the Deep Burn Project is on once-through burning of transuranics (TRU) in light-water reactors (LWRs). The fuel form is called Fully-Ceramic Micro-encapsulated (FCM) fuel, a concept that borrows the tri-isotropic (TRISO) fuel particle design from high-temperature reactor technology. In the Deep Burn LWR (DB-LWR) concept, these fuel particles are pressed into compacts using SiC matrix material and loaded into fuel pins for use in conventional LWRs. The TRU loading comes from the spent fuel of a conventional LWR after 5 years of cooling. Unit cell and assembly calculations have been performed using the DRAGON-4 code tomore » assess the physics attributes of TRU-only FCM fuel in an LWR lattice. Depletion calculations assuming an infinite lattice condition were performed with calculations of various reactivity coefficients performed at each step. Unit cells and assemblies containing typical UO2 and mixed oxide (MOX) fuel were analyzed in the same way to provide a baseline against which to compare the TRU-only FCM fuel. Then, assembly calculations were performed evaluating the performance of heterogeneous arrangements of TRU-only FCM fuel pins along with UO2 pins.« less

Design of compact and ultra efficient aspherical lenses for extended Lambertian sources in two-dimensional geometry

PubMed Central

Wu, Rengmao; Hua, Hong; Benítez, Pablo; Miñano, Juan C.; Liang, Rongguang

2016-01-01

The energy efficiency and compactness of an illumination system are two main concerns in illumination design for extended sources. In this paper, we present two methods to design compact, ultra efficient aspherical lenses for extended Lambertian sources in two-dimensional geometry. The light rays are directed by using two aspherical surfaces in the first method and one aspherical surface along with an optimized parabola in the second method. The principles and procedures of each design method are introduced in detail. Three examples are presented to demonstrate the effectiveness of these two methods in terms of performance and capacity in designing compact, ultra efficient aspherical lenses. The comparisons made between the two proposed methods indicate that the second method is much simpler and easier to be implemented, and has an excellent extensibility to three-dimensional designs. PMID:29092336

Submersion criticality safety of tungsten-rhenium urania cermet fuel for space propulsion and power applications

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

A.E. Craft; R. C. O'Brien; S. D. Howe

Nuclear thermal rockets are the preferred propulsion technology for a manned mission to Mars, and tungsten–uranium oxide cermet fuels could provide significant performance and cost advantages for nuclear thermal rockets. A nuclear reactor intended for use in space must remain subcritical before and during launch, and must remain subcritical in launch abort scenarios where the reactor falls back to Earth and becomes submerged in terrestrial materials (including seawater, wet sand, or dry sand). Submersion increases reflection of neutrons and also thermalizes the neutron spectrum, which typically increases the reactivity of the core. This effect is typically very significant for compact,more » fast-spectrum reactors. This paper provides a submersion criticality safety analysis for a representative tungsten/uranium oxide fueled reactor with a range of fuel compositions. Each submersion case considers both the rhenium content in the matrix alloy and the uranium oxide volume fraction in the cermet. The inclusion of rhenium significantly improves the submersion criticality safety of the reactor. While increased uranium oxide content increases the reactivity of the core, it does not significantly affect the submersion behavior of the reactor. There is no significant difference in submersion behavior between reactors with rhenium distributed within the cermet matrix and reactors with a rhenium clad in the coolant channels. The combination of the flooding of the coolant channels in submersion scenarios and the presence of a significant amount of spectral shift absorbers (i.e. high rhenium concentration) further decreases reactivity for short reactor cores compared to longer cores.« less

Improving the oxidation resistance of 316L stainless steel in simulated pressurized water reactor primary water by electropolishing treatment

NASA Astrophysics Data System (ADS)

Han, Guangdong; Lu, Zhanpeng; Ru, Xiangkun; Chen, Junjie; Xiao, Qian; Tian, Yongwu

2015-12-01

The oxidation behavior of 316L stainless steel specimens after emery paper grounding, mechanical polishing, and electropolishing were investigated in simulated pressurized water reactor primary water at 310 °C for 120 and 500 h. Electropolishing afforded improved oxidation resistance especially during the early immersion stages. Duplex oxide films comprising a coarse Fe-rich outer layer and a fine Cr-rich inner layer formed on all specimens after 500 h of immersion. Only a compact layer was observed on the electropolished specimen after 120 h of immersion. The enrichment of chromium in the electropolished layer contributed to the passivity and protectiveness of the specimen.

Understanding and optimizing the dual excipient functionality of sodium lauryl sulfate in tablet formulation of poorly water soluble drug: wetting and lubrication.

PubMed

Aljaberi, Ahmad; Chatterji, Ashish; Dong, Zedong; Shah, Navnit H; Malick, Waseem; Singhal, Dharmendra; Sandhu, Harpreet K

2013-01-01

To evaluate and optimize sodium lauryl sulfate (SLS) and magnesium stearate (Mg.St) levels, with respect to dissolution and compaction, in a high dose, poorly soluble drug tablet formulation. A model poorly soluble drug was formulated using high shear aqueous granulation. A D-optimal design was used to evaluate and model the effect of granulation conditions, size of milling screen, SLS and Mg.St levels on tablet compaction and ejection. The compaction profiles were generated using a Presster(©) compaction simulator. Dissolution of the kernels was performed using a USP dissolution apparatus II and intrinsic dissolution was determined using a stationary disk system. Unlike kernels dissolution which failed to discriminate between tablets prepared with various SLS contents, the intrinsic dissolution rate showed that a SLS level of 0.57% was sufficient to achieve the required release profile while having minimal effect on compaction. The formulation factors that affect tablet compaction and ejection were identified and satisfactorily modeled. The design space of best factor setting to achieve optimal compaction and ejection properties was successfully constructed by RSM analysis. A systematic study design helped identify the critical factors and provided means to optimize the functionality of key excipient to design robust drug product.

Modeling the shear rate and pressure drop in a hydrodynamic cavitation reactor with experimental validation based on KI decomposition studies.

PubMed

Badve, Mandar P; Alpar, Tibor; Pandit, Aniruddha B; Gogate, Parag R; Csoka, Levente

2015-01-01

A mathematical model describing the shear rate and pressure variation in a complex flow field created in a hydrodynamic cavitation reactor (stator and rotor assembly) has been depicted in the present study. The design of the reactor is such that the rotor is provided with surface indentations and cavitational events are expected to occur on the surface of the rotor as well as within the indentations. The flow characteristics of the fluid have been investigated on the basis of high accuracy compact difference schemes and Navier-Stokes method. The evolution of streamlining structures during rotation, pressure field and shear rate of a Newtonian fluid flow have been numerically established. The simulation results suggest that the characteristics of shear rate and pressure area are quite different based on the magnitude of the rotation velocity of the rotor. It was observed that area of the high shear zone at the indentation leading edge shrinks with an increase in the rotational speed of the rotor, although the magnitude of the shear rate increases linearly. It is therefore concluded that higher rotational speeds of the rotor, tends to stabilize the flow, which in turn results into less cavitational activity compared to that observed around 2200-2500RPM. Experiments were carried out with initial concentration of KI as 2000ppm. Maximum of 50ppm of iodine liberation was observed at 2200RPM. Experimental as well as simulation results indicate that the maximum cavitational activity can be seen when rotation speed is around 2200-2500RPM. Copyright © 2014 Elsevier B.V. All rights reserved.

The QUASAR facility

NASA Astrophysics Data System (ADS)

Gates, David

2013-10-01

The QUAsi-Axisymmetric Research (QUASAR) stellarator is a new facility which can solve two critical problems for fusion, disruptions and steady-state, and which provides new insights into the role of magnetic symmetry in plasma confinement. If constructed it will be the only quasi-axisymmetric stellarator in the world. The innovative principle of quasi-axisymmetry (QA) will be used in QUASAR to study how ``tokamak-like'' systems can be made: 1) Disruption-free, 2) Steady-state with low recirculating power, while preserving or improving upon features of axisymmetric tokamaks, such as 1) Stable at high pressure simultaneous with 2) High confinement (similar to tokamaks), and 3) Scalable to a compact reactor Stellarator research is critical to fusion research in order to establish the physics basis for a magnetic confinement device that can operate efficiently in steady-state, without disruptions at reactor-relevant parameters. The two large stellarator experiments - LHD in Japan and W7-X under construction in Germany are pioneering facilities capable of developing 3D physics understanding at large scale and for very long pulses. The QUASAR design is unique in being QA and optimized for confinement, stability, and moderate aspect ratio (4.5). It projects to a reactor with a major radius of ~8 m similar to advanced tokamak concepts. It is striking that (a) the EU DEMO is a pulsed (~2.5 hour) tokamak with major R ~ 9 m and (b) the ITER physics scenarios do not presume steady-state behavior. Accordingly, QUASAR fills a critical gap in the world stellarator program. This work supported by DoE Contract No. DEAC02-76CH03073.

Nuclear design of a very-low-activation fusion reactor

NASA Astrophysics Data System (ADS)

Cheng, E. T.; Hopkins, G. R.

1983-06-01

The nuclear design aspects of using very-low-activation materials, such as SiC, MgO, and aluminum for fusion-reactor first wall, blanket, and shield applications were investigated. In addition to the advantage of very-low radioactive inventory, it was found that the very-low-activation fusion reactor can also offer an adequate tritium-breeding ratio and substantial amount of blanket nuclear heating as a conventional-material-structured reactor does. The most-stringent design constraint found in a very-low-activation fusion reactor is the limited space available in the inboard region of a Tokamak concept for shielding to protect the superconducting toroidal field coil. A reference design was developed which mitigates the constraint by adopting a removable tungsten shield design that retains the inboard dimensions and gives the same shield performance as the reference STARFIRE Tokamak reactor design.

10 CFR 72.120 - General considerations.

Code of Federal Regulations, 2010 CFR

2010-01-01

... NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE General Design... reactor-related GTCC waste in an ISFSI or to store spent fuel, high-level radioactive waste, or reactor... be designed to store spent fuel and/or solid reactor-related GTCC waste. (1) Reactor-related GTCC...

Design of a 25-kWe Surface Reactor System Based on SNAP Reactor Technologies

NASA Astrophysics Data System (ADS)

Dixon, David D.; Hiatt, Matthew T.; Poston, David I.; Kapernick, Richard J.

2006-01-01

A Hastelloy-X clad, sodium-potassium (NaK-78) cooled, moderated spectrum reactor using uranium zirconium hydride (UZrH) fuel based on the SNAP program reactors is a promising design for use in surface power systems. This paper presents a 98 kWth reactor for a power system the uses multiple Stirling engines to produce 25 kWe-net for 5 years. The design utilizes a pin type geometry containing UZrHx fuel clad with Hastelloy-X and NaK-78 flowing around the pins as coolant. A compelling feature of this design is its use of 49.9% enriched U, allowing it to be classified as a category III-D attractiveness and reducing facility costs relative to highly-enriched space reactor concepts. Presented below are both the design and an analysis of this reactor's criticality under various safety and operations scenarios.

DOE/NNSA perspective safeguard by design: GEN III/III+ light water reactors and beyond

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

Pan, Paul Y

2010-12-10

An overview of key issues relevant to safeguards by design (SBD) for GEN III/IV nuclear reactors is provided. Lessons learned from construction of typical GEN III+ water reactors with respect to SBD are highlighted. Details of SBD for safeguards guidance development for GEN III/III+ light water reactors are developed and reported. This paper also identifies technical challenges to extend SBD including proliferation resistance methodologies to other GEN III/III+ reactors (except HWRs) and GEN IV reactors because of their immaturity in designs.

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

Pavlou, A. T.; Betzler, B. R.; Burke, T. P.

Uncertainties in the composition and fabrication of fuel compacts for the Fort St. Vrain (FSV) high temperature gas reactor have been studied by performing eigenvalue sensitivity studies that represent the key uncertainties for the FSV neutronic analysis. The uncertainties for the TRISO fuel kernels were addressed by developing a suite of models for an 'average' FSV fuel compact that models the fuel as (1) a mixture of two different TRISO fuel particles representing fissile and fertile kernels, (2) a mixture of four different TRISO fuel particles representing small and large fissile kernels and small and large fertile kernels and (3)more » a stochastic mixture of the four types of fuel particles where every kernel has its diameter sampled from a continuous probability density function. All of the discrete diameter and continuous diameter fuel models were constrained to have the same fuel loadings and packing fractions. For the non-stochastic discrete diameter cases, the MCNP compact model arranged the TRISO fuel particles on a hexagonal honeycomb lattice. This lattice-based fuel compact was compared to a stochastic compact where the locations (and kernel diameters for the continuous diameter cases) of the fuel particles were randomly sampled. Partial core configurations were modeled by stacking compacts into fuel columns containing graphite. The differences in eigenvalues between the lattice-based and stochastic models were small but the runtime of the lattice-based fuel model was roughly 20 times shorter than with the stochastic-based fuel model. (authors)« less

REVIEW OF POWER AND HEAT REACTOR DESIGNS. Domestic and Foreign

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

Appleby, E.R., comp

1963-10-01

Unclassified information from domestic and foreign literature from January 1952 through September 1963 is compiled. Design characteristics and current information on the status of the individual designs are given, along with references for the associated literature. SNAP systems, proposed reactors, and chemonuclear and test reactors with characteristics similar to power reactors are included. The designs are indexed by name, location, type, and some special characteristics. (D.C.W.)

Influence of implant collar design on stress and strain distribution in the crestal compact bone: a three-dimensional finite element analysis.

PubMed

Shen, Wan-Ling; Chen, Chen-Sheng; Hsu, Ming-Lun

2010-01-01

To evaluate the influence of implant collar geometry on the distribution of stress and strain in the crestal compact bone contiguous to an implant collar for four types of bone under axial and oblique loads. Finite element models of threaded implants with three kinds of implant collar designs (divergent, straight, and convergent) with their corresponding suprastructures embedded in the posterior mandible were created with ANSYS software. Eight different test conditions incorporating four types of bone (orthotropic and effectively isotropic in part 1 and high and low densities in part 2) under separate 100-N axial and 35.6-degree oblique forces were created to investigate the stress and strain distributions in the crestal compact bone around the implant collars. In all eight conditions, the divergent collar demonstrated the lowest maximum von Mises and principal stresses and strains in the crestal compact bone contiguous to the implant collar, followed by the straight and convergent collars. The oblique load induced higher peak values than the axial load. The orthotropic design amplified and increased the pathologic microstrains and tensile stresses in the crestal compact bone compared to the effectively isotropic design, especially in models with a convergent collar design. In part 2 of the study, the maximum von Mises stresses and strains increased with a decrease in the cancellous bone density. Under oblique loading, the convergent and straight collars showed pathologic microstrain values as well as excessive ultimate tensile stresses in the orthotropic bone model with low-density cancellous bone. Within the limitations, it was concluded that stress and strain distributions in the adjacent compact bone are influenced by the implant collar design. The divergent implant collar design was associated with the lowest stress and strain concentrations in the crestal compact bone.

Performance improvement of magnetized coaxial plasma gun by magnetic circuit on a bias coil

NASA Astrophysics Data System (ADS)

Edo, Takahiro; Matsumoto, Tadafumi; Asai, Tomohiko; Kamino, Yasuhiro; Inomoto, Michiaki; Gota, Hiroshi

2016-10-01

A magnetized coaxial plasmoid accelerator has been utilized for compact torus (CT) injection to refuel into fusion reactor core plasma. Recently, CT injection experiments have been conducted on the C-2/C-2U facility at Tri Alpha Energy. In the series of experiments successful refueling, i.e. increased particle inventory of field-reversed configuration (FRC) plasma, has been observed. In order to improve the performance of CT injector and to refuel in the upgraded FRC device, called C-2W, with higher confinement magnetic field, magnetic circuit consisting of magnetic material onto a bias magnetic coil is currently being tested at Nihon University. Numerical work suggests that the optimized bias magnetic field distribution realizes the increased injection velocity because of higher conversion efficiency of Lorenz self force to kinetic energy. Details of the magnetic circuit design as well as results of the test experiment and field calculations will be presented and discussed.

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

Redi, M.H.; Mynick, H.E.; Suewattana, M.

Hamiltonian coordinate, guiding-center code calculations of the confinement of suprathermal ions in quasi-axisymmetric stellarator (QAS) designs have been carried out to evaluate the attractiveness of compact configurations which are optimized for ballooning stability. A new stellarator particle-following code is used to predict ion loss rates and particle confinement for thermal and neutral beam ions in a small experiment with R = 145 cm, B = 1-2 T and for alpha particles in a reactor-size device. In contrast to tokamaks, it is found that high edge poloidal flux has limited value in improving ion confinement in QAS, since collisional pitch-angle scatteringmore » drives ions into ripple wells and stochastic field regions, where they are quickly lost. The necessity for reduced stellarator ripple fields is emphasized. The high neutral beam ion loss predicted for these configurations suggests that more interesting physics could be explored with an experiment of less constrained size and magnetic field geometry.« less

The Goals and Status of SoLid Experiment

NASA Astrophysics Data System (ADS)

Park, Jaewon

2016-09-01

SoLid is a short baseline sterile neutrino oscillation search experiment using the BR2 compact core reactor in Belgium. Ruling out or confirming sterile neutrino is one of main interests in the neutrino physics field. Highly segmented scintillator cube detector with 6LiF:ZnS(Ag) neutron screen provides high purity neutron tagging by pulse shape discrimination (PSD), and capture position identification. These capabilities from this novel detector are critical to isolate neutrino interactions in a high background environment. The prototype detector (SM1) provides important feedback for validating the performance of the detector design. Recent results from SM1 will be presented. Construction of the SoLid Phase-1 detector is underway. The three-ton detector with three years running will allow us to reach the sterile neutrino exclusion limit of sin2 2 θ < 0 . 03 at Δm2 2eV2 at the 99% confidence level.

Compact, Lightweight Electromagnetic Pump for Liquid Metal

NASA Technical Reports Server (NTRS)

Godfroy, Thomas; Palzin, Kurt

2010-01-01

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

A small, 1400 K, reactor for Brayton space power systems.

NASA Technical Reports Server (NTRS)

Lantz, E.; Mayo, W.

1972-01-01

An investigation was conducted to determine minimum dimensions and minimum weight obtainable in a design for a reactor using uranium-233 nitride or plutonium-239 nitride as fuel. Such a reactor had been considered by Krasner et al. (1971). Present space power status is discussed, together with questions of reactor design and power distribution in the reactor. The characteristics of various reactor types are compared, giving attention also to a zirconium hydride reactor.

Alternative approaches to fusion. [reactor design and reactor physics for Tokamak fusion reactors

NASA Technical Reports Server (NTRS)

Roth, R. J.

1976-01-01

The limitations of the Tokamak fusion reactor concept are discussed and various other fusion reactor concepts are considered that employ the containment of thermonuclear plasmas by magnetic fields (i.e., stellarators). Progress made in the containment of plasmas in toroidal devices is reported. Reactor design concepts are illustrated. The possibility of using fusion reactors as a power source in interplanetary space travel and electric power plants is briefly examined.

AGR-2 Irradiation Test Final As-Run Report, Rev 2

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

Collin, Blaise P.

2014-08-01

This document presents the as-run analysis of the AGR-2 irradiation experiment. AGR-2 is the second of the planned irradiations for the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. Funding for this program is provided by the U.S. Department of Energy as part of the Very High Temperature Reactor (VHTR) Technical Development Office (TDO) program. The objectives of the AGR-2 experiment are to: (a) Irradiate UCO (uranium oxycarbide) and UO 2 (uranium dioxide) fuel produced in a large coater. Fuel attributes are based on results obtained from the AGR-1 test and other project activities. (b) Provide irradiated fuel samplesmore » for post-irradiation experiment (PIE) and safety testing. (c) Support the development of an understanding of the relationship between fuel fabrication processes, fuel product properties, and irradiation performance. The primary objective of the test was to irradiate both UCO and UO 2 TRISO (tri-structural isotropic) fuel produced from prototypic scale equipment to obtain normal operation and accident condition fuel performance data. The UCO compacts were subjected to a range of burnups and temperatures typical of anticipated prismatic reactor service conditions in three capsules. The test train also includes compacts containing UO 2 particles produced independently by the United States, South Africa, and France in three separate capsules. The range of burnups and temperatures in these capsules were typical of anticipated pebble bed reactor service conditions. The results discussed in this report pertain only to U.S. produced fuel. In order to achieve the test objectives, the AGR-2 experiment was irradiated in the B-12 position of the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) for a total irradiation duration of 559.2 effective full power days (EFPD). Irradiation began on June 22, 2010, and ended on October 16, 2013, spanning 12 ATR power cycles and approximately three and a half calendar years. The test contained six independently controlled and monitored capsules. Each U.S. capsule contained 12 compacts of either UCO or UO2 AGR coated fuel. No fuel particles failed during the AGR-2 irradiation. Final burnup values on a per compact basis ranged from 7.26 to 13.15% FIMA (fissions per initial heavy-metal atom) for UCO fuel, and 9.01 to 10.69% FIMA for UO 2 fuel, while fast fluence values ranged from 1.94 to 3.47 x 10 25 n/m 2 (E >0.18 MeV) for UCO fuel, and from 3.05 to 3.53 x 10 25 n/m 2 (E >0.18 MeV) for UO 2 fuel. Time-average volume-average (TAVA) temperatures on a capsule basis at the end of irradiation ranged from 987°C in Capsule 6 to 1296°C in Capsule 2 for UCO, and from 996 to 1062°C in UO 2-fueled Capsule 3. By the end of the irradiation, all of the installed thermocouples (TCs) had failed. Fission product release-to-birth (R/B) ratios were quite low. In the UCO capsules, R/B values during the first three cycles were below 10 -6 with the exception of the hotter Capsule 2, in which the R/Bs reached 2 x 10 -6. In the UO 2 capsule (Capsule 3), the R/B values during the first three cycles were below 10 -7. R/B values for all following cycles are not reliable due to gas flow and cross talk issues.« less

Application of a Virtual Reactivity Feedback Control Loop in Non-Nuclear Testing of a Fast Spectrum Reactor

NASA Technical Reports Server (NTRS)

Bragg-Sitton, Shannon M.; Forsbacka, Matthew

2004-01-01

For a compact, fast-spectrum reactor, reactivity feedback is dominated by core deformation at elevated temperature. Given the use of accurate deformation measurement techniques, it is possible to simulate nuclear feedback in non-nuclear electrically heated reactor tests. Implementation of simulated reactivity feedback in response to measured deflection is being tested at the NASA Marshall Space Flight Center Early Flight Fission Test Facility (EFF-TF). During tests of the SAFE-100 reactor prototype, core deflection was monitored using a high resolution camera. "virtual" reactivity feedback was accomplished by applying the results of Monte Carlo calculations (MCNPX) to core deflection measurements; the computational analysis was used to establish the reactivity worth of van'ous core deformations. The power delivered to the SAFE-100 prototype was then dusted accordingly via kinetics calculations, The work presented in this paper will demonstrate virtual reactivity feedback as core power was increased from 1 kilowatt(sub t), to 10 kilowatts(sub t), held approximately constant at 10 kilowatts (sub t), and then allowed to decrease based on the negative thermal reactivity coefficient.

Chasing the light sterile neutrino with the STEREO detector

NASA Astrophysics Data System (ADS)

Minotti, A.

2017-09-01

The standard three-family neutrino oscillation model is challenged by a number of observations, such as the reactor antineutrino anomaly (RAA), that can be explained by the existence of sterile neutrinos at the eV mass scale. The STEREO experiment detects {\\bar ν _e} produced in the 58.3MW Th compact core of the ILL research reactor via inverse beta decay (IBD) interactions in a liquid scintillator. Using 6 identical target cells, STEREO compares {\\bar ν _e} energy spectra at different baselines in order to observe possible distortions due to short-baseline oscillations toward eV sterile neutrinos. IBD events are effectively singled out from γ radiation by selecting events with a two-fold coincidence that is typical of an IBD interaction. External background is reduced by means of layers of shielding material. A Cherenkov veto allows to partially remove background produced by cosmic muons, and the remaining component is measured in reactor-off periods and subtracted statistically. If no evidence of sterile neutrinos after the full statistics of 6 reactor cycles is gathered, STEREO is expected to fully exclude the RAA allowed region.

10 CFR Appendix A to Part 52 - Design Certification Rule for the U.S. Advanced Boiling Water Reactor

Code of Federal Regulations, 2010 CFR

2010-01-01

... Water Reactor A Appendix A to Part 52 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSES... Rule for the U.S. Advanced Boiling Water Reactor I. Introduction Appendix A constitutes the standard design certification for the U.S. Advanced Boiling Water Reactor (ABWR) design, in accordance with 10 CFR...

