Loss-of-flow-without-scram tests in Experimental Breeder Reactor-II and comparison with pretest predictions

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

Chang, L.K.; Mohr, D.; Planchon, H.P.

This article discusses a series of successful loss-of-flow-without-scram tests conducted in Experimental Breeder Reactor-II (EBR-II), a metal-fueled, sodium-cooled fast reactor. These May 1985 tests demonstrated the capability of the EBR to reduce reactor power passively during a loss of flow and to maintain reactor temperatures within bounds without any reliance on an active safety system. The tests were run from reduced power to ensure that temperatures could be maintained well below the fuel-clad eutectic temperature. Good agreement was found between selected test data and pretest predictions made with the EBR-II system analysis code NATDEMO and the hot channel analysis codemore » HOTCHAN. The article also discusses safety assessments of the tests as well as modifications required on the EBR-II reactor safety system for conducting required on the EBR-II reactor safety system for the conducting the tests.« less

Disposition of fuel elements from the Aberdeen and Sandia pulse reactor (SPR-II) assemblies

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

Mckerley, Bill; Bustamante, Jacqueline M; Costa, David A

2010-01-01

We describe the disposition of fuel from the Aberdeen (APR) and the Sandia Pulse Reactors (SPR-II) which were used to provide intense neutron bursts for radiation effects testing. The enriched Uranium - 10% Molybdenum fuel from these reactors was shipped to the Los Alamos National Laboratory (LANL) for size reduction prior to shipment to the Savannah River Site (SRS) for final disposition in the H Canyon facility. The Shipper/Receiver Agreements (SRA), intra-DOE interfaces, criticality safety evaluations, safety and quality requirements and key materials management issues required for the successful completion of this project will be presented. This work is inmore » support of the DOE Consolidation and Disposition program. Sandia National Laboratories (SNL) has operated pulse nuclear reactor research facilities for the Department of Energy since 1961. The Sandia Pulse Reactor (SPR-II) was a bare metal Godiva-type reactor. The reactor facilities have been used for research and development of nuclear and non-nuclear weapon systems, advanced nuclear reactors, reactor safety, simulation sources and energy related programs. The SPR-II was a fast burst reactor, designed and constructed by SNL that became operational in 1967. The SPR-ll core was a solid-metal fuel enriched to 93% {sup 235}U. The uranium was alloyed with 10 weight percent molybdenum to ensure the phase stabilization of the fuel. The core consisted of six fuel plates divided into two assemblies of three plates each. Figure 1 shows a cutaway diagram of the SPR-II Reactor with its decoupling shroud. NNSA charged Sandia with removing its category 1 and 2 special nuclear material by the end of 2008. The main impetus for this activity was based on NNSA Administrator Tom D'Agostino's six focus areas to reenergize NNSA's nuclear material consolidation and disposition efforts. For example, the removal of SPR-II from SNL to DAF was part of this undertaking. This project was in support of NNSA's efforts to

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

NASA Astrophysics Data System (ADS)

Brown, Juliana; Burgos, William

2010-05-01

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

Installation of automatic control at experimental breeder reactor II

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

Larson, H.A.; Booty, W.F.; Chick, D.R.

1985-08-01

The Experimental Breeder Reactor II (EBR-II) has been modified to permit automatic control capability. Necessary mechanical and electrical changes were made on a regular control rod position; motor, gears, and controller were replaced. A digital computer system was installed that has the programming capability for varied power profiles. The modifications permit transient testing at EBR-II. Experiments were run that increased power linearly as much as 4 MW/s (16% of initial power of 25 MW(thermal)/s), held power constant, and decreased power at a rate no slower than the increase rate. Thus the performance of the automatic control algorithm, the mechanical andmore » electrical control equipment, and the qualifications of the driver fuel for future power change experiments were all demonstrated.« less

Study of fusion product effects in field-reversed mirrors

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

Driemeyer, D.E.

1980-01-01

The effect of fusion products (fps) on Field-Reversed Mirror (FRM) reactor concepts has been evaluated through the development of two new computer models. The first code (MCFRM) treats fps as test particles in a fixed background plasma, which is represented as a fluid. MCFRM includes a Monte Carlo treatment of Coulomb scattering and thus provides an accurate treatment of fp behavior even at lower energies where pitch-angle scattering becomes important. The second code (FRMOD) is a steady-state, globally averaged, two-fluid (ion and electron), point model of the FRM plasma that incorporates fp heating and ash buildup values which are consistentmore » with the MCFRM calculations. These codes have been used extensively in the development of an advanced-fuel FRM reactor design (SAFFIRE). A Catalyzed-D version of the plant is also discussed along with an investigation of the steady-state energy distribution of fps in the FRM. User guides for the two computer codes are also included.« less

The high-resolution time-of-flight spectrometer TOFTOF

NASA Astrophysics Data System (ADS)

Unruh, Tobias; Neuhaus, Jürgen; Petry, Winfried

2007-10-01

The TOFTOF spectrometer is a multi-disc chopper time-of-flight spectrometer for cold neutrons at the research neutron source Heinz Maier-Leibnitz (FRM II). After five reactor cycles of routine operation the characteristics of the instrument are reported in this article. The spectrometer features an excellent signal to background ratio due to its remote position in the neutron guide hall, an elaborated shielding concept and an s-shaped curved primary neutron guide which acts i.a. as a neutron velocity filter. The spectrometer is fed with neutrons from the undermoderated cold neutron source of the FRM II leading to a total neutron flux of ˜1010n/cm2/s in the continuous white beam at the sample position distributed over a continuous and particularly broad wavelength spectrum. A high energy resolution is achieved by the use of high speed chopper discs made of carbon-fiber-reinforced plastic. In the combination of intensity, resolution and signal to background ratio the spectrometer offers new scientific prospects in the fields of inelastic and quasielastic neutron scattering.

Feasibility of Ground Testing a Moon and Mars Surface Power Reactor in EBR-II

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

Sheryl Morton; Carl Baily; Tom Hill

Ground testing of a surface fission power system would be necessary to verify the design and validate reactor performance to support safe and sustained human exploration of the Moon and Mars. The Idaho National Laboratory (INL) has several facilities that could be adapted to support a ground test. This paper focuses on the feasibility of ground testing at the Experimental Breeder Reactor II (EBR-II) facility and using other INL existing infrastructure to support such a test. This brief study concludes that the INL EBR-II facility and supporting infrastructure are a viable option for ground testing the surface power system. Itmore » provides features and attributes that offer advantages to locating and performing ground testing at this site, and it could support the National Aeronautics and Space Administration schedules for human exploration of the Moon. This study used the initial concept examined by the U.S. Department of Energy Inter-laboratory Design and Analysis Support Team for surface power, a lowtemperature, liquid-metal, three-loop Brayton power system. With some facility modification, the EBR-II can safely house a test chamber and perform long-term testing of the space reactor power system. The INL infrastructure is available to receive and provide bonded storage for special nuclear materials. Facilities adjacent to EBR-II can provide the clean room environment needed to assemble and store the test article assembly, disassemble the power system at the conclusion of testing, and perform posttest examination. Capability for waste disposal is also available at the INL.« less

Feasibility of Ground Testing a Moon and Mars Surface Power Reactor in EBR-II

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

Morton, Sheryl L.; Baily, Carl E.; Hill, Thomas J.

Ground testing of a surface fission power system would be necessary to verify the design and validate reactor performance to support safe and sustained human exploration of the Moon and Mars. The Idaho National Laboratory (INL) has several facilities that could be adapted to support a ground test. This paper focuses on the feasibility of ground testing at the Experimental Breeder Reactor II (EBR-II) facility and using other INL existing infrastructure to support such a test. This brief study concludes that the INL EBR-II facility and supporting infrastructure are a viable option for ground testing the surface power system. Itmore » provides features and attributes that offer advantages to locating and performing ground testing at this site, and it could support the National Aeronautics and Space Administration schedules for human exploration of the Moon. This study used the initial concept examined by the U.S. Department of Energy Inter-laboratory Design and Analysis Support Team for surface power, a low-temperature, liquid-metal, three-loop Brayton power system. With some facility modification, the EBR-II can safely house a test chamber and perform long-term testing of the space reactor power system. The INL infrastructure is available to receive and provide bonded storage for special nuclear materials. Facilities adjacent to EBR-II can provide the clean room environment needed to assemble and store the test article assembly, disassemble the power system at the conclusion of testing, and perform posttest examination. Capability for waste disposal is also available at the INL.« less

Feasibility of Ground Testing a Moon and Mars Surface Power Reactor in EBR-II

NASA Astrophysics Data System (ADS)

Morton, Sheryl L.; Baily, Carl E.; Hill, Thomas J.; Werner, James E.

2006-01-01

Ground testing of a surface fission power system would be necessary to verify the design and validate reactor performance to support safe and sustained human exploration of the Moon and Mars. The Idaho National Laboratory (INL) has several facilities that could be adapted to support a ground test. This paper focuses on the feasibility of ground testing at the Experimental Breeder Reactor II (EBR-II) facility and using other INL existing infrastructure to support such a test. This brief study concludes that the INL EBR-II facility and supporting infrastructure are a viable option for ground testing the surface power system. It provides features and attributes that offer advantages to locating and performing ground testing at this site, and it could support the National Aeronautics and Space Administration schedules for human exploration of the Moon. This study used the initial concept examined by the U.S. Department of Energy Inter-laboratory Design and Analysis Support Team for surface power, a low-temperature, liquid-metal, three-loop Brayton power system. With some facility modification, the EBR-II can safely house a test chamber and perform long-term testing of the space reactor power system. The INL infrastructure is available to receive and provide bonded storage for special nuclear materials. Facilities adjacent to EBR-II can provide the clean room environment needed to assemble and store the test article assembly, disassemble the power system at the conclusion of testing, and perform posttest examination. Capability for waste disposal is also available at the INL.

BESAFE II: Accident safety analysis code for MFE reactor designs

NASA Astrophysics Data System (ADS)

Sevigny, Lawrence Michael

The viability of controlled thermonuclear fusion as an alternative energy source hinges on its desirability from an economic and an environmental and safety standpoint. It is the latter which is the focus of this thesis. For magnetic fusion energy (MFE) devices, the safety concerns equate to a design's behavior during a worst-case accident scenario which is the loss of coolant accident (LOCA). In this dissertation, we examine the behavior of MFE devices during a LOCA and how this behavior relates to the safety characteristics of the machine; in particular the acute, whole-body, early dose. In doing so, we have produced an accident safety code, BESAFE II, now available to the fusion reactor design community. The Appendix constitutes the User's Manual for BESAFE II. The theory behind early dose calculations including the mobilization of activation products is presented in Chapter 2. Since mobilization of activation products is a strong function of temperature, it becomes necessary to calculate the thermal response of a design during a LOCA in order to determine the fraction of the activation products which are mobilized and thus become the source for the dose. The code BESAFE II is designed to determine the temperature history of each region of a design and determine the resulting mobilization of activation products at each point in time during the LOCA. The BESAFE II methodology is discussed in Chapter 4, followed by demonstrations of its use for two reference design cases: a PCA-Li tokamak and a SiC-He tokamak. Of these two cases, it is shown that the SiC-He tokamak is a better design from an accident safety standpoint than the PCA-Li tokamak. It is also found that doses derived from temperature-dependent mobilization data are different than those predicted using set mobilization categories such as those that involve Piet fractions. This demonstrates the need for more experimental data on fusion materials. The possibility for future improvements and modifications

Status report on the cold neutron source of the Garching neutron research facility FRM-II

NASA Astrophysics Data System (ADS)

Gobrecht, K.; Gutsmiedl, E.; Scheuer, A.

2002-01-01

The new high flux research reactor of the Technical University of Munich (Technische Universität München, TUM) will be equipped with a cold neutron source (CNS). The centre of the CNS will be located in the D 2O-reflector tank at 400 mm from the reactor core axis close to the thermal neutron flux maximum. The power of 4500 W developed by the nuclear heating in the 16 l of liquid deuterium at 25 K, and in the structures, is evacuated by a two-phase thermal siphon avoiding film boiling and flooding. The thermal siphon is a single tube with counter current flow. It is inclined by 10° from vertical, and optimised for a deuterium flow rate of 14 g/s. Optimisation of structure design and material, as well as safety aspects will be discussed. Those parts of the structure, which are exposed to high thermal neutron flux, are made from Zircaloy 4 and 6061T6 aluminium. Structure failure due to embrittlement of the structure material under high rapid neutron flux is very improbable during the lifetime of the CNS (30 years). Double, in pile even triple, containment with inert gas liner guarantees lack of explosion risk and of tritium contamination to the environment. Adding a few percent of hydrogen (H 2) to the deuterium (D 2) will improve the moderating properties of our relatively small moderator volume. Nearly all of the hydrogen is bound in the form of HD molecules. A long-term change of the hydrogen content in the deuterium is avoided by storing the mixture not in a gas buffer volume but as a metal hydride at low pressure. The metal hydride storage system contains two getter beds, one with 250 kg of LaCo 3Ni 2, the other one with 150 kg of ZrCo 0.8Ni 0.2. Each bed can take the total gas inventory, both beds together can absorb the total gas inventory in <6 min at a pressure <3 bar. The new reactor will have 13 beam tubes, 4 of which are looking at the CNS, including two for very cold (VCN) and ultra-cold neutron (UCN) production. The latter will take place in the

Applying design principles to fusion reactor configurations for propulsion in space

NASA Technical Reports Server (NTRS)

Carpenter, Scott A.; Deveny, Marc E.; Schulze, Norman R.

1993-01-01

The application of fusion power to space propulsion requires rethinking the engineering-design solution to controlled-fusion energy. Whereas the unit cost of electricity (COE) drives the engineering-design solution for utility-based fusion reactor configurations; initial mass to low earth orbit (IMLEO), specific jet power (kW(thrust)/kg(engine)), and reusability drive the engineering-design solution for successful application of fusion power to space propulsion. We applied three design principles (DP's) to adapt and optimize three candidate-terrestrial-fusion-reactor configurations for propulsion in space. The three design principles are: provide maximum direct access to space for waste radiation, operate components as passive radiators to minimize cooling-system mass, and optimize the plasma fuel, fuel mix, and temperature for best specific jet power. The three candidate terrestrial fusion reactor configurations are: the thermal barrier tandem mirror (TBTM), field reversed mirror (FRM), and levitated dipole field (LDF). The resulting three candidate space fusion propulsion systems have their IMLEO minimized and their specific jet power and reusability maximized. We performed a preliminary rating of these configurations and concluded that the leading engineering-design solution to space fusion propulsion is a modified TBTM that we call the Mirror Fusion Propulsion System (MFPS).