10 CFR Appendix A to Part 52 - Design Certification Rule for the U.S. Advanced Boiling Water Reactor

Code of Federal Regulations, 2011 CFR

2011-01-01

... Water Reactor A Appendix A to Part 52 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSES... Rule for the U.S. Advanced Boiling Water Reactor I. Introduction Appendix A constitutes the standard design certification for the U.S. Advanced Boiling Water Reactor (ABWR) design, in accordance with 10 CFR...

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

Ceramography of Irradiated tristructural isotropic (TRISO) Fuel from the AGR-2 Experiment

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

Rice, Francine Joyce; Stempien, John Dennis

2016-09-01

Ceramography was performed on cross sections from four tristructural isotropic (TRISO) coated particle fuel compacts taken from the AGR-2 experiment, which was irradiated between June 2010 and October 2013 in the Advanced Test Reactor (ATR). The fuel compacts examined in this study contained TRISO-coated particles with either uranium oxide (UO2) kernels or uranium oxide/uranium carbide (UCO) kernels that were irradiated to final burnup values between 9.0 and 11.1% FIMA. These examinations are intended to explore kernel and coating morphology evolution during irradiation. This includes kernel porosity, swelling, and migration, and irradiation-induced coating fracture and separation. Variations in behavior within amore » specific cross section, which could be related to temperature or burnup gradients within the fuel compact, are also explored. The criteria for categorizing post-irradiation particle morphologies developed for AGR-1 ceramographic exams, was applied to the particles in the AGR-2 compacts particles examined. Results are compared with similar investigations performed as part of the earlier AGR-1 irradiation experiment. This paper presents the results of the AGR-2 examinations and discusses the key implications for fuel irradiation performance.« less

Spark-plasma-sintering magnetic field assisted compaction of Co{sub 80}Ni{sub 20} nanowires for anisotropic ferromagnetic bulk materials

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

Ouar, Nassima; Schoenstein, Frédéric; Mercone, Silvana

We developed a two-step process showing the way for sintering anisotropic nanostructured bulk ferromagnetic materials. A new reactor has been optimized allowing the synthesis of several grams per batch of nanopowders via a polyol soft chemistry route. The feasibility of the scale-up has been successfully demonstrated for Co{sub 80}Ni{sub 20} nanowires and a massic yield of ∼97% was obtained. The thus obtained nanowires show an average diameter of ∼6 nm and a length of ∼270 nm. A new bottom-up strategy allowed us to compact the powder into a bulk nanostructured system. We used a spark-plasma-sintering technique under uniaxial compression andmore » low temperature assisted by a permanent magnetic field of 1 T. A macroscopic pellet of partially aligned nanowire arrays has been easily obtained. This showed optimized coercive properties along the direction of the magnetic field applied during compaction (i.e., the nanowires' direction)« less

Tower Shielding Reactor II design and operation report: Vol. 2. Safety Analysis

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

Holland, L. B.; Kolb, J. O.

1970-01-01

Information on the Tower Shielding Reactor II is contained in the TSR-II Design and Operation Report and in the Tower Shielding Facility Manual. The TSR-II Design and Operating Report consists of three volumes. Volume 1 is Descriptions of the Tower Shielding Reactor II and Facility; Volume 2 is Safety analysis of the Tower Shielding Reactor II; and Volume 3 is the Assembly and Testing of the Tower Shielding Reactor II Control Mechanism Housing.

Analysis of the Gas Core Actinide Transmutation Reactor (GCATR)

NASA Technical Reports Server (NTRS)

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

1977-01-01

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

Princeton Plasma Physics Laboratory: Annual report, October 1, 1986--September 30, 1987

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

Not Available

1987-01-01

This report contains papers on the following topics: Principle Parameters Achieved in Experimental Devices (FY87); Tokamak Fusion Test Reactor; Princeton Beta Experiment-Modification; S-1 Spheromak; Current-Drive Experiment; X-Ray Laser Studies; Theoretical Division; Tokamak Modeling; Compact Ignition Tokamak; Engineering Department; Project Planning and Safety Office; Quality Assurance and Reliability; Administrative Operations; and PPPL Patent Invention Disclosures (FY87).

Coil Design for Low Aspect Ratio Stellarators

NASA Astrophysics Data System (ADS)

Miner, W. H., Jr.; Valanju, P. M.; Wiley, J. C.; Hirshman, S. P.; Whitson, J. C.

1998-11-01

Two compact stellarator designs have recently been under investigation because of their potential as a reactor featuring steady-state, disruption-free operation, low recirculating power and good confinement and beta. Both quasi-axisymmetric (QA) equilibria and quasi-omnigenous (QO) equilibria have been obtained by using the 3-D MHD equilibrium code VMEC. In order to build an experiment, coil sets must be obtained that are compatable with these equilibria. We have been using both the NESCOIL(Merkel, P., Nucl. Fus. 27, 5 (1987) 867.) code and the COILOPT code to find coilsets for both of these types of equilibria. We are considering three types of coil configurations. The first is a combination of modular coils and vertical field coils. The second configuration is a combination of toroidal field coils, vertical field coils and saddle coils. A third configuration is a combination of modular coils and a single helical winding. The quality of each coil set will be evaluated by computing its magnetic field and using that as input to VMEC in free boundary mode to see how accurately the original equilibrium can be reconstructed.

Development of toroid-type HTS DC reactor series for HVDC system

NASA Astrophysics Data System (ADS)

Kim, Kwangmin; Go, Byeong-Soo; Park, Hea-chul; Kim, Sung-kyu; Kim, Seokho; Lee, Sangjin; Oh, Yunsang; Park, Minwon; Yu, In-Keun

2015-11-01

This paper describes design specifications and performance of a toroid-type high-temperature superconducting (HTS) DC reactor. The first phase operation targets of the HTS DC reactor were 400 mH and 400 A. The authors have developed a real HTS DC reactor system during the last three years. The HTS DC reactor was designed using 2G GdBCO HTS wires. The HTS coils of the toroid-type DC reactor magnet were made in the form of a D-shape. The electromagnetic performance of the toroid-type HTS DC reactor magnet was analyzed using the finite element method program. A conduction cooling method was adopted for reactor magnet cooling. The total system has been successfully developed and tested in connection with LCC type HVDC system. Now, the authors are studying a 400 mH, kA class toroid-type HTS DC reactor for the next phase research. The 1500 A class DC reactor system was designed using layered 13 mm GdBCO 2G HTS wire. The expected operating temperature is under 30 K. These fundamental data obtained through both works will usefully be applied to design a real toroid-type HTS DC reactor for grid application.

Design and implementation of the ESL compact range underhung bridge crane

NASA Technical Reports Server (NTRS)

Shamansky, H. T.; Dominek, A. K.; Burnside, W. D.

1987-01-01

As the indoor compact range technology has continued to increase, the need to handle larger and heavier targets has also increased. This need for target lifting and handling prompted the feasibility study of the use of an underhung bridge crane to be installed in the ESL (ElectroScience Laboratory, Ohio State University) compact range. This report documents both the design of the underhung bridge crane that was installed and the implementation of the design in the actual installation of the crane.

A compact inflow control device for simulating flight fan noise

NASA Technical Reports Server (NTRS)

Homyak, L.; Mcardle, J. G.; Heidelberg, L. J.

1983-01-01

Inflow control device (ICD's) of various shapes and sizes have been used to simulate inflight fan tone noise during ground static tests. A small, simple inexpensive ICD design was optimized from previous design and fabrication techniques. This compact two-fan-diameter ICD exhibits satisfactory acoustic performance characteristics without causing noise attenuation or redirection. In addition, it generates no important new noise sources. Design and construction details of the compact ICD are discussed and acoustic performance test results are presented.

Hydrogen production by reforming of liquid hydrocarbons in a membrane reactor for portable power generation-Experimental studies

NASA Astrophysics Data System (ADS)

Damle, Ashok S.

One of the most promising technologies for lightweight, compact, portable power generation is proton exchange membrane (PEM) fuel cells. PEM fuel cells, however, require a source of pure hydrogen. Steam reforming of hydrocarbons in an integrated membrane reactor has potential to provide pure hydrogen in a compact system. Continuous separation of product hydrogen from the reforming gas mixture is expected to increase the yield of hydrogen significantly as predicted by model simulations. In the laboratory-scale experimental studies reported here steam reforming of liquid hydrocarbon fuels, butane, methanol and Clearlite ® was conducted to produce pure hydrogen in a single step membrane reformer using commercially available Pd-Ag foil membranes and reforming/WGS catalysts. All of the experimental results demonstrated increase in hydrocarbon conversion due to hydrogen separation when compared with the hydrocarbon conversion without any hydrogen separation. Increase in hydrogen recovery was also shown to result in corresponding increase in hydrocarbon conversion in these studies demonstrating the basic concept. The experiments also provided insight into the effect of individual variables such as pressure, temperature, gas space velocity, and steam to carbon ratio. Steam reforming of butane was found to be limited by reaction kinetics for the experimental conditions used: catalysts used, average gas space velocity, and the reactor characteristics of surface area to volume ratio. Steam reforming of methanol in the presence of only WGS catalyst on the other hand indicated that the membrane reactor performance was limited by membrane permeation, especially at lower temperatures and lower feed pressures due to slower reconstitution of CO and H 2 into methane thus maintaining high hydrogen partial pressures in the reacting gas mixture. The limited amount of data collected with steam reforming of Clearlite ® indicated very good match between theoretical predictions and experimental results indicating that the underlying assumption of the simple model of conversion of hydrocarbons to CO and H 2 followed by equilibrium reconstitution to methane appears to be reasonable one.

Design and evaluation of experimental ceramic automobile thermal reactors

NASA Technical Reports Server (NTRS)

Stone, P. L.; Blankenship, C. P.

1974-01-01

The paper summarizes the results obtained in an exploratory evaluation of ceramics for automobile thermal reactors. Candidate ceramic materials were evaluated in several reactor designs using both engine dynamometer and vehicle road tests. Silicon carbide contained in a corrugated metal support structure exhibited the best performance, lasting 1100 hours in engine dynamometer tests and for more than 38,600 kilimeters (24,000 miles) in vehicle road tests. Although reactors containing glass-ceramic components did not perform as well as silicon carbide, the glass-ceramics still offer good potential for reactor use with improved reactor designs.

Design and evaluation of experimental ceramic automobile thermal reactors

NASA Technical Reports Server (NTRS)

Stone, P. L.; Blankenship, C. P.

1974-01-01

The results obtained in an exploratory evaluation of ceramics for automobile thermal reactors are summarized. Candidate ceramic materials were evaluated in several reactor designs by using both engine-dynamometer and vehicle road tests. Silicon carbide contained in a corrugated-metal support structure exhibited the best performance, lasting 1100 hr in engine-dynamometer tests and more than 38,600 km (24000 miles) in vehicle road tests. Although reactors containing glass-ceramic components did not perform as well as those containing silicon carbide, the glass-ceramics still offer good potential for reactor use with improved reactor designs.

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

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

Sienicki, James J.; Grandy, Christopher

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

High-throughput spectrometer designs in a compact form-factor: principles and applications

NASA Astrophysics Data System (ADS)

Norton, S. M.

2013-05-01

Many compact, portable Raman spectrometers have entered the market in the past few years with applications in narcotics and hazardous material identification, as well as verification applications in pharmaceuticals and security screening. Often, the required compact form-factor has forced designers to sacrifice throughput and sensitivity for portability and low-cost. We will show that a volume phase holographic (VPH)-based spectrometer design can achieve superior throughput and thus sensitivity over conventional Czerny-Turner reflective designs. We will look in depth at the factors influencing throughput and sensitivity and illustrate specific VPH-based spectrometer examples that highlight these design principles.

Initial results from safety testing of US AGR-2 irradiation test fuel

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

Morris, Robert Noel; Hunn, John D.; Baldwin, Charles A.

Two cylindrical compacts containing tristructural isotropic (TRISO)-coated particles with kernels that contained a mixture of uranium carbide and uranium oxide (UCO) and two compacts with UO 2-kernel TRISO particles have undergone 1600°C safety testing. These compacts were irradiated in the US Advanced Gas Reactor Fuel Development and Qualification Program's second irradiation test (AGR-2). The time-dependent releases of several radioisotopes ( 110mAg, 134Cs, 137Cs, 154Eu, 155Eu, 90Sr, and 85Kr) were monitored while heating the fuel specimens to 1600°C in flowing helium for 300 h. The UCO compacts behaved similarly to previously reported 1600°C-safety-tested UCO compacts from the AGR-1 irradiation. No failedmore » TRISO or failed SiC were detected (based on krypton and cesium release), and cesium release through intact SiC was very low. Release behavior of silver, europium, and strontium appeared to be dominated by inventory originally released through intact coating layers during irradiation but retained in the compact matrix until it was released during safety testing. Both UO 2 compacts exhibited cesium release from multiple particles whose SiC failed during the safety test. Europium and strontium release from these two UO 2 compacts appeared to be dominated by release from the particles with failed SiC. Silver release was characteristically like the release from the UCO compacts in that an initial release of the majority of silver trapped in the matrix occurred during ramping to 1600°C. However, additional silver release was observed later in the safety testing due to the UO 2 TRISO with failed SiC. Failure of the SiC layer in the UO 2 fuel appears to have been dominated by CO corrosion, as opposed to the palladium degradation observed in AGR-1 UCO fuel.« less

Initial results from safety testing of US AGR-2 irradiation test fuel

DOE PAGES

Morris, Robert Noel; Hunn, John D.; Baldwin, Charles A.; ...

2017-08-18

Two cylindrical compacts containing tristructural isotropic (TRISO)-coated particles with kernels that contained a mixture of uranium carbide and uranium oxide (UCO) and two compacts with UO 2-kernel TRISO particles have undergone 1600°C safety testing. These compacts were irradiated in the US Advanced Gas Reactor Fuel Development and Qualification Program's second irradiation test (AGR-2). The time-dependent releases of several radioisotopes ( 110mAg, 134Cs, 137Cs, 154Eu, 155Eu, 90Sr, and 85Kr) were monitored while heating the fuel specimens to 1600°C in flowing helium for 300 h. The UCO compacts behaved similarly to previously reported 1600°C-safety-tested UCO compacts from the AGR-1 irradiation. No failedmore » TRISO or failed SiC were detected (based on krypton and cesium release), and cesium release through intact SiC was very low. Release behavior of silver, europium, and strontium appeared to be dominated by inventory originally released through intact coating layers during irradiation but retained in the compact matrix until it was released during safety testing. Both UO 2 compacts exhibited cesium release from multiple particles whose SiC failed during the safety test. Europium and strontium release from these two UO 2 compacts appeared to be dominated by release from the particles with failed SiC. Silver release was characteristically like the release from the UCO compacts in that an initial release of the majority of silver trapped in the matrix occurred during ramping to 1600°C. However, additional silver release was observed later in the safety testing due to the UO 2 TRISO with failed SiC. Failure of the SiC layer in the UO 2 fuel appears to have been dominated by CO corrosion, as opposed to the palladium degradation observed in AGR-1 UCO fuel.« less

A systematic reactor design approach for the synthesis of active pharmaceutical ingredients.

PubMed

Emenike, Victor N; Schenkendorf, René; Krewer, Ulrike

2018-05-01

Today's highly competitive pharmaceutical industry is in dire need of an accelerated transition from the drug development phase to the drug production phase. At the heart of this transition are chemical reactors that facilitate the synthesis of active pharmaceutical ingredients (APIs) and whose design can affect subsequent processing steps. Inspired by this challenge, we present a model-based approach for systematic reactor design. The proposed concept is based on the elementary process functions (EPF) methodology to select an optimal reactor configuration from existing state-of-the-art reactor types or can possibly lead to the design of novel reactors. As a conceptual study, this work summarizes the essential steps in adapting the EPF approach to optimal reactor design problems in the field of API syntheses. Practically, the nucleophilic aromatic substitution of 2,4-difluoronitrobenzene was analyzed as a case study of pharmaceutical relevance. Here, a small-scale tubular coil reactor with controlled heating was identified as the optimal set-up reducing the residence time by 33% in comparison to literature values. Copyright © 2017 Elsevier B.V. All rights reserved.

Micro-focused Small Angle Neutron Scattering and Imaging for Science and Engineering Using RTP--A Preliminary Study

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

Mohamed, Abdul Aziz; Al Rashid Megat Ahmad, Megat Harun; Md Idris, Faridah

2010-01-05

Malaysian Nuclear Agency's (Nuclear Malaysia) Small Angle Neutron Scattering (SANS) facility--(MYSANS)--is utilizing low flux of thermal neutron at the agency's 1 MW TRIGA reactor. As the design nature of the 8 m SANS facility can allow object resolution in the range between 5 and 80 nm to be obtained. It can be used to study alloys, ceramics and polymers in certain area of problems that relate to samples containing strong scatterers or contrast. The current SANS system at Malaysian Nuclear Agency is only capable to measure Q in limited range with a PSD (128x128) fixed at 4 m from themore » sample. The existing reactor hall that incorporate this MYSANS facility has a layout that prohibits the rebuilding of MYSANS therefore the position between the wavelength selector (HOPG) and sample and the PSD cannot be increased for wider Q range. The flux of the neutron at current sample holder is very low which around 10{sup 3} n/cm{sup 2}/sec. Thus it is important to rebuild the MYSANS to maximize the utilization of neutron. Over the years, the facility has undergone maintenance and some changes have been made. Modification on secondary shutter and control has been carried out to improve the safety level of the instrument. A compact micro-focus SANS method can suit this objective together with an improve cryostat system. This paper will explain some design concept and approaches in achieving higher flux and the modification needs to establish the micro-focused SANS.« less

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.

Inherently Safe and Long-Life Fission Power System for Lunar Outposts

NASA Astrophysics Data System (ADS)

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

Power requirements for future lunar outposts, of 10's to 100's kWe, can be fulfilled using nuclear reactor power systems. In addition to the long life and operation reliability, safety is paramount in all phases, including fabrication and assembly, launch, emplacement below grade on the lunar surface, operation, post-operation decay heat removal and long-term storage and eventual retrieval. This paper introduces the Solid Core-Sectored Compact Reactor (SC-SCoRe) and power system with static components and no single point failures. They ensure reliable continuous operation for ~21 years and fulfill the safety requirements. The SC-SCoRe nominally generates 1.0 MWth at liquid NaK-56 coolant inlet and exit temperatures of 850 K and 900 K and the power system provides 38 kWe at high DC voltage using SiGe thermoelectric (TE) conversion assemblies. In case of a loss of coolant or cooling in a reactor core sector, the power system continues to operate; generating ~4 kWe to the outpost for emergency life support needs. The post-operation storage of the reactor below grade on the lunar surface is a safe and practical choice. The total radioactivity in the reactor drops from ~1 million Ci, immediately at shutdown, to below 164 Ci after 300 years of storage. At such time, the reactor is retrieved safely with no contamination or environmental concerns.

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

Federal Register 2010, 2011, 2012, 2013, 2014

2011-03-25

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

Scoping study for compact high-field superconducting net energy tokamaks

NASA Astrophysics Data System (ADS)

Mumgaard, R. T.; Greenwald, M.; Freidberg, J. P.; Wolfe, S. M.; Hartwig, Z. S.; Brunner, D.; Sorbom, B. N.; Whyte, D. G.

2016-10-01

The continued development and commercialization of high temperature superconductors (HTS) may enable the construction of compact, net-energy tokamaks. HTS, in contrast to present generation low temperature superconductors, offers improved performance in high magnetic fields, higher current density, stronger materials, higher temperature operation, and simplified assembly. Using HTS along with community-consensus confinement physics (H98 =1) may make it possible to achieve net-energy (Q>1) or burning plasma conditions (Q>5) in DIII-D or ASDEX-U sized, conventional aspect ratio tokamaks. It is shown that, by operating at high plasma current and density enabled by the high magnetic field (B>10T), the required triple products may be achieved at plasma volumes under 20m3, major radii under 2m, with external heating powers under 40MW. This is at the scale of existing devices operated by laboratories, universities and companies. The trade-offs in the core heating, divertor heat exhaust, sustainment, stability, and proximity to known plasma physics limits are discussed in the context of the present tokamak experience base and the requirements for future devices. The resulting HTS-based design space is compared and contrasted to previous studies on high-field copper experiments with similar missions. The physics exploration conducted with such HTS devices could decrease the real and perceived risks of ITER exploitation, and aid in quickly developing commercially-applicable tokamak pilot plants and reactors.

10 CFR Appendix N to Part 52 - Standardization of Nuclear Power Plant Designs: Combined Licenses To Construct and Operate...

Code of Federal Regulations, 2011 CFR

2011-01-01

... Licenses To Construct and Operate Nuclear Power Reactors of Identical Design at Multiple Sites N Appendix N... Designs: Combined Licenses To Construct and Operate Nuclear Power Reactors of Identical Design at Multiple... construct and operate nuclear power reactors of identical design (“common design”) to be located at multiple...

UV-laser-based longitudinal illuminated diffuser (LID) incorporating diffractive and Lambertian reflectance for the disinfection of beverages

NASA Astrophysics Data System (ADS)

Lizotte, Todd

2010-08-01

A novel laser beam shaping system was designed to demonstrate the potential of using high power UV laser sources for large scale disinfection of liquids used in the production of food products, such as juices, beer, milk and other beverage types. The design incorporates a patented assembly of optical components including a diffractive beam splitting/shaping element and a faceted pyramidal or conically shaped Lambertian diffuser made from a compression molded PTFE compounds. When properly sintered to an appropriate density, as an example between 1.10 and 1.40 grams per cubic centimeter, the compressed PTFE compounds show a ~99% reflectance at wavelengths ranging from 300 nm to 1500 nm, and a ~98.5% refection of wavelengths from 250 nm to 2000 nm [1]. The unique diffuser configuration also benefits from the fact that the PTFE compounds do not degrade when exposed to ultraviolet radiation as do barium sulfate materials and silver or aluminized mirror coatings [2]. These components are contained within a hermetically sealed quartz tube. Once assembled a laser beam is directed through one end of the tube. This window takes the form of a computer generated diffractive splitter or other diffractive shaper element to split the laser beam into a series of spot beamlets, circular rings or other geometric shapes. As each of the split beamlets or rings cascade downward, they illuminate various points along the tapered PTFE cone or faceted pyramidal form. As they strike the surface they each diffuse in a Lambertian reflectance pattern creating a pseudo-uniform circumferential illuminator along the length of the quartz tube enclosing the assembly. The compact tubular structure termed Longitudinal Illuminated Diffuser (LID) provides a unique UV disinfection source that can be placed within a centrifugal reactor or a pipe based reactor chamber. This paper will review the overall design principle, key component design parameters, preliminary analytic and bench operational testing results.

Investigation of Isotopically Tailored Boron in Advanced Fission and Fusion Reactor Systems.

NASA Astrophysics Data System (ADS)

Domaszek, Gerald Raymond

This research examines the use of B^ {11}, in the form of metallic boron and boron carbide, as a moderating and reflecting material. An examination of the neutronic characteristics of the B ^{11} isotope of boron has revealed that B^{11} has neutron scattering and absorption cross sections favorably comparable to those of Be^9 and C^ {12}. Preliminary analysis of the neutronics of B ^{11} were performed by conducting one dimensional transport calculations on an infinite slab of varying thickness. Beryllium is the best of the three materials in reflecting neutrons due primarily to the contribution from (n,2n) reactions. Tailored neutron energy beam transmission experiments were carried out to experimentally verify the predicted neutronic characteristics of B^{11 }. To further examine the neutron moderating and reflecting characteristics of B^{11 }, the energy dependent neutron flux was measured as a function of position in an exponential pile constructed of B_4C isotopically enriched to 98.5 percent B^{11}. After the experimental verification of the neutronic behavior of B^{11}, further design studies were conducted using metallic boron and boron carbide enriched in the B^{11 } isotope. The use of materials isotopically enriched in B^{11} as a liner in the first wall/blanket of a magnetic confinement fusion reactor demonstrated acceptable tritium regeneration in the lithium blanket. Analysis of the effect of contaminant levels of B^{10} showed that B^{10} contents of less than 1 percent in metallic boron produced negligible adverse effects on the tritium breeding. A comparison of the effectiveness of graphite and B^{11}_4C when used as moderators in a reactor fueled with natural uranium has shown that the maximum k_infty for a given fuel rod design is approximately the same for both materials. Approximately half the volume of the moderator is required when B^{11 }_4C is substituted for graphite to obtain essentially the same K_infty . An analysis of the effectiveness of various materials as reflector control elements for a compact space reactor has shown that B^{11} is neutronically superior to graphite in these applications. Metallic boron and boron carbide isotopically enriched in B^{11} have been demonstrated to be neutronically acceptable for varied applications in advanced reactor systems. B^ {11} has been shown to be superior in performance to graphite. While only somewhat inferior to beryllium as neutron multipliers, B^ {11} and B^{11} _4C have safety, supply and cost advantage over beryllium. (Abstract shortened with permission of author.).