Adaptive control method for core power control in TRIGA Mark II reactor

NASA Astrophysics Data System (ADS)

Sabri Minhat, Mohd; Selamat, Hazlina; Subha, Nurul Adilla Mohd

2018-01-01

The 1MWth Reactor TRIGA PUSPATI (RTP) Mark II type has undergone more than 35 years of operation. The existing core power control uses feedback control algorithm (FCA). It is challenging to keep the core power stable at the desired value within acceptable error bands to meet the safety demand of RTP due to the sensitivity of nuclear research reactor operation. Currently, the system is not satisfied with power tracking performance and can be improved. Therefore, a new design core power control is very important to improve the current performance in tracking and regulate reactor power by control the movement of control rods. In this paper, the adaptive controller and focus on Model Reference Adaptive Control (MRAC) and Self-Tuning Control (STC) were applied to the control of the core power. The model for core power control was based on mathematical models of the reactor core, adaptive controller model, and control rods selection programming. The mathematical models of the reactor core were based on point kinetics model, thermal hydraulic models, and reactivity models. The adaptive control model was presented using Lyapunov method to ensure stable close loop system and STC Generalised Minimum Variance (GMV) Controller was not necessary to know the exact plant transfer function in designing the core power control. The performance between proposed adaptive control and FCA will be compared via computer simulation and analysed the simulation results manifest the effectiveness and the good performance of the proposed control method for core power control.

High-flux PGAA for milligram-weight samples

NASA Astrophysics Data System (ADS)

Kudejova, P.; Révay, Z.; Kleszcz, K.; Genreith, C.; Rossbach, M.

2015-05-01

With the high-intensity cold neutron flux available at the Prompt Gamma Activation Analysis (PGAA) instrument of the research reactor FRM II at the Heinz Maier-Leibnitz Zentrum (MLZ), samples with a weight of 1 mg or even less can be investigated for their elemental compositions using the (n,γ) capture reaction. In such cases, the typical sample packing material for PGAA experiments made of 25 μm thick PTFE foil (ca. 80 mg) can be orders of magnitude more massive than the sample weight itself. Proper choice of the packing material and measuring conditions are then of the highest importance [1].

Experimental power density distribution benchmark in the TRIGA Mark II reactor

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

Snoj, L.; Stancar, Z.; Radulovic, V.

2012-07-01

In order to improve the power calibration process and to benchmark the existing computational model of the TRIGA Mark II reactor at the Josef Stefan Inst. (JSI), a bilateral project was started as part of the agreement between the French Commissariat a l'energie atomique et aux energies alternatives (CEA) and the Ministry of higher education, science and technology of Slovenia. One of the objectives of the project was to analyze and improve the power calibration process of the JSI TRIGA reactor (procedural improvement and uncertainty reduction) by using absolutely calibrated CEA fission chambers (FCs). This is one of the fewmore » available power density distribution benchmarks for testing not only the fission rate distribution but also the absolute values of the fission rates. Our preliminary calculations indicate that the total experimental uncertainty of the measured reaction rate is sufficiently low that the experiments could be considered as benchmark experiments. (authors)« less

DIissolution of low enriched uranium from the experimental breeder reactor-II fuel stored at the Idaho National Laboratory

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

Daniel, G.; Rudisill, T.; Almond, P.

The Idaho National Laboratory (INL) is actively engaged in the development of electrochemical processing technology for the treatment of fast reactor fuels using irradiated fuel from the Experimental Breeder Reactor-II (EBR-II) as the primary test material. The research and development (R&D) activities generate a low enriched uranium (LEU) metal product from the electrorefining of the EBR-II fuel and the subsequent consolidation and removal of chloride salts by the cathode processor. The LEU metal ingots from past R&D activities are currently stored at INL awaiting disposition. One potential disposition pathway is the shipment of the ingots to the Savannah River Sitemore » (SRS) for dissolution in H-Canyon. Carbon steel cans containing the LEU metal would be loaded into reusable charging bundles in the H-Canyon Crane Maintenance Area and charged to the 6.4D or 6.1D dissolver. The LEU dissolution would be accomplished as the final charge in a dissolver batch (following the dissolution of multiple charges of spent nuclear fuel (SNF)). The solution would then be purified and the 235U enrichment downblended to allow use of the U in commercial reactor fuel. To support this potential disposition path, the Savannah River National Laboratory (SRNL) developed a dissolution flowsheet for the LEU using samples of the material received from INL.« less

Fe(II) oxidation during acid mine drainage neutralization in a pilot-scale Sequencing Batch Reactor.

PubMed

Zvimba, J N; Mathye, M; Vadapalli, V R K; Swanepoel, H; Bologo, L

2013-01-01

This study investigated Fe(II) oxidation during acid mine drainage (AMD) neutralization using CaCO3 in a pilot-scale Sequencing Batch Reactor (SBR) of hydraulic retention time (HRT) of 90 min and sludge retention time (SRT) of 360 min in the presence of air. The removal kinetics of Fe(II), of initial concentration 1,033 ± 0 mg/L, from AMD through oxidation to Fe(III) was observed to depend on both pH and suspended solids, resulting in Fe(II) levels of 679 ± 32, 242 ± 64, 46 ± 16 and 28 ± 0 mg/L recorded after cycles 1, 2, 3 and 4 respectively, with complete Fe(II) oxidation only achieved after complete neutralization of AMD. Generally, it takes 30 min to completely oxidize Fe(II) during cycle 4, suggesting that further optimization of SBR operation based on both pH and suspended solids manipulation can result in significant reduction of the number of cycles required to achieve acceptable Fe(II) oxidation for removal as ferric hydroxide. Overall, complete removal of Fe(II) during AMD neutralization is attractive as it promotes recovery of better quality waste gypsum, key to downstream gypsum beneficiation for recovery of valuables, thereby enabling some treatment-cost recovery and prevention of environmental pollution from dumping of sludge into landfills.

Radio frequency elevator for a pulsed positron beam

NASA Astrophysics Data System (ADS)

Dickmann, Marcel; Mitteneder, Johannes; Kögel, Gottfried; Egger, Werner; Sperr, Peter; Ackermann, Ulrich; Piochacz, Christian; Dollinger, Günther

2016-06-01

An elevator increases the potential energy of a particle beam with respect to ground potential without any alteration of kinetic energy and other beam parameters. This elevator is necessary for the implementation of the Munich Scanning Positron Microscope (SPM) at the intense positron source NEPOMUC at the research reactor FRM II in Munich. The principles of the rf elevator for pure electrostatically guided positrons are described. Measurements of beam quality behind the elevator are reported, which confirm that after the implementation of elevator and SPM at NEPOMUC the SPM can be operated at a considerably improved resolution (~ 0.3 μm) and event rate (~3.7 kHz) compared to the laboratory based β+-source.

Corrections on energy spectrum and scatterings for fast neutron radiography at NECTAR facility

NASA Astrophysics Data System (ADS)

Liu, Shu-Quan; Bücherl, Thomas; Li, Hang; Zou, Yu-Bin; Lu, Yuan-Rong; Guo, Zhi-Yu

2013-11-01

Distortions caused by the neutron spectrum and scattered neutrons are major problems in fast neutron radiography and should be considered for improving the image quality. This paper puts emphasis on the removal of these image distortions and deviations for fast neutron radiography performed at the NECTAR facility of the research reactor FRM- II in Technische Universität München (TUM), Germany. The NECTAR energy spectrum is analyzed and established to modify the influence caused by the neutron spectrum, and the Point Scattered Function (PScF) simulated by the Monte-Carlo program MCNPX is used to evaluate scattering effects from the object and improve image quality. Good analysis results prove the sound effects of the above two corrections.

EBR-II Data Digitization

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

Yoon, Su-Jong; Rabiti, Cristian; Sackett, John

2014-08-01

1. Objectives To produce a validation database out of those recorded signals it will be necessary also to identify the documents need to reconstruct the status of reactor at the time of the beginning of the recordings. This should comprehends the core loading specification (assemblies type and location and burn-up) along with this data the assemblies drawings and the core drawings will be identified. The first task of the project will be identify the location of the sensors, with respect the reactor plant layout, and the physical quantities recorded by the Experimental Breeder Reactor-II (EBR-II) data acquisition system. This firstmore » task will allow guiding and prioritizing the selection of drawings needed to numerically reproduce those signals. 1.1 Scopes and Deliverables The deliverables of this project are the list of sensors in EBR-II system, the identification of storing location of those sensors, identification of a core isotopic composition at the moment of the start of system recording. Information of the sensors in EBR-II reactor system was summarized from the EBR-II system design descriptions listed in Section 1.2.« less

Technical Information on the Carbonation of the EBR-II Reactor, Summary Report Part 2: Application to EBR-II Primary Sodium System and Related Systems

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

Steven R. Sherman; Collin J. Knight

Residual sodium is defined as sodium metal that remains behind in pipes, vessels, and tanks after the bulk sodium metal has been melted and drained from such components. The residual sodium has the same chemical properties as bulk sodium, and differs from bulk sodium only in the thickness of the sodium deposit. Typically, sodium is considered residual when the thickness of the deposit is less than 5-6 cm. This residual sodium must be removed or deactivated when a pipe, vessel, system, or entire reactor is permanently taken out of service, in order to make the component or system safer and/ormore » to comply with decontamination and decomissioning regulations. As an alternative to the established residual sodium deactivation techniques (steam-and-nitrogen, wet vapor nitrogen, etc.), a technique involving the use of moisture and carbon dioxide has been developed. With this technique, sodium metal is converted into sodium bicarbonate by reacting it with humid carbon dioxide. Hydrogen is emitted as a by-product. This technique was first developed in the laboratory by exposing sodium samples to humidifed carbon dioxide under controlled conditions, and then demonstrated on a larger scale by treating residual sodium within the Experimental Breeder Reactor II (EBR-II) secondary cooling system, followed by the primary cooling system, respectively. The EBR-II facility is located at the Idaho National Laboratory (INL) in southeastern Idaho, USA. This report is Part 2 of a two-part report. This second report provides a supplement to the first report and describes the application of the humdidified carbon dioxide technique ("carbonation") to the EBR-II primary tank, primary cover gas systems, and the intermediate heat exchanger. Future treatment plans are also provided.« less

Technical Information on the Carbonation of the EBR-II Reactor, Summary Report Part 1: Laboratory Experiments and Application to EBR-II Secondary Sodium System

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

Steven R. Sherman

Residual sodium is defined as sodium metal that remains behind in pipes, vessels, and tanks after the bulk sodium metal has been melted and drained from such components. The residual sodium has the same chemical properties as bulk sodium, and differs from bulk sodium only in the thickness of the sodium deposit. Typically, sodium is considered residual when the thickness of the deposit is less than 5-6 cm. This residual sodium must be removed or deactivated when a pipe, vessel, system, or entire reactor is permanently taken out of service, in order to make the component or system safer and/ormore » to comply with decommissioning regulations. As an alternative to the established residual sodium deactivation techniques (steam-and-nitrogen, wet vapor nitrogen, etc.), a technique involving the use of moisture and carbon dioxide has been developed. With this technique, sodium metal is converted into sodium bicarbonate by reacting it with humid carbon dioxide. Hydrogen is emitted as a by-product. This technique was first developed in the laboratory by exposing sodium samples to humidified carbon dioxide under controlled conditions, and then demonstrated on a larger scale by treating residual sodium within the Experimental Breeder Reactor II (EBR-II) secondary cooling system, followed by the primary cooling system, respectively. The EBR-II facility is located at the Idaho National Laboratory (INL) in southeastern Idaho, U.S.A. This report is Part 1 of a two-part report. It is divided into three sections. The first section describes the chemistry of carbon dioxide-water-sodium reactions. The second section covers the laboratory experiments that were conducted in order to develop the residual sodium deactivation process. The third section discusses the application of the deactivation process to the treatment of residual sodium within the EBR-II secondary sodium cooling system. Part 2 of the report, under separate cover, describes the application of the technique to

Analysis of JSI TRIGA MARK II reactor physical parameters calculated with TRIPOLI and MCNP.

PubMed

Henry, R; Tiselj, I; Snoj, L

2015-03-01

New computational model of the JSI TRIGA Mark II research reactor was built for TRIPOLI computer code and compared with existing MCNP code model. The same modelling assumptions were used in order to check the differences of the mathematical models of both Monte Carlo codes. Differences between the TRIPOLI and MCNP predictions of keff were up to 100pcm. Further validation was performed with analyses of the normalized reaction rates and computations of kinetic parameters for various core configurations. Copyright © 2014 Elsevier Ltd. All rights reserved.

78 FR 35990 - All Operating Boiling-Water Reactor Licensees With Mark I And Mark II Containments; Docket Nos...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-06-14

... NUCLEAR REGULATORY COMMISSION [NRC-2013-0128] All Operating Boiling-Water Reactor Licensees With Mark I And Mark II Containments; Docket Nos. (As Shown In Attachment 1), License Nos. (As Shown In Attachment 1), EA-13-109; Order Modifying Licenses With Regard to Reliable Hardened Containment Vents Capable of Operation Under Severe Accident...

Fabrication and testing of a 4-node micro-pocket fission detector array for the Kansas State University TRIGA Mk. II research nuclear reactor

NASA Astrophysics Data System (ADS)

Reichenberger, Michael A.; Nichols, Daniel M.; Stevenson, Sarah R.; Swope, Tanner M.; Hilger, Caden W.; Unruh, Troy C.; McGregor, Douglas S.; Roberts, Jeremy A.

2017-08-01

Advancements in nuclear reactor core modeling and computational capability have encouraged further development of in-core neutron sensors. Micro-Pocket Fission Detectors (MPFDs) have been fabricated and tested previously, but successful testing of these prior detectors was limited to single-node operation with specialized designs. Described in this work is a modular, four-node MPFD array fabricated and tested at Kansas State University (KSU). The four sensor nodes were equally spaced to span the length of the fuel-region of the KSU TRIGA Mk. II research nuclear reactor core. The encapsulated array was filled with argon gas, serving as an ionization medium in the small cavities of the MPFDs. The unified design improved device ruggedness and simplified construction over previous designs. A 0.315-in. (8-mm) penetration in the upper grid plate of the KSU TRIGA Mk. II research nuclear reactor was used to deploy the array between fuel elements in the core. The MPFD array was coupled to an electronic support system which has been developed to support pulse-mode operation. Neutron-induced pulses were observed on all four sensor channels. Stable device operation was confirmed by testing under steady-state reactor conditions. Each of the four sensors in the array responded to changes in reactor power between 10 kWth and full power (750 kWth). Reactor power transients were observed in real-time including positive transients with periods of 5, 15, and 30 s. Finally, manual reactor power oscillations were observed in real-time.

Generation of an activation map for decommissioning planning of the Berlin Experimental Reactor-II

NASA Astrophysics Data System (ADS)

Lapins, Janis; Guilliard, Nicole; Bernnat, Wolfgang

2017-09-01

The BER-II is an experimental facility with 10 MW that was operated since 1974. Its planned operation will end in 2019. To support the decommissioning planning, a map with the overall distribution of relevant radionuclides has to be created according to the state of the art. In this paper, a procedure to create these 3-d maps using a combination of MCNP and deterministic methods is presented. With this approach, an activation analysis is performed for the whole reactor geometry including the most remote parts of the concrete shielding.

Feasibility of processing the experimental breeder reactor-II driver fuel from the Idaho National Laboratory through Savannah River Site's H-Canyon facility

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

Magoulas, V. E.

Savannah River National Laboratory (SRNL) was requested to evaluate the potential to receive and process the Idaho National Laboratory (INL) uranium (U) recovered from the Experimental Breeder Reactor II (EBR-II) driver fuel through the Savannah River Site’s (SRS) H-Canyon as a way to disposition the material. INL recovers the uranium from the sodium bonded metallic fuel irradiated in the EBR-II reactor using an electrorefining process. There were two compositions of EBR-II driver fuel. The early generation fuel was U-5Fs, which consisted of 95% U metal alloyed with 5% noble metal elements “fissium” (2.5% molybdenum, 2.0% ruthenium, 0.3% rhodium, 0.1% palladium,more » and 0.1% zirconium), while the later generation was U-10Zr which was 90% U metal alloyed with 10% zirconium. A potential concern during the H-Canyon nitric acid dissolution process of the U metal containing zirconium (Zr) is the explosive behavior that has been reported for alloys of these materials. For this reason, this evaluation was focused on the ability to process the lower Zr content materials, the U-5Fs material.« less

Study of the surface contamination of copper with the improved positron annihilation-induced Auger electron spectrometer at NEPOMUC

NASA Astrophysics Data System (ADS)

Mayer, J.; Hugenschmidt, C.; Schreckenbach, K.

2008-10-01

The high intensity positron source NEPOMUC at the FRM-II in Munich enables measurement times for positron annihilation-induced Auger electron spectroscopy (PAES) of only 2.4 h/spectrum, in contrast to usual lab beams with measurement times up to several days. The high electron background due to surrounding experiments in the experimental hall of the FRM-II has been eliminated and hence background free experiments have become possible. Due to this, the signal to noise ratio has been enhanced to 4.5:1, compared to 1:3 with EAES. In addition, a long-term measurement has been performed in order to observe the contamination of a polycrystalline copper foil at 150 °C.

FERRET-SAND II physics-dosimetry analysis for N Reactor Pressure Tubes 2954, 3053 and 1165 using a WIMS calculated input spectrum

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

McElroy, W.N.; Kellogg, L.S.; Matsumoto, W.Y.

1988-05-01

This report is in response to a request from Westinghouse Hanford Company (WHC) that the PNL National Dosimetry Center (NDC) perform physics-dosimetry analyses (E > MeV) for N Reactor Pressure Tubes 2954 and 3053. As a result of these analyses, and recommendations for additional studies, two physics-dosimetry re-evaluations for Pressure Tube 1165 were also accomplished. The primary objective of Pacific Northwest Laboratories' (PNL) National Dosimetry Center (NDC) physics-dosimetry work for N Reactor was to provide FERRET-SAND II physics-dosimetry results to assist in the assessment of neutron radiation-induced changes in the physical and mechanical properties of N Reactor pressure tubes. 15more » refs., 6 figs., 5 tabs.« less

Search for the permanent electric dipole moment of 129Xe

NASA Astrophysics Data System (ADS)

Sachdeva, Natasha; Chupp, Timothy; Gong, Fei; Babcock, Earl; Salhi, Zahir; Burghoff, Martin; Fan, Isaac; Killian, Wolfgang; Knappe-Grüneberg, Silvia; Schabel, Allard; Seifert, Frank; Trahms, Lutz; Voigt, Jens; Degenkolb, Skyler; Fierlinger, Peter; Krägeloh, Eva; Lins, Tobias; Marino, Michael; Meinel, Jonas; Niessen, Benjamin; Stuiber, Stefan; Terrano, William; Kuchler, Florian; Singh, Jaideep

2017-09-01

CP-violation in Beyond-the-Standard-Model physics, necessary to explain the baryon asymmetry, gives rise to permanent electric dipole moments (EDMs). EDM measurements of the neutron, electron, paramagnetic and diamagnetic atoms constrain CP-violating parameters. The current limit for the 129Xe EDM is 6 ×10-27 e . cm (95 % CL). The HeXeEDM experiment at FRM-II (Munich Research Reactor) and BMSR-2 (Berlin Magnetically Shielded Room) uses a stable magnetic field in a magnetically shielded room and 3He comagnetometer with potential to improve the limit by two orders of magnitude. Polarized 3He and 129Xe free precession is detected with SQUID magnetometers in the presence of applied electric and magnetic fields. Conclusions from recent measurements will be presented.

Implementation of k0-INAA standardisation at ITU TRIGA Mark II research reactor, Turkey based on k0-IAEA software

NASA Astrophysics Data System (ADS)

Esen, Ayse Nur; Haciyakupoglu, Sevilay

2016-02-01

The purpose of this study is to test the applicability of k0-INAA method at the Istanbul Technical University TRIGA Mark II research reactor. The neutron spectrum parameters such as epithermal neutron flux distribution parameter (α), thermal to epithermal neutron flux ratio (f) and thermal neutron flux (φth) were determined at the central irradiation channel of the ITU TRIGA Mark II research reactor using bare triple-monitor method. HPGe detector calibrations and calculations were carried out by k0-IAEA software. The α, f and φth values were calculated to be -0.009, 15.4 and 7.92·1012 cm-2 s-1, respectively. NIST SRM 1633b coal fly ash and intercomparison samples consisting of clay and sandy soil samples were used to evaluate the validity of the method. For selected elements, the statistical evaluation of the analysis results was carried out by z-score test. A good agreement between certified/reported and experimental values was obtained.

Adaptation of anaerobic cultures of E scherichia coli  K‐12 in response to environmental trimethylamine‐N‐oxide

PubMed Central

Denby, Katie J.; Rolfe, Matthew D.; Crick, Ellen; Sanguinetti, Guido; Poole, Robert K.

2015-01-01

Summary Systematic analyses of transcriptional and metabolic changes occurring when E scherichia coli  K‐12 switches from fermentative growth to anaerobic respiratory growth with trimethylamine‐N‐oxide (TMAO) as the terminal electron acceptor revealed: (i) the induction of torCAD, but not genes encoding alternative TMAO reductases; (ii) transient expression of frmRAB, encoding formaldehyde dehydrogenase; and (iii) downregulation of copper resistance genes. Simultaneous inference of 167 transcription factor (TF) activities implied that transcriptional re‐programming was mediated by 20 TFs, including the transient inactivation of the two‐component system ArcBA; a prediction validated by direct measurement of phosphorylated ArcA. Induction of frmRAB, detection of dimethylamine in culture medium and formaldehyde production when cell‐free extracts were incubated with TMAO suggested the presence of TMAO demethylase activity. Accordingly, the viability of an frmRAB mutant was compromised upon exposure to TMAO. Downregulation of genes involved in copper resistance could be accounted for by TMAO inhibition of Cu(II) reduction. The simplest interpretation of the data is that during adaptation to the presence of environmental TMAO, anaerobic fermentative cultures of E . coli respond by activating the TorTSR regulatory system with consequent induction of TMAO reductase activity, resulting in net oxidation of menaquinone and inhibition of Cu(II) reduction, responses that are sensed by ArcBA and CusRS respectively. PMID:25471524

Instrument developments and recent scientific highlights at the J-NSE

NASA Astrophysics Data System (ADS)

Ivanova, Oxana; Pasini, Stefano; Monkenbusch, Michael; Holderer, Olaf

2017-06-01

The J-NSE neutron spin echo spectrometer faces now 10 years of successful user operation at the FRM II research reactor at the Heinz Maier-Leibnitz Zentrum (MLZ). We present scientific highlights and instrumental developments of the last decade, for example the development of grazing incidence neutron spin echo spectroscopy (GINSES) at the J-NSE and investigations of the dynamics at solid-liquid interfaces with this new option. Polymers in confinement have been a prominent topic, as well as the internal dynamics of proteins. The scientific questions also triggered instrumental developments such as a new polarizer and a new neutron guide concept. Finally, the future of the J-NSE will be addressed with a short presentation of the current upgrade program with superconducting main coils with reduced intrinsic field integral inhomogeneity.

Importance of reduced sulfur for the equilibrium chemistry and kinetics of Fe(II), Co(II) and Ni(II) supplemented to semi-continuous stirred tank biogas reactors fed with stillage.

PubMed

Shakeri Yekta, Sepehr; Lindmark, Amanda; Skyllberg, Ulf; Danielsson, Asa; Svensson, Bo H

2014-03-30

The objective of the present study was to assess major chemical reactions and chemical forms contributing to solubility and speciation of Fe(II), Co(II), and Ni(II) during anaerobic digestion of sulfur (S)-rich stillage in semi-continuous stirred tank biogas reactors (SCSTR). These metals are essential supplements for efficient and stable performance of stillage-fed SCSTR. In particular, the influence of reduced inorganic and organic S species on kinetics and thermodynamics of the metals and their partitioning between aqueous and solid phases were investigated. Solid phase S speciation was determined by use of S K-edge X-ray absorption near-edge spectroscopy. Results demonstrated that the solubility and speciation of supplemented Fe were controlled by precipitation of FeS(s) and formation of the aqueous complexes of Fe-sulfide and Fe-thiol. The relatively high solubility of Co (∼ 20% of total Co content) was attributed to the formation of compounds other than Co-sulfide and Co-thiol, presumably of microbial origin. Nickel had lower solubility than Co and its speciation was regulated by interactions with FeS(s) (e.g. co-precipitation, adsorption, and ion substitution) in addition to precipitation/dissolution of discrete NiS(s) phase and formation of aqueous Ni-sulfide complexes. Copyright © 2014 Elsevier B.V. All rights reserved.

Paving the Road for Modern Particle Therapy - What Can We Learn from the Experience Gained with Fast Neutron Therapy in Munich?

PubMed

Specht, Hanno M; Neff, Teresa; Reuschel, Waltraud; Wagner, Franz M; Kampfer, Severin; Wilkens, Jan J; Petry, Winfried; Combs, Stephanie E

2015-01-01

While neutron therapy was a highly topical subject in the 70s and 80s, today there are only a few remaining facilities offering fast neutron therapy (FNT). Nevertheless, up to today more than 30,000 patients were treated with neutron therapy. For some indications like salivary gland tumors and malignant melanoma, there is clinical evidence that the addition of FNT leads to superior local control compared to photon treatment alone. FNT was available in Munich from 1985 until 2000 at the Reactor Neutron Therapy (RENT) facility. Patient treatment continued at the new research reactor FRM II in 2007 under improved treatment conditions, and today it can still be offered to selected patients as an individual treatment option. As there is a growing interest in high-linear energy transfer (LET) therapy with new hadron therapy centers emerging around the globe, the clinical data generated by neutron therapy might help to develop biologically driven treatment planning algorithms. Also FNT might experience its resurgence as a combinational partner of modern immunotherapies.

Paving the Road for Modern Particle Therapy – What Can We Learn from the Experience Gained with Fast Neutron Therapy in Munich?

PubMed Central

Specht, Hanno M.; Neff, Teresa; Reuschel, Waltraud; Wagner, Franz M.; Kampfer, Severin; Wilkens, Jan J.; Petry, Winfried; Combs, Stephanie E.

2015-01-01

While neutron therapy was a highly topical subject in the 70s and 80s, today there are only a few remaining facilities offering fast neutron therapy (FNT). Nevertheless, up to today more than 30,000 patients were treated with neutron therapy. For some indications like salivary gland tumors and malignant melanoma, there is clinical evidence that the addition of FNT leads to superior local control compared to photon treatment alone. FNT was available in Munich from 1985 until 2000 at the Reactor Neutron Therapy (RENT) facility. Patient treatment continued at the new research reactor FRM II in 2007 under improved treatment conditions, and today it can still be offered to selected patients as an individual treatment option. As there is a growing interest in high-linear energy transfer (LET) therapy with new hadron therapy centers emerging around the globe, the clinical data generated by neutron therapy might help to develop biologically driven treatment planning algorithms. Also FNT might experience its resurgence as a combinational partner of modern immunotherapies. PMID:26640777

Adaptation of anaerobic cultures of Escherichia coli K-12 in response to environmental trimethylamine-N-oxide.

PubMed

Denby, Katie J; Rolfe, Matthew D; Crick, Ellen; Sanguinetti, Guido; Poole, Robert K; Green, Jeffrey

2015-07-01

Systematic analyses of transcriptional and metabolic changes occurring when Escherichia coli K-12 switches from fermentative growth to anaerobic respiratory growth with trimethylamine-N-oxide (TMAO) as the terminal electron acceptor revealed: (i) the induction of torCAD, but not genes encoding alternative TMAO reductases; (ii) transient expression of frmRAB, encoding formaldehyde dehydrogenase; and (iii) downregulation of copper resistance genes. Simultaneous inference of 167 transcription factor (TF) activities implied that transcriptional re-programming was mediated by 20 TFs, including the transient inactivation of the two-component system ArcBA; a prediction validated by direct measurement of phosphorylated ArcA. Induction of frmRAB, detection of dimethylamine in culture medium and formaldehyde production when cell-free extracts were incubated with TMAO suggested the presence of TMAO demethylase activity. Accordingly, the viability of an frmRAB mutant was compromised upon exposure to TMAO. Downregulation of genes involved in copper resistance could be accounted for by TMAO inhibition of Cu(II) reduction. The simplest interpretation of the data is that during adaptation to the presence of environmental TMAO, anaerobic fermentative cultures of E. coli respond by activating the TorTSR regulatory system with consequent induction of TMAO reductase activity, resulting in net oxidation of menaquinone and inhibition of Cu(II) reduction, responses that are sensed by ArcBA and CusRS respectively. © 2014 Society for Applied Microbiology and John Wiley & Sons Ltd.

Testing the applicability of the k0-NAA method at the MINT's TRIGA MARK II reactor

NASA Astrophysics Data System (ADS)

Siong, Wee Boon; Dung, Ho Manh; Wood, Ab. Khalik; Salim, Nazaratul Ashifa Abd.; Elias, Md. Suhaimi

2006-08-01

The Analytical Chemistry Laboratory at MINT is using the NAA technique since 1980s and is the only laboratory in Malaysia equipped with a research reactor, namely the TRIGA MARK II. Throughout the years the development of NAA technique has been very encouraging and was made applicable to a wide range of samples. At present, the k0 method has become the preferred standardization method of NAA ( k0-NAA) due to its multi-elemental analysis capability without using standards. Additionally, the k0 method describes NAA in physically and mathematically understandable definitions and is very suitable for computer evaluation. Eventually, the k0-NAA method has been adopted by MINT in 2003, in collaboration with the Nuclear Research Institute (NRI), Vietnam. The reactor neutron parameters ( α and f) for the pneumatic transfer system and for the rotary rack at various locations, as well as the detector efficiencies were determined. After calibration of the reactor and the detectors, the implemented k0 method was validated by analyzing some certified reference materials (including IAEA Soil 7, NIST 1633a, NIST 1632c, NIST 1646a and IAEA 140/TM). The analysis results of the CRMs showed an average u score well below the threshold value of 2 with a precision of better than ±10% for most of the elemental concentrations obtained, validating herewith the introduction of the k0-NAA method at the MINT.

Verification of MCNP simulation of neutron flux parameters at TRIGA MK II reactor of Malaysia.

PubMed

Yavar, A R; Khalafi, H; Kasesaz, Y; Sarmani, S; Yahaya, R; Wood, A K; Khoo, K S

2012-10-01

A 3-D model for 1 MW TRIGA Mark II research reactor was simulated. Neutron flux parameters were calculated using MCNP-4C code and were compared with experimental results obtained by k(0)-INAA and absolute method. The average values of φ(th),φ(epi), and φ(fast) by MCNP code were (2.19±0.03)×10(12) cm(-2)s(-1), (1.26±0.02)×10(11) cm(-2)s(-1) and (3.33±0.02)×10(10) cm(-2)s(-1), respectively. These average values were consistent with the experimental results obtained by k(0)-INAA. The findings show a good agreement between MCNP code results and experimental results. Copyright © 2012 Elsevier Ltd. All rights reserved.