Exploratory development of a glass ceramic automobile thermal reactor. [anti-pollution devices

NASA Technical Reports Server (NTRS)

Gould, R. E.; Petticrew, R. W.

1973-01-01

This report summarizes the design, fabrication and test results obtained for glass-ceramic (CER-VIT) automotive thermal reactors. Several reactor designs were evaluated using both engine-dynamometer and vehicle road tests. A maximum reactor life of about 330 hours was achieved in engine-dynamometer tests with peak gas temperatures of about 1065 C (1950 F). Reactor failures were mechanically induced. No evidence of chemical degradation was observed. It was concluded that to be useful for longer times, the CER-VIT parts would require a mounting system that was an improvement over those tested in this program. A reactor employing such a system was designed and fabricated.

Advanced Exploration Systems Logistics Reduction and Repurposing Trash-to-Gas and Heat Melt Compactor KSC

NASA Technical Reports Server (NTRS)

Caraccio, Anne J.; Layne, Andrew; Hummerick, Mary

2013-01-01

Topics covered: 1. Project Structure 2. "Trash to Gas" 3. "Smashing Trash! The Heat Melt Compactor" 4. "Heat Melt Compaction as an Effective Treatment for Eliminating Microorganisms from Solid Waste" Thermal degradation of trash reduces volume while creating water, carbon dioxide and ash. CO2 can be fed to Sabatier reactor for CH4 production to fuel LOX/LCH4 ascent vehicle. Optimal performance: HFWS, full temperature ramp to 500-600 C. Tar challenges exist. Catalysis: Dolomag did eliminate allene byproducts from the product stream. 2nd Gen Reactor Studies. Targeting power, mass, time efficiency. Gas separation, Catalysis to reduce tar formation. Microgravity effects. Downselect in August will determine where we should spend time optimizing the technology.

40 CFR 258.40 - Design criteria.

Code of Federal Regulations, 2013 CFR

2013-07-01

... component must consist of at least a two-foot layer of compacted soil with a hydraulic conductivity of no... compacted soil com-ponent. (c) When approving a design that complies with paragraph (a)(1) of this section...

40 CFR 258.40 - Design criteria.

Code of Federal Regulations, 2014 CFR

2014-07-01

... component must consist of at least a two-foot layer of compacted soil with a hydraulic conductivity of no... compacted soil com-ponent. (c) When approving a design that complies with paragraph (a)(1) of this section...

40 CFR 258.40 - Design criteria.

Code of Federal Regulations, 2011 CFR

2011-07-01

... component must consist of at least a two-foot layer of compacted soil with a hydraulic conductivity of no... compacted soil com-ponent. (c) When approving a design that complies with paragraph (a)(1) of this section...

40 CFR 258.40 - Design criteria.

Code of Federal Regulations, 2010 CFR

2010-07-01

... component must consist of at least a two-foot layer of compacted soil with a hydraulic conductivity of no... compacted soil com-ponent. (c) When approving a design that complies with paragraph (a)(1) of this section...

40 CFR 258.40 - Design criteria.

Code of Federal Regulations, 2012 CFR

2012-07-01

... component must consist of at least a two-foot layer of compacted soil with a hydraulic conductivity of no... compacted soil com-ponent. (c) When approving a design that complies with paragraph (a)(1) of this section...

Moving bed reactor setup to study complex gas-solid reactions.

PubMed

Gupta, Puneet; Velazquez-Vargas, Luis G; Valentine, Charles; Fan, Liang-Shih

2007-08-01

A moving bed scale reactor setup for studying complex gas-solid reactions has been designed in order to obtain kinetic data for scale-up purpose. In this bench scale reactor setup, gas and solid reactants can be contacted in a cocurrent and countercurrent manner at high temperatures. Gas and solid sampling can be performed through the reactor bed with their composition profiles determined at steady state. The reactor setup can be used to evaluate and corroborate model parameters accounting for intrinsic reaction rates in both simple and complex gas-solid reaction systems. The moving bed design allows experimentation over a variety of gas and solid compositions in a single experiment unlike differential bed reactors where the gas composition is usually fixed. The data obtained from the reactor can also be used for direct scale-up of designs for moving bed reactors.

NUCLEAR REACTOR FUEL ELEMENTS AND METHOD OF PREPARATION

DOEpatents

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

1963-07-01

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

Shielding Analysis of a Small Compact Space Nuclear Reactor

DTIC Science & Technology

1987-08-01

RESPONSE) =4, MAXWELLIAN FISSION SPECTRUM (ILNTEGRAL RESPONSE) =5, LOS ALAMOS FISSION SPECTRUM, 1982 (INTEGRAL RESPONSE) =6, VITAMIN C NEUTRON SPECTRUM...Appendices Appendix A: Calculations of Effective Radii.. A-1 Appendix B: Atom Density Calculations for FEMPlD and FEMP2D ................ B-I Appendix C ...FEMPID and FEM22D Data........... C -i Appendix D: Energy Group Definition .......... D-I Appendix E: Transport Equation, Legendr4 Polynomial

Effects of Beryllium and Compaction Pressure on the Thermal Diffusivity of Uranium Dioxide Fuel Pellets

NASA Astrophysics Data System (ADS)

Camarano, D. M.; Mansur, F. A.; Santos, A. M. M.; Ferraz, W. B.; Ferreira, R. A. N.

2017-09-01

In nuclear reactors, the performance of uranium dioxide (UO2) fuel is strongly dependent on the thermal conductivity, which directly affects the fuel pellet temperature, the fission gas release and the fuel rod mechanical behavior during reactor operation. The use of additives to improve UO2 fuel performance has been investigated, and beryllium oxide (BeO) appears as a suitable additive because of its high thermal conductivity and excellent chemical compatibility with UO2. In this paper, UO2-BeO pellets were manufactured by mechanical mixing, pressing and sintering processes varying the BeO contents and compaction pressures. Pellets with BeO contents of 2 wt%, 3 wt%, 5 wt% and 7 wt% BeO were pressed at 400 MPa, 500 MPa and 600 MPa. The laser flash method was applied to determine the thermal diffusivity, and the results showed that the thermal diffusivity tends to increase with BeO content. Comparing thermal diffusivity results of UO2 with UO2-BeO pellets, it was observed that there was an increase in thermal diffusivity of at least 18 % for the UO2-2 wt% BeO pellet pressed at 400 MPa. The maximum relative expanded uncertainty (coverage factor k = 2) of the thermal diffusivity measurements was estimated to be 9 %.

Navy Nuclear-Powered Surface Ships: Background, Issues, and Options for Congress

DTIC Science & Technology

2010-09-29

to design a smaller scale version of a naval pressurized water reactor , or to design a new reactor type potentially using a thorium liquid salt...integrated nuclear power system capable of use on destroyer- sized vessels either using a pressurized water reactor or a thorium liquid salt reactor ...nuclear reactors for Navy surface ships. The text of Section 246 is as follows: SEC. 246. STUDY ON THORIUM -LIQUID FUELED REACTORS FOR NAVAL FORCES

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.

10 CFR 2.1115 - Designation of issues for adjudicatory hearing.

Code of Federal Regulations, 2010 CFR

2010-01-01

... at Civilian Nuclear Power Reactors § 2.1115 Designation of issues for adjudicatory hearing. (a) After... reactor already licensed to operate at the site, or any civilian nuclear power reactor for which a... the issuance of a construction permit or operating license for a civilian nuclear power reactor at...

10 CFR Appendix N to Part 50 - Standardization of Nuclear Power Plant Designs: Permits To Construct and Licenses To Operate...

Code of Federal Regulations, 2011 CFR

2011-01-01

... Construct and Licenses To Operate Nuclear Power Reactors of Identical Design at Multiple Sites N Appendix N... Construct and Licenses To Operate Nuclear Power Reactors of Identical Design at Multiple Sites Section 101... nuclear power reactors of essentially the same design to be located at different sites. 1 1 If the design...

10 CFR Appendix N to Part 50 - Standardization of Nuclear Power Plant Designs: Permits To Construct and Licenses To Operate...

Code of Federal Regulations, 2012 CFR

2012-01-01

... Construct and Licenses To Operate Nuclear Power Reactors of Identical Design at Multiple Sites N Appendix N... Construct and Licenses To Operate Nuclear Power Reactors of Identical Design at Multiple Sites Section 101... nuclear power reactors of essentially the same design to be located at different sites. 1 1 If the design...

10 CFR Appendix N to Part 50 - Standardization of Nuclear Power Plant Designs: Permits To Construct and Licenses To Operate...

Code of Federal Regulations, 2013 CFR

2013-01-01

... Construct and Licenses To Operate Nuclear Power Reactors of Identical Design at Multiple Sites N Appendix N... Construct and Licenses To Operate Nuclear Power Reactors of Identical Design at Multiple Sites Section 101... nuclear power reactors of essentially the same design to be located at different sites. 1 1 If the design...

10 CFR Appendix A to Part 725 - Categories of Restricted Data Available

Code of Federal Regulations, 2011 CFR

2011-01-01

... centrifuge or gaseous diffusion processes. b. Design, construction, and operation of any plant, facility or..., design, criticality studies and operation of reactors, reactor systems and reactor components. d... aqueous lithium hydroxide solution in packed columns. Not included is information regarding plant design...

10 CFR Appendix A to Part 725 - Categories of Restricted Data Available

Code of Federal Regulations, 2012 CFR

2012-01-01

... centrifuge or gaseous diffusion processes. b. Design, construction, and operation of any plant, facility or..., design, criticality studies and operation of reactors, reactor systems and reactor components. d... aqueous lithium hydroxide solution in packed columns. Not included is information regarding plant design...

10 CFR Appendix A to Part 725 - Categories of Restricted Data Available

Code of Federal Regulations, 2010 CFR

2010-01-01

... centrifuge or gaseous diffusion processes. b. Design, construction, and operation of any plant, facility or..., design, criticality studies and operation of reactors, reactor systems and reactor components. d... aqueous lithium hydroxide solution in packed columns. Not included is information regarding plant design...

10 CFR Appendix A to Part 725 - Categories of Restricted Data Available

Code of Federal Regulations, 2014 CFR

2014-01-01

... centrifuge or gaseous diffusion processes. b. Design, construction, and operation of any plant, facility or..., design, criticality studies and operation of reactors, reactor systems and reactor components. d... aqueous lithium hydroxide solution in packed columns. Not included is information regarding plant design...

POWER-BURST FACILITY (PBF) CONCEPTUAL DESIGN

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

Wasserman, A.A.; Johnson, S.O.; Heffner, R.E.

1963-06-21

A description is presented of the conceptual design of a high- performance, pulsed reactor called the Power Burst Facility (PBF). This reactor is designed to generate power bursts with initial asymptotic periods as short as 1 msec, producing energy releases large enough to destroy entire fuel subassemblies placed in a capsule or flow loop mounted in the reactor, all without damage to the reactor itself. It will be used primarily to evaluate the consequences and hazards of very rapid destructive accidents in reactors representing the entire range of current nuclear technology as applied to power generation, propulsion, and testing. Itmore » will also be used to carry out detailed studies of nondestructive reactivity feedback mechanisms in the shortperiod domain. The facility was designed to be sufficiently flexible to accommodate future cores of even more advanced design. The design for the first reactor core is based upon proven technology; hence, completion of the final design of this core will involve no significant development delays. Construction of the PBF is proposed to begin in September 1984, and is expected to take approximately 20 months to complete. (auth)« less

The 5-kwe reactor thermoelectric system summary

NASA Technical Reports Server (NTRS)

Vanosdol, J. H. (Editor)

1973-01-01

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

Radiography with cosmic-ray and compact accelerator muons; Exploring inner-structure of large-scale objects and landforms

PubMed Central

NAGAMINE, Kanetada

2016-01-01

Cosmic-ray muons (CRM) arriving from the sky on the surface of the earth are now known to be used as radiography purposes to explore the inner-structure of large-scale objects and landforms, ranging in thickness from meter to kilometers scale, such as volcanic mountains, blast furnaces, nuclear reactors etc. At the same time, by using muons produced by compact accelerators (CAM), advanced radiography can be realized for objects with a thickness in the sub-millimeter to meter range, with additional exploration capability such as element identification and bio-chemical analysis. In the present report, principles, methods and specific research examples of CRM transmission radiography are summarized after which, principles, methods and perspective views of the future CAM radiography are described. PMID:27725469

Radiography with cosmic-ray and compact accelerator muons; Exploring inner-structure of large-scale objects and landforms.

PubMed

Nagamine, Kanetada

2016-01-01

Cosmic-ray muons (CRM) arriving from the sky on the surface of the earth are now known to be used as radiography purposes to explore the inner-structure of large-scale objects and landforms, ranging in thickness from meter to kilometers scale, such as volcanic mountains, blast furnaces, nuclear reactors etc. At the same time, by using muons produced by compact accelerators (CAM), advanced radiography can be realized for objects with a thickness in the sub-millimeter to meter range, with additional exploration capability such as element identification and bio-chemical analysis. In the present report, principles, methods and specific research examples of CRM transmission radiography are summarized after which, principles, methods and perspective views of the future CAM radiography are described.

Application and Development of Microstructured Solid-State Neutron Detectors

NASA Astrophysics Data System (ADS)

Weltz, Adam D.

Neutron detectors are useful for a number of applications, including the identification of nuclear weapons, radiation dosimetry, and nuclear reactor monitoring, among others. Microstructured solid-state neutron detectors (SSNDs) developed at RPI have the potential to reinvent a variety of neutron detection systems due to their compact size, zero bias requirement, competitive thermal neutron detection efficiency (up to 29%), low gamma sensitivity (below the PNNL recommendation of 10-6 corresponding to a 10 mR/hr gamma exposure), and scalability to large surface areas with a single preamplifier (<20% loss in relative efficiency from 1 to 16 cm2). These microstructured SSNDs have semiconducting substrate etched with a repeated, three-dimensional microstructure of high aspect ratio holes filled with 10B. MCNP simulations optimized the dimensions of each microstructure geometry for each detector application, improving the overall performance. This thesis outlines the development of multiple, novel neutron detection applications using microstructured SSNDs developed at RPI. The Directional and Spectral Neutron Detection System (DSNDS) is a modular and portable system that uses rings of microstructured SSNDs embedded in polyethylene in order to gather real-time information about the directionality and spectrum of an unidentified neutron source. This system can be used to identify the presence of diverted special nuclear material (SNM), determine its position, and gather spectral information in real-time. The compact and scalable zero-bias SSNDs allow for customization and modularity of the detector array, which provides design flexibility and enhanced portability. Additionally, a real-time personal neutron dosimeter is a wearable device that uses a combination of fast and thermal microstructured SSNDs in order to determine an individual's neutron dose rate. This system demonstrates that neutron detection systems utilizing microstructured SSNDs are applicable for personal neutron dosimetry. The development of these systems using the compact, zero-bias microstructured SSNDs lays the groundwork for a new generation of neutron detection tools, outlines the challenges and design considerations associated with the implementation of these devices, and demonstrates the value that these detectors bring to the future of neutron detection systems.

Safety approach to the selection of design criteria for the CRBRP reactor refueling system

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

Meisl, C J; Berg, G E; Sharkey, N F

1979-01-01

The selection of safety design criteria for Liquid Metal Fast Breeder Reactor (LMFBR) refueling systems required the extrapolation of regulations and guidelines intended for Light Water Reactor refueling systems and was encumbered by the lack of benefit from a commercially licensed predecessor other than Fermi. The overall approach and underlying logic are described for developing safety design criteria for the reactor refueling system (RRS) of the Clinch River Breeder Reactor Plant (CRBRP). The complete selection process used to establish the criteria is presented, from the definition of safety functions to the finalization of safety design criteria in the appropriate documents.more » The process steps are illustrated by examples.« less

Grey water characteristics and treatment options for rural areas in Jordan.

PubMed

Halalsheh, M; Dalahmeh, S; Sayed, M; Suleiman, W; Shareef, M; Mansour, M; Safi, M

2008-09-01

Low water consumption in rural areas in Jordan had resulted in the production of concentrated grey water. Average COD, BOD and TSS values were 2568mg/l, 1056mg/l and 845mg/l, respectively. The average grey water generation was measured to be 14L/c.d. Three different treatment options were selected based on certain criterions, and discussed in this article. The examined treatment systems are septic tank followed by intermittent sand filter; septic tank followed by wetlands; and UASB-hybrid reactor. Advantages and disadvantages of each system are presented. It was concluded that UASB-hybrid reactor would be the most suitable treatment option in terms of compactness and simplicity in operation. The volume of UASB-hybrid reactor was calculated to be 0.268m(3) with a surface area of 0.138m(2) for each house having 10 inhabitants on average. Produced effluent is expected to meet Jordanian standards set for reclaimed water reuse in irrigating fruit trees.

AGR-2 and AGR-3/4 Release-to-Birth Ratio Data Analysis

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

Pham, Binh T.; Einerson, Jeffrey J.; Scates, Dawn M.

A series of Advanced Gas Reactor (AGR) irradiation tests is being conducted in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) in support of development and qualification of tristructural isotropic (TRISO) low enriched fuel used in the High Temperature Gas-cooled Reactor (HTGR). Each AGR test consists of multiple independently controlled and monitored capsules containing fuel compacts placed in a graphite cylinder shrouded by a steel shell. These capsules are instrumented with thermocouples embedded in the graphite enabling temperature control. AGR configuration and irradiation conditions are based on prismatic HTGR technology that is distinguished primarily through use of heliummore » coolant, a low-power-density ceramic core capable of withstanding very high temperatures, and TRISO coated particle fuel. Thus, these tests provide valuable irradiation performance data to support fuel process development, qualify fuel for normal operating conditions, and support development and validation of fuel performance and fission product transport models and codes.« less

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

Furuta, H.; Imura, A.; Furuta, Y.

Recently, technique of Gadolinium loaded liquid scintillator (Gd-LS) for reactor neutrino oscillation experiments has attracted attention as a monitor of reactor operation and 'nuclear Gain (GA)' for IAEA safeguards. For the practical use, R and D of the 1 ton class compact detector, which is measurable above ground, is necessary. Especially, it is important to reduce much amount of fast neutron background induced by cosmic muons with data analysis for the measurement above ground. We developed a prototype of the Gd-LS detector with 200 L of the target volume, which has Pulse Shape Discrimination (PSD) ability for the fast neutronmore » reduction with data analysis. Usually, it is well known that it is difficult to keep high fast neutron reduction power of PSD with the large volume size such as the neutrino reactor monitor. We evaluated the PSD ability of our prototype with real fast neutrons induced by the muons in our laboratory above ground, and we could confirm to keep the high fast neutron reduction power with even our large detector size. (authors)« less

Particulate fuel bed tests

NASA Astrophysics Data System (ADS)

Horn, F. L.; Powell, J. R.; Savino, J. M.

Gas-cooled reactors using packed beds of small-diameter, coated fuel particles have been proposed for compact, high-power systems. To test the thermal-hydraulic performance of the particulate reactor fuel under simulated reactor conditions, a bed of 800-micrometer diameter particles was heated by its electrical resistance current and cooled by flowing helium gas. The specific resistance of the bed composed of pyrocarbon-coated particles was measured at several temperatures, and found to be 0.09 ohm-cm at 1273 K and 0.06 ohm-cm at 1600 K. The maximum bed power density reached was 1500 W/cu cm at 1500 K. The pressure drop followed the packed-bed correlation, typically 100,000 Pa/cm. The various frit materials used to contain the bed were also tested to 2000 K in helium and hydrogen to determine their properties and reactions with the fuel. Rhenium metal, zirconium carbide, and zirconium oxide appeared to be the best candidate materials, while tungsten and tungsten-rhenium lost mass and strength.

Safety Testing of AGR-2 UCO Compacts 5-2-2, 2-2-2, and 5-4-1

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

Hunn, John D.; Morris, Robert Noel; Baldwin, Charles A.

2016-08-01

Post-irradiation examination (PIE) is being performed on tristructural-isotropic (TRISO) coated-particle fuel compacts from the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program second irradiation experiment (AGR-2). This effort builds upon the understanding acquired throughout the AGR-1 PIE campaign, and is establishing a database for the different AGR-2 fuel designs. The AGR-2 irradiation experiment included TRISO fuel particles coated at BWX Technologies (BWXT) with a 150-mm-diameter engineering-scale coater. Two coating batches were tested in the AGR-2 irradiation experiment. Batch 93085 had 508-μm-diameter uranium dioxide (UO 2) kernels. Batch 93073 had 427-μm-diameter UCO kernels, which is a kernel design where somemore » of the uranium oxide is converted to uranium carbide during fabrication to provide a getter for oxygen liberated during fission and limit CO production. Fabrication and property data for the AGR-2 coating batches have been compiled and compared to those for AGR-1. The AGR-2 TRISO coatings were most like the AGR-1 Variant 3 TRISO deposited in the 50-mm-diameter ORNL lab-scale coater. In both cases argon-dilution of the hydrogen and methyltrichlorosilane coating gas mixture employed to deposit the SiC was used to produce a finer-grain, more equiaxed SiC microstructure. In addition to the fact that AGR-1 fuel had smaller, 350-μm-diameter UCO kernels, notable differences in the TRISO particle properties included the pyrocarbon anisotropy, which was slightly higher in the particles coated in the engineering-scale coater, and the exposed kernel defect fraction, which was higher for AGR-2 fuel due to the detected presence of particles with impact damage introduced during TRISO particle handling.« less

SECOND TARGET STATION MODERATOR PERFORMANCE WITH A ROTATING TARGET

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

Remec, Igor; Gallmeier, Franz X; Rennich, Mark J

2016-01-01

Oak Ridge National Laboratory manages and operates the Spallation Neutron Source and the High Flux Isotope Reactor, two of the world's most advanced neutron scattering facilities. Both facilities are funded by the U.S. Department of Energy, Office of Science, Office of Basic Energy Science, and are available to researchers from all over the world. Delivering cutting edge science requires continuous improvements and development of the facilities and instruments. The SNS was designed from the outset to accommodate an additional target station, or Second Target Station (STS), and an upgraded accelerator feeding proton beams to STS and the existing First Targetmore » Station (FTS). Upgrade of the accelerator and the design and construction of STS are being proposed. The presently considered STS configuration is driven with short (<1 s) proton pulses at 10 Hz repetition rate and 467 kW proton beam power, and is optimized for high intensity and high resolution long wavelength neutron applications. STS will allow installation of 22 beamlines and will expand and complement the current national neutron scattering capabilities. In 2015 the STS studies were performed for a compact tungsten target; first a stationary tungsten plate target was analyzed to considerable details and then dropped in favor of a rotating target. For both target options the proton beam footprint as small as acceptable from mechanical and heat removal aspects is required to arrive at a compact-volume neutron production zone in the target, which is essential for tight coupling of target and moderators and for achieving high-intensity peak neutron fluxes. This paper will present recent STS work with the emphasis on neutronics and moderator performance.« less

Does Compact Development Increase or Reduce Traffic Congestion?

DOT National Transportation Integrated Search

2017-10-01

From years of research, we know that compact development that is dense, diverse, well-designed, etc. produces fewer vehicle miles traveled (VMT) than sprawling development. But compact development also concentrates origins and destinations. No one ha...

Determination of the Sensitivity of the Antineutrino Probe for Reactor Core Monitoring

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

Cormon, S.; Fallot, M., E-mail: fallot@subatech.in2p3.fr; Bui, V.-M.

This paper presents a feasibility study of the use of the detection of reactor-antineutrinos (ν{sup ¯}{sub e}) for non proliferation purpose. To proceed, we have started to study different reactor designs with our simulation tools. We use a package called MCNP Utility for Reactor Evolution (MURE), initially developed by CNRS/IN2P3 labs to study Generation IV reactors. The MURE package has been coupled to fission product beta decay nuclear databases for studying reactor antineutrino emission. This method is the only one able to predict the antineutrino emission from future reactor cores, which don't use the thermal fission of {sup 235}U, {supmore » 239}Pu and {sup 241}Pu. It is also the only way to include off-equilibrium effects, due to neutron captures and time evolution of the fission product concentrations during a reactor cycle. We will present here the first predictions of antineutrino energy spectra from innovative reactor designs (Generation IV reactors). We will then discuss a summary of our results of non-proliferation scenarios involving the latter reactor designs, taking into account reactor physics constraints.« less

Demonstration of Robustness and Integrated Operation of a Series-Bosch System

NASA Technical Reports Server (NTRS)

Abney, Morgan B.; Mansell, Matthew J.; Stanley, Christine; Barnett, Bill; Junaedi, Christian; Vilekar, Saurabh A.; Ryan, Kent

2016-01-01

Manned missions beyond low Earth orbit will require highly robust, reliable, and maintainable life support systems that maximize recycling of water and oxygen. Bosch technology is one option to maximize oxygen recovery, in the form of water, from metabolically-produced carbon dioxide (CO2). A two stage approach to Bosch, called Series-Bosch, reduces metabolic CO2 with hydrogen (H2) to produce water and solid carbon using two reactors: a Reverse Water-Gas Shift (RWGS) reactor and a carbon formation (CF) reactor. Previous development efforts demonstrated the stand-alone performance of a NASA-designed RWGS reactor designed for robustness against carbon formation, two membrane separators intended to maximize single pass conversion of reactants, and a batch CF reactor with both transit and surface catalysts. In the past year, Precision Combustion, Inc. (PCI) developed and delivered a RWGS reactor for testing at NASA. The reactor design was based on their patented Microlith® technology and was first evaluated under a Phase I Small Business Innovative Research (SBIR) effort in 2010. The RWGS reactor was recently evaluated at NASA to compare its performance and operating conditions with NASA's RWGS reactor. The test results will be provided in this paper. Separately, in 2015, a semi-continuous CF reactor was designed and fabricated at NASA based on the results from batch CF reactor testing. The batch CF reactor and the semi-continuous CF reactor were individually integrated with an upstream RWGS reactor to demonstrate the system operation and to evaluate performance. Here, we compare the performance and robustness to carbon formation of both RWGS reactors. We report the results of the integrated operation of a Series-Bosch system and we discuss the technology readiness level.

Mixing and Compaction Temperatures for Hot Mix Asphalt Concrete

DOT National Transportation Integrated Search

2000-01-01

According to Superpave mixture design, gyratory specimens are mixed and compacted at equiviscous binder temperatures corresponding to viscosities of 0.17 and 0.28 Pa's, respectively. These were the values previously used in the Marshal mix design met...

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

DOE PAGES

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

2016-12-21

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

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

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

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

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

Design Study of Modular Nuclear Power Plant with Small Long Life Gas Cooled Fast Reactors Utilizing MOX Fuel

NASA Astrophysics Data System (ADS)

Ilham, Muhammad; Su'ud, Zaki

2017-01-01

Growing energy needed due to increasing of the world’s population encourages development of technology and science of nuclear power plant in its safety and security. In this research, it will be explained about design study of modular fast reactor with helium gas cooling (GCFR) small long life reactor, which can be operated over 20 years. It had been conducted about neutronic design GCFR with Mixed Oxide (UO2-PuO2) fuel in range of 100-200 MWth NPPs of power and 50-60% of fuel fraction variation with cylindrical pin cell and cylindrical balance of reactor core geometry. Calculation method used SRAC-CITATION code. The obtained results are the effective multiplication factor and density value of core reactor power (with geometry optimalization) to obtain optimum design core reactor power, whereas the obtained of optimum core reactor power is 200 MWth with 55% of fuel fraction and 9-13% of percentages.

The hybrid reactor project based on the straight field line mirror concept

NASA Astrophysics Data System (ADS)

Ågren, O.; Noack, K.; Moiseenko, V. E.; Hagnestâl, A.; Källne, J.; Anglart, H.

2012-06-01

The straight field line mirror (SFLM) concept is aiming towards a steady-state compact fusion neutron source. Besides the possibility for steady state operation for a year or more, the geometry is chosen to avoid high loads on materials and plasma facing components. A comparatively small fusion hybrid device with "semi-poor" plasma confinement (with a low fusion Q factor) may be developed for industrial transmutation and energy production from spent nuclear fuel. This opportunity arises from a large fission to fusion energy multiplication ratio, Qr = Pfis/Pfus>>1. The upper bound on Qr is primarily determined by geometry and reactor safety. For the SFLM, the upper bound is Qr≈150, corresponding to a neutron multiplicity of keff=0.97. Power production in a mirror hybrid is predicted for a substantially lower electron temperature than the requirement Te≈10 keV for a fusion reactor. Power production in the SFLM seems possible with Q≈0.15, which is 10 times lower than typically anticipated for hybrids (and 100 times smaller than required for a fusion reactor). This relaxes plasma confinement demands, and broadens the range for use of plasmas with supra-thermal ions in hybrid reactors. The SFLM concept is based on a mirror machine stabilized by qudrupolar magnetic fields and large expander tanks beyond the confinement region. The purpose of the expander tanks is to distribute axial plasma loss flow over a sufficiently large area so that the receiving plates can withstand the heat. Plasma stability is not relying on a plasma flow into the expander regions. With a suppressed plasma flow into the expander tanks, a possibility arise for higher electron temperature. A brief presentation will be given on basic theory for the SFLM with plasma stability and electron temperature issues, RF heating computations with sloshing ion formation, neutron transport computations with reactor safety margins and material load estimates, magnetic coil designs as well as a discussion on the implications of the geometry for possible diagnostics. Reactor safety issues are addressed and a vertical orientation of the device could assist passive coolant circulation. Specific attention is put to a device with a 25 m long confinement region and 40 cm plasma radius in the mid-plane. In an optimal case (keff = 0.97) with a fusion power of only 10 MW, such a device may be capable of producing a power of 1.5 GWth.

Design and fabrication of metal briquette machine for shop floor

NASA Astrophysics Data System (ADS)

Pramod, R.; Kumar, G. B. Veeresh; Prashanth B., N.

2017-07-01

Efforts have to be taken to ensure efficient waste management system in shop floors, with minimum utilization of space and energy when it comes to disposing metal chips formed during machining processes. The salvaging of junk metallic chips and the us e of scrap are important for the economic production of a steelworks. For this purpose, we have fabricated a metal chip compaction machine, which can compact the metal chips into small briquettes. The project started with the survey of chips formed in shop floors and the practices involved in waste management. Study was done on the requirements for a better compaction. The heating chamber was designed taking into consideration the temperature required for an easy compaction of the metal chips. The power source for compaction and the pneumatic design for mechanism was done following the appropriate calculations regarding the air pressure provided and thrust required. The processes were tested under different conditions and found effective. The fabrication of the machine has been explained in detail and the results have been discussed.

Preparation of Simulated LBL Defects for Round Robin Experiment

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

Gerczak, Tyler J.; Baldwin, Charles A.; Hunn, John D.

2016-01-01

A critical characteristic of the TRISO fuel design is its ability to retain fission products. During reactor operation, the TRISO layers act as barriers to release of fission products not stabilized in the kernel. Each component of the TRISO particle and compact construction plays a unique role in retaining select fission products, and layer performance is often interrelated. The IPyC, SiC, and OPyC layers are barriers to the release of fission product gases such as Kr and Xe. The SiC layer provides the primary barrier to release of metallic fission products not retained in the kernel, as transport across themore » SiC layer is rate limiting due to the greater permeability of the IPyC and OPyC layers to many metallic fission products. These attributes allow intact TRISO coatings to successfully retain most fission products released from the kernel, with the majority of released fission products during operation being due to defective, damaged, or failed coatings. This dominant release of fission products from compromised particles contributes to the overall source term in reactor; causing safety and maintenance concerns and limiting the lifetime of the fuel. Under these considerations, an understanding of the nature and frequency of compromised particles is an important part of predicting the expected fission product release and ensuring safe and efficient operation.« less

Modeling and Real-Time Process Monitoring of Organometallic Chemical Vapor Deposition of III-V Phosphides and Nitrides at Low and High Pressure

NASA Technical Reports Server (NTRS)

Bachmann, K. J.; Cardelino, B. H.; Moore, C. E.; Cardelino, C. A.; Sukidi, N.; McCall, S.

1999-01-01

The purpose of this paper is to review modeling and real-time monitoring by robust methods of reflectance spectroscopy of organometallic chemical vapor deposition (OMCVD) processes in extreme regimes of pressure. The merits of p-polarized reflectance spectroscopy under the conditions of chemical beam epitaxy (CBE) and of internal transmission spectroscopy and principal angle spectroscopy at high pressure are assessed. In order to extend OMCVD to materials that exhibit large thermal decomposition pressure at their optimum growth temperature we have designed and built a differentially-pressure-controlled (DCP) OMCVD reactor for use at pressures greater than or equal to 6 atm. We also describe a compact hard-shell (CHS) reactor for extending the pressure range to 100 atm. At such very high pressure the decomposition of source vapors occurs in the vapor phase, and is coupled to flow dynamics and transport. Rate constants for homogeneous gas phase reactions can be predicted based on a combination of first principles and semi-empirical calculations. The pressure dependence of unimolecular rate constants is described by RRKM theory, but requires variational and anharmonicity corrections not included in presently available calculations with the exception of ammonia decomposition. Commercial codes that include chemical reactions and transport exist, but do not adequately cover at present the kinetics of heteroepitaxial crystal growth.

Development of Nanomaterials for Nuclear Energetics

NASA Astrophysics Data System (ADS)

Petrunin, V. F.

Structure and properties peculiarities of the nanocrystalline powders give the opportunity to design new and to develop a modernization of nuclear energy industry materials. It was shown experimentally, that addition of 5-10% uranium dioxide nanocrystalline powder to traditional coarse powder allows to decrease the sintering temperature or to increase the fuel tablets size of grain. Similar perspectives for the technology of neutron absorbing tablets of control-rod modernization are shown by nanopowder of dysprosium hafnate changing instead now using boron carbide. It is powders in nanocrystalline state get an opportunity to sinter them and to receive compact tablet with 8,2-8,4 g/cm2 density for automatic defence system of nuclear reactor. Resource of dysprosium hafnate ceramics can be 18-20 years instead 4-5 years for boron carbide. To step up the radiation-damage stability of fuel element jacket material was suggested to strengthen a heat-resistant ferrite-martensite steel by Y2O3 nanocrystalline powder addition. Nanopowder with size of particles 560 nm and crystallite size 9 nm was prepeared by chemical coprecipitation method. To make lighter the container for transport and provisional disposal of exposed fuel from nuclear reactor a new boron-aluminium alloy called as boral was developed. This composite armed with nanopowders of boron-containing materials and heavy metals oxides can replace succesburnt-up corrosion-resistant steels.

Heated-Pressure-Ball Monopropellant Rocket Engine

NASA Technical Reports Server (NTRS)

Greene, William D.

2005-01-01

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

Calculated criticality for sup 235 U/graphite systems using the VIM Monte Carlo code

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

Collins, P.J.; Grasseschi, G.L.; Olsen, D.N.

1992-01-01

Calculations for highly enriched uranium and graphite systems gained renewed interest recently for the new production modular high-temperature gas-cooled reactor (MHTGR). Experiments to validate the physics calculations for these systems are being prepared for the Transient Reactor Test Facility (TREAT) reactor at Argonne National Laboratory (ANL-West) and in the Compact Nuclear Power Source facility at Los Alamos National Laboratory. The continuous-energy Monte Carlo code VIM, or equivalently the MCNP code, can utilize fully detailed models of the MHTGR and serve as benchmarks for the approximate multigroup methods necessary in full reactor calculations. Validation of these codes and their associated nuclearmore » data did not exist for highly enriched {sup 235}U/graphite systems. Experimental data, used in development of more approximate methods, dates back to the 1960s. The authors have selected two independent sets of experiments for calculation with the VIM code. The carbon-to-uranium (C/U) ratios encompass the range of 2,000, representative of the new production MHTGR, to the ratio of 10,000 in the fuel of TREAT. Calculations used the ENDF/B-V data.« less

MACHINING TEST SPECIMENS FROM HARVESTED ZION RPV SEGMENTS

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

Nanstad, Randy K; Rosseel, Thomas M; Sokolov, Mikhail A

2015-01-01

The decommissioning of the Zion Nuclear Generating Station (NGS) in Zion, Illinois, presents a special and timely opportunity for developing a better understanding of materials degradation and other issues associated with extending the lifetime of existing nuclear power plants (NPPs) beyond 60 years of service. In support of extended service and current operations of the US nuclear reactor fleet, the Oak Ridge National Laboratory (ORNL), through the Department of Energy (DOE), Light Water Reactor Sustainability (LWRS) Program, is coordinating and contracting with Zion Solutions, LLC, a subsidiary of Energy Solutions, an international nuclear services company, the selective procurement of materials,more » structures, components, and other items of interest from the decommissioned reactors. In this paper, we will discuss the acquisition of segments of the Zion Unit 2 Reactor Pressure Vessel (RPV), cutting these segments into blocks from the beltline and upper vertical welds and plate material and machining those blocks into mechanical (Charpy, compact tension, and tensile) test specimens and coupons for microstructural (TEM, SEM, APT, SANS and nano indention) characterization. Access to service-irradiated RPV welds and plate sections will allow through wall attenuation studies to be performed, which will be used to assess current radiation damage models [1].« less

Off-design performance of a chemical looping combustion (CLC) combined cycle: effects of ambient temperature

NASA Astrophysics Data System (ADS)

Chi, Jinling; Wang, Bo; Zhang, Shijie; Xiao, Yunhan

2010-02-01

The present work investigates the influence of ambient temperature on the steady-state off-design thermodynamic performance of a chemical looping combustion (CLC) combined cycle. A sensitivity analysis of the CLC reactor system was conducted, which shows that the parameters that influence the temperatures of the CLC reactors most are the flow rate and temperature of air entering the air reactor. For the ambient temperature variation, three off-design control strategies have been assumed and compared: 1) without any Inlet Guide Vane (IGV) control, 2) IGV control to maintain air reactor temperature and 3) IGV control to maintain constant fuel reactor temperature, aside from fuel flow rate adjusting. Results indicate that, compared with the conventional combined cycle, due to the requirement of pressure balance at outlet of the two CLC reactors, CLC combined cycle shows completely different off-design thermodynamic characteristics regardless of the control strategy adopted. For the first control strategy, temperatures of the two CLC reactors both rise obviously as ambient temperature increases. IGV control adopted by the second and the third strategy has the effect to maintain one of the two reactors' temperatures at design condition when ambient temperature is above design point. Compare with the second strategy, the third would induce more severe decrease of efficiency and output power of the CLC combined cycle.

The design and fabrication of a prototype trash compacting unit. [for long duration space missions

NASA Technical Reports Server (NTRS)

1973-01-01

A prototype trash compactor, that is compatible with the anticipated requirements of future long-term space missions, is described. Preliminary problem definition studies were conducted to identify typical types and quantities of waste materials to be expected from a typical mission. Bench-scale compaction tests were then conducted on typical waste materials to determine force/compaction curves. These data were used to design a boilerplate compactor that was fabricated to prove the feasibility of the basic design concept. A final design was then prepared from which the deliverable unit was fabricated. Design concepts are presented for suggested further development of the compactor, including a version that is capable of handling wet biodegradable wastes.

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

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

The fluoride salt-cooled high-temperature reactor (FHR) demonstration reactor (DR) is a concept for a salt-cooled reactor with 100 megawatts of thermal output (MWt). It would use tristructural-isotropic (TRISO) particle fuel within prismatic graphite blocks. FLiBe (2 LiF-BeF 2) is the reference primary coolant. The FHR DR is designed to be small, simple, and affordable. Development of the FHR DR is a necessary intermediate step to enable near-term commercial FHRs. Lower risk technologies are purposely included in the initial FHR DR design to ensure that the reactor can be built, licensed, and operated within an acceptable budget and schedule. These technologiesmore » include TRISO particle fuel, replaceable core structural material, the use of that same material for the primary and intermediate loops, and tube-and-shell primary-to-intermediate heat exchangers. Several preconceptual and conceptual design efforts that have been conducted on FHR concepts bear a significant influence on the FHR DR design. Specific designs include the Oak Ridge National Laboratory (ORNL) advanced high-temperature reactor (AHTR) with 3400/1500 MWt/megawatts of electric output (MWe), as well as a 125 MWt small modular AHTR (SmAHTR) from ORNL. Other important examples are the Mk1 pebble bed FHR (PB-FHR) concept from the University of California, Berkeley (UCB), and an FHR test reactor design developed at the Massachusetts Institute of Technology (MIT). The MIT FHR test reactor is based on a prismatic fuel platform and is directly relevant to the present FHR DR design effort. These FHR concepts are based on reasonable assumptions for credible commercial prototypes. The FHR DR concept also directly benefits from the operating experience of the Molten Salt Reactor Experiment (MSRE), as well as the detailed design efforts for a large molten salt reactor concept and its breeder variant, the Molten Salt Breeder Reactor. The FHR DR technology is most representative of the 3400 MWt AHTR concept, and it will demonstrate key operational features of that design. The FHR DR will be closely scaled to the SmAHTR concept in power and flows, so any technologies demonstrated will be directly applicable to a reactor concept of that size. The FHR DR is not a commercial prototype design, but rather a DR that serves a cost and risk mitigation function for a later commercial prototype. It is expected to have a limited operational lifetime compared to a commercial plant. It is designed to be a low-cost reactor compared to more mature advanced prototype DRs. A primary reason to build the FHR DR is to learn about salt reactor technologies and demonstrate solutions to remaining technical gaps.« less

Roller Compacted Concrete over Soil Cement under Accelerated Loading : Research Project Capsule

DOT National Transportation Integrated Search

2012-09-01

Roller compacted concrete (RCC) is a stiff , zero-slump concrete mixture placed with modifi ed : asphalt paving equipment and compacted by vibratory rollers. Properly designed RCC mixes : can achieve outstanding compressive strengths similar to those...

Lessons Learned about Liquid Metal Reactors from FFTF Experience

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

Wootan, David W.; Casella, Andrew M.; Omberg, Ronald P.

2016-09-20

The Fast Flux Test Facility (FFTF) is the most recent liquid-metal reactor (LMR) to operate in the United States, from 1982 to 1992. FFTF is located on the DOE Hanford Site near Richland, Washington. The 400-MWt sodium-cooled, low-pressure, high-temperature, fast-neutron flux, nuclear fission test reactor was designed specifically to irradiate Liquid Metal Fast Breeder Reactor (LMFBR) fuel and components in prototypical temperature and flux conditions. FFTF played a key role in LMFBR development and testing activities. The reactor provided extensive capability for in-core irradiation testing, including eight core positions that could be used with independent instrumentation for the test specimens.more » In addition to irradiation testing capabilities, FFTF provided long-term testing and evaluation of plant components and systems for LMFBRs. The FFTF was highly successful and demonstrated outstanding performance during its nearly 10 years of operation. The technology employed in designing and constructing this reactor, as well as information obtained from tests conducted during its operation, can significantly influence the development of new advanced reactor designs in the areas of plant system and component design, component fabrication, fuel design and performance, prototype testing, site construction, and reactor operations. The FFTF complex included the reactor, as well as equipment and structures for heat removal, containment, core component handling and examination, instrumentation and control, and for supplying utilities and other essential services. The FFTF Plant was designed using a “system” concept. All drawings, specifications and other engineering documentation were organized by these systems. Efforts have been made to preserve important lessons learned during the nearly 10 years of reactor operation. A brief summary of Lessons Learned in the following areas will be discussed: Acceptance and Startup Testing of FFTF FFTF Cycle Reports« less

High-Temperature Gas-Cooled Test Reactor Point Design

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

Sterbentz, James William; Bayless, Paul David; Nelson, Lee Orville

2016-04-01

A point design has been developed for a 200 MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched UCO fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technological readiness level, licensing approach and costs.

Initial examination of fuel compacts and TRISO particles from the US AGR-2 irradiation test

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

Hunn, John D.; Baldwin, Charles A.; Montgomery, Fred C.

Post-irradiation examination was completed on two as-irradiated compacts from the US Advanced Gas Reactor Fuel Development and Qualification Program’s second irradiation test. These compacts were selected for examination because there were indications that they may have contained particles that released cesium through a failed or defective SiC layer. The coated particles were recovered from these compacts by electrolytic deconsolidation of the surrounding graphitic matrix in nitric acid. The leach-burn-leach (LBL) process was used to dissolve and analyze exposed metallic elements (actinides and fission products), and each particle was individually surveyed for relative cesium retention with the Irradiated Microsphere Gamma Analyzermore » (IMGA). Data from IMGA and LBL examinations provided information on fission product release during irradiation and whether any specific particles had below-average retention that could be related to coating layer defects or radiation-induced degradation. A few selected normal-retention particles and six with abnormally-low cesium inventory were analyzed using X-ray tomography to produce three-dimensional images of the internal coating structure. Four of the low-cesium particles had obviously damaged or degraded SiC, and X-ray imaging was able to guide subsequent grinding and polishing to expose the regions of interest for analysis by optical and electron microscopy. Additional particles from each compact were also sectioned and examined to study the overall radiation-induced microstructural changes in the kernel and coating layers.« less

Initial examination of fuel compacts and TRISO particles from the US AGR-2 irradiation test

DOE PAGES

Hunn, John D.; Baldwin, Charles A.; Montgomery, Fred C.; ...

2017-10-21

Post-irradiation examination was completed on two as-irradiated compacts from the US Advanced Gas Reactor Fuel Development and Qualification Program’s second irradiation test. These compacts were selected for examination because there were indications that they may have contained particles that released cesium through a failed or defective SiC layer. The coated particles were recovered from these compacts by electrolytic deconsolidation of the surrounding graphitic matrix in nitric acid. The leach-burn-leach (LBL) process was used to dissolve and analyze exposed metallic elements (actinides and fission products), and each particle was individually surveyed for relative cesium retention with the Irradiated Microsphere Gamma Analyzermore » (IMGA). Data from IMGA and LBL examinations provided information on fission product release during irradiation and whether any specific particles had below-average retention that could be related to coating layer defects or radiation-induced degradation. A few selected normal-retention particles and six with abnormally-low cesium inventory were analyzed using X-ray tomography to produce three-dimensional images of the internal coating structure. Four of the low-cesium particles had obviously damaged or degraded SiC, and X-ray imaging was able to guide subsequent grinding and polishing to expose the regions of interest for analysis by optical and electron microscopy. Additional particles from each compact were also sectioned and examined to study the overall radiation-induced microstructural changes in the kernel and coating layers.« less

A simple, space constrained NIRIM type reactor for chemical vapour deposition of diamond

NASA Astrophysics Data System (ADS)

Thomas, Evan L. H.; Ginés, Laia; Mandal, Soumen; Klemencic, Georgina M.; Williams, Oliver A.

2018-03-01

In this paper the design of a simple, space constrained chemical vapour deposition reactor for diamond growth is detailed. Based on the design by NIRIM, the reactor is composed of a quartz discharge tube placed within a 2.45 GHz waveguide to create the conditions required for metastable growth of diamond. Utilising largely off-the-shelf components and a modular design, the reactor allows for easy modification, repair, and cleaning between growth runs. The elements of the reactor design are laid out with the CAD files, parts list, and control files made easily available to enable replication. Finally, the quality of nanocrystalline diamond films produced are studied with SEM and Raman spectroscopy, with the observation of clear faceting and a large diamond fraction suggesting the design offers deposition of diamond with minimal complexity.

Reactor application of an improved bundle divertor

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

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

1978-11-01

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

Advanced reactor design study. Assessing nonbackfittable concepts for improving uranium utilization in light water reactors

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

Fleischman, R.M.; Goldsmith, S.; Newman, D.F.

1981-09-01

The objective of the Advanced Reactor Design Study (ARDS) is to identify and evaluate nonbackfittable concepts for improving uranium utilization in light water reactors (LWRs). The results of this study provide a basis for selecting and demonstrating specific nonbackfittable concepts that have good potential for implementation. Lead responsibility for managing the study was assigned to the Pacific Northwest Laboratory (PNL). Nonbackfittable concepts for improving uranium utilization in LWRs on the once-through fuel cycle were selected separately for PWRs and BWRs due to basic differences in the way specific concepts apply to those plants. Nonbackfittable concepts are those that are toomore » costly to incorporate in existing plants, and thus, could only be economically incorporated in new reactor designs or plants in very early stages of construction. Essential results of the Advanced Reactor Design Study are summarized.« less

Degradation of the Herbicide (2,4-Dichlorophenoxyacetic Acid) Using a Photoreactor with Exciplex Lamps

NASA Astrophysics Data System (ADS)

Vershinin, N. O.; Sokolova, I. V.; Tchaikovskaya, O. N.

2013-09-01

We present the results of tests of a compact flow-through reactor for neutralization of a broad class of persistent toxic compounds. As the toxicant we used the herbicide 2,4-dichlorophenoxyacetic acid, and we used exciplex lamps with different emission wave lengths (λ ~ 222 nm and 172 nm). We show the experimental decrease in the amount of organic compounds vs. irradiation time as obtained from the absorption spectra.