Interim waste storage for the Integral Fast Reactor

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

Benedict, R.W.; Phipps, R.D.; Condiff, D.W.

1991-01-01

The Integral Fast Reactor (IFR), which Argonne National Laboratory is developing, is an innovative liquid metal breeder reactor that uses metallic fuel and has a close coupled fuel recovery process. A pyrochemical process is used to separate the fission products from the actinide elements. These actinides are used to make new fuel for the reactor. As part of the overall IFR development program, Argonne has refurbished an existing Fuel Cycle Facility at ANL-West and is installing new equipment to demonstrate the remote reprocessing and fabrication of fuel for the Experimental Breeder Reactor II (EBR-II). During this demonstration the wastes thatmore » are produced will be treated and packaged to produce waste forms that would be typical of future commercial operations. These future waste forms would, assuming Argonne development goals are fulfilled, be essentially free of long half-life transuranic isotopes. Promising early results indicate that actinide extraction processes can be developed to strip these isotopes from waste stream and return them to the IFR type reactors for fissioning. 1 fig.« less

A Review of Gas-Cooled Reactor Concepts for SDI Applications

DTIC Science & Technology

1989-08-01

710 program .) Wire- Core Reactor (proposed by Rockwell). The wire- core reactor utilizes thin fuel wires woven between spacer wires to form an open...reactor is based on results of developmental studies of nuclear rocket propulsion systems. The reactor core is made up of annular fuel assemblies of...XE Addendum to Volume II. NERVA Fuel Development , Westinghouse Astronuclear Laboratory, TNR-230, July 15’ 1972. J I8- Rover Program Reactor Tests

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

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

Olson, D.L.

1992-01-01

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

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

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

Olson, D.L.

1992-05-01

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

77 FR 69900 - Advisory Committee on Reactor Safeguards; Notice of Meeting

Federal Register 2010, 2011, 2012, 2013, 2014

2012-11-21

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards; Notice of Meeting In... Advisory Committee on Reactor Safeguards (ACRS) will hold a meeting on December 6-8, 2012, 11545 Rockville... Recommendations (SECY-12-0064), (3) Venting Systems for Boiling Water Reactors (BWRs) with Mark I and Mark II...

Surface investigation of Si(1 0 0), Cu, Cu on Si(1 0 0), and Au on Cu with positron annihilation induced Auger-electron spectroscopy

NASA Astrophysics Data System (ADS)

Hugenschmidt, C.; Mayer, J.; Schreckenbach, K.

2007-06-01

The surfaces of polycrystalline Cu, Au-coated Cu, Si(1 0 0) and of Si(1 0 0) coated with 1.5 monolayer Cu were investigated with positron annihilation induced Auger-electron spectroscopy (PAES). Since the electron background has been reduced considerably we observed the Cu M 2,3VV-Auger transition on a copper surface within only three hours which is the shortest acquisition time reported so far for PAES. In order to demonstrate PAES' high potential the Auger-yield, the signal-to-background ratio as well as the surface selectivity were compared with accompanying EAES-measurements quantitatively. A more efficient electron energy analyzer for the present PAES setup would lead to an additional efficiency gain of more than two orders of magnitude. The presented measurements were performed at the low-energy positron beam of high intensity NEPOMUC at the research reactor FRM II.

Validation of absolute axial neutron flux distribution calculations with MCNP with 197Au(n,γ)198Au reaction rate distribution measurements at the JSI TRIGA Mark II reactor.

PubMed

Radulović, Vladimir; Štancar, Žiga; Snoj, Luka; Trkov, Andrej

2014-02-01

The calculation of axial neutron flux distributions with the MCNP code at the JSI TRIGA Mark II reactor has been validated with experimental measurements of the (197)Au(n,γ)(198)Au reaction rate. The calculated absolute reaction rate values, scaled according to the reactor power and corrected for the flux redistribution effect, are in good agreement with the experimental results. The effect of different cross-section libraries on the calculations has been investigated and shown to be minor. Copyright © 2013 Elsevier Ltd. All rights reserved.

52. ARAII. Support piers for SL1 reactor building. September 5, ...

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

52. ARA-II. Support piers for SL-1 reactor building. September 5, 1957. Ineel photo no. 57-4398. Photographer: Ken Mansfield. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

The Upgrade of the Neutron Induced Positron Source NEPOMUC

NASA Astrophysics Data System (ADS)

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

2013-06-01

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

Neutron measurement at the thermal column of the Malaysian Triga Mark II reactor using gold foil activation method and TLD

NASA Astrophysics Data System (ADS)

Shalbi, Safwan; Salleh, Wan Norhayati Wan; Mohamad Idris, Faridah; Aliff Ashraff Rosdi, Muhammad; Syahir Sarkawi, Muhammad; Liyana Jamsari, Nur; Nasir, Nur Aishah Mohd

2018-01-01

In order to design facilities for boron neutron capture therapy (BNCT), the neutron measurement must be considered to obtain the optimal design of BNCT facility such as collimator and shielding. The previous feasibility study showed that the thermal column could generate higher thermal neutrons yield for BNCT application at the TRIGA MARK II reactor. Currently, the facility for BNCT are planned to be developed at thermal column. Thus, the main objective was focused on the thermal neutron and epithermal neutron flux measurement at the thermal column. In this measurement, pure gold and cadmium were used as a filter to obtain the thermal and epithermal neutron fluxes from inside and outside of the thermal column door of the 200kW reactor power using a gold foil activation method. The results were compared with neutron fluxes using TLD 600 and TLD 700. The outcome of this work will become the benchmark for the design of BNCT collimator and the shielding

The status of ABWR-II development

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

Hiroyuki, Okada; Hideya Kitamura; Kumiaki, Moriya

This paper reports on the current development status of the ABWR-II project, a next generation reactor design based on the ABWR. In the early 90's, a program to develop the next generation reactor for the 21. century was launched, at a time when the first ABWR was still under construction. At the initial stage of this project, development of a 'user friendly' plant design was the primary objective. Thus, the main focus was placed on selecting a design with features promoting ease of operation and maintenance. Meanwhile, the circumstances surrounding the Japanese nuclear power industry changed. The delay of FBRmore » development and the deregulation of the power generation market have significantly boosted the role of light water reactors, and accelerated the need to improve LWR economics. For these reasons, economic competitiveness became an overriding objective in the development of the ABWR-II, with no less importance placed on achieving the highest standards of safety. Several new features were adopted to enhance economic performance: 1700 MW electric output, large fuel bundles, simplified MSIV, large capacity SRV. An output of 1700 MWe was selected for compatibility with the Japanese power grid, and with consideration of current reactor pressure vessel manufacturing capability. Large fuel bundles will contribute to a shortened refueling outage period and a reduction of CRDs. For enhanced safety, the reference design implements a modified ECCS with four subdivision RHR, a diversified power source incorporating gas turbine generators (GTG), an advanced RCIC (ARCIC) and passive heat removal systems consisting of a passive containment cooling system (PCCS) and a passive reactor cooling system (PRCS). The modified ECCS configuration also enables on-line maintenance. While current reactors rely on complex accident management (AM) procedures, implemented by operators in the event of a serious accident, the ABWR-II incorporated severe accident countermeasures at the

Accumulation of radioactive corrosion products on steel surfaces of VVER-type nuclear reactors. II. 60Co

NASA Astrophysics Data System (ADS)

Varga, Kálmán; Hirschberg, Gábor; Németh, Zoltán; Myburg, Gerrit; Schunk, János; Tilky, Péter

2001-10-01

In the case of intact fuel claddings, the predominant source of radioactivity in the primary circuits of water-cooled nuclear reactors is the activation of corrosion products in the core. The most important corrosion product radionuclides in the primary coolant of pressurized water reactors (PWRs) are 60Co, 58Co, 51Cr, 54Mn, 59Fe (as well as 110mAg in some Soviet-made VVER-type reactor). The second part of this series is focused on the complex studies of the formation and build-up of 60Co-containing species on an austenitic stainless steel type 08X18H10T (GOST 5632-61) and magnetite-covered carbon steel often to be used in Soviet-planned VVERs. The kinetics and mechanism of the cobalt accumulation were studied by a combination (coupling) of an in situ radiotracer method and voltammetry in a model solution of the primary circuit coolant. In addition, independent techniques such as X-ray photoelectron spectroscopic (XPS) and ICP-OES are also used to analyze the chemical state of Co species in the passive layer formed on stainless steel as well as the chemical composition of model solution. The experimental results have revealed that: (i) The passive behavior of the austenitic stainless steel at open-circuit conditions, the slightly alkaline pH and the reducing water chemistry can be considered to be optimal to minimize the 60Co contamination. (ii) The highly potential dependent deposition of various Co-oxides at E>1.10 V (vs. RHE) offers a unique possibility to elaborate a novel electrochemical method for the decrease or removal of cobalt traces from borate-containing coolants contaminated with 60Co and/or 58Co radionuclides.

Generating unstructured nuclear reactor core meshes in parallel

DOE PAGES

Jain, Rajeev; Tautges, Timothy J.

2014-10-24

Recent advances in supercomputers and parallel solver techniques have enabled users to run large simulations problems using millions of processors. Techniques for multiphysics nuclear reactor core simulations are under active development in several countries. Most of these techniques require large unstructured meshes that can be hard to generate in a standalone desktop computers because of high memory requirements, limited processing power, and other complexities. We have previously reported on a hierarchical lattice-based approach for generating reactor core meshes. Here, we describe efforts to exploit coarse-grained parallelism during reactor assembly and reactor core mesh generation processes. We highlight several reactor coremore » examples including a very high temperature reactor, a full-core model of the Korean MONJU reactor, a ¼ pressurized water reactor core, the fast reactor Experimental Breeder Reactor-II core with a XX09 assembly, and an advanced breeder test reactor core. The times required to generate large mesh models, along with speedups obtained from running these problems in parallel, are reported. A graphical user interface to the tools described here has also been developed.« less

KINETICS OF LOW SOURCE REACTOR STARTUPS. PART II

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

hurwitz, H. Jr.; MacMillan, D.B.; Smith, J.H.

1962-06-01

A computational technique is described for computation of the probability distribution of power level for a low source reactor startup. The technique uses a mathematical model, for the time-dependent probability distribution of neutron and precursor concentration, having finite neutron lifetime, one group of delayed neutron precursors, and no spatial dependence. Results obtained by the technique are given. (auth)

NECTAR—A fission neutron radiography and tomography facility

NASA Astrophysics Data System (ADS)

Bücherl, T.; Lierse von Gostomski, Ch.; Breitkreutz, H.; Jungwirth, M.; Wagner, F. M.

2011-09-01

NECTAR (Neutron Computerized Tomography and Radiography) is a versatile facility for radiographic and tomographic investigations as well as for neutron activation experiments using fission neutrons. The radiation sources for this facility are two plates of highly enriched uranium situated in the moderator vessel in FRM II. Thermal neutrons originating from the main fuel element of the reactor generate in these plates fast neutrons. These can escape through a horizontal beam tube without moderation. The beam can be filtered and manipulated in order to reduce the accompanying gamma radiation and to match the specific experimental tasks. A summary of the main parameters required for experimental set-up and (quantitative) data evaluation is presented. The (measured) spectra of the neutron and gamma radiations are shown along with the effect of different filters on their behavior. The neutron and gamma fluxes, dose rates, L/ D-ratios, etc. and the main parameters of the actually used detection systems for neutron imaging are given, too.

Surface and bulk investigations at the high intensity positron beam facility NEPOMUC

NASA Astrophysics Data System (ADS)

Hugenschmidt, C.; Dollinger, G.; Egger, W.; Kögel, G.; Löwe, B.; Mayer, J.; Pikart, P.; Piochacz, C.; Repper, R.; Schreckenbach, K.; Sperr, P.; Stadlbauer, M.

2008-10-01

The NEutron-induced POsitron source MUniCh (NEPOMUC) at the research reactor FRM II delivers a low-energy positron beam ( E = 15-1000 eV) of high intensity in the range between 4 × 10 7 and 5 × 10 8 moderated positrons per second. At present four experimental facilities are in operation at NEPOMUC: a coincident Doppler-broadening spectrometer (CDBS) for defect spectroscopy and investigations of the chemical vicinity of defects, a positron annihilation-induced Auger-electron spectrometer (PAES) for surface studies and an apparatus for the production of the negatively charged positronium ion Ps -. Recently, the pulsed low-energy positron system (PLEPS) has been connected to the NEPOMUC beam line, and first positron lifetime spectra were recorded within short measurement times. A positron remoderation unit which is operated with a tungsten single crystal in back reflection geometry has been implemented in order to improve the beam brilliance. An overview of NEPOMUC's status, experimental results and recent developments at the running spectrometers are presented.

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

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

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

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

Research on soybean protein wastewater treatment by the integrated two-phase anaerobic reactor

PubMed Central

Yu, Yaqin

2015-01-01

The start-up tests of treating soybean protein wastewater by the integrated two-phase anaerobic reactor were studied. The results showed that the soybean protein wastewater could be successfully processed around 30 days when running under the situation of dosing seed sludge with the influent of approximately 2000 mg/L and an HRT of 40 h. When the start-up was finished, the removal rate of COD by the reactor was about 80%. In the zone I, biogas mainly revealed carbon dioxide (CO2) and hydrogen (H2). Methane was the main component in the zone 2 which ranged from 53% to 59% with an average of 55%. The methane content in biogas increased from the zone I to II. It indicated that the methane-producing capacity of the anaerobic sludge increased. It was found that the uniquely designed two-phase integrated anaerobic reactor played a key role in treating soybean protein wastewater. The acidogenic fermentation bacteria dominated in the zone I, while methanogen became dominant in the zone II. It realized the relatively effective separation of hydrolysis acidification and methanogenesis process in the reactor, which was benefit to promote a more reasonable space distribution of the microbial communities in the reactor. There were some differences between the activities of the sludge in the two reaction zones of the integrated two-phase anaerobic reactor. The activity of protease was higher in the reaction zone I. And the coenzyme F420 in the reaction zone II was twice than that in the reaction zone I, which indicated that the activity of the methanogens was stronger in the reaction zone II. PMID:26288554

EBR-II high-ramp transients under computer control

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

Forrester, R.J.; Larson, H.A.; Christensen, L.J.

1983-01-01

During reactor run 122, EBR-II was subjected to 13 computer-controlled overpower transients at ramps of 4 MWt/s to qualify the facility and fuel for transient testing of LMFBR oxide fuels as part of the EBR-II operational-reliability-testing (ORT) program. A computer-controlled automatic control-rod drive system (ACRDS), designed by EBR-II personnel, permitted automatic control on demand power during the transients.

TOPAZ II Anti-Criticality Device Rapid Prototype

NASA Astrophysics Data System (ADS)

Campbell, Donald R.; Otting, William D.

1994-07-01

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

Immobilization of catalase on electrospun PVA/PA6-Cu(II) nanofibrous membrane for the development of efficient and reusable enzyme membrane reactor.