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

Plasma Source Development

NASA Astrophysics Data System (ADS)

Walker, Jonathan; Heinrich, Jonathon; Font, Gabriel; Ebersohn, Frans; Garrett, Michael

2017-10-01

A 100 kW class lanthanum-hexaboride plasma source is under continuing development for the Lockheed Martin Compact Fusion Reactor program. The current experiment, T4B, has become a test bed for plasma source operation with the goal of creating a high density plasma target for neutral beam heating. We present operation and performance of different plasma source geometries, results of plasma source coupling, and future plasma source development plans. ©2017 Lockheed Martin Corporation. All Rights Reserved.

Compact hohlraum configuration with parallel planar-wire-array x-ray sources at the 1.7-MA Zebra generator.

PubMed

Kantsyrev, V L; Chuvatin, A S; Rudakov, L I; Velikovich, A L; Shrestha, I K; Esaulov, A A; Safronova, A S; Shlyaptseva, V V; Osborne, G C; Astanovitsky, A L; Weller, M E; Stafford, A; Schultz, K A; Cooper, M C; Cuneo, M E; Jones, B; Vesey, R A

2014-12-01

A compact Z-pinch x-ray hohlraum design with parallel-driven x-ray sources is experimentally demonstrated in a configuration with a central target and tailored shine shields at a 1.7-MA Zebra generator. Driving in parallel two magnetically decoupled compact double-planar-wire Z pinches has demonstrated the generation of synchronized x-ray bursts that correlated well in time with x-ray emission from a central reemission target. Good agreement between simulated and measured hohlraum radiation temperature of the central target is shown. The advantages of compact hohlraum design applications for multi-MA facilities are discussed.

A Compact Multilayer Diplexer in LTCC Substrate Using LPF with Multiple Attenuation Poles and Wideband BPF

NASA Astrophysics Data System (ADS)

Oshima, Shinpei; Wada, Kouji; Murata, Ryuji; Shimakata, Yukihiro

Recently, compact wideband BPFs for UWB system are studied actively. In this paper we propose a compact diplexer in LTCC substrate for UWB system and 2.4GHz wireless systems. Firstly, a wideband BPF for UWB system and an LPF with multiple attenuation poles for 2.4GHz wireless systems are described. Secondly, we design matching circuits of a common port to keep basic performance of both the BPF and the LPF. Thirdly, in accordance with the result of the study, we design a compact diplexer in LTCC substrate. Finally, we verify the effectiveness of proposed method by experiments.

Nitrate removal from high strength nitrate-bearing wastes in granular sludge sequencing batch reactors.

PubMed

Krishna Mohan, Tulasi Venkata; Renu, Kadali; Nancharaiah, Yarlagadda Venkata; Satya Sai, Pedapati Murali; Venugopalan, Vayalam Purath

2016-02-01

A 6-L sequencing batch reactor (SBR) was operated for development of granular sludge capable of denitrification of high strength nitrates. Complete and stable denitrification of up to 5420 mg L(-1) nitrate-N (2710 mg L(-1) nitrate-N in reactor) was achieved by feeding simulated nitrate waste at a C/N ratio of 3. Compact and dense denitrifying granular sludge with relatively stable microbial community was developed during reactor operation. Accumulation of large amounts of nitrite due to incomplete denitrification occurred when the SBR was fed with 5420 mg L(-1) NO3-N at a C/N ratio of 2. Complete denitrification could not be achieved at this C/N ratio, even after one week of reactor operation as the nitrite levels continued to accumulate. In order to improve denitrification performance, the reactor was fed with nitrate concentrations of 1354 mg L(-1), while keeping C/N ratio at 2. Subsequently, nitrate concentration in the feed was increased in a step-wise manner to establish complete denitrification of 5420 mg L(-1) NO3-N at a C/N ratio of 2. The results show that substrate concentration plays an important role in denitrification of high strength nitrate by influencing nitrite accumulation. Complete denitrification of high strength nitrates can be achieved at lower substrate concentrations, by an appropriate acclimatization strategy. Copyright © 2015 The Society for Biotechnology, Japan. Published by Elsevier B.V. All rights reserved.

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 .

Innovative conception and performance evaluation of a compact on-site treatment system.

PubMed

Sousa, V P; Chernicharo, C A L

2006-01-01

The purpose of this study was to develop a new configuration for a compact on-site treatment system, which could become an attractive alternative, from technical, economic, social and environmental viewpoints, to the technologies that are currently employed. The treatment unit consists of a cylindrical tank, where half of the volume is used as a modified septic tank and the other half is divided between an anaerobic hybrid reactor and a trickling filter. An intermittent feeding system was used, with minimum, mean and maximum flowrate settings (Qmin = 0.25l.s(-1), Qmean = 0.50l.s(-1) and Qmax = 1.00l.s(-1)), to reflect the actual operating conditions of a compact on-site treatment system serving a typical dwelling. An average 24-hour hydraulic detention time was used, corresponding to a flowrate of 750l.d(-1). High removal efficiencies and low concentrations of COD, BOD and TSS in the final effluent were achieved, even when the unit was exposed to hydraulic loading peaks during feeding periods at maximum flowrate.

Compact NE213 neutron spectrometer with high energy resolution for fusion applications

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

Zimbal, A.; Reginatto, M.; Schuhmacher, H.

Neutron spectrometry is a tool for obtaining important information on the fuel ion composition, velocity distribution and temperature of fusion plasmas. A compact NE213 liquid scintillator, fully characterized at Physikalisch-Technische Bundesanstalt, was installed and operated at the Joint European Torus (JET) during two experimental campaigns (C8-2002 and trace tritium experiment-TTE 2003). The results show that this system can operate in a real fusion experiment as a neutron (1.5 MeV

Thorium fueled reactor

NASA Astrophysics Data System (ADS)

Sipaun, S.

2017-01-01

Current development in thorium fueled reactors shows that they can be designed to operate in the fast or thermal spectrum. The thorium/uranium fuel cycle converts fertile thorium-232 into fissile uranium-233, which fissions and releases energy. This paper analyses the characteristics of thorium fueled reactors and discusses the thermal reactor option. It is found that thorium fuel can be utilized in molten salt reactors through many configurations and designs. A balanced assessment on the feasibility of adopting one reactor technology versus another could lead to optimized benefits of having thorium resource.

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

DTIC Science & Technology

1980-08-01

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

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

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

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

Thermomechanical analysis of fast-burst reactors

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

Miller, J.D.

1994-08-01

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

A spatially resolved retarding field energy analyzer design suitable for uniformity analysis across the surface of a semiconductor wafer

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

Sharma, S., E-mail: shailesh.sharma6@mail.dcu.ie; National Centre for Plasma Science and Technology, Dublin City University, Glasnevin, Dublin 9; Gahan, D., E-mail: david.gahan@impedans.com

2014-04-15

A novel retarding field energy analyzer design capable of measuring the spatial uniformity of the ion energy and ion flux across the surface of a semiconductor wafer is presented. The design consists of 13 individual, compact-sized, analyzers, all of which are multiplexed and controlled by a single acquisition unit. The analyzers were tested to have less than 2% variability from unit to unit due to tight manufacturing tolerances. The main sensor assembly consists of a 300 mm disk to mimic a semiconductor wafer and the plasma sampling orifices of each sensor are flush with disk surface. This device is placedmore » directly on top of the rf biased electrode, at the wafer location, in an industrial capacitively coupled plasma reactor without the need for any modification to the electrode structure. The ion energy distribution, average ion energy, and average ion flux were measured at the 13 locations over the surface of the powered electrode to determine the degree of spatial nonuniformity. The ion energy and ion flux are shown to vary by approximately 20% and 5%, respectively, across the surface of the electrode for the range of conditions investigated in this study.« less

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

Paul Demkowicz; Lance Cole; Scott Ploger

The AGR-1 irradiation experiment ended on November 6, 2009, after 620 effective full power days in the Advanced Test Reactor, achieving a peak burnup of 19.6% FIMA. The test train was shipped to the Materials and Fuels Complex in March 2010 for post-irradiation examination. The first PIE activities included non-destructive examination of the test train, followed by disassembly of the test train and individual capsules and detailed inspection of the capsule contents, including the fuel compacts and the graphite fuel holders. Dimensional measurements of the compacts, graphite holders, and steel capsules shells were performed using a custom vision measurement systemmore » (for outer diameters and lengths) and conventional bore gauges (for inner diameters). Gamma spectrometry of the intact test train gave a preliminary look at the condition of the interior components. No evidence of damage to compacts or graphite components was evident from the isotopic and gross gamma scans. Neutron radiography of the intact Capsule 2 showed a high degree of detail of interior components and confirmed the observation that there was no major damage to the capsule. Disassembly of the capsules was initiated using procedures qualified during out-of-cell mockup testing. Difficulties were encountered during capsule disassembly due to irradiation-induced changes in some of the capsule components’ properties, including embrittled niobium and molybdenum parts that were susceptible to fracture and swelling of the graphite fuel holders that affected their removal from the capsule shells. This required various improvised modifications to the disassembly procedure to avoid damage to the fuel compacts. Ultimately the capsule disassembly was successful and only one compact from Capsule 4 (out of 72 total in the test train) sustained damage during the disassembly process, along with the associated graphite holder. The compacts were generally in very good condition upon removal. Only relatively minor damage or markings were visible using high resolution photographic inspection. Compact dimensional measurements indicated diametrical shrinkage of 0.9 to 1. 4%, and length shrinkage of 0.2 to 1.1%. The shrinkage was somewhat dependent on compact location within each capsule and within the test train. Compacts exhibited a maximum diametrical shrinkage at a fast neutron fluence of approximately 3×1021 n/cm2. A multivariate statistical analysis indicates that fast neutron fluence as well as compact position in the test train influence compact shrinkage.« less

Digital computer study of nuclear reactor thermal transients during startup of 60-kWe Brayton power conversion system

NASA Technical Reports Server (NTRS)

Jefferies, K. S.; Tew, R. C.

1974-01-01

A digital computer study was made of reactor thermal transients during startup of the Brayton power conversion loop of a 60-kWe reactor Brayton power system. A startup procedure requiring the least Brayton system complication was tried first; this procedure caused violations of design limits on key reactor variables. Several modifications of this procedure were then found which caused no design limit violations. These modifications involved: (1) using a slower rate of increase in gas flow; (2) increasing the initial reactor power level to make the reactor respond faster; and (3) appropriate reactor control drum manipulation during the startup transient.

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.

Reactor Operations Monitoring System

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

Hart, M.M.

1989-01-01

The Reactor Operations Monitoring System (ROMS) is a VME based, parallel processor data acquisition and safety action system designed by the Equipment Engineering Section and Reactor Engineering Department of the Savannah River Site. The ROMS will be analyzing over 8 million signal samples per minute. Sixty-eight microprocessors are used in the ROMS in order to achieve a real-time data analysis. The ROMS is composed of multiple computer subsystems. Four redundant computer subsystems monitor 600 temperatures with 2400 thermocouples. Two computer subsystems share the monitoring of 600 reactor coolant flows. Additional computer subsystems are dedicated to monitoring 400 signals from assortedmore » process sensors. Data from these computer subsystems are transferred to two redundant process display computer subsystems which present process information to reactor operators and to reactor control computers. The ROMS is also designed to carry out safety functions based on its analysis of process data. The safety functions include initiating a reactor scram (shutdown), the injection of neutron poison, and the loadshed of selected equipment. A complete development Reactor Operations Monitoring System has been built. It is located in the Program Development Center at the Savannah River Site and is currently being used by the Reactor Engineering Department in software development. The Equipment Engineering Section is designing and fabricating the process interface hardware. Upon proof of hardware and design concept, orders will be placed for the final five systems located in the three reactor areas, the reactor training simulator, and the hardware maintenance center.« less

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

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

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

2017-03-01

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

Reactor Design and Decommissioning - An Overview of International Activities in Post Fukushima Era1 - 12396

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

Devgun, Jas S.; Laraia, Michele; Pescatore, Claudio

Accidents at the Fukushima Dai-ichi reactors as a result of the devastating earthquake and tsunami of March 11, 2011 have not only dampened the nuclear renaissance but have also initiated a re-examination of the design and safety features for the existing and planned nuclear reactors. Even though failures of some of the key site features at Fukushima can be attributed to events that in the past would have been considered as beyond the design basis, the industry as well as the regulatory authorities are analyzing what features, especially passive features, should be designed into the new reactor designs to minimizemore » the potential for catastrophic failures. It is also recognized that since the design of the Fukushima BWR reactors which were commissioned in 1971, many advanced safety features are now a part of the newer reactor designs. As the recovery efforts at the Fukushima site are still underway, decisions with respect to the dismantlement and decommissioning of the damaged reactors and structures have not yet been finalized. As it was with Three Mile Island, it could take several decades for dismantlement, decommissioning and clean up, and the project poses especially tough challenges. Near-term assessments have been issued by several organizations, including the IAEA, the USNRC and others. Results of such investigations will lead to additional improvements in system and site design measures including strengthening of the anti-tsunami defenses, more defense-in-depth features in reactor design, and better response planning and preparation involving reactor sites. The question also arises what would the effect be on the decommissioning scene worldwide, and what would the effect be on the new reactors when they are eventually retired and dismantled. This paper provides an overview of the US and international activities related to recovery and decommissioning including the decommissioning features in the reactor design process and examines these from a new perspective in the post Fukushima -accident era. Accidents at the Fukushima Daiichi reactors in the aftermath of the devastating earthquake and tsunami of March 11, 2011 have slowed down the nuclear renaissance world-wide and may have accelerated decommissioning either because some countries have decided to halt or reduce nuclear, or because the new safety requirements may reduce life-time extensions. Even in countries such as the UK and France that favor nuclear energy production existing nuclear sites are more likely to be chosen as sites for future NPPs. Even as the site recovery efforts continue at Fukushima and any decommissioning decisions are farther into the future, the accidents have focused attention on the reactor designs in general and specifically on the Fukushima type BWRs. The regulatory authorities in many countries have initiated a re-examination of the design of the systems, structures and components and considerations of the capability of the station to cope with beyond-design basis events. Enhancements to SSCs and site features for the existing reactors and the reactors that will be built will also impact the decommissioning phase activities. The newer reactor designs of today not only have enhanced safety features but also take into consideration the features that will facilitate future decommissioning. Lessons learned from past management and operation of reactors as well as the lessons from decommissioning are incorporated into the new designs. However, in the post-Fukushima era, the emphasis on beyond-design-basis capability may lead to significant changes in SSCs, which eventually will also have impact on the decommissioning phase. Additionally, where some countries decide to phase out the nuclear power, many reactors may enter the decommissioning phase in the coming decade. While the formal updating and expanding of existing guidance documents for accident cleanup and decommissioning would benefit by waiting until the Fukushima project has progressed sufficiently for that experience to be reliably interpreted, the development of structured on-line sharing of information and especially the creation of an on-line compendium of methods, tools, and techniques by which damaged fuel and other unique situations have been addressed can be addressed sooner and maintained as new problems and solutions arise and are resolved. The IAEA's new 'WEB 2.0 tool' CONNECT is expected to play a significant role in this and related information-sharing activities. The trend in some countries such as the United States has been to re-license the existing reactors for additional twenty years, beyond the original design life. Given the advances in technology over the past four decades, and considering that the newer designs incorporate significant improvements in safety systems, it may not be economical or technically feasible to retrofit enhancements into some of the older reactors. In such cases, the reactors may be retired from service and decommissioned. Overall, the energy demand in the world continues to rise, with sharp increases in the Asian countries, and nuclear power's role in the world's energy supply is expected to continue. Events at Fukushima have led to a re-examination on many fronts, including reactor design and regulatory requirements. Further changes may occur in these areas in the post-Fukushima era. These changes in turn will also impact the world-wide decommissioning scene and the decommissioning phase of the future reactors. (authors)« less

Summary of NR Program Prometheus Efforts

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

J Ashcroft; C Eshelman

2006-02-08

The Naval Reactors Program led work on the development of a reactor plant system for the Prometheus space reactor program. The work centered on a 200 kWe electric reactor plant with a 15-20 year mission applicable to nuclear electric propulsion (NEP). After a review of all reactor and energy conversion alternatives, a direct gas Brayton reactor plant was selected for further development. The work performed subsequent to this selection included preliminary nuclear reactor and reactor plant design, development of instrumentation and control techniques, modeling reactor plant operational features, development and testing of core and plant material options, and development ofmore » an overall project plan. Prior to restructuring of the program, substantial progress had been made on defining reference plant operating conditions, defining reactor mechanical, thermal and nuclear performance, understanding the capabilities and uncertainties provided by material alternatives, and planning non-nuclear and nuclear system testing. The mission requirements for the envisioned NEP missions cannot be accommodated with existing reactor technologies. Therefore concurrent design, development and testing would be needed to deliver a functional reactor system. Fuel and material performance beyond the current state of the art is needed. There is very little national infrastructure available for fast reactor nuclear testing and associated materials development and testing. Surface mission requirements may be different enough to warrant different reactor design approaches and development of a generic multi-purpose reactor requires substantial sacrifice in performance capability for each mission.« less

Generation III reactors safety requirements and the design solutions

NASA Astrophysics Data System (ADS)

Felten, P.

2009-03-01

Nuclear energy's public acceptance, and hence its development, depends on its safety. As a reactor designer, we will first briefly remind the basic safety principles of nuclear reactors' design. We will then show how the industry, and in particular Areva with its EPR, made design evolution in the wake of the Three Miles Island accident in 1979. In particular, for this new generation of reactors, severe accidents are taken into account beyond the standard design basis accidents. Today, Areva's EPR meets all so-called "generation III" safety requirements and was licensed by several nuclear safety authorities in the world. Many innovative solutions are integrated in the EPR, some of which will be introduced here.

Small low mass advanced PBR's for propulsion

NASA Astrophysics Data System (ADS)

Powell, J. R.; Todosow, M.; Ludewig, H.

1993-10-01

The advanced Particle Bed Reactor (PBR) to be described in this paper is characterized by relatively low power, and low cost, while still maintaining competition values for thrust/weight, specific impulse and operating times. In order to retain competitive values for the thrust/weight ratio while reducing the reactor size, it is necessary to change the basic reactor layout, by incorporating new concepts. The new reactor design concept is termed SIRIUS (Small Lightweight Reactor Integral Propulsion System). The following modifications are proposed for the reactor design to be discussed in this paper: Pre-heater (U-235 included in Moderator); Hy-C (Hydride/De-hydride for Reactor Control); Afterburner (U-235 impregnated into Hot Frit); and Hy-S (Hydride Spike Inside Hot Frit). Each of the modifications will be briefly discussed below, with benefits, technical issues, design approach, and risk levels addressed. The paper discusses conceptual assumptions, feasibility analysis, mass estimates, and information needs.

Induction Heating Model of Cermet Fuel Element Environmental Test (CFEET)

NASA Technical Reports Server (NTRS)

Gomez, Carlos F.; Bradley, D. E.; Cavender, D. P.; Mireles, O. R.; Hickman, R. R.; Trent, D.; Stewart, E.

2013-01-01

Deep space missions with large payloads require high specific impulse and relatively high thrust to achieve mission goals in reasonable time frames. Nuclear Thermal Rockets (NTR) are capable of producing a high specific impulse by employing heat produced by a fission reactor to heat and therefore accelerate hydrogen through a rocket nozzle providing thrust. Fuel element temperatures are very high (up to 3000 K) and hydrogen is highly reactive with most materials at high temperatures. Data covering the effects of high-temperature hydrogen exposure on fuel elements are limited. The primary concern is the mechanical failure of fuel elements due to large thermal gradients; therefore, high-melting-point ceramics-metallic matrix composites (cermets) are one of the fuels under consideration as part of the Nuclear Cryogenic Propulsion Stage (NCPS) Advance Exploration System (AES) technology project at the Marshall Space Flight Center. The purpose of testing and analytical modeling is to determine their ability to survive and maintain thermal performance in a prototypical NTR reactor environment of exposure to hydrogen at very high temperatures and obtain data to assess the properties of the non-nuclear support materials. The fission process and the resulting heating performance are well known and do not require that active fissile material to be integrated in this testing. A small-scale test bed; Compact Fuel Element Environmental Tester (CFEET), designed to heat fuel element samples via induction heating and expose samples to hydrogen is being developed at MSFC to assist in optimal material and manufacturing process selection without utilizing fissile material. This paper details the analytical approach to help design and optimize the test bed using COMSOL Multiphysics for predicting thermal gradients induced by electromagnetic heating (Induction heating) and Thermal Desktop for radiation calculations.

Innovative power conversion system for the French SFR prototype, ASTRID

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

Cachon, L.; Biscarrat, C.; Morin, F.

2012-07-01

In the framework of the French Act of 28 June 2006 about nuclear materials and waste management, the prototype ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration), foreseen in operation by the 20's, will have to demonstrate not only the minor actinide transmutation capability, but also the progress made in Sodium Fast Reactor (SFR) technology on an industrial scale, by qualifying innovative options. Some of these options still require improvements, especially in the field of operability and safety. In fact, one of the main issues with the standard steam/water Power Conversion System (PCS) of SFR is the fast and energeticmore » chemical reaction between water and sodium, which could occur in steam generators in case of tube failure. To manage the sodium/water reaction, one way consists in minimizing the impact of such event: hence studies are carried out on steam generator design, improvement of the physical knowledge of this phenomenon, development of numerical simulation to predict the reaction onset and consequences, and associated detection improvement. On the other hand, the other way consists in eliminating sodium/water reaction. In this frame, the CEA contribution to the feasibility evaluation of an alternative innovative PCS (replacing steam/water by 180 bar pressurised nitrogen) is focused on the following main topics: - The parametric study leading to nitrogen selection: the thermodynamic cycle efficiency optimisation on Brayton cycles is performed with several gases at different pressures. - The design of innovative compact heat exchangers for the gas loop: here the key points are the nuclear codification associated with inspection capability, the innovative welding process and the thermal-hydraulic and thermal-mechanic optimisations. After a general introduction of the ASTRID project, this paper presents in detail these different feasibility studies being led on the innovative gas PCS for an SFR. (authors)« less

Simplifying Chemical Reactor Design by using Molar Quantities Instead of Fractional Conversion.

ERIC Educational Resources Information Center

Brown, Lee F.; Falconer, John L.

1987-01-01

Explains the advantages of using molar quantities in chemical reactor design. Advocates the use of differential versions of reactor mass balances rather than the integrated forms. Provides specific examples and cases to illustrate the principles. (ML)

Control console replacement at the WPI Reactor

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

Not Available

1992-01-01

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

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

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

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

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

Room temperature micro-hydrogen-generator

NASA Astrophysics Data System (ADS)

Gervasio, Don; Tasic, Sonja; Zenhausern, Frederic

A new compact and cost-effective hydrogen-gas generator has been made that is well suited for supplying hydrogen to a fuel-cell for providing base electrical power to hand-carried appliances. This hydrogen-generator operates at room temperature, ambient pressure and is orientation-independent. The hydrogen-gas is generated by the heterogeneous catalytic hydrolysis of aqueous alkaline borohydride solution as it flows into a micro-reactor. This reactor has a membrane as one wall. Using the membrane keeps the liquid in the reactor, but allows the hydrogen-gas to pass out of the reactor to a fuel-cell anode. Aqueous alkaline 30 wt% borohydride solution is safe and promotes long application life, because this solution is non-toxic, non-flammable, and is a high energy-density (≥2200 W-h per liter or per kilogram) hydrogen-storage solution. The hydrogen is released from this storage-solution only when it passes over the solid catalyst surface in the reactor, so controlling the flow of the solution over the catalyst controls the rate of hydrogen-gas generation. This allows hydrogen generation to be matched to hydrogen consumption in the fuel-cell, so there is virtually no free hydrogen-gas during power generation. A hydrogen-generator scaled for a system to provide about 10 W electrical power is described here. However, the technology is expected to be scalable for systems providing power spanning from 1 W to kW levels.

Comparison of some characteristics of aerobic granules and sludge flocs from sequencing batch reactors.

PubMed

Li, J; Garny, K; Neu, T; He, M; Lindenblatt, C; Horn, H

2007-01-01

Physical, chemical and biological characteristics were investigated for aerobic granules and sludge flocs from three laboratory-scale sequencing batch reactors (SBRs). One reactor was operated as normal SBR (N-SBR) and two reactors were operated as granular SBRs (G-SBR1 and G-SBR2). G-SBR1 was inoculated with activated sludge and G-SBR2 with granules from the municipal wastewater plant in Garching (Germany). The following major parameters and functions were measured and compared between the three reactors: morphology, settling velocity, specific gravity (SG), sludge volume index (SVI), specific oxygen uptake rate (SOUR), distribution of the volume fraction of extracellular polymeric substances (EPS) and bacteria, organic carbon and nitrogen removal. Compared with sludge flocs, granular sludge had excellent settling properties, good solid-liquid separation, high biomass concentration, simultaneous nitrification and denitrification. Aerobic granular sludge does not have a higher microbial activity and there are some problems including higher effluent suspended solids, lower ratio of VSS/SS and no nitrification at the beginning of cultivation. Measurement with CLSM and additional image analysis showed that EPS glycoconjugates build one main fraction inside the granules. The aerobic granules from G-SBR1 prove to be heavier, smaller and have a higher microbial activity compared with G-SBR2. Furthermore, the granules were more compact, with lower SVI and less filamentous bacteria.