PubMed

Feng, Quan; Zhao, Yong; Wei, Anfang; Li, Changlong; Wei, Qufu; Fong, Hao

2014-09-02

In this study, a mat/membrane consisting of overlaid PVA/PA6-Cu(II) composite nanofibers was prepared via the electrospinning technique followed by coordination/chelation with Cu(II) ions; an enzyme of catalase (CAT) was then immobilized onto the PVA/PA6-Cu(II) nanofibrous membrane. The amount of immobilized catalase reached a high value of 64 ± 4.6 mg/g, while the kinetic parameters (Vmax and Km) of enzyme were 3774 μmol/mg·min and 41.13 mM, respectively. Furthermore, the thermal stability and storage stability of immobilized catalase were improved significantly. Thereafter, a plug-flow type of immobilized enzyme membrane reactor (IEMR) was assembled from the PVA/PA6-Cu(II)-CAT membrane. With the increase of operational pressure from 0.02 to 0.2 MPa, the flux value of IEMR increased from 0.20 ± 0.02 to 0.76 ± 0.04 L/m(2)·min, whereas the conversion ratio of H2O2 decreased slightly from 92 ± 2.5% to 87 ± 2.1%. After 5 repeating cycles, the production capacity of IEMR was merely decreased from 0.144 ± 0.006 to 0.102 ± 0.004 mol/m(2)·min. These results indicated that the assembled IEMR possessed high productivity and excellent reusability, suggesting that the IEMR based on electrospun PVA/PA6-Cu(II) nanofibrous membrane might have great potential for various applications, particularly those related to environmental protection.

The RERTR Program status and progress

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

Travelli, A.

1995-12-01

The progress of the Reduced Enrichment Research and Test Reactor (RERTR) Program is described. The major events, findings, and activities of 1995 are reviewed after a brief summary of the results which the RERTR Program had achieved by the end of 1994. The revelation that Iraq was on the verge of developing a nuclear weapon at the time of the Gulf War, and that it was planning to do so by extracting HEU from the fuel of its research reactors, has given new impetus and urgency to the RERTR commitment of eliminating HEU use in research and test reactors worldwide.more » Development of advanced LEU research reactor fuels is scheduled to begin in October 1995. The Russian RERTR program, which aims to develop and demonstrate within the next five years the technical means needed to convert Russian-supplied research reactors to LEU fuels, is now in operation. A Statement of Intent was signed by high US and Chinese officials, endorsing cooperative activities between the RERTR program and Chinese laboratories involved in similar activities. Joint studies of LEU technical feasibility were completed for the SAFARI-I reactor in South Africa and for the ANS reactor in the US. A new study has been initiated for the FRM-II reactor in Germany. Significant progress was made on several aspects of producing {sup 99}Mo from fission targets utilizing LEU instead of HEU. A cooperation agreements is in place with the Indonesian BATAN. The first prototypical irradiation of an LEU metal-foil target for {sup 99}Mo production was accomplished in Indonesia. The TR-2 reactor, in Turkey, began conversion. SAPHIR, in Switzerland, was shut down. LEU fuel fabrication has begun for the conversion of two more US reactors. Twelve foreign reactors and nine domestic reactors have been fully converted. Approximately 60 % of the work required to eliminate the use of HEU in US-supplied research reactors has been accomplished.« less

The Angra Project: Monitoring Nuclear Reactors with Antineutrino Detectors

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

Anjos, J. C.; Barbosa, A. F.; Lima, H. P. Jr.

2010-03-30

We present the status of the Angra Neutrino project, describing the development of an antineutrino detector aimed at monitoring nuclear reactor activity. The experiment will take place at the Brazilian nuclear power plant located in Angra dos Reis. The Angra II reactor, with 4 GW of thermal power, will be used as a source of antineutrinos. A water Cherenkov detector will be placed above ground in a commercial container outside the reactor containment, about 30 m from the reactor core. With a detector of one ton scale a few thousand antineutrino interactions per day are expected. We intend, in amore » first step, to use the measured neutrino event rate to monitor the on--off status and the thermal power delivered by the reactor. In addition to the safeguards issues the project will provide an alternative tool to have an independent measurement of the reactor power.« less

The Angra Project: Monitoring Nuclear Reactors with Antineutrino Detectors

NASA Astrophysics Data System (ADS)

Anjos, J. C.; Barbosa, A. F.; Bezerra, T. J. C.; Chimenti, P.; Gonzalez, L. F. G.; Kemp, E.; de Oliveira, M. A. Leigui; Lima, H. P.; Lima, R. M.; Nunokawa, H.

2010-03-01

We present the status of the Angra Neutrino project, describing the development of an antineutrino detector aimed at monitoring nuclear reactor activity. The experiment will take place at the Brazilian nuclear power plant located in Angra dos Reis. The Angra II reactor, with 4 GW of thermal power, will be used as a source of antineutrinos. A water Cherenkov detector will be placed above ground in a commercial container outside the reactor containment, about 30 m from the reactor core. With a detector of one ton scale a few thousand antineutrino interactions per day are expected. We intend, in a first step, to use the measured neutrino event rate to monitor the on—off status and the thermal power delivered by the reactor. In addition to the safeguards issues the project will provide an alternative tool to have an independent measurement of the reactor power.

Promoting flood risk reduction: The role of insurance in Germany and England

NASA Astrophysics Data System (ADS)

Surminski, Swenja; Thieken, Annegret H.

2017-10-01

Improving society's ability to prepare for, respond to and recover from flooding requires integrated, anticipatory flood risk management (FRM). However, most countries still focus their efforts on responding to flooding events if and when they occur rather than addressing their current and future vulnerability to flooding. Flood insurance is one mechanism that could promote a more ex ante approach to risk by supporting risk reduction activities. This paper uses an adapted version of Easton's System Theory to investigate the role of insurance for FRM in Germany and England. We introduce an anticipatory FRM framework, which allows flood insurance to be considered as part of a broader policy field. We analyze if and how flood insurance can catalyze a change toward a more anticipatory approach to FRM. In particular we consider insurance's role in influencing five key components of anticipatory FRM: risk knowledge, prevention through better planning, property-level protection measures, structural protection and preparedness (for response). We find that in both countries FRM is still a reactive, event-driven process, while anticipatory FRM remains underdeveloped. Collaboration between insurers and FRM decision-makers has already been successful, for example in improving risk knowledge and awareness, while in other areas insurance acts as a disincentive for more risk reduction action. In both countries there is evidence that insurance can play a significant role in encouraging anticipatory FRM, but this remains underutilized. Effective collaboration between insurers and government should not be seen as a cost, but as an investment to secure future insurability through flood resilience.

Oak Ridge National Laboratory Support of Non-light Water Reactor Technologies: Capabilities Assessment for NRC Near-term Implementation Action Plans for Non-light Water Reactors

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

Belles, Randy; Jain, Prashant K.; Powers, Jeffrey J.

The Oak Ridge National Laboratory (ORNL) has a rich history of support for light water reactor (LWR) and non-LWR technologies. The ORNL history involves operation of 13 reactors at ORNL including the graphite reactor dating back to World War II, two aqueous homogeneous reactors, two molten salt reactors (MSRs), a fast-burst health physics reactor, and seven LWRs. Operation of the High Flux Isotope Reactor (HFIR) has been ongoing since 1965. Expertise exists amongst the ORNL staff to provide non-LWR training; support evaluation of non-LWR licensing and safety issues; perform modeling and simulation using advanced computational tools; run laboratory experiments usingmore » equipment such as the liquid salt component test facility; and perform in-depth fuel performance and thermal-hydraulic technology reviews using a vast suite of computer codes and tools. Summaries of this expertise are included in this paper.« less

The multi-purpose three-axis spectrometer (TAS) MIRA at FRM II

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

Georgii, Robert; Weber, Tobias; Brandl, Georg

The cold-neutron three-axis spectrometer MIRA is an instrument optimized for low-energy excitations. Its excellent intrinsic $Q$-resolution makes it ideal for studying incommensurate magnetic systems (elastic and inelastic). MIRA is at the forefront of using advanced neutron focusing optics such as elliptic guides, which enable the investigation of small samples under extreme conditions. Another advantage of MIRA is the modular assembly allowing for instrumental adaption to the needs of the experiment within a few hours. The development of new methods such as the spin-echo technique MIEZE is another important application at MIRA. Finally, scientific topics include the investigation of complex inter-metallicmore » alloys and spectroscopy on incommensurate magnetic structures.« less

The multi-purpose three-axis spectrometer (TAS) MIRA at FRM II

NASA Astrophysics Data System (ADS)

Georgii, R.; Weber, T.; Brandl, G.; Skoulatos, M.; Janoschek, M.; Mühlbauer, S.; Pfleiderer, C.; Böni, P.

2018-02-01

The cold-neutron three-axis spectrometer MIRA is an instrument optimized for low-energy excitations. Its excellent intrinsic Q-resolution makes it ideal for studying incommensurate magnetic systems (elastic and inelastic). MIRA is at the forefront of using advanced neutron focusing optics such as elliptic guides, which enable the investigation of small samples under extreme conditions. Another advantage of MIRA is the modular assembly allowing for instrumental adaption to the needs of the experiment within a few hours. The development of new methods such as the spin-echo technique MIEZE is another important application at MIRA. Scientific topics include the investigation of complex inter-metallic alloys and spectroscopy on incommensurate magnetic structures.

Developing the European Center of Competence on VVER-Type Nuclear Power Reactors

ERIC Educational Resources Information Center

Geraskin, Nikolay; Pironkov, Lyubomir; Kulikov, Evgeny; Glebov, Vasily

2017-01-01

This paper presents the results of the European educational projects CORONA and CORONA-II which are dedicated to preserving and further developing nuclear knowledge and competencies in the area of VVER-type nuclear power reactors technologies (Water-Water Energetic Reactor, WWER or VVER). The development of the European Center of Competence for…

TRIGA Mark II nuclear reactor facility. Final report, 1 July 1980--30 June 1995

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

Ryan, B.C.

1997-05-01

This report is a final culmination of activities funded through the Department of Energy`s (DOE) University Reactor Sharing Program, Grant DE-FG02-80ER10273, during the period 1 July 1980 through 30 June 1995. Progress reports have been periodically issued to the DOE, namely the Reactor Facility Annual Reports C00-2082/2219-7 through C00-2082/10723-21, which are contained as an appendix to this report. Due to the extent of time covered by this grant, summary tables are presented. Table 1 lists the fiscal year financial obligations of the grant. As listed in the original grant proposals, the DOE grant financed 70% of project costs, namely themore » total amount spent of these projects minus materials costs and technical support. Thus the bulk of funds was spent directly on reactor operations. With the exception of a few years, spending was in excess of the grant amount. As shown in Tables 2 and 3, the Reactor Sharing grant funded a immense number of research projects in nuclear engineering, geology, animal science, chemistry, anthropology, veterinary medicine, and many other fields. A list of these users is provided. Out of the average 3000 visitors per year, some groups participated in classes involving the reactor such as Boy Scout Merit Badge classes, teacher`s workshops, and summer internships. A large number of these projects met the requirements for the Reactor Sharing grant, but were funded by the University instead.« less

Performance of low smeared density sodium-cooled fast reactor metal fuel

NASA Astrophysics Data System (ADS)

Porter, D. L.; Chichester, H. J. M.; Medvedev, P. G.; Hayes, S. L.; Teague, M. C.

2015-10-01

An experiment was performed in the Experimental Breeder Rector-II (EBR-II) in the 1990s to show that metallic fast reactor fuel could be used in reactors with a single, once-through core. To prove the long duration, high burnup, high neutron exposure capability an experiment where the fuel pin was designed with a very large fission gas plenum and very low fuel smeared density (SD). The experiment, X496, operated to only 8.3 at.% burnup because the EBR-II reactor was scheduled for shut-down at that time. Many of the examinations of the fuel pins only funded recently with the resurgence of reactor designs using very high-burnup fuel. The results showed that, despite the low smeared density of 59% the fuel swelled radially to contact the cladding, fission gas release appeared to be slightly higher than demonstrated in conventional 75%SD fuel tests and axial growth was about the same as 75% SD fuel. There were axial positions in some of the fuel pins which showed evidence of fuel restructuring and an absence of fission products with low melting points and gaseous precursors (Cs and Rb). A model to investigate whether these areas may have overheated due to a loss of bond sodium indicates that it is a possible explanation for the fuel restructuring and something to be considered for fuel performance modeling of low SD fuel.

Monte Carol-based validation of neutronic methodology for EBR-II analyses

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

Liaw, J.R.; Finck, P.J.

1993-01-01

The continuous-energy Monte Carlo code VIM (Ref. 1) has been validated extensively over the years against fast critical experiments and other neutronic analysis codes. A high degree of confidence in VIM for predicting reactor physics parameters has been firmly established. This paper presents a numerical validation of two conventional multigroup neutronic analysis codes, DIF3D (Ref. 4) and VARIANT (Ref. 5), against VIM for two Experimental Breeder Reactor II (EBR-II) core loadings in detailed three-dimensional hexagonal-z geometry. The DIF3D code is based on nodal diffusion theory, and it is used in calculations for day-today reactor operations, whereas the VARIANT code ismore » based on nodal transport theory and is used with increasing frequency for specific applications. Both DIF3D and VARIANT rely on multigroup cross sections generated from ENDF/B-V by the ETOE-2/MC[sup 2]-II/SDX (Ref. 6) code package. Hence, this study also validates the multigroup cross-section processing methodology against the continuous-energy approach used in VIM.« less

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

PubMed

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

2006-08-25

Upflow anaerobic sludge blanket (UASB) reactor that was seeded with anaerobic sludge acclimated to chlorophenols was used to investigate the feasibility of anaerobic biotreatment of synthetic wastewater containing pentachlorophenol (PCP) with additional sucrose as carbon source. Two sets of UASB reactors were operated at one time. But the seeded sludge for the two reactors was different and Reactor I was seeded with the sludge that was acclimated to PCP completely for half a year, and Reactor II was seeded with the mixed sludge that was acclimated for half a year to PCP, 4-CP, 3-CP or 2-CP, respectively. The degradation of PCP and the operation fee treating the wastewater are affected by the concentration of MEDS (microorganism easily degradable substrate). So the confirmation of the suitable ratio of [COD] and [PCP] was the key factor of treating the wastewater containing PCP economically and efficiently. During the experiment, the synthetic wastewater with 180.0 mg L(-1) PCP and 1250-10000 mg L(-1) COD could be treated steadily in the experimental Reactor I. The removal efficiency of PCP was more than 99.5% and the removal efficiency of COD was up to 90%. [PCP] (concentration of PCP) in effluent was less than 0.5 mg L(-1). [PCP] in influent could affect proper [COD] (concentration of COD) range in influent that was required for maintenance of steady running of the experimental reactor with a hydraulic retention time (HRT) from 20 to 22 h. [PCP] in influent would directly affect the necessary [COD] in influent when the UASB reactor ran normally and treated the wastewater containing PCP. When [PCP] was 100.4, 151.6 and 180.8 mg L(-1) in influent, respectively, [COD] in influent had to be controlled about 1250-7500, 2500-5000 and 5000 mg L(-1) to maintain the UASB reactor steady running normally and contemporarily ensure that [COD] and [PCP] in effluent were less than 300 and 0.5 mg L(-1), respectively. With the increase of [PCP] in influent, the range of variation

Experimental investigation into fast pyrolysis of biomass using an entrained-flow reactor

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

Bohn, M.; Benham, C.

1981-02-01

Pyrolysis experiments were performed using 30 and 90cm entrained-flow reactors, with steam as a carrier gas and two different feedstocks - wheat straw and powdered material drived from municipal solid waste (ECO-II TM). Reactor wall temperature was varied from 700/sup 0/ to 1400/sup 0/C. Gas composition data from the ECO-II tests were comparable to previously reported data but ethylene yield appeared to vary with reactor wall temperature and residence time. The important conclusion from the wheat straw tests is that olefin yields are about one half that obtained from ECO-II. Evidence was found that high olefin yields from ECO-II aremore » due to the presence of plastics in the feedstock. Batch experiments were run on wheat straw using a Pyroprobe/sup TM/. The samples were heated at a high rate (20,000/sup 0/ C/sec) to 1000/sup 0/ and held at 1000/sup 0/C for a variable period of time from 0.05 to 4.95s. For times up to 0.15s volume fractions of ethylene, propylene, and methane increase while that of carbon dioxide decreases. Subsequently, only carbon monoxide and hydrogen are produced. The change may be related to poor thermal contact and suggests caution in using the Pyroprobe.« less

Neutron Radiography, Tomography, and Diffraction of Commercial Lithium-ion Polymer Batteries

NASA Astrophysics Data System (ADS)

Butler, Leslie G.; Lehmann, Eberhard H.; Schillinger, Burkhard

Imaging an intact, commercial battery as it cycles and wears is proved possible with neutron imaging. The wavelength range of imaging neutrons corresponds nicely with crystallographic dimensions of the electrochemically active species and the metal elec- trodes are relatively transparent. The time scale of charge/discharge cycling is well matched to dynamic tomography as performed with a golden ratio based projection angle ordering. The hydrogen content does create scatter which tends to blur internal struc- ture. In this report, three neutron experiments will be described: 3D images of charged and discharged batteries were obtained with monochromatic neutrons at the FRM II reactor. 2D images (PSI) of fresh and worn batteries as a function of charge state may show a new wear pattern. In situ neutron diffraction (SNS) of the intact battery provides more information about the concentrations of electrochemical species within the battery as a function of charge state and wear. The combination of 2D imaging, 3D imaging, and diffraction data show how neutron imaging can contribute to battery development and wear monitoring.

Spreading dynamics of forget-remember mechanism

NASA Astrophysics Data System (ADS)

Deng, Shengfeng; Li, Wei

2017-04-01

We study extensively the forget-remember mechanism (FRM) for message spreading, originally introduced in Eur. Phys. J. B 62, 247 (2008), 10.1140/epjb/e2008-00139-4. The freedom of specifying forget-remember functions governing the FRM can enrich the spreading dynamics to a very large extent. The master equation is derived for describing the FRM dynamics. By applying the mean field techniques, we have shown how the steady states can be reached under certain conditions, which agrees well with the Monte Carlo simulations. The distributions of forget and remember times can be explicitly given when the forget-remember functions take linear or exponential forms, which might shed some light on understanding the temporal nature of diseases like flu. For time-dependent FRM there is an epidemic threshold related to the FRM parameters. We have proven that the mean field critical transmissibility for the SIS model and the critical transmissibility for the SIR model are the lower and the the upper bounds of the critical transmissibility for the FRM model, respectively.

Removal of Total Coliforms, Thermotolerant Coliforms, and Helminth Eggs in Swine Production Wastewater Treated in Anaerobic and Aerobic Reactors

PubMed Central

Zacarias Sylvestre, Silvia Helena; Lux Hoppe, Estevam Guilherme; de Oliveira, Roberto Alves

2014-01-01

The present work evaluated the performance of two treatment systems in reducing indicators of biological contamination in swine production wastewater. System I consisted of two upflow anaerobic sludge blanket (UASB) reactors, with 510 and 209 L in volume, being serially arranged. System II consisted of a UASB reactor, anaerobic filter, trickling filter, and decanter, being also organized in series, with volumes of 300, 190, 250, and 150 L, respectively. Hydraulic retention times (HRT) applied in the first UASB reactors were 40, 30, 20, and 11 h in systems I and II. The average removal efficiencies of total and thermotolerant coliforms in system I were 92.92% to 99.50% and 94.29% to 99.56%, respectively, and increased in system II to 99.45% to 99.91% and 99.52% to 99.93%, respectively. Average removal rates of helminth eggs in system I were 96.44% to 99.11%, reaching 100% as in system II. In reactor sludge, the counts of total and thermotolerant coliforms ranged between 105 and 109 MPN (100 mL)−1, while helminth eggs ranged from 0.86 to 9.27 eggs g−1 TS. PMID:24812560

Combustion of a Pb(II)-loaded olive tree pruning used as biosorbent.

PubMed

Ronda, A; Della Zassa, M; Martín-Lara, M A; Calero, M; Canu, P

2016-05-05

The olive tree pruning is a specific agroindustrial waste that can be successfully used as adsorbent, to remove Pb(II) from contaminated wastewater. Its final incineration has been studied in a thermobalance and in a laboratory flow reactor. The study aims at evaluating the fate of Pb during combustion, at two different scales of investigation. The flow reactor can treat samples approximately 10(2) larger than the conventional TGA. A detailed characterization of the raw and Pb(II)-loaded waste, before and after combustion is presented, including analysis of gas and solids products. The Pb(II)-loaded olive tree pruning has been prepared by a previous biosorption step in a lead solution, reaching a concentration of lead of 2.3 wt%. Several characterizations of the ashes and the mass balances proved that after the combustion, all the lead presents in the waste remained in ashes. Combustion in a flow reactor produced results consistent with those obtained in the thermobalance. It is thus confirmed that the combustion of Pb(II)-loaded olive tree pruning is a viable option to use it after the biosorption process. The Pb contained in the solid remained in the ashes, preventing possible environmental hazards. Copyright © 2016 Elsevier B.V. All rights reserved.

Pre-treatment processes of Azolla filiculoides to remove Pb(II), Cd(II), Ni(II) and Zn(II) from aqueous solution in the batch and fixed-bed reactors.

PubMed

Khosravi, Morteza; Rakhshaee, Roohan; Ganji, Masuod Taghi

2005-12-09

Intact and treated biomass can remove heavy metals from water and wastewater. This study examined the ability of the activated, semi-intact and inactivated Azolla filiculoides (a small water fern) to remove Pb(2+), Cd(2+), Ni(2+) and Zn(2+) from the aqueous solution. The maximum uptake capacities of these metal ions using the activated Azolla filiculoides by NaOH at pH 10.5 +/- 0.2 and then CaCl(2)/MgCl(2)/NaCl with total concentration of 2 M (2:1:1 mole ratio) in the separate batch reactors were obtained about 271, 111, 71 and 60 mg/g (dry Azolla), respectively. The obtained capacities of maximum adsorption for these kinds of the pre-treated Azolla in the fixed-bed reactors (N(o)) were also very close to the values obtained for the batch reactors (Q(max)). On the other hand, it was shown that HCl, CH(3)OH, C(2)H(5)OH, FeCl(2), SrCl(2), BaCl(2) and AlCl(3) in the pre-treatment processes decreased the ability of Azolla to remove the heavy metals in comparison to the semi-intact Azolla, considerably. The kinetic studies showed that the heavy metals uptake by the activated Azolla was done more rapid than those for the semi-intact Azolla.

Preliminary Options Assessment of Versatile Irradiation Test Reactor

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

Sen, Ramazan Sonat

The objective of this report is to summarize the work undertaken at INL from April 2016 to January 2017 and aimed at analyzing some options for designing and building a versatile test reactor; the scope of work was agreed upon with DOE-NE. Section 2 presents some results related to KNK II and PRISM Mod A. Section 3 presents some alternatives to the VCTR presented in [ ] as well as a neutronic parametric study to assess the minimum power requirement needed for a 235U metal fueled fast test reactor capable to generate a fast (>100 keV) flux of 4.0 xmore » 1015 n /cm2-s at the test location. Section 4 presents some results regarding a fundamental characteristic of test reactors, namely displacement per atom (dpa) in test samples. Section 5 presents the INL assessment of the ANL fast test reactor design FASTER. Section 6 presents a summary.« less

EBR-II and TREAT Digitization Project

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

Griffith, George W.; Rabiti, Cristian

2015-09-01

Digitizing the technical drawings for EBR-II and TREAT provides multiple benefits. Moving the scanned or hard copy drawings to modern 3-D CAD (Computer Aided Drawing) format saves data that could be lost over time. The 3-D drawings produce models that can interface with other drawings to make complex assemblies. The 3-D CAD format can also include detailed material properties and parametric coding that can tie critical dimensions together allowing easier modification. Creating the new files from the old drawings has found multiple inconsistencies that are being flagged or corrected improving understanding of the reactor(s).

Toward a Mechanistic Source Term in Advanced Reactors: A Review of Past U.S. SFR Incidents, Experiments, and Analyses

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

Bucknor, Matthew; Brunett, Acacia J.; Grabaskas, David

In 2015, as part of a Regulatory Technology Development Plan (RTDP) effort for sodium-cooled fast reactors (SFRs), Argonne National Laboratory investigated the current state of knowledge of source term development for a metal-fueled, pool-type SFR. This paper provides a summary of past domestic metal-fueled SFR incidents and experiments and highlights information relevant to source term estimations that were gathered as part of the RTDP effort. The incidents described in this paper include fuel pin failures at the Sodium Reactor Experiment (SRE) facility in July of 1959, the Fermi I meltdown that occurred in October of 1966, and the repeated meltingmore » of a fuel element within an experimental capsule at the Experimental Breeder Reactor II (EBR-II) from November 1967 to May 1968. The experiments described in this paper include the Run-Beyond-Cladding-Breach tests that were performed at EBR-II in 1985 and a series of severe transient overpower tests conducted at the Transient Reactor Test Facility (TREAT) in the mid-1980s.« less

Performance and microbial communities of Mn(II)-based autotrophic denitrification in a Moving Bed Biofilm Reactor (MBBR).

PubMed

Su, Jun Feng; Luo, Xian Xin; Wei, Li; Ma, Fang; Zheng, Sheng Chen; Shao, Si Cheng

2016-07-01

In this study, Mn(II) as electron donor was tested for the effects on denitrification in the MBBR under the conditions of initial nitrate concentration (10mgL(-1), 30mgL(-1), 50mgL(-1)), pH (5, 6, 7) and hydraulic retention time (HRT) (4h, 8h, 12h) which conducted by response surface methodology (RSM), the results demonstrated that the highest nitrate removal efficiency was occurred under the conditions of initial nitrate concentration of 47.64mgL(-1), HRT of 11.96h and pH 5.21. Analysis of SEM and flow cytometry suggested that microorganisms were immobilized on the Yu Long plastic carrier media successfully before the reactor began to operate. Furthermore, high-throughput sequencing was employed to characterize and compare the community compositions and structures of MBBR under the optimum conditions, the results showed that Pseudomonas sp. SZF15 was the dominant contributor for effective removal of nitrate in the MBBR. Copyright © 2016 Elsevier Ltd. All rights reserved.

Tokamak experimental power reactor conceptual design. Volume II

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

Not Available

1976-08-01

Volume II contains the following appendices: (1) summary of EPR design parameters, (2) impurity control, (3) plasma computational models, (4) structural support system, (5) materials considerations for the primary energy conversion system, (6) magnetics, (7) neutronics penetration analysis, (8) first wall stress analysis, (9) enrichment of isotopes of hydrogen by cryogenic distillation, and (10) noncircular plasma considerations. (MOW)

Engineering Porous Polymer Hollow Fiber Microfluidic Reactors for Sustainable C-H Functionalization.

PubMed

He, Yingxin; Rezaei, Fateme; Kapila, Shubhender; Rownaghi, Ali A

2017-05-17

Highly hydrophilic and solvent-stable porous polyamide-imide (PAI) hollow fibers were created by cross-linking of bare PAI hollow fibers with 3-aminopropyl trimethoxysilane (APS). The APS-grafted PAI hollow fibers were then functionalized with salicylic aldehyde for binding catalytically active Pd(II) ions through a covalent postmodification method. The catalytic activity of the composite hollow fiber microfluidic reactors (Pd(II) immobilized APS-grafted PAI hollow fibers) was tested via heterogeneous Heck coupling reaction of aryl halides under both batch and continuous-flow reactions in polar aprotic solvents at high temperature (120 °C) and low operating pressure. X-ray photoelectron spectroscopy (XPS) and inductively coupled plasma (ICP) analyses of the starting and recycled composite hollow fibers indicated that the fibers contain very similar loadings of Pd(II), implying no degree of catalyst leaching from the hollow fibers during reaction. The composite hollow fiber microfluidic reactors showed long-term stability and strong control over the leaching of Pd species.

Calculation to experiment comparison of SPND signals in various nuclear reactor environments

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

Barbot, Loic; Radulovic, Vladimir; Fourmentel, Damien

2015-07-01

In the perspective of irradiation experiments in the future Jules Horowitz Reactor (JHR), the Instrumentation Sensors and Dosimetry Laboratory of CEA Cadarache (France) is developing a numerical tool for SPND design, simulation and operation. In the frame of the SPND numerical tool qualification, dedicated experiments have been performed both in the Slovenian TRIGA Mark II reactor (JSI) and very recently in the French CEA Saclay OSIRIS reactor, as well as a test of two detectors in the core of the Polish MARIA reactor (NCBJ). A full description of experimental set-ups and neutron-gamma calculations schemes are provided in the first partmore » of the paper. Calculation to experiment comparison of the various SPNDs in the different reactors is thoroughly described and discussed in the second part. Presented comparisons show promising final results. (authors)« less

Copper (II) Removal In Anaerobic Continuous Column Reactor System By Using Sulfate Reducing Bacteria

NASA Astrophysics Data System (ADS)

Bilgin, A.; Jaffe, P. R.

2017-12-01

Copper is an essential element for the synthesis of the number of electrons carrying proteins and the enzymes. However, it has a high level of toxicity. In this study; it is aimed to treat copper heavy metal in anaerobic environment by using anaerobic continuous column reactor. Sulfate reducing bacteria culture was obtained in anaerobic medium using enrichment culture method. The column reactor experiments were carried out with bacterial culture obtained from soil by culture enrichment method. The system is operated with continuous feeding and as parallel. In the first rector, only sand was used as packing material. The first column reactor was only fed with the bacteria nutrient media. The same solution was passed through the second reactor, and copper solution removal was investigated by continuously feeding 15-600 mg/L of copper solution at the feeding inlet in the second reactor. When the experiment was carried out by adding the 10 mg/L of initial copper concentration, copper removal in the rate of 45-75% was obtained. In order to determine the use of carbon source during copper removal of mixed bacterial cultures in anaerobic conditions, total organic carbon TOC analysis was used to calculate the change in carbon content, and it was calculated to be between 28% and 75%. When the amount of sulphate is examined, it was observed that it changed between 28-46%. During the copper removal, the amounts of sulphate and carbon moles were equalized and more sulfate was added by changing the nutrient media in order to determine the consumption of sulphate or carbon. Accordingly, when the concentration of added sulphate is increased, it is calculated that between 35-57% of sulphate is spent. In this system, copper concentration of up to 15-600 mg / L were studied.