Design and Development of a Prototype Permanent Magnet for Focusing/Defocusing for Electron-Ion Colliders

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

Wagner, Bob

Electron-ion colliders (EIC) have been identified as an ideal tool to study the next frontier of nuclear physics – the gluon force that holds the building blocks of matter together, and which is a fundamental component of the theory of Quantum Chromodynamics (QCD). Future electron-ion colliders under consideration can be based on the Energy Recovery Linac (ERL) architecture. The beam lines for this architecture could be built of the newly developed Non-Scaling Fixed Field Alternating Gradient (NS FFAG) structure, so that they can transfer multiple energies within the same aperture. This structure allows for the use of compact, economical quadupolemore » permanent magnets. In this SBIR, we propose to design and to manufacture prototype quadrupole permanent magnets of focusing/defocusing combined function for use in this beam line. For our SBIR project, we proposed to design and build the focusing/defocusing quadrupole with a gradient strength of 50 T/m and with a beam gap of 16mm. The proposed permanent magnet material is SmCo because of its higher radiation resistance as compared to NdBFe2. The use of permanent magnets will reduce the overall cost. For Phase I, we took a recent design by Dr. Dejan Trbojevic, and reran Tosca code on the design to optimize the iron yoke with respect to the thickness of SmCo. We then fabricated one prototype focusing/defocusing combined function quadruple and measured field quality dG/Go. Our plan for Phase II is that, based on our Phase I prototype experience, we shall improve the design and fabricate a production quadruple, and design and incorporate coils for skew dipoles and normal quadrupole correctors, etc. In addition, we shall fabricate enough quadrupoles for one cell. The development of quadrupole permanent magnets is of fundamental importance for there application in the future electron-ion colliders. This accelerator structure will also advance the development of muon accelerators and allow for the development of compact, simplified, less expensive proton accelerators which will promote their use in areas such as proton cancer therapy, and for high-power proton drivers for tritium and neutron production, waste transmutation, driving a sub-critical nuclear reactor to produce energy, cargo contain inspection, and radioisotope production. Proton cancer therapy has been identified as a particularly attractive and viable commercial application for the immediate future.« less

Investigation of internal elements impaction on particles circulation in a fluidized bed reactor

NASA Astrophysics Data System (ADS)

Solovev, S. A.; Soloveva, O. V.; Antipin, A. V.; Shamsutdinov, E. V.

2018-01-01

A numerical study of the fluidized bed apparatus in the presence of various internal elements is carried out. A chemical reaction for temperature-dependent processes with heat absorption is considered. The task of incoming heated catalyst granules to the reactor is investigated. The main emphasis is focused on the circulation flows of the catalyst particles, heating of the reactor, and the efficiency of the chemical reaction. The analysis of the impact of various design elements on the efficiency of the reactor is carried out. The influence of feeding heated catalyst device design on the effectiveness of whole reactor heating is educed. The influence of the presence of fine particles on the efficiency of the reaction for different reactor design features is also educed.

Design and progress report for compact cryocooled sapphire oscillator 'VCSO'

NASA Technical Reports Server (NTRS)

Dick, G. John; Wang, Rabi T.; Tjoelker, Robert L.

2005-01-01

We report on the development of a compact cryocooled sapphiere oscillator 'VCSO', designed as a higher-performance replacement for ultra-stable quartz oscillators in local oscillator, cleanup, and flywheel applications in the frequency generation and distribution subsystems of NASA's Deep Space Network (DSN).

Nuclear Reactors. Revised.

ERIC Educational Resources Information Center

Hogerton, John F.

This publication is one of a series of information booklets for the general public published by the United States Atomic Energy Commission. Among the topics discussed are: How Reactors Work; Reactor Design; Research, Teaching, and Materials Testing; Reactors (Research, Teaching and Materials); Production Reactors; Reactors for Electric Power…

Modeling and Design Optimization of Multifunctional Membrane Reactors for Direct Methane Aromatization

PubMed Central

Fouty, Nicholas J.; Carrasco, Juan C.; Lima, Fernando V.

2017-01-01

Due to the recent increase of natural gas production in the U.S., utilizing natural gas for higher-value chemicals has become imperative. Direct methane aromatization (DMA) is a promising process used to convert methane to benzene, but it is limited by low conversion of methane and rapid catalyst deactivation by coking. Past work has shown that membrane separation of the hydrogen produced in the DMA reactions can dramatically increase the methane conversion by shifting the equilibrium toward the products, but it also increases coke production. Oxygen introduction into the system has been shown to inhibit this coke production while not inhibiting the benzene production. This paper introduces a novel mathematical model and design to employ both methods in a multifunctional membrane reactor to push the DMA process into further viability. Multifunctional membrane reactors, in this case, are reactors where two different separations occur using two differently selective membranes, on which no systems studies have been found. The proposed multifunctional membrane design incorporates a hydrogen-selective membrane on the outer wall of the reaction zone, and an inner tube filled with airflow surrounded by an oxygen-selective membrane in the middle of the reactor. The design is shown to increase conversion via hydrogen removal by around 100%, and decrease coke production via oxygen addition by 10% when compared to a tubular reactor without any membranes. Optimization studies are performed to determine the best reactor design based on methane conversion, along with coke and benzene production. The obtained optimal design considers a small reactor (length = 25 cm, diameter of reaction tube = 0.7 cm) to subvert coke production and consumption of the product benzene as well as a high permeance (0.01 mol/s·m2·atm1/4) through the hydrogen-permeable membrane. This modeling and design approach sets the stage for guiding further development of multifunctional membrane reactor models and designs for natural gas utilization and other chemical reaction systems. PMID:28850068

High Temperature Heat Exchanger Design and Fabrication for Systems with Large Pressure Differentials

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

Chordia, Lalit; Portnoff, Marc A.; Green, Ed

The project’s main purpose was to design, build and test a compact heat exchanger for supercritical carbon dioxide (sCO 2) power cycle recuperators. The compact recuperator is required to operate at high temperature and high pressure differentials, 169 bar (~2,500 psi), between streams of sCO 2. Additional project tasks included building a hot air-to-sCO 2 Heater heat exchanger (HX) and design, build and operate a test loop to characterize the recuperator and heater heat exchangers. A novel counter-current microtube recuperator was built to meet the high temperature high differential pressure criteria and tested. The compact HX design also incorporated amore » number of features that optimize material use, improved reliability and reduced cost. The air-to-sCO 2 Heater HX utilized a cross flow, counter-current, micro-tubular design. This compact HX design was incorporated into the test loop and exceeded design expectations. The test loop design to characterize the prototype Brayton power cycle HXs was assembled, commissioned and operated during the program. Both the prototype recuperator and Heater HXs were characterized. Measured results for the recuperator confirmed the predictions of the heat transfer models developed during the project. Heater HX data analysis is ongoing.« less

Design options for a bunsen reactor.

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

Moore, Robert Charles

2013-10-01

This work is being performed for Matt Channon Consulting as part of the Sandia National Laboratories New Mexico Small Business Assistance Program (NMSBA). Matt Channon Consulting has requested Sandia's assistance in the design of a chemical Bunsen reactor for the reaction of SO2, I2 and H2O to produce H2SO4 and HI with a SO2 feed rate to the reactor of 50 kg/hour. Based on this value, an assumed reactor efficiency of 33%, and kinetic data from the literature, a plug flow reactor approximately 1%E2%80%9D diameter and and 12 inches long would be needed to meet the specification of the project.more » Because the Bunsen reaction is exothermic, heat in the amount of approximately 128,000 kJ/hr would need to be removed using a cooling jacket placed around the tubular reactor. The available literature information on Bunsen reactor design and operation, certain support equipment needed for process operation and a design that meet the specification of Matt Channon Consulting are presented.« less

A Parametric Sizing Model for Molten Regolith Electrolysis Reactors to Produce Oxygen from Lunar Regolith

NASA Technical Reports Server (NTRS)

Schreiner, Samuel S.; Dominguez, Jesus A.; Sibille, Laurent; Hoffman, Jeffrey A.

2015-01-01

We present a parametric sizing model for a Molten Electrolysis Reactor that produces oxygen and molten metals from lunar regolith. The model has a foundation of regolith material properties validated using data from Apollo samples and simulants. A multiphysics simulation of an MRE reactor is developed and leveraged to generate a vast database of reactor performance and design trends. A novel design methodology is created which utilizes this database to parametrically design an MRE reactor that 1) can sustain the required mass of molten regolith, current, and operating temperature to meet the desired oxygen production level, 2) can operate for long durations via joule heated, cold wall operation in which molten regolith does not touch the reactor side walls, 3) can support a range of electrode separations to enable operational flexibility. Mass, power, and performance estimates for an MRE reactor are presented for a range of oxygen production levels. The effects of several design variables are explored, including operating temperature, regolith type/composition, batch time, and the degree of operational flexibility.

Compact Polarimetry Potentials

NASA Technical Reports Server (NTRS)

Truong-Loi, My-Linh; Dubois-Fernandez, Pascale; Pottier, Eric

2011-01-01

The goal of this study is to show the potential of a compact-pol SAR system for vegetation applications. Compact-pol concept has been suggested to minimize the system design while maximize the information and is declined as the ?/4, ?/2 and hybrid modes. In this paper, the applications such as biomass and vegetation height estimates are first presented, then, the equivalence between compact-pol data simulated from full-pol data and compact-pol data processed from raw data as such is shown. Finally, a calibration procedure using external targets is proposed.

ENGINEERING AND CONSTRUCTING THE HALLAM NUCLEAR POWER FACILITY REACTOR STRUCTURE

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

Mahlmeister, J E; Haberer, W V; Casey, D F

1960-12-15

The Hallam Nuclear Power Facility reactor structure, including the cavity liner, is described, and the design philosophy and special design requirements which were developed during the preliminary and final engineering phases of the project are explained. The structure was designed for 600 deg F inlet and 1000 deg F outlet operating sodium temperatures and fabricated of austenitic and ferritic stainless steels. Support for the reactor core components and adequate containment for biological safeguards were readily provided even though quite conservative design philosophy was used. The calculated operating characteristics, including heat generation, temperature distributions and stress levels for full-power operation, aremore » summarized. Ship fabrication and field installation experiences are also briefly related. Results of this project have established that the sodium graphite reactor permits practical and economical fabrication and field erection procedures; considerably higher operating design temperatures are believed possible without radical design changes. Also, larger reactor structures can be similarly constructed for higher capacity (300 to 1000 Mwe) nuclear power plants. (auth)« less

Introduction to Chemical Engineering Reactor Analysis: A Web-Based Reactor Design Game

ERIC Educational Resources Information Center

Orbey, Nese; Clay, Molly; Russell, T.W. Fraser

2014-01-01

An approach to explain chemical engineering through a Web-based interactive game design was developed and used with college freshman and junior/senior high school students. The goal of this approach was to demonstrate how to model a lab-scale experiment, and use the results to design and operate a chemical reactor. The game incorporates both…

A Framework for Human Performance Criteria for Advanced Reactor Operational Concepts

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

Jacques V Hugo; David I Gertman; Jeffrey C Joe

2014-08-01

This report supports the determination of new Operational Concept models needed in support of the operational design of new reactors. The objective of this research is to establish the technical bases for human performance and human performance criteria frameworks, models, and guidance for operational concepts for advanced reactor designs. The report includes a discussion of operating principles for advanced reactors, the human performance issues and requirements for human performance based upon work domain analysis and current regulatory requirements, and a description of general human performance criteria. The major findings and key observations to date are that there is some operatingmore » experience that informs operational concepts for baseline designs for SFR and HGTRs, with the Experimental Breeder Reactor-II (EBR-II) as a best-case predecessor design. This report summarizes the theoretical and operational foundations for the development of a framework and model for human performance criteria that will influence the development of future Operational Concepts. The report also highlights issues associated with advanced reactor design and clarifies and codifies the identified aspects of technology and operating scenarios.« less

Design of a Compact Quad-Channel Diplexer

NASA Astrophysics Data System (ADS)

Xu, Jin

2016-01-01

This paper presents a compact quad-channel diplexer by using two asymmetrical coupling shorted stub loaded stepped-impedance (SSLSIR) dual-band bandpass filters (DB-BPFs) to replace two single-band BPFs in a traditional BPF-based diplexer. Part of its impedance matching circuit is implemented by using a three-element lowpass T-network to acquire the desired phase shift. Detailed design procedures are given to guide the diplexer design. The fabricated quad-channel diplexer occupies a compact circuit area of 0.168λg×0.136λg. High band-to-band isolation and wide stopband performance are achieved. Good agreement is shown between the simulated and measured results.

Enzyme reactor design under thermal inactivation.

PubMed

Illanes, Andrés; Wilson, Lorena

2003-01-01

Temperature is a very relevant variable for any bioprocess. Temperature optimization of bioreactor operation is a key aspect for process economics. This is especially true for enzyme-catalyzed processes, because enzymes are complex, unstable catalysts whose technological potential relies on their operational stability. Enzyme reactor design is presented with a special emphasis on the effect of thermal inactivation. Enzyme thermal inactivation is a very complex process from a mechanistic point of view. However, for the purpose of enzyme reactor design, it has been oversimplified frequently, considering one-stage first-order kinetics of inactivation and data gathered under nonreactive conditions that poorly represent the actual conditions within the reactor. More complex mechanisms are frequent, especially in the case of immobilized enzymes, and most important is the effect of catalytic modulators (substrates and products) on enzyme stability under operation conditions. This review focuses primarily on reactor design and operation under modulated thermal inactivation. It also presents a scheme for bioreactor temperature optimization, based on validated temperature-explicit functions for all the kinetic and inactivation parameters involved. More conventional enzyme reactor design is presented merely as a background for the purpose of highlighting the need for a deeper insight into enzyme inactivation for proper bioreactor design.

Strategy Guideline. Compact Air Distribution Systems

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

Burdick, Arlan

2013-06-01

This guideline discusses the benefits and challenges of using a compact air distribution system to handle the reduced loads and reduced air volume needed to condition the space within an energy efficient home. The decision criteria for a compact air distribution system must be determined early in the whole-house design process, considering both supply and return air design. However, careful installation of a compact air distribution system can result in lower material costs from smaller equipment, shorter duct runs, and fewer outlets; increased installation efficiencies, including ease of fitting the system into conditioned space; lower loads on a better balancedmore » HVAC system, and overall improved energy efficiency of the home.« less

An inexpensive and reliable monitoring station design for use with lightweight, compact data loggers

Treesearch

Ronald S., Jr. Zalesny; Adam H. Wiese; Edmund O. Bauer; William L., Jr. Headlee; Richard B. Hall; A. Assibi Mahama; Jill A. Zalesny

2007-01-01

We designed, constructed, and field-tested an inexpensive and reliable monitoring station that can be used with lightweight, compact data loggers. We feel this design, improved three times over 6 yr, could benefit anyone in nursery or field settings interested in acquiring environmental data. We provide step-by-step instructions on the construction of the monitoring...

Design study of an ultra-compact superconducting cyclotron for isotope production

NASA Astrophysics Data System (ADS)

Smirnov, V.; Vorozhtsov, S.; Vincent, J.

2014-11-01

A 12.5 MeV, 25 μA, proton compact superconducting cyclotron for medical isotope production has been designed and is currently in fabrication. The machine is initially aimed at producing 13N ammonia for Positron Emission Tomography (PET) cardiology applications. With an ultra-compact size and cost-effective price point, this system will offer clinicians unprecedented access to the preferred radiopharmaceutical isotope for cardiac PET imaging. A systems approach that carefully balanced the subsystem requirements coupled to precise beam dynamics calculations was followed. The system is designed to irradiate a liquid target internal to the cyclotron and to minimize the need for radiation shielding. The main parameters of the cyclotron, its design, and principal steps of the development work are presented here.

Relaunch of the Interactive Plasma Physics Educational Experience (IPPEX)

NASA Astrophysics Data System (ADS)

Dominguez, A.; Rusaitis, L.; Zwicker, A.; Stotler, D. P.

2015-11-01

In the late 1990's PPPL's Science Education Department developed an innovative online site called the Interactive Plasma Physics Educational Experience (IPPEX). It featured (among other modules) two Java based applications which simulated tokamak physics: A steady state tokamak (SST) and a time dependent tokamak (TDT). The physics underlying the SST and the TDT are based on the ASPECT code which is a global power balance code developed to evaluate the performance of fusion reactor designs. We have relaunched the IPPEX site with updated modules and functionalities: The site itself is now dynamic on all platforms. The graphic design of the site has been modified to current standards. The virtual tokamak programming has been redone in Javascript, taking advantage of the speed and compactness of the code. The GUI of the tokamak has been completely redesigned, including more intuitive representations of changes in the plasma, e.g., particles moving along magnetic field lines. The use of GPU accelerated computation provides accurate and smooth visual representations of the plasma. We will present the current version of IPPEX as well near term plans of incorporating real time NSTX-U data into the simulation.

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

Impact of Various Compaction Equipment on Hot-Mix Asphalt (HMA) Design in Ohio.

DOT National Transportation Integrated Search

1998-09-01

The study was initiated to examine the effect of different laboratory compaction devices on density, air voids, and optimum asphalt binder content for mixes designed for heavy volume traffic that are currently used in Ohio. A total of six and twelve ...

Concept of DT fuel cycle for a fusion neutron source

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

Anan'ev, S.; Spitsyn, A.V.; Kuteev, B.V.

2015-03-15

A concept of DT-fusion neutron source (FNS) with the neutron yield higher than 10{sup 18} neutrons per second is under design in Russia. Such a FNS is of interest for many applications: 1) basic and applied research (neutron scattering, etc); 2) testing the structural materials for fusion reactors; 3) control of sub-critical nuclear systems and 4) nuclear waste processing (including transmutation of minor actinides). This paper describes the fuel cycle concept of a compact fusion neutron source based on a small spherical tokamak (FNS-ST) with a MW range of DT fusion power and considers the key physics issues of thismore » device. The major and minor radii are ∼0.5 and ∼0.3 m, magnetic field ∼1.5 T, heating power less than 15 MW and plasma current 1-2 MA. The system provides the fuel mixture with equal fractions of D and T (D:T = 1:1) for all FNS technology systems. (authors)« less

Coupling of a 2.5 kW steam reformer with a 1 kW el PEM fuel cell

NASA Astrophysics Data System (ADS)

Mathiak, J.; Heinzel, A.; Roes, J.; Kalk, Th.; Kraus, H.; Brandt, H.

The University of Duisburg-Essen has developed a compact multi-fuel steam reformer suitable for natural gas, propane and butane. This steam reformer was combined with a polymer electrolyte membrane fuel cell (PEM FC) and a system test of the process chain was performed. The fuel processor comprises a prereformer step, a primary reformer, water gas shift reactors, a steam generator, internal heat exchangers in order to achieve an optimised heat integration and an external burner for heat supply as well as a preferential oxidation step (PROX) as CO purification. The fuel processor is designed to deliver a thermal hydrogen power output from 500 W to 2.5 kW. The PEM fuel cell stack provides about 1 kW electrical power. In the following paper experimental results of measurements of the single components PEM fuel cell and fuel processor as well as results of the coupling of both to form a process chain are presented.

An Ultrasonic Compactor for Oil and Gas Exploration

NASA Astrophysics Data System (ADS)

Feeney, Andrew; Sikaneta, Sakalima; Harkness, Patrick; Lucas, Margaret

The Badger Explorer is a rig-less oil and gas exploration tool which drills into the subsea environment to collect geological data. Drill spoil is transported from the front end of the system to the rear, where the material is compacted. Motivated by the need to develop a highly efficient compaction system, an ultrasonic compactor for application with granular geological materials encountered in subsea environments is designed and fabricated as part of this study. The finite element method is used to design a compactor configuration suitable for subsea exploration, consisting of a vibrating ultrasonic horn called a resonant compactor head, which operates in a longitudinal mode at 20 kHz, driven by a Langevin piezoelectric transducer. A simplified version of the compactor is also designed, due to its ease of incorporating in a lab-based experimental rig, in order to demonstrate enhanced compaction using ultrasonics. Numerical analysis of this simplified compactor system is supported with experimental characterisation using laser Doppler vibrometry. Compaction testing is then conducted on granular geological material, showing that compaction can be enhanced through the use of an ultrasonic compactor.

Two-step design method for highly compact three-dimensional freeform optical system for LED surface light source.

PubMed

Mao, Xianglong; Li, Hongtao; Han, Yanjun; Luo, Yi

2014-10-20

Designing an illumination system for a surface light source with a strict compactness requirement is quite challenging, especially for the general three-dimensional (3D) case. In accordance with the two key features of an expected illumination distribution, i.e., a well-controlled boundary and a precise illumination pattern, a two-step design method is proposed in this paper for highly compact 3D freeform illumination systems. In the first step, a target shape scaling strategy is combined with an iterative feedback modification algorithm to generate an optimized freeform optical system with a well-controlled boundary of the target distribution. In the second step, a set of selected radii of the system obtained in the first step are optimized to further improve the illuminating quality within the target region. The method is quite flexible and effective to design highly compact optical systems with almost no restriction on the shape of the desired target field. As examples, three highly compact freeform lenses with ratio of center height h of the lens and the maximum dimension D of the source ≤ 2.5:1 are designed for LED surface light sources to form a uniform illumination distribution on a rectangular, a cross-shaped and a complex cross pierced target plane respectively. High light control efficiency of η > 0.7 as well as low relative standard illumination deviation of RSD < 0.07 is obtained simultaneously for all the three design examples.

Electrical Capacitance Volume Tomography for the Packed Bed Reactor ISS Flight Experiment

NASA Technical Reports Server (NTRS)

Marashdeh, Qussai; Motil, Brian; Wang, Aining; Liang-Shih, Fan

2013-01-01

Fixed packed bed reactors are compact, require minimum power and maintenance to operate, and are highly reliable. These features make this technology a highly desirable unit operation for long duration life support systems in space. NASA is developing an ISS experiment to address this technology with particular focus on water reclamation and air revitalization. Earlier research and development efforts funded by NASA have resulted in two hydrodynamic models which require validation with appropriate instrumentation in an extended microgravity environment. To validate these models, the instantaneous distribution of the gas and liquid phases must be measured.Electrical Capacitance Volume Tomography (ECVT) is a non-invasive imaging technology recently developed for multi-phase flow applications. It is based on distributing flexible capacitance plates on the peripheral of a flow column and collecting real-time measurements of inter-electrode capacitances. Capacitance measurements here are directly related to dielectric constant distribution, a physical property that is also related to material distribution in the imaging domain. Reconstruction algorithms are employed to map volume images of dielectric distribution in the imaging domain, which is in turn related to phase distribution. ECVT is suitable for imaging interacting materials of different dielectric constants, typical in multi-phase flow systems. ECVT is being used extensively for measuring flow variables in various gas-liquid and gas-solid flow systems. Recent application of ECVT include flows in risers and exit regions of circulating fluidized beds, gas-liquid and gas-solid bubble columns, trickle beds, and slurry bubble columns. ECVT is also used to validate flow models and CFD simulations. The technology is uniquely qualified for imaging phase concentrations in packed bed reactors for the ISS flight experiments as it exhibits favorable features of compact size, low profile sensors, high imaging speed, and flexibility to fit around columns of various shapes and sizes. ECVT is also safer than other commonly used imaging modalities as it operates in the range of low frequencies (1 MHz) and does not radiate radioactive energy. In this effort, ECVT is being used to image flow parameters in a packed bed reactor for an ISS flight experiment.

Polymeric behavior evaluation of PVP K30-poloxamer binary carrier for solid dispersed nisoldipine by experimental design.

PubMed

Kyaw Oo, May; Mandal, Uttam K; Chatterjee, Bappaditya

2017-02-01

High melting point polymeric carrier without plasticizer is unacceptable for solid dispersion (SD) by melting method. Combined polymer-plasticizer carrier significantly affects drug solubility and tableting property of SD. To evaluate and optimize the combined effect of a binary carrier consisting PVP K30 and poloxamer 188, on nisoldipine solubility and tensile strength of amorphous SD compact (SD compact ) by experimental design. SD of nisoldpine (SD nisol ) was prepared by melt mixing with different PVP K30 and poloxamer amount. A 3 2 factorial design was employed using nisoldipine solubility and tensile strength of SD compact as response variables. Statistical optimization by design expert software, and SD nisol characterization using ATR FTIR, DSC and microscopy were done. PVP K30:poloxamer, at a ratio of 3.73:6.63, was selected as the optimized combination of binary polymeric carrier resulting nisoldipine solubility of 115 μg/mL and tensile strength of 1.19 N/m 2 . PVP K30 had significant positive effect on both responses. Increase in poloxamer concentration after a certain level decreased nisoldipine solubility and tensile strength of SD compact . An optimized PVP K30-poloxamer binary composition for SD carrier was developed. Tensile strength of SD compact can be considered as a response for experimental design to optimize SD.