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

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

Experiences in utilization of research reactors in Yugoslavia

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

Copic, M.; Gabrovsek, Z.; Pop-Jordanov, J.

1971-06-15

The nuclear institutes in Yugoslavia possess three research reactors. Since 1958, two heavy-water reactors have been in operation at the 'Boris Kidric' Institute, a zero-power reactor RB and a 6. 5-MW reactor RA. At the Jozef Stefan Institute, a 250-kW TRIGA Mark II reactor has been operating since 1966. All reactors are equipped with the necessary experimental facilities. The main activities based on these reactors are: (1) fundamental research in solid-state and nuclear physics; (2) R and D activities related to nuclear power program; and (3) radioisotope production. In fundamental physics, inelastic neutron scattering and diffraction phenomena are studied bymore » means of the neutron beam tubes and applied to investigations of the structures of solids and liquids. Valuable results are also obtained in n - γ reaction studies. Experiments connected with the fuel -element development program, owing to the characteristics of the existing reactors, are limited to determination of the fuel element parameters, to studies on the purity of uranium, and to a small number of capsule irradiations. All three reactors are also used for the verification of different methods applied in the analysis of power reactors, particularly concerning neutron flux distributions, the optimization of reactor core configurations and the shielding effects. An appreciable irradiation space in the reactors is reserved for isotope production. Fruitful international co-operation has been established in all these activities, on the basis of either bilateral or multilateral arrangements. The paper gives a critical analysis of the utilization of research reactors in a developing country such as Yugoslavia. The investments in and the operational costs of research reactors are compared with the benefits obtained in different areas of reactor application. The impact on the general scientific, technological and educational level in the country is also considered. In particular, an attempt is made ro

Performance of low smeared density sodium-cooled fast reactor metal fuel

DOE PAGES

Porter, D. L.; H. J. M. Chichester; Medvedev, P. G.; ...

2015-06-17

An experiment was performed in the Experimental Breeder Rector-II (EBR-II) in the 1990s to show that metallic fast reactor fuel could be used in reactors with a single, once-through core. To prove the long duration, high burnup, high neutron exposure capability an experiment where the fuel pin was designed with a very large fission gas plenum and very low fuel smeared density (SD). The experiment, X496, operated to only 8.3 at. % burnup because the EBR-II reactor was scheduled for shut-down at that time. Many of the examinations of the fuel pins only funded recently with the resurgence of reactormore » designs using very high-burnup fuel. The results showed that, despite the low smeared density of 59% the fuel swelled radially to contact the cladding, fission gas release appeared to be slightly higher than demonstrated in conventional 75%SD fuel tests and axial growth was about the same as 75% SD fuel. There were axial positions in some of the fuel pins which showed evidence of fuel restructuring and an absence of fission products with low metaling points and gaseous precursors (Cs and Rb). Lastly, a model to investigate whether these areas may have overheated due to a loss of bond sodium indicates that it is a possible explanation for the fuel restructuring and something to be considered for fuel performance modeling of low SD fuel.« less

Deep-Earth reactor: Nuclear fission, helium, and the geomagnetic field

PubMed Central

Hollenbach, D. F.; Herndon, J. M.

2001-01-01

Geomagnetic field reversals and changes in intensity are understandable from an energy standpoint as natural consequences of intermittent and/or variable nuclear fission chain reactions deep within the Earth. Moreover, deep-Earth production of helium, having 3He/4He ratios within the range observed from deep-mantle sources, is demonstrated to be a consequence of nuclear fission. Numerical simulations of a planetary-scale geo-reactor were made by using the SCALE sequence of codes. The results clearly demonstrate that such a geo-reactor (i) would function as a fast-neutron fuel breeder reactor; (ii) could, under appropriate conditions, operate over the entire period of geologic time; and (iii) would function in such a manner as to yield variable and/or intermittent output power. PMID:11562483

Deep-Earth reactor: nuclear fission, helium, and the geomagnetic field.

PubMed

Hollenbach, D F; Herndon, J M

2001-09-25

Geomagnetic field reversals and changes in intensity are understandable from an energy standpoint as natural consequences of intermittent and/or variable nuclear fission chain reactions deep within the Earth. Moreover, deep-Earth production of helium, having (3)He/(4)He ratios within the range observed from deep-mantle sources, is demonstrated to be a consequence of nuclear fission. Numerical simulations of a planetary-scale geo-reactor were made by using the SCALE sequence of codes. The results clearly demonstrate that such a geo-reactor (i) would function as a fast-neutron fuel breeder reactor; (ii) could, under appropriate conditions, operate over the entire period of geologic time; and (iii) would function in such a manner as to yield variable and/or intermittent output power.

Production assurance program strategy for N Reactor balance of plant systems

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

House, R.D.; Bitten, E.J.; Keenan, J.P.

1986-03-18

A production assurance program has been established for N Reactor, a dual purpose reactor plant, operated to produce special nuclear materials and steam for electricity. N Reactor, which began operation in December 1963, is now approaching the end of its design life. This paper describes the two phase program for Balance of Plant (BOP) systems. The Phase I evaluation has been completed and indications are that the lifetime of systems and components could be extended by implementing appropriate surveillance, operations and maintenance strategies. In Phase II, a thorough evaluation of components and systems is underway and action items are beingmore » identified which will allow component and system extended operation.« less

Modeling of the HiPco process for carbon nanotube production. II. Reactor-scale analysis

NASA Technical Reports Server (NTRS)

Gokcen, Tahir; Dateo, Christopher E.; Meyyappan, M.

2002-01-01

The high-pressure carbon monoxide (HiPco) process, developed at Rice University, has been reported to produce single-walled carbon nanotubes from gas-phase reactions of iron carbonyl in carbon monoxide at high pressures (10-100 atm). Computational modeling is used here to develop an understanding of the HiPco process. A detailed kinetic model of the HiPco process that includes of the precursor, decomposition metal cluster formation and growth, and carbon nanotube growth was developed in the previous article (Part I). Decomposition of precursor molecules is necessary to initiate metal cluster formation. The metal clusters serve as catalysts for carbon nanotube growth. The diameter of metal clusters and number of atoms in these clusters are some of the essential information for predicting carbon nanotube formation and growth, which is then modeled by the Boudouard reaction with metal catalysts. Based on the detailed model simulations, a reduced kinetic model was also developed in Part I for use in reactor-scale flowfield calculations. Here this reduced kinetic model is integrated with a two-dimensional axisymmetric reactor flow model to predict reactor performance. Carbon nanotube growth is examined with respect to several process variables (peripheral jet temperature, reactor pressure, and Fe(CO)5 concentration) with the use of the axisymmetric model, and the computed results are compared with existing experimental data. The model yields most of the qualitative trends observed in the experiments and helps to understanding the fundamental processes in HiPco carbon nanotube production.

Fuzzy regression modeling for tool performance prediction and degradation detection.

PubMed

Li, X; Er, M J; Lim, B S; Zhou, J H; Gan, O P; Rutkowski, L

2010-10-01

In this paper, the viability of using Fuzzy-Rule-Based Regression Modeling (FRM) algorithm for tool performance and degradation detection is investigated. The FRM is developed based on a multi-layered fuzzy-rule-based hybrid system with Multiple Regression Models (MRM) embedded into a fuzzy logic inference engine that employs Self Organizing Maps (SOM) for clustering. The FRM converts a complex nonlinear problem to a simplified linear format in order to further increase the accuracy in prediction and rate of convergence. The efficacy of the proposed FRM is tested through a case study - namely to predict the remaining useful life of a ball nose milling cutter during a dry machining process of hardened tool steel with a hardness of 52-54 HRc. A comparative study is further made between four predictive models using the same set of experimental data. It is shown that the FRM is superior as compared with conventional MRM, Back Propagation Neural Networks (BPNN) and Radial Basis Function Networks (RBFN) in terms of prediction accuracy and learning speed.

Small space reactor power systems for unmanned solar system exploration missions

NASA Technical Reports Server (NTRS)

Bloomfield, Harvey S.

1987-01-01

A preliminary feasibility study of the application of small nuclear reactor space power systems to the Mariner Mark II Cassini spacecraft/mission was conducted. The purpose of the study was to identify and assess the technology and performance issues associated with the reactor power system/spacecraft/mission integration. The Cassini mission was selected because study of the Saturn system was identified as a high priority outer planet exploration objective. Reactor power systems applied to this mission were evaluated for two different uses. First, a very small 1 kWe reactor power system was used as an RTG replacement for the nominal spacecraft mission science payload power requirements while still retaining the spacecraft's usual bipropellant chemical propulsion system. The second use of reactor power involved the additional replacement of the chemical propulsion system with a small reactor power system and an electric propulsion system. The study also provides an examination of potential applications for the additional power available for scientific data collection. The reactor power system characteristics utilized in the study were based on a parametric mass model that was developed specifically for these low power applications. The model was generated following a neutronic safety and operational feasibility assessment of six small reactor concepts solicited from U.S. industry. This assessment provided the validation of reactor safety for all mission phases and generatad the reactor mass and dimensional data needed for the system mass model.

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

Monte Carlo modelling of TRIGA research reactor

NASA Astrophysics Data System (ADS)

El Bakkari, B.; Nacir, B.; El Bardouni, T.; El Younoussi, C.; Merroun, O.; Htet, A.; Boulaich, Y.; Zoubair, M.; Boukhal, H.; Chakir, M.

2010-10-01

The Moroccan 2 MW TRIGA MARK II research reactor at Centre des Etudes Nucléaires de la Maâmora (CENM) achieved initial criticality on May 2, 2007. The reactor is designed to effectively implement the various fields of basic nuclear research, manpower training, and production of radioisotopes for their use in agriculture, industry, and medicine. This study deals with the neutronic analysis of the 2-MW TRIGA MARK II research reactor at CENM and validation of the results by comparisons with the experimental, operational, and available final safety analysis report (FSAR) values. The study was prepared in collaboration between the Laboratory of Radiation and Nuclear Systems (ERSN-LMR) from Faculty of Sciences of Tetuan (Morocco) and CENM. The 3-D continuous energy Monte Carlo code MCNP (version 5) was used to develop a versatile and accurate full model of the TRIGA core. The model represents in detailed all components of the core with literally no physical approximation. Continuous energy cross-section data from the more recent nuclear data evaluations (ENDF/B-VI.8, ENDF/B-VII.0, JEFF-3.1, and JENDL-3.3) as well as S( α, β) thermal neutron scattering functions distributed with the MCNP code were used. The cross-section libraries were generated by using the NJOY99 system updated to its more recent patch file "up259". The consistency and accuracy of both the Monte Carlo simulation and neutron transport physics were established by benchmarking the TRIGA experiments. Core excess reactivity, total and integral control rods worth as well as power peaking factors were used in the validation process. Results of calculations are analysed and discussed.

BOILING WATER REACTOR TECHNOLOGY STATUS OF THE ART REPORT. VOLUME II. WATER CHEMISTRY AND CORROSION

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

Breden, C.R.

1963-02-01

Information concerning the corrosive effects of water in power reactor moderator-coolant systems is presented. The information is based on investigations reported in the unclassified literature believed to be fairly complete to 1959, but less complete since then. The material is presented in sections on water decomposition, water chemistry, materials corrosion, corrosion product deposits, and radioactivity. It is noted that the report is presented as a part of a continuing program in development of less expensive materials for use in reactors. (J.R.D.)

A reagent-free tubular biofilm reactor for on-line determination of biochemical oxygen demand.

PubMed

Liu, Changyu; Zhao, Huijun; Gao, Shan; Jia, Jianbo; Zhao, Limin; Yong, Daming; Dong, Shaojun

2013-07-15

We reported a reagent-free tubular biofilm reactor (BFR) based analytical system for rapid online biochemical oxygen demand (BOD) determination. The BFR was cultivated using microbial seeds from activated sludge. It only needs tap water to operate and does not require any chemical reagent. The analytical performance of this reagent-free BFR system was found to be equal to or better than the BFR system operated using phosphate buffer saline (PBS) and high purity deionized water. The system can readily achieve a limit of detection of 0.25 mg O2 L(-1), possessing superior reproducibility, and long-term operational and storage stability. More importantly, we confirmed for the first time that the BFR system is capable of tolerating common toxicants found in wastewaters, such as 3,5-dichlorophenol and Zn(II), Cr(VI), Cd(II), Cu(II), Pb(II), Mn(II) and Ni(II), enabling the method to be applied to a wide range of wastewaters. The sloughing and clogging are the important attributes affecting the operational stability, hence, the reliability of most online wastewater monitoring systems, which can be effectively avoided, benefiting from the tubular geometry of the reactor and high flow rate conditions. These advantages, coupled with simplicity in device, convenience in operation and minimal maintenance, make such a reagent-free BFR analytical system promising for practical BOD online determination. Copyright © 2013 Elsevier B.V. All rights reserved.

Eugene P. Wigner's Visionary Contributions to Generations-I through IV Fission Reactors

NASA Astrophysics Data System (ADS)

Carré, Frank

2014-09-01

Among Europe's greatest scientists who fled to Britain and America in the 1930s, Eugene P. Wigner made instrumental advances in reactor physics, reactor design and technology, and spent nuclear fuel processing for both purposes of developing atomic weapons during world-war II and nuclear power afterwards. Wigner who had training in chemical engineering and self-education in physics first gained recognition for his remarkable articles and books on applications of Group theory to Quantum mechanics, Solid state physics and other topics that opened new branches of Physics.

EBR-II Static Neutronic Calculations by PHISICS / MCNP6 codes

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

Paolo Balestra; Carlo Parisi; Andrea Alfonsi

2016-02-01

The International Atomic Energy Agency (IAEA) launched a Coordinated Research Project (CRP) on the Shutdown Heat Removal Tests (SHRT) performed in the '80s at the Experimental fast Breeder Reactor EBR-II, USA. The scope of the CRP is to improve and validate the simulation tools for the study and the design of the liquid metal cooled fast reactors. Moreover, training of the next generation of fast reactor analysts is being also considered the other scope of the CRP. In this framework, a static neutronic model was developed, using state-of-the art neutron transport codes like SCALE/PHISICS (deterministic solution) and MCNP6 (stochastic solution).more » Comparison between both solutions is briefly illustrated in this summary.« less

II. Electrodeposition/removal of nickel in a spouted electrochemical reactor.

PubMed

Grimshaw, Pengpeng; Calo, Joseph M; Shirvanian, Pezhman A; Hradil, George

2011-08-17

An investigation is presented of nickel electrodeposition from acidic solutions in a cylindrical spouted electrochemical reactor. The effects of solution pH, temperature, and applied current on nickel removal/recovery rate, current efficiency, and corrosion rate of deposited nickel on the cathodic particles were explored under galvanostatic operation. Nitrogen sparging was used to decrease the dissolved oxygen concentration in the electrolyte in order to reduce the nickel corrosion rate, thereby increasing the nickel electrowinning rate and current efficiency. A numerical model of electrodeposition, including corrosion and mass transfer in the particulate cathode moving bed, is presented that describes the behavior of the experimental net nickel electrodeposition data quite well.