Compact propane fuel processor for auxiliary power unit application

NASA Astrophysics Data System (ADS)

Dokupil, M.; Spitta, C.; Mathiak, J.; Beckhaus, P.; Heinzel, A.

With focus on mobile applications a fuel cell auxiliary power unit (APU) using liquefied petroleum gas (LPG) is currently being developed at the Centre for Fuel Cell Technology (Zentrum für BrennstoffzellenTechnik, ZBT gGmbH). The system is consisting of an integrated compact and lightweight fuel processor and a low temperature PEM fuel cell for an electric power output of 300 W. This article is presenting the current status of development of the fuel processor which is designed for a nominal hydrogen output of 1 k Wth,H2 within a load range from 50 to 120%. A modular setup was chosen defining a reformer/burner module and a CO-purification module. Based on the performance specifications, thermodynamic simulations, benchmarking and selection of catalysts the modules have been developed and characterised simultaneously and then assembled to the complete fuel processor. Automated operation results in a cold startup time of about 25 min for nominal load and carbon monoxide output concentrations below 50 ppm for steady state and dynamic operation. Also fast transient response of the fuel processor at load changes with low fluctuations of the reformate gas composition have been achieved. Beside the development of the main reactors the transfer of the fuel processor to an autonomous system is of major concern. Hence, concepts for packaging have been developed resulting in a volume of 7 l and a weight of 3 kg. Further a selection of peripheral components has been tested and evaluated regarding to the substitution of the laboratory equipment.

Determining the compactive effort required to model pavement voids using the Corps of Engineers gyratory testing machine.

DOT National Transportation Integrated Search

1997-11-01

Various agencies have used the Corps of Engineers gyratory testing machine (GTM) to design and test asphalt mixes. Materials properties such as shear strength and strain are measured during the compaction process. However, a compaction process duplic...

10 CFR 52.167 - Issuance of manufacturing license.

Code of Federal Regulations, 2010 CFR

2010-01-01

... proposed reactor(s) can be incorporated into a nuclear power plant and operated at sites having... design and manufacture the proposed nuclear power reactor(s); (5) The proposed inspections, tests... the construction of a nuclear power facility using the manufactured reactor(s). (2) A holder of a...

Phenomena Important in Molten Salt Reactor Simulations

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

Diamond, David J.; Brown, Nicholas R.; Denning, Richard

The U.S. Nuclear Regulatory Commission (NRC) is preparing for the future licensing of advanced reactors that will be very different from current light water reactors. Part of the NRC preparation strategy is to identify the simulation tools that will be used for confirmatory safety analysis of normal operation and abnormal situations in those reactors. This report advances that strategy for reactors that will use molten salts (MSRs). This includes reactors with the fuel within the salt as well as reactors using solid fuel. Although both types are discussed in this report, the emphasis is on those reactors with liquid fuelmore » because of the perception that solid-fuel MSRs will be significantly easier to simulate. These liquid-fuel reactors include thermal and fast neutron spectrum alternatives. The specific designs discussed in the report are a subset of many designs being considered in the U.S. and elsewhere but they are considered the most likely to submit information to the NRC in the near future. The objective herein, is to understand the design of proposed molten salt reactors, how they will operate under normal or transient/accident conditions, and what will be the corresponding modeling needs of simulation tools that consider neutronics, heat transfer, fluid dynamics, and material composition changes in the molten salt. These tools will enable the NRC to eventually carry out confirmatory analyses that examine the validity and accuracy of applicant’s calculations and help determine the margin of safety in plant design.« less

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

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

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

2012-07-01

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

AP1000{sup R} design robustness against extreme external events - Seismic, flooding, and aircraft crash

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

Pfister, A.; Goossen, C.; Coogler, K.

2012-07-01

Both the International Atomic Energy Agency (IAEA) and the U.S. Nuclear Regulatory Commission (NRC) require existing and new nuclear power plants to conduct plant assessments to demonstrate the unit's ability to withstand external hazards. The events that occurred at the Fukushima-Dai-ichi nuclear power station demonstrated the importance of designing a nuclear power plant with the ability to protect the plant against extreme external hazards. The innovative design of the AP1000{sup R} nuclear power plant provides unparalleled protection against catastrophic external events which can lead to extensive infrastructure damage and place the plant in an extended abnormal situation. The AP1000 plantmore » is an 1100-MWe pressurized water reactor with passive safety features and extensive plant simplifications that enhance construction, operation, maintenance and safety. The plant's compact safety related footprint and protection provided by its robust nuclear island structures prevent significant damage to systems, structures, and components required to safely shutdown the plant and maintain core and spent fuel pool cooling and containment integrity following extreme external events. The AP1000 nuclear power plant has been extensively analyzed and reviewed to demonstrate that it's nuclear island design and plant layout provide protection against both design basis and extreme beyond design basis external hazards such as extreme seismic events, external flooding that exceeds the maximum probable flood limit, and malicious aircraft impact. The AP1000 nuclear power plant uses fail safe passive features to mitigate design basis accidents. The passive safety systems are designed to function without safety-grade support systems (such as AC power, component cooling water, service water, compressed air or HVAC). The plant has been designed to protect systems, structures, and components critical to placing the reactor in a safe shutdown condition within the steel containment vessel which is further surrounded by a substantial 'steel concrete' composite shield building. The containment vessel is not affected by external flooding, and the shield building design provides hazard protection beyond that provided by a comparable reinforced concrete structure. The intent of this paper is to demonstrate the robustness of the AP1000 design against extreme events. The paper will focus on the plants ability to withstand extreme external events such as beyond design basis flooding, seismic events, and malicious aircraft impact. The paper will highlight the robustness of the AP1000 nuclear island design including the protection provided by the unique AP1000 composite shield building. (authors)« less

Conceptual design study of the moderate size superconducting spherical tokamak power plant

NASA Astrophysics Data System (ADS)

Gi, Keii; Ono, Yasushi; Nakamura, Makoto; Someya, Youji; Utoh, Hiroyasu; Tobita, Kenji; Ono, Masayuki

2015-06-01

A new conceptual design of the superconducting spherical tokamak (ST) power plant was proposed as an attractive choice for tokamak fusion reactors. We reassessed a possibility of the ST as a power plant using the conservative reactor engineering constraints often used for the conventional tokamak reactor design. An extensive parameters scan which covers all ranges of feasible superconducting ST reactors was completed, and five constraints which include already achieved plasma magnetohydrodynamic (MHD) and confinement parameters in ST experiments were established for the purpose of choosing the optimum operation point. Based on comparison with the estimated future energy costs of electricity (COEs) in Japan, cost-effective ST reactors can be designed if their COEs are smaller than 120 mills kW-1 h-1 (2013). We selected the optimized design point: A = 2.0 and Rp = 5.4 m after considering the maintenance scheme and TF ripple. A self-consistent free-boundary MHD equilibrium and poloidal field coil configuration of the ST reactor were designed by modifying the neutral beam injection system and plasma profiles. The MHD stability of the equilibrium was analysed and a ramp-up scenario was considered for ensuring the new ST design. The optimized moderate-size ST power plant conceptual design realizes realistic plasma and fusion engineering parameters keeping its economic competitiveness against existing energy sources in Japan.

A Basic LEGO Reactor Design for the Provision of Lunar Surface Power

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

John Darrell Bess

2008-06-01

A final design has been established for a basic Lunar Evolutionary Growth-Optimized (LEGO) Reactor using current and near-term technologies. The LEGO Reactor is a modular, fast-fission, heatpipe-cooled, clustered-reactor system for lunar-surface power generation. The reactor is divided into subcritical units that can be safely launched with lunar shipments from Earth, and then emplaced directly into holes drilled into the lunar regolith to form a critical reactor assembly. The regolith would not just provide radiation shielding, but serve as neutron-reflector material as well. The reactor subunits are to be manufactured using proven and tested materials for use in radiation environments, suchmore » as uranium-dioxide fuel, stainless-steel cladding and structural support, and liquid-sodium heatpipes. The LEGO Reactor system promotes reliability, safety, and ease of manufacture and testing at the cost of an increase in launch mass per overall rated power level and a reduction in neutron economy when compared to a single-reactor system. A single unshielded LEGO Reactor subunit has an estimated mass of approximately 448 kg and provides approximately 5 kWe. The overall envelope for a single subunit with fully extended radiator panels has a height of 8.77 m and a diameter of 0.50 m. Six subunits could provide sufficient power generation throughout the initial stages of establishing a lunar outpost. Portions of the reactor may be neutronically decoupled to allow for reduced power production during unmanned periods of base operations. During later stages of lunar-base development, additional subunits may be emplaced and coupled into the existing LEGO Reactor network, subject to lunar base power demand. Improvements in reactor control methods, fuel form and matrix, shielding, as well as power conversion and heat rejection techniques can help generate an even more competitive LEGO Reactor design. Further modifications in the design could provide power generative opportunities for use on other extraterrestrial surfaces.« less

HIGH-TEMPERATURE SAFETY TESTING OF IRRADIATED AGR-1 TRISO FUEL

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

Stempien, John D.; Demkowicz, Paul A.; Reber, Edward L.

High-Temperature Safety Testing of Irradiated AGR-1 TRISO Fuel John D. Stempien, Paul A. Demkowicz, Edward L. Reber, and Cad L. Christensen Idaho National Laboratory, P.O. Box 1625 Idaho Falls, ID 83415, USA Corresponding Author: john.stempien@inl.gov, +1-208-526-8410 Two new safety tests of irradiated tristructural isotropic (TRISO) coated particle fuel have been completed in the Fuel Accident Condition Simulator (FACS) furnace at the Idaho National Laboratory (INL). In the first test, three fuel compacts from the first Advanced Gas Reactor irradiation experiment (AGR-1) were simultaneously heated in the FACS furnace. Prior to safety testing, each compact was irradiated in the Advanced Testmore » Reactor to a burnup of approximately 15 % fissions per initial metal atom (FIMA), a fast fluence of 3×1025 n/m2 (E > 0.18 MeV), and a time-average volume-average (TAVA) irradiation temperature of about 1020 °C. In order to simulate a core-conduction cool-down event, a temperature-versus-time profile having a peak temperature of 1700 °C was programmed into the FACS furnace controllers. Gaseous fission products (i.e., Kr-85) were carried to the Fission Gas Monitoring System (FGMS) by a helium sweep gas and captured in cold traps featuring online gamma counting. By the end of the test, a total of 3.9% of an average particle’s inventory of Kr-85 was detected in the FGMS traps. Such a low Kr-85 activity indicates that no TRISO failures (failure of all three TRISO layers) occurred during the test. If released from the compacts, condensable fission products (e.g., Ag-110m, Cs-134, Cs-137, Eu-154, Eu-155, and Sr-90) were collected on condensation plates fitted to the end of the cold finger in the FACS furnace. These condensation plates were then analyzed for fission products. In the second test, five loose UCO fuel kernels, obtained from deconsolidated particles from an irradiated AGR-1 compact, were heated in the FACS furnace to a peak temperature of 1600 °C. This test had two primary goals. First, the test was intended to assess the retention of fission products in loose kernels without the effects of the other TRISO layers (buffer, IPyC, SiC, and OPyC) or the graphitic matrix material comprising the compact. Second, this test served as an evaluation of the FACS fission product condensation plate collection efficiency.« less

Design and analysis of a nuclear reactor core for innovative small light water reactors

NASA Astrophysics Data System (ADS)

Soldatov, Alexey I.

In order to address the energy needs of developing countries and remote communities, Oregon State University has proposed the Multi-Application Small Light Water Reactor (MASLWR) design. In order to achieve five years of operation without refueling, use of 8% enriched fuel is necessary. This dissertation is focused on core design issues related with increased fuel enrichment (8.0%) and specific MASLWR operational conditions (such as lower operational pressure and temperature, and increased leakage due to small core). Neutron physics calculations are performed with the commercial nuclear industry tools CASMO-4 and SIMULATE-3, developed by Studsvik Scandpower Inc. The first set of results are generated from infinite lattice level calculations with CASMO-4, and focus on evaluation of the principal differences between standard PWR fuel and MASLWR fuel. Chapter 4-1 covers aspects of fuel isotopic composition changes with burnup, evaluation of kinetic parameters and reactivity coefficients. Chapter 4-2 discusses gadolinium self-shielding and shadowing effects, and subsequent impacts on power generation peaking and Reactor Control System shadowing. The second aspect of the research is dedicated to core design issues, such as reflector design (chapter 4-3), burnable absorber distribution and programmed fuel burnup and fuel use strategy (chapter 4-4). This section also includes discussion of the parameters important for safety and evaluation of Reactor Control System options for the proposed core design. An evaluation of the sensitivity of the proposed design to uncertainty in calculated parameters is presented in chapter 4-5. The results presented in this dissertation cover a new area of reactor design and operational parameters, and may be applicable to other small and large pressurized water reactor designs.

Control console replacement at the WPI Reactor. [Final report

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

Not Available

1992-12-31

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

Applicability of 100kWe-class of space reactor power systems to NASA manned space station missions

NASA Technical Reports Server (NTRS)

Silverman, S. W.; Willenberg, H. J.; Robertson, C.

1985-01-01

An assessment is made of a manned space station operating with sufficiently high power demands to require a multihundred kilowatt range electrical power system. The nuclear reactor is a competitor for supplying this power level. Load levels were selected at 150kWe and 300kWe. Interactions among the reactor electrical power system, the manned space station, the space transportation system, and the mission were evaluated. The reactor shield and the conversion equipment were assumed to be in different positions with respect to the station; on board, tethered, and on a free flyer platform. Mission analyses showed that the free flyer concept resulted in unacceptable costs and technical problems. The tethered reactor providing power to an electrolyzer for regenerative fuel cells on the space station, results in a minimum weight shield and can be designed to release the reactor power section so that it moves to a high altitude orbit where the decay period is at least 300 years. Placing the reactor on the station, on a structural boom is an attractive design, but heavier than the long tethered reactor design because of the shield weight for manned activity near the reactor.

Space Nuclear Power Plant Pre-Conceptual Design Report, For Information

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

B. Levine

2006-01-27

This letter transmits, for information, the Project Prometheus Space Nuclear Power Plant (SNPP) Pre-Conceptual Design Report completed by the Naval Reactors Prime Contractor Team (NRPCT). This report documents the work pertaining to the Reactor Module, which includes integration of the space nuclear reactor with the reactor radiation shield, energy conversion, and instrumentation and control segments. This document also describes integration of the Reactor Module with the Heat Rejection segment, the Power Conditioning and Distribution subsystem (which comprise the SNPP), and the remainder of the Prometheus spaceship.

Multi-phase CFD modeling of solid sorbent carbon capture system

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

Ryan, E. M.; DeCroix, D.; Breault, R.

2013-07-01

Computational fluid dynamics (CFD) simulations are used to investigate a low temperature post-combustion carbon capture reactor. The CFD models are based on a small scale solid sorbent carbon capture reactor design from ADA-ES and Southern Company. The reactor is a fluidized bed design based on a silica-supported amine sorbent. CFD models using both Eulerian–Eulerian and Eulerian–Lagrangian multi-phase modeling methods are developed to investigate the hydrodynamics and adsorption of carbon dioxide in the reactor. Models developed in both FLUENT® and BARRACUDA are presented to explore the strengths and weaknesses of state of the art CFD codes for modeling multi-phase carbon capturemore » reactors. The results of the simulations show that the FLUENT® Eulerian–Lagrangian simulations (DDPM) are unstable for the given reactor design; while the BARRACUDA Eulerian–Lagrangian model is able to simulate the system given appropriate simplifying assumptions. FLUENT® Eulerian–Eulerian simulations also provide a stable solution for the carbon capture reactor given the appropriate simplifying assumptions.« less

Multi-Phase CFD Modeling of Solid Sorbent Carbon Capture System

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

Ryan, Emily M.; DeCroix, David; Breault, Ronald W.

2013-07-30

Computational fluid dynamics (CFD) simulations are used to investigate a low temperature post-combustion carbon capture reactor. The CFD models are based on a small scale solid sorbent carbon capture reactor design from ADA-ES and Southern Company. The reactor is a fluidized bed design based on a silica-supported amine sorbent. CFD models using both Eulerian-Eulerian and Eulerian-Lagrangian multi-phase modeling methods are developed to investigate the hydrodynamics and adsorption of carbon dioxide in the reactor. Models developed in both FLUENT® and BARRACUDA are presented to explore the strengths and weaknesses of state of the art CFD codes for modeling multi-phase carbon capturemore » reactors. The results of the simulations show that the FLUENT® Eulerian-Lagrangian simulations (DDPM) are unstable for the given reactor design; while the BARRACUDA Eulerian-Lagrangian model is able to simulate the system given appropriate simplifying assumptions. FLUENT® Eulerian-Eulerian simulations also provide a stable solution for the carbon capture reactor given the appropriate simplifying assumptions.« less

Advancement of Systems Designs and Key Engineering Technologies for Materials Based Hydrogen Storage

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

van Hassel, Bart A.

UTRC lead the development of the Simulink Framework model that enables a comparison of different hydrogen storage systems on a common basis. The Simulink Framework model was disseminated on the www.HSECoE.org website that is hosted by NREL. UTRC contributed to a better understanding of the safety aspects of the proposed hydrogen storage systems. UTRC also participated in the Failure Mode and Effect Analysis of both the chemical- and the adsorbent-based hydrogen storage system during Phase 2 of the Hydrogen Storage Engineering Center of Excellence. UTRC designed a hydrogen storage system with a reversible metal hydride material in a compacted formmore » for light-duty vehicles with a 5.6 kg H2 storage capacity, giving it a 300 miles range. It contains a heat exchanger that enables efficient cooling of the metal hydride material during hydrogen absorption in order to meet the 3.3 minute refueling time target. It has been shown through computation that the kinetics of hydrogen absorption of Ti-catalyzed NaAlH4 was ultimately limiting the rate of hydrogen absorption to 85% of the material capacity in 3.3 minutes. An inverse analysis was performed in order to determine the material property requirements in order for a metal hydride based hydrogen storage system to meet the DOE targets. Work on metal hydride storage systems was halted after the Phase 1 to Phase 2 review due to the lack of metal hydride materials with the required material properties. UTRC contributed to the design of a chemical hydrogen storage system by developing an adsorbent for removing the impurity ammonia from the hydrogen gas, by developing a system to meter the transport of Ammonia Borane (AB) powder to a thermolysis reactor, and by developing a gas-liquid-separator (GLS) for the separation of hydrogen gas from AB slurry in silicone oil. Stripping impurities from hydrogen gas is essential for a long life of the fuel cell system on board of a vehicle. Work on solid transport of AB was halted after the Phase 1 to Phase 2 review in favor of studying the slurry-form of AB as it appeared to be difficult to transport a solid form of AB through the thermolysis reactor. UTRC demonstrated the operation of a compact GLS in the laboratory at a scale that would be required for the actual automotive application. The GLS met the targets for weight and volume. UTRC also reported about the unresolved issue associated with the high vapor pressure of fluids that could be used for making a slurry-form of AB. Work on the GLS was halted after the Phase 2 to Phase 3 review as the off-board regeneration efficiency of the spent AB was below the DOE target of 60%. UTRC contributed to the design of an adsorbent-based hydrogen storage system through measurements of the thermal conductivity of a compacted form of Metal Organic Framework (MOF) number 5 and through the development and sizing of a particulate filter. Thermal conductivity is important for the design of the modular adsorbent tank insert (MATI), as developed by Oregon State University (OSU), in order to enable a rapid refueling process. Stringent hydrogen quality requirements can only be met with an efficient particulate filtration system. UTRC developed a method to size the particulate filter by taking into account the effect of the pressure drop on the hydrogen adsorption process in the tank. UTRC raised awareness about the potential use of materials-based H2 storage systems in applications outside the traditional light-duty vehicle market segment by presenting at several conferences about niche application opportunities in Unmanned Aerial Vehicles (UAV), Autonomous Underwater Vehicles (AUV), portable power and others.« less

10 CFR 72.120 - General considerations.

Code of Federal Regulations, 2012 CFR

2012-01-01

... NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE General Design... reactor-related GTCC waste in an ISFSI or to store spent fuel, high-level radioactive waste, or reactor-related GTCC waste in an MRS must include the design criteria for the proposed storage installation. These...

10 CFR 72.120 - General considerations.

Code of Federal Regulations, 2011 CFR

2011-01-01

... NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE General Design... reactor-related GTCC waste in an ISFSI or to store spent fuel, high-level radioactive waste, or reactor-related GTCC waste in an MRS must include the design criteria for the proposed storage installation. These...

HORIZONTAL BOILING REACTOR SYSTEM

DOEpatents

Treshow, M.

1958-11-18

Reactors of the boiling water type are described wherein water serves both as the moderator and coolant. The reactor system consists essentially of a horizontal pressure vessel divided into two compartments by a weir, a thermal neutronic reactor core having vertical coolant passages and designed to use water as a moderator-coolant posltioned in one compartment, means for removing live steam from the other compartment and means for conveying feed-water and water from the steam compartment to the reactor compartment. The system further includes auxiliary apparatus to utilize the steam for driving a turbine and returning the condensate to the feed-water inlet of the reactor. The entire system is designed so that the reactor is self-regulating and has self-limiting power and self-limiting pressure features.

Potential Applications of Zeolite Membranes in Reaction Coupling Separation Processes

PubMed Central

Daramola, Michael O.; Aransiola, Elizabeth F.; Ojumu, Tunde V.

2012-01-01

Future production of chemicals (e.g., fine and specialty chemicals) in industry is faced with the challenge of limited material and energy resources. However, process intensification might play a significant role in alleviating this problem. A vision of process intensification through multifunctional reactors has stimulated research on membrane-based reactive separation processes, in which membrane separation and catalytic reaction occur simultaneously in one unit. These processes are rather attractive applications because they are potentially compact, less capital intensive, and have lower processing costs than traditional processes. Therefore this review discusses the progress and potential applications that have occurred in the field of zeolite membrane reactors during the last few years. The aim of this article is to update researchers in the field of process intensification and also provoke their thoughts on further research efforts to explore and exploit the potential applications of zeolite membrane reactors in industry. Further evaluation of this technology for industrial acceptability is essential in this regard. Therefore, studies such as techno-economical feasibility, optimization and scale-up are of the utmost importance.

A laser scanning system for metrology and viewing in ITER

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

Spampinato, P.T.; Barry, R.E.; Menon, M.M.

1996-05-01

The construction and operation of a next-generation fusion reactor will require metrology to achieve and verify precise alignment of plasma-facing components and inspection in the reactor vessel. The system must be compatible with the vessel environment of high gamma radiation (10{sup 4} Gy/h), ultra-high-vacuum (10{sup {minus}8} torr), and elevated temperature (200 C). The high radiation requires that the system be remotely deployed. A coherent frequency modulated laser radar-based system will be integrated with a remotely operated deployment mechanism to meet these requirements. The metrology/viewing system consists of a compact laser transceiver optics module which is linked through fiber optics tomore » the laser source and imaging units that are located outside of a biological shield. The deployment mechanism will be a mast-like positioning system. Radiation-damage tests will be conducted on critical sensor components at Oak Ridge National Laboratory to determine threshold damage levels and effects on data transmission. This paper identifies the requirements for International Thermonuclear Experimental Reactor metrology and viewing and describes a remotely operated precision ranging and surface mapping system.« less

Assessment for advanced fuel cycle options in CANDU

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

Morreale, A.C.; Luxat, J.C.; Friedlander, Y.

2013-07-01

The possible options for advanced fuel cycles in CANDU reactors including actinide burning options and thorium cycles were explored and are feasible options to increase the efficiency of uranium utilization and help close the fuel cycle. The actinide burning TRUMOX approach uses a mixed oxide fuel of reprocessed transuranic actinides from PWR spent fuel blended with natural uranium in the CANDU-900 reactor. This system reduced actinide content by 35% and decreased natural uranium consumption by 24% over a PWR once through cycle. The thorium cycles evaluated used two CANDU-900 units, a generator and a burner unit along with a drivermore » fuel feedstock. The driver fuels included plutonium reprocessed from PWR, from CANDU and low enriched uranium (LEU). All three cycles were effective options and reduced natural uranium consumption over a PWR once through cycle. The LEU driven system saw the largest reduction with a 94% savings while the plutonium driven cycles achieved 75% savings for PWR and 87% for CANDU. The high neutron economy, online fuelling and flexible compact fuel make the CANDU system an ideal reactor platform for many advanced fuel cycles.« less

Effect of TiO2 nanoparticles on aerobic granulation of algal-bacterial symbiosis system and nutrients removal from synthetic wastewater.

PubMed

Li, Bing; Huang, Wenli; Zhang, Chao; Feng, Sisi; Zhang, Zhenya; Lei, Zhongfang; Sugiura, Norio

2015-01-01

The influence of TiO2 nanoparticles (TiO2-NPs) (10-50mg/L) on aerobic granulation of algal-bacterial symbiosis system was investigated by using two identical sequencing batch reactors (SBRs). Although little adverse effect was observed on their nitritation efficiency (98-100% in both reactors), algal-bacterial granules in the control SBR (Rc) gradually lost stability mainly brought about by algae growth. TiO2-NPs addition to RT was found to enhance the granulation process achieving stable and compact algal-bacterial granules with remarkably improved nitratation thus little nitrite accumulation in RT when influent TiO2-NPs⩾30mg/L. Despite almost similar organics and phosphorus removals obtained in both reactors, the stably high nitratation efficiency in addition to much stable granular structure in RT suggests that TiO2-NPs addition might be a promising remedy for the long-term operation of algal-bacterial granular system, most probably attributable to the stimulated excretion of extracellular polymeric substances and less filamentous TM7. Copyright © 2015 Elsevier Ltd. All rights reserved.

Long Distance Reactor Antineutrino Flux Monitoring

NASA Astrophysics Data System (ADS)

Dazeley, Steven; Bergevin, Marc; Bernstein, Adam

2015-10-01

The feasibility of antineutrino detection as an unambiguous and unshieldable way to detect the presence of distant nuclear reactors has been studied. While KamLAND provided a proof of concept for long distance antineutrino detection, the feasibility of detecting single reactors at distances greater than 100 km has not yet been established. Even larger detectors than KamLAND would be required for such a project. Considerations such as light attenuation, environmental impact and cost, which favor water as a detection medium, become more important as detectors get larger. We have studied both the sensitivity of water based detection media as a monitoring tool, and the scientific impact such detectors might provide. A next generation water based detector may be able to contribute to important questions in neutrino physics, such as supernova neutrinos, sterile neutrino oscillations, and non standard electroweak interactions (using a nearby compact accelerator source), while also providing a highly sensitive, and inherently unshieldable reactor monitoring tool to the non proliferation community. In this talk I will present the predicted performance of an experimental non proliferation and high-energy physics program. Lawrence Livermore National Laboratory is operated by Lawrence Livermore National Security, LLC, for the U.S. Department of Energy, National Nuclear Security Administration under Contract DE-AC52-07NA27344. Release number LLNL-ABS-674192.

Preliminary Evaluation of Convective Heat Transfer in a Water Shield for a Surface Power Reactor

NASA Technical Reports Server (NTRS)

Pearson J. Boise; Reid, Robert S.

2007-01-01

As part of the Vision for Space Exploration, the end of the next decade will bring man back to the surface of the moon. A crucial issue for the establishment of human presence on the moon will be the availability of compact power sources. This presence could require greater than 10's of kWt's in follow on years. Nuclear reactors are well suited to meet the needs for power generation on the lunar or Martian surface. Radiation shielding is a key component of any surface power reactor system. Several competing concepts exist for lightweight, safe, robust shielding systems such as a water shield, lithium hydride (LiH), and boron carbide. Water offers several potential advantages, including reduced cost, reduced technical risk, and reduced mass. Water has not typically been considered for space reactor applications because of the need for gravity to fix the location of any vapor that could form radiation streaming paths. The water shield concept relies on the predictions of passive circulation of the shield water by natural convection to adequately cool the shield. This prediction needs to be experimentally evaluated, especially for shields with complex geometries. NASA Marshall Space Flight Center has developed the experience and facilities necessary to do this evaluation in its Early Flight Fission - Test Facility (EFF-TF).

Experimental Evaluation of a Water Shield for a Surface Power Reactor

NASA Technical Reports Server (NTRS)

Pearson, J. Boise; Reid, Robert S.

2006-01-01

As part of the Vision for Space Exploration the end of the next decade will bring man back to the surface of the moon. One of the most critical issues for the establishment of human presence on the moon will be the availability of compact power sources. The establishment of man on the moon will require power from greater than 10's of kWt's in follow on years. Nuclear reactors are extremely we11 suited to meet the needs for power generation on the lunar or Martian surface. reactor system. Several competing concepts exist for lightweight, safe, robust shielding systems such as a water shield, lithium hydride (LiH), Boron Carbide, and others. Water offers several potential advantages, including reduced cost, reduced technical risk, and reduced mass. Water has not typically been considered for space reactor applications because of the need for gravity to remove the potential for radiation streaming paths. The water shield concept relies on predictions of passive circulation of the shield water by natural convection to adequately cool the shield. This prediction needs to be experimentally evaluated, especially for shields with complex geometries. MSFC has developed the experience and fac necessary to do this evaluation in the Early Flight Fission - Test Facility (EFF-TF).

Analytical design and performance studies of nuclear furnace tests of small nuclear light bulb models

NASA Technical Reports Server (NTRS)

Latham, T. S.; Rodgers, R. J.

1972-01-01

Analytical studies were continued to identify the design and performance characteristics of a small-scale model of a nuclear light bulb unit cell suitable for testing in a nuclear furnace reactor. Emphasis was placed on calculating performance characteristics based on detailed radiant heat transfer analyses, on designing the test assembly for ease of insertion, connection, and withdrawal at the reactor test cell, and on determining instrumentation and test effluent handling requirements. In addition, a review of candidate test reactors for future nuclear light bulb in-reactor tests was conducted.

Design, prototyping, and testing of a compact superconducting double quarter wave crab cavity

NASA Astrophysics Data System (ADS)

Xiao, Binping; Alberty, Luis; Belomestnykh, Sergey; Ben-Zvi, Ilan; Calaga, Rama; Cullen, Chris; Capatina, Ofelia; Hammons, Lee; Li, Zenghai; Marques, Carlos; Skaritka, John; Verdu-Andres, Silvia; Wu, Qiong

2015-04-01

We proposed a novel design for a compact superconducting crab cavity with a double quarter wave (DQWCC) shape. After fabrication and surface treatments, this niobium proof-of-principle cavity was tested cryogenically in a vertical cryostat. The cavity is extremely compact yet has a low frequency of 400 MHz, an essential property for service in the Large Hadron Collider luminosity upgrade. The cavity's electromagnetic properties are well suited for this demanding task. The demonstrated deflecting voltage of 4.6 MV is well above the required 3.34 MV for a crab cavity in the future High Luminosity LHC. In this paper, we present the design, prototyping, and results from testing the DQWCC.

Neutronic design studies of a conceptual DCLL fusion reactor for a DEMO and a commercial power plant

NASA Astrophysics Data System (ADS)

Palermo, I.; Veredas, G.; Gómez-Ros, J. M.; Sanz, J.; Ibarra, A.

2016-01-01

Neutronic analyses or, more widely, nuclear analyses have been performed for the development of a dual-coolant He/LiPb (DCLL) conceptual design reactor. A detailed three-dimensional (3D) model has been examined and optimized. The design is based on the plasma parameters and functional materials of the power plant conceptual studies (PPCS) model C. The initial radial-build for the detailed model has been determined according to the dimensions established in a previous work on an equivalent simplified homogenized reactor model. For optimization purposes, the initial specifications established over the simplified model have been refined on the detailed 3D design, modifying material and dimension of breeding blanket, shield and vacuum vessel in order to fulfil the priority requirements of a fusion reactor in terms of the fundamental neutronic responses. Tritium breeding ratio, energy multiplication factor, radiation limits in the TF coils, helium production and displacements per atom (dpa) have been calculated in order to demonstrate the functionality and viability of the reactor design in guaranteeing tritium self-sufficiency, power efficiency, plasma confinement, and re-weldability and structural integrity of the components. The paper describes the neutronic design improvements of the DCLL reactor, obtaining results for both DEMO and power plant operational scenarios.

SP-100 reactor with Brayton conversion for lunar surface applications

NASA Technical Reports Server (NTRS)

Mason, Lee S.; Rodriguez, Carlos D.; Mckissock, Barbara I.; Hanlon, James C.; Mansfield, Brian C.

1992-01-01

Examined here is the potential for integrating Brayton-cycle power conversion with the SP-100 reactor for lunar surface power system applications. Two designs were characterized and modeled. The first design integrates a 100-kWe SP-100 Brayton power system with a lunar lander. This system is intended to meet early lunar mission power needs while minimizing on-site installation requirements. Man-rated radiation protection is provided by an integral multilayer, cylindrical lithium hydride/tungsten (LiH/W) shield encircling the reactor vessel. Design emphasis is on ease of deployment, safety, and reliability, while utilizing relatively near-term technology. The second design combines Brayton conversion with the SP-100 reactor in a erectable 550-kWe powerplant concept intended to satisfy later-phase lunar base power requirements. This system capitalizes on experience gained from operating the initial 100-kWe module and incorporates some technology improvements. For this system, the reactor is emplaced in a lunar regolith excavation to provide man-rated shielding, and the Brayton engines and radiators are mounted on the lunar surface and extend radially from the central reactor. Design emphasis is on performance, safety, long life, and operational flexibility.

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

DOE PAGES

Melin, Alexander M.; Kisner, Roger A.

2018-04-03

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

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

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

Melin, Alexander M.; Kisner, Roger A.

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

New Trends in the Design, Cost, Construction of the Modern School Building.

ERIC Educational Resources Information Center

Allied Masonry Council, McLean, VA.

The compact school, generally defined as a brick structure with a flexible interior and natural light admission of skylights, domes, clearstories and interior courtyards, emerged from the new educational programs. Evaluation of the compact school design includes--(1) appraisals and reactions to the physical environment, (2) explanations of the…

An inexpensive compact automatic camera system for wildlife research

Treesearch

William R. Danielson; Richard M. DeGraaf; Todd K. Fuller

1996-01-01

This paper describes the design, conversion, and deployment of a reliable, compact, automatic multiple-exposure photographic system that was used to photograph nest predation events. This system may be the most versatile yet described in the literature because of its simplicity, portability, and dependability. The system was very reliable because it was designed around...

A Project to Design and Build Compact Heat Exchangers

ERIC Educational Resources Information Center

Davis, Richard A.

2005-01-01

Students designed and manufactured compact, shell-and-tube heat exchangers in a project-based learning exercise integrated with our heat transfer course. The heat exchangers were constructed from common building materials available at home improvement centers. The cost of materials for a device was less than $20. The project gave students…

SGC Tests for Influence of Material Composition on Compaction Characteristic of Asphalt Mixtures

PubMed Central

Chen, Qun

2013-01-01

Compaction characteristic of the surface layer asphalt mixture (13-type gradation mixture) was studied using Superpave gyratory compactor (SGC) simulative compaction tests. Based on analysis of densification curve of gyratory compaction, influence rules of the contents of mineral aggregates of all sizes and asphalt on compaction characteristic of asphalt mixtures were obtained. SGC Tests show that, for the mixture with a bigger content of asphalt, its density increases faster, that there is an optimal amount of fine aggregates for optimal compaction and that an appropriate amount of mineral powder will improve workability of mixtures, but overmuch mineral powder will make mixtures dry and hard. Conclusions based on SGC tests can provide basis for how to adjust material composition for improving compaction performance of asphalt mixtures, and for the designed asphalt mixture, its compaction performance can be predicted through these conclusions, which also contributes to the choice of compaction schemes. PMID:23818830

SGC tests for influence of material composition on compaction characteristic of asphalt mixtures.

PubMed

Chen, Qun; Li, Yuzhi

2013-01-01

Compaction characteristic of the surface layer asphalt mixture (13-type gradation mixture) was studied using Superpave gyratory compactor (SGC) simulative compaction tests. Based on analysis of densification curve of gyratory compaction, influence rules of the contents of mineral aggregates of all sizes and asphalt on compaction characteristic of asphalt mixtures were obtained. SGC Tests show that, for the mixture with a bigger content of asphalt, its density increases faster, that there is an optimal amount of fine aggregates for optimal compaction and that an appropriate amount of mineral powder will improve workability of mixtures, but overmuch mineral powder will make mixtures dry and hard. Conclusions based on SGC tests can provide basis for how to adjust material composition for improving compaction performance of asphalt mixtures, and for the designed asphalt mixture, its compaction performance can be predicted through these conclusions, which also contributes to the choice of compaction schemes.

Emerging needs for mobile nuclear powerplants

NASA Technical Reports Server (NTRS)

Anderson, J. L.

1972-01-01

Incentives for broadening the present role of civilian nuclear power to include mobile nuclear power plants that are compact, lightweight, and safe are examined. Specifically discussed is the growing importance of: (1) a new international cargo transportation capability, and (2) the capability for development of resources in previously remote regions of the earth including the oceans and the Arctic. This report surveys present and potential systems (vehicles, remote stations, and machines) that would both provide these capabilities and require enough power to justify using mobile nuclear reactor power plants.

Advanced Catalysis Technologies: Lanthanum Cerium Manganese Hexaaluminate Combustion Catalysts for Flat Plate Reactor for Compact Steam Reformers

DTIC Science & Technology

2008-12-01

to decompose the urea into carbon dioxide and ammonia. This increased the pH and caused sol condensation. The mixture was calcined in air at 550°C...propane to carbon dioxide and water. Its high manganese content provides a higher intrinsic activity than the other catalysts and thus the lowest...lean natural gas turbines in order to reduce NOx emissions to reforming catalyst to convert diesel and kerosene to hydrogen rich gases. Unlike

Reforming of natural gas—hydrogen generation for small scale stationary fuel cell systems

NASA Astrophysics Data System (ADS)

Heinzel, A.; Vogel, B.; Hübner, P.

The reforming of natural gas to produce hydrogen for fuel cells is described, including the basic concepts (steam reforming or autothermal reforming) and the mechanisms of the chemical reactions. Experimental work has been done with a compact steam reformer, and a prototype of an experimental reactor for autothermal reforming was tested, both containing a Pt-catalyst on metallic substrate. Experimental results on the steam reforming system and a comparison of the steam reforming process with the autothermal process are given.

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

NASA Astrophysics Data System (ADS)

Qvist, Staffan Alexander

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

Mission-Based Funding Compacts with Public Universities. Go8 Backgrounder 6

ERIC Educational Resources Information Center

Group of Eight (NJ1), 2008

2008-01-01

This Go8 Backgrounder explores the possible uses of compacts in government financing of university activities, examines their potential costs and benefits, and outlines principles for their design and implementation. The Government has committed to compacts as an element of its future funding arrangements with public universities but has not yet…

Fluid mechanics relevant to flow through pretreatment of cellulosic biomass.

PubMed

Archambault-Léger, Véronique; Lynd, Lee R

2014-04-01

The present study investigates fluid mechanical properties of cellulosic feedstocks relevant to flow through (FT) pretreatment for biological conversion of cellulosic biomass. The results inform identifying conditions for which FT pretreatment can be implemented in a practical context. Measurements of pressure drop across packed beds, viscous compaction and water absorption are reported for milled and not milled sugarcane bagasse, switchgrass and poplar, and important factors impacting viscous flow are deduced. Using biomass knife-milled to pass through a 2mm sieve, the observed pressure drop was highest for bagasse, intermediate for switchgrass and lowest for poplar. The highest pressure drop was associated with the presence of more fine particles, greater viscous compaction and the degree of water absorption. Using bagasse without particle size reduction, the instability of the reactor during pretreatment above 140kg/m(3) sets an upper bound on the allowable concentration for continuous stable flow. Copyright © 2014. Published by Elsevier Ltd.

Research opportunities with compact accelerator-driven neutron sources

NASA Astrophysics Data System (ADS)

Anderson, I. S.; Andreani, C.; Carpenter, J. M.; Festa, G.; Gorini, G.; Loong, C.-K.; Senesi, R.

2016-10-01

Since the discovery of the neutron in 1932 neutron beams have been used in a very broad range of applications, As an aging fleet of nuclear reactor sources is retired the use of compact accelerator-driven neutron sources (CANS) is becoming more prevalent. CANS are playing a significant and expanding role in research and development in science and engineering, as well as in education and training. In the realm of multidisciplinary applications, CANS offer opportunities over a wide range of technical utilization, from interrogation of civil structures to medical therapy to cultural heritage study. This paper aims to provide the first comprehensive overview of the history, current status of operation, and ongoing development of CANS worldwide. The basic physics and engineering regarding neutron production by accelerators, target-moderator systems, and beam line instrumentation are introduced, followed by an extensive discussion of various evolving applications currently exploited at CANS.

Analysis of Technology for Compact Coherent Lidar

NASA Technical Reports Server (NTRS)

Amzajerdian, Farzin

1997-01-01

In view of the recent advances in the area of solid state and semiconductor lasers has created new possibilities for the development of compact and reliable coherent lidars for a wide range of applications. These applications include: Automated Rendezvous and Capture, wind shear and clear air turbulence detection, aircraft wake vortex detection, and automobile collision avoidance. The work performed by the UAH personnel under this Delivery Order, concentrated on design and analyses of a compact coherent lidar system capable of measuring range and velocity of hard targets, and providing air mass velocity data. The following is the scope of this work. a. Investigate various laser sources and optical signal detection configurations in support of a compact and lightweight coherent laser radar to be developed for precision range and velocity measurements of hard and fuzzy targets. Through interaction with MSFC engineers, the most suitable laser source and signal detection technique that can provide a reliable compact and lightweight laser radar design will be selected. b. Analyze and specify the coherent laser radar system configuration and assist with its optical and electronic design efforts. Develop a system design including its optical layout design. Specify all optical components and provide the general requirements of the electronic subsystems including laser beam modulator and demodulator drivers, detector electronic interface, and the signal processor. c. Perform a thorough performance analysis to predict the system measurement range and accuracy. This analysis will utilize various coherent laser radar sensitivity formulations and different target models.

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

Harmony, S.C.; Steiner, J.L.; Stumpf, H.J.

The PIUS advanced reactor is a 640-MWe pressurized water reactor developed by Asea Brown Boveri (ABB). A unique feature of the PIUS concept is the absence of mechanical control and shutdown rods. Reactivity is controlled by coolant boron concentration and the temperature of the moderator coolant. As part of the preapplication and eventual design certification process, advanced reactor applicants are required to submit neutronic and thermal-hydraulic safety analyses over a sufficient range of normal operation, transient conditions, and specified accident sequences. Los Alamos is supporting the US Nuclear Regulatory Commission`s preapplication review of the PIUS reactor. A fully one-dimensional modelmore » of the PIUS reactor has been developed for the Transient Reactor Analysis Code, TRACPF1/MOD2. Early in 1992, ABB submitted a Supplemental Information Package describing recent design modifications. An important feature of the PIUS Supplement design was the addition of an active scram system that will function for most transient and accident conditions. A one-dimensional Transient Reactor Analysis Code baseline calculation of the PIUS Supplement design were performed for a break in the main steam line at the outlet nozzle of the loop 3 steam generator. Sensitivity studies were performed to explore the robustness of the PIUS concept to severe off-normal conditions following a main steam line break. The sensitivity study results provide insights into the robustness of the design.« less

NRC ARDC Guidance Support Status Report

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

Holbrook, Mark R.

This report provides a summary that reflects the progress and status of proposed regulatory design criteria for advanced non-light water reactor (LWR) designs in accordance with the Level 3 milestone M3AT-17IN2001013 in work package AT-17IN200101. These criteria have been designated as advanced reactor design criteria (ARDC) and they provide guidance to future applicants for addressing the general design criteria (GDC) that are currently applied specifically to LWR designs. This report provides a summary of Phase 2 activities related to the various tasks associated with ARDC development and the subsequent development of ARDC regulatory guidance for sodium fast reactor (SFR) andmore » modular high-temperature gas-cooled reactor (HTGR) designs. Status Report Organization: Section 2 discusses the origin of the GDC and their application to LWRs. Section 3 addresses the objective of this initiative and how it benefits the advanced non-LWR reactor vendors. Section 4 discusses the scope and structure of the initiative. Section 5 provides background on the U.S. Department of Energy (DOE) ARDC team’s original development of the proposed ARDC that were submitted to the NRC for consideration. Section 6 provides a summary of recent ARDC Phase 2 activities. Appendices A through E document the DOE ARDC team’s public comments on various sections of the NRC’s draft regulatory guide DG–1330, “Guidance for Developing Principal Design Criteria for Non-Light Water Reactors.”« less

Autonomous Control of Space Nuclear Reactors

NASA Technical Reports Server (NTRS)

Merk, John

2013-01-01

Nuclear reactors to support future robotic and manned missions impose new and innovative technological requirements for their control and protection instrumentation. Long-duration surface missions necessitate reliable autonomous operation, and manned missions impose added requirements for failsafe reactor protection. There is a need for an advanced instrumentation and control system for space-nuclear reactors that addresses both aspects of autonomous operation and safety. The Reactor Instrumentation and Control System (RICS) consists of two functionally independent systems: the Reactor Protection System (RPS) and the Supervision and Control System (SCS). Through these two systems, the RICS both supervises and controls a nuclear reactor during normal operational states, as well as monitors the operation of the reactor and, upon sensing a system anomaly, automatically takes the appropriate actions to prevent an unsafe or potentially unsafe condition from occurring. The RPS encompasses all electrical and mechanical devices and circuitry, from sensors to actuation device output terminals. The SCS contains a comprehensive data acquisition system to measure continuously different groups of variables consisting of primary measurement elements, transmitters, or conditioning modules. These reactor control variables can be categorized into two groups: those directly related to the behavior of the core (known as nuclear variables) and those related to secondary systems (known as process variables). Reliable closed-loop reactor control is achieved by processing the acquired variables and actuating the appropriate device drivers to maintain the reactor in a safe operating state. The SCS must prevent a deviation from the reactor nominal conditions by managing limitation functions in order to avoid RPS actions. The RICS has four identical redundancies that comply with physical separation, electrical isolation, and functional independence. This architecture complies with the safety requirements of a nuclear reactor and provides high availability to the host system. The RICS is intended to interface with a host computer (the computer of the spacecraft where the reactor is mounted). The RICS leverages the safety features inherent in Earth-based reactors and also integrates the wide range neutron detector (WRND). A neutron detector provides the input that allows the RICS to do its job. The RICS is based on proven technology currently in use at a nuclear research facility. In its most basic form, the RICS is a ruggedized, compact data-acquisition and control system that could be adapted to support a wide variety of harsh environments. As such, the RICS could be a useful instrument outside the scope of a nuclear reactor, including military applications where failsafe data acquisition and control is required with stringent size, weight, and power constraints.

PRESSURIZED WATER REACTOR PROGRAM TECHNICAL PROGRESS REPORT FOR THE PERIOD MAY 5, 1955 TO JUNE 16, 1955

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

None

The current PWR plant and core parameters are listed. Resign requirements are briefly summarized for a radiation monitoring system, a fuel handling water system, a coolant purification system, an electrical power distribution system, and component shielding. Results of studies on thermal bowing and stressing of UO/sub 2/ are reported. A graph is presented of reactor power vs. reactor flow for various hot channel conditions. Development of U-- Mo and U-Nb alloys has been stopped because of the recent selection of UO/sub 2/ fuel material for the PWR core and blanket. The fabrication characteristics of UO/sub 2/ powders are being studied.more » Seamless Zircaloy-2 tubing has been tested to determine elastic limits, bursting pressures, and corrosion resistance. Fabrication techniques and tests for corrosion and defects in Zircaloy-clad U-Mo and UO/sub 2/ fuel rods are described. The preparation of UO/sub 2/ by various methods is being studied to determine which method produces a material most suitable for PWR fuel elements. The stability of UO/sub 2/ compacts in high temperature water and steam is being determined. Surface area and density measurements have been performed on samples of UO/sub 2/ powder prepared by various methods. Revelopment work on U-- Mo and U--Nb alloys has included studies of the effect on corrosion behavior of additions to the test water, additions to the alloys, homogenization of the alloys, annealing times, cladding, and fabrication techniques. Data are presented on relaxation in spring materials after exposure to a corrosive environment. Results are reported from loop and autoclave tests on fission product and crud deposition. Results of irradiation and corrosion testing of clad and unclad U--Mo and U-Nh alloys are described. The UO/sub 2/ irradiation program has included studies of dimensional changes, release of fission gases, and activity in the water surrounding the samples. A review of the methods of calculating reactor physics parameters has been completed, and the established procedures have been applied to determination of PWR reference design parameters. Critical experiments and primary loop shielding analyses are described. (D.E.B.)« less