II. Electrodeposition/removal of nickel in a spouted electrochemical reactor

PubMed Central

Grimshaw, Pengpeng; Calo, Joseph M.; Shirvanian, Pezhman A.; Hradil, George

2011-01-01

An investigation is presented of nickel electrodeposition from acidic solutions in a cylindrical spouted electrochemical reactor. The effects of solution pH, temperature, and applied current on nickel removal/recovery rate, current efficiency, and corrosion rate of deposited nickel on the cathodic particles were explored under galvanostatic operation. Nitrogen sparging was used to decrease the dissolved oxygen concentration in the electrolyte in order to reduce the nickel corrosion rate, thereby increasing the nickel electrowinning rate and current efficiency. A numerical model of electrodeposition, including corrosion and mass transfer in the particulate cathode moving bed, is presented that describes the behavior of the experimental net nickel electrodeposition data quite well. PMID:22039317

Nonenzymic spectrophotometric determination of potential poison ivy cross-reactors.

PubMed

Quattrone, A J

1977-03-01

I describe an inexpensive, nonenzymic analytical system for prescreening substances that might cross-react as Rhus toxing (e.g., poison ivy, poison oak, and sumac allergens) on human skin. By spectrophotometric assay after incubation with an oxidizing mixture of Cu(II)ammine complex and ammonium persulfate, I could accurately and reproducibly determine o-quinoidal products of several potential synthetic cross-reactors and native poison ivy allergen, and could distinguish these from catecholamines, resorcinol, p-hydroquinone, and a closely related phenol. A good correlation was obtained between this nonenzymic technique and an enzymic assay. This Cu(II)ammine/persulfate oxidative assay, however, is inexpensive and obviates any spectral interference from enzymic proteins.

General layout of reactor and control areas upon advent of ...

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

General layout of reactor and control areas upon advent of power burst facility (PBF). Shows relationship of PBF to SPERT-I, -II, -III, and -IV. Ebasco Services 1205-PER/PBF-U-102. Date: July 1965. INEEL index no. 761-0100-00-205-123006 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

Nuclear reactor construction with bottom supported reactor vessel

DOEpatents

Sharbaugh, John E.

1987-01-01

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

Globally linearized control on diabatic continuous stirred tank reactor: a case study.

PubMed

Jana, Amiya Kumar; Samanta, Amar Nath; Ganguly, Saibal

2005-07-01

This paper focuses on the promise of globally linearized control (GLC) structure in the realm of strongly nonlinear reactor system control. The proposed nonlinear control strategy is comprised of: (i) an input-output linearizing state feedback law (transformer), (ii) a state observer, and (iii) an external linear controller. The synthesis of discrete-time GLC controller for single-input single-output diabatic continuous stirred tank reactor (DCSTR) has been studied first, followed by the synthesis of feedforward/feedback controller for the same reactor having dead time in process as well as in disturbance. Subsequently, the multivariable GLC structure has been designed and then applied on multi-input multi-output DCSTR system. The simulation study shows high quality performance of the derived nonlinear controllers. The better-performed GLC in conjunction with reduced-order observer has been compared with the conventional proportional integral controller on the example reactor and superior performance has been achieved by the proposed GLC control scheme.

Developing the European Center of Competence on VVER-type nuclear power reactors

NASA Astrophysics Data System (ADS)

Geraskin, Nikolay; Pironkov, Lyubomir; Kulikov, Evgeny; Glebov, Vasily

2017-09-01

This paper presents the results of the European educational projects CORONA and CORONA-II which are dedicated to preserving and further developing nuclear knowledge and competencies in the area of VVER-type nuclear power reactors technologies (Water-Water Energetic Reactor, WWER or VVER). The development of the European Center of Competence for VVER-technology is focused on master's degree programmes. The specifics of a systematic approach to training in the area of VVER-type nuclear power reactors technologies are analysed. This paper discusses enhancement of the training opportunities of the European Center that have arisen from advances in methodology and distance education. With a special attention paid to the European Nuclear Education Network (ENEN), the possibilities of further development of the international cooperation between European countries and educational institutions are examined.

Assessment of treatment efficacy and sebosuppressive effect of fractional radiofrequency microneedle on acne vulgaris.

PubMed

Lee, Kyung Real; Lee, Eo Gin; Lee, Hee Jung; Yoon, Moon Soo

2013-12-01

A minimally invasive fractional radiofrequency microneedle (FRM) device has been used in skin rejuvenation and acne scars, and a recent pilot study demonstrated the positive therapeutic effect on acne. We evaluated the efficacy of FRM device for acne vulgaris in Asians and conducted objective measurement to assess its effect on sebum production. Twenty Korean patients with acne vulgaris received a single full-face FRM treatment. Outcome assessments included standardized photography, physician's global assessment, patient's satisfaction scores, acne lesion count, and objective measurements of casual sebum level (CSL) and sebum excretion rate (SER). They were evaluated at baseline and 2, 4, 8 weeks after the treatment. After a single FRM treatment, the CSL and the SER showed 30-60% and 70-80% reduction, respectively, at week 2 (P < 0.01), and remained below the baseline level until week 8. Physician's global improvement scores for acne severity and acne lesion count also revealed clinical improvement with maximum efficacy at week 2, but returned to the baseline in most patients by week 8. Patients' satisfaction scores (0-4) were above 2 on average, and adverse effects were minimal. This prospective study demonstrated the sebosuppressive effect from a single FRM treatment, but its therapeutic efficacy in acne requires further evaluation. © 2013 Wiley Periodicals, Inc.

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

First steps towards real-time radiography at the NECTAR facility

NASA Astrophysics Data System (ADS)

Bücherl, T.; Wagner, F. M.; v. Gostomski, Ch. Lierse

2009-06-01

The beam tube SR10 at Forschungsneutronenquelle Heinz Maier-Leibnitz (FRM II) provides an intense beam of fission neutrons for medical application (MEDAPP) and for radiography and tomography of technical and other objects (NECTAR). The high neutron flux of up to 9.8E+07 cm -2 s -1 (depending on filters and collimation) with a mean energy of about 1.9 MeV at the sample position at the NECTAR facility prompted an experimental feasibility study to investigate the potential for real-time (RT) radiography.

Novel type of neutron polarization analysis using the multianalyzer-equipment of the three-axes spectrometer PUMA

NASA Astrophysics Data System (ADS)

Schwesig, Steffen; Maity, Avishek; Sobolev, Oleg; Ziegler, Fabian; Eckold, Götz

2018-01-01

The combination of polarization analysis and multianalyzer system available at the three axes spectrometer PUMA@FRM II allows the simultaneous determination of both spin states of the scattered neutrons and the absolute value of the polarization. The present paper describes the technical details along with the basic formalism used for the precise calibration. Moreover, the performance of this method is illustrated by several test experiments including first polarized inelastic studies of the magnetic excitations of CuO in the multiferroic and the uniaxial antiferromagnetic phases.

Comparison of SAND-II and FERRET

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

Wootan, D.W.; Schmittroth, F.

1981-01-01

A comparison was made of the advantages and disadvantages of two codes, SAND-II and FERRET, for determining the neutron flux spectrum and uncertainty from experimental dosimeter measurements as anticipated in the FFTF Reactor Characterization Program. This comparison involved an examination of the methodology and the operational performance of each code. The merits of each code were identified with respect to theoretical basis, directness of method, solution uniqueness, subjective influences, and sensitivity to various input parameters.

Rebuilding the Brookhaven high flux beam reactor: A feasibility study

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

Brynda, W.J.; Passell, L.; Rorer, D.C.

1995-01-01

After nearly thirty years of operation, Brookhaven`s High Flux Beam Reactor (HFBR) is still one of the world`s premier steady-state neutron sources. A major center for condensed matter studies, it currently supports fifteen separate beamlines conducting research in fields as diverse as crystallography, solid-state, nuclear and surface physics, polymer physics and structural biology and will very likely be able to do so for perhaps another decade. But beyond that point the HFBR will be running on borrowed time. Unless appropriate remedial action is taken, progressive radiation-induced embrittlement problems will eventually shut it down. Recognizing the HFBR`s value as a nationalmore » scientific resource, members of the Laboratory`s scientific and reactor operations staffs began earlier this year to consider what could be done both to extend its useful life and to assure that it continues to provide state-of-the-art research facilities for the scientific community. This report summarizes the findings of that study. It addresses two basic issues: (i) identification and replacement of lifetime-limiting components and (ii) modifications and additions that could expand and enhance the reactor`s research capabilities.« less

Metal ion promoted hydrogels for bovine serum albumin adsorption: Cu(II) and Co(II) chelated poly[(N-vinylimidazole)-maleic acid].

PubMed

Pekel, Nursel; Salih, Bekir; Güven, Olgun

2005-05-10

Poly[(N-vinylimidazole)-maleic acid] (poly(VIm-MA)), copolymeric hydrogels were prepared by gamma-irradiating ternary mixtures of N-vinylimidazole-maleic acid-water in a (60)Co-gamma source. Cu(II) and Co(II) ions were chelated within the gels at pH=5.0. The maximum adsorption capacity of the gels were 3.71 mmol/g dry gel for Cu(II) and 1.25 mmol/g dry gel for Co(II) at pH=5.0. The swelling ratios of the gels were 1200% for poly(VIm-MA), 60 and 45% for Cu(II) and Co(II)-chelated poly(VIm-MA) gels at pH=5.0 in acetate buffer solution. These affinity gels with different swelling ratios for plain poly(VIm-MA), Cu(II)-, and Co(II)-chelated poly(VIm-MA), in acetate and phosphate buffers were used in the bovine serum albumin (BSA) adsorption/desorption studies in batch reactor. The maximum BSA adsorption capacities of the gels were 0.38 g/g dry gel for plain, 0.88 g/g dry gel for Cu(II)-chelated poly(VIm-MA) and 1.05 g/g dry gel for Co(II)-chelated poly(VIm-MA) gels. Adsorption capacity of BSA by the gels was reduced dramatically by increasing the ionic strength adjusted with NaCl. More than 95% of BSA were desorbed in 10 h in desorption medium containing 0.1M of EDTA for metal ion-chelated gels at pH=4.7.

PHISICS/RELAP5-3D RESULTS FOR EXERCISES II-1 AND II-2 OF THE OECD/NEA MHTGR-350 BENCHMARK

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

Strydom, Gerhard

2016-03-01

The Idaho National Laboratory (INL) Advanced Reactor Technologies (ART) High-Temperature Gas-Cooled Reactor (HTGR) Methods group currently leads the Modular High-Temperature Gas-Cooled Reactor (MHTGR) 350 benchmark. The benchmark consists of a set of lattice-depletion, steady-state, and transient problems that can be used by HTGR simulation groups to assess the performance of their code suites. The paper summarizes the results obtained for the first two transient exercises defined for Phase II of the benchmark. The Parallel and Highly Innovative Simulation for INL Code System (PHISICS), coupled with the INL system code RELAP5-3D, was used to generate the results for the Depressurized Conductionmore » Cooldown (DCC) (exercise II-1a) and Pressurized Conduction Cooldown (PCC) (exercise II-2) transients. These exercises require the time-dependent simulation of coupled neutronics and thermal-hydraulics phenomena, and utilize the steady-state solution previously obtained for exercise I-3 of Phase I. This paper also includes a comparison of the benchmark results obtained with a traditional system code “ring” model against a more detailed “block” model that include kinetics feedback on an individual block level and thermal feedbacks on a triangular sub-mesh. The higher spatial fidelity that can be obtained by the block model is illustrated with comparisons of the maximum fuel temperatures, especially in the case of natural convection conditions that dominate the DCC and PCC events. Differences up to 125 K (or 10%) were observed between the ring and block model predictions of the DCC transient, mostly due to the block model’s capability of tracking individual block decay powers and more detailed helium flow distributions. In general, the block model only required DCC and PCC calculation times twice as long as the ring models, and it therefore seems that the additional development and calculation time required for the block model could be worth the gain that can

Field Evaluation of Portable and Central Site PM Samplers Emphasizing Additive and Differential Mass Concentration Estimates

EPA Science Inventory

The US Environmental Protection Agency (EPA) published a National Ambient Air Quality Standard (NAAQS) and the accompanying Federal Reference Method (FRM) for PM₁₀ in 1987. The EPA revised the particle standards and FRM in 1997 to include PM_2.5. In 2005, EPA...

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.

The Thermal Neutron Beam Option for NECTAR at MLZ

NASA Astrophysics Data System (ADS)

Mühlbauer, M. J.; Bücherl, T.; Genreith, C.; Knapp, M.; Schulz, M.; Söllradl, S.; Wagner, F. M.; Ehrenberg, H.

The beam port SR10 at the neutron source FRM II of Heinz Maier-Leibnitz Zentrum (MLZ) is equipped with a moveable assembly of two uranium plates, which can be placed in front of the entrance window of the beam tube via remote control. With these plates placed in their operating position the thermal neutron spectrum produced by the neutron source FRM II is converted to fission neutrons with 1.9 MeV of mean energy. This fission neutron spectrum is routinely used for medical applications at the irradiation facility MEDAPP, for neutron radiography and tomography experiments at the facility NECTAR and for materials testing. If, however, the uranium plates are in their stand-by position far off the tip of the beam tube and the so-called permanent filter for thermal neutrons is removed, thermal neutrons originating from the moderator tank enter the beam tube and a thermal spectrum becomes available for irradiation or activation of samples. By installing a temporary flight tube the beam may be used for thermal neutron radiography and tomography experiments at NECTAR. The thermal neutron beam option not only adds a pure thermal neutron spectrum to the energy ranges available for neutron imaging at MLZ instruments but it also is an unique possibility to combine two quite different neutron energy ranges at a single instrument including their respective advantages. The thermal neutron beam option for NECTAR is funded by BMBF in frame of research project 05K16VK3.

Particle-Sizing System for Scanning Electron Microscope Images of Solid- Propellant Combustion Exhaust

DTIC Science & Technology

1991-03-01

bl2[3]; 1* ditto */ char b13(61, b1413]; /* unused at the moment ~ char bl5(3], b16 (3], b17[31, b18[3]; /* YIN input buffers * char bl9[3], b20[3...gbool(SET,frm,13,wn.10, 1, ’ALL: Enable HELP screens (Y,N] fatrib,’ - ,&HELPLVL,bl5,NSTR,NSTR); bool2YN(DOSEQ, b16 ); wngbool(SET,frm,14,wn,ll, l,’SEMEX...CLIP, TAG and SIZE without asking[Y.N): fatrib,’ ’ ,&DOSEQ, bl6 ,NSTR.NSTR): bool2YN(DFINVERT,b17); wngbool(SET,frm,15,wn,12, 1, ACQUIRE: Complement

Supporting the information domains of fall-risk management in home care via health information technology.

PubMed

Alhuwail, Dari; Koru, Güneş; Mills, Mary Etta

2016-01-01

In the United States, home care clinicians often start the episode of care devoid of relevant fall-risk information. By collecting and analyzing qualitative data from 30 clinicians in one home health agency, this case study aimed to understand how the currently adopted information technology solutions supported the clinicians' fall-risk management (FRM) information domains, and explored opportunities to adopt other solutions to better support FRM. The currently adopted electronic health record system and fall-reporting application served only some information domains with a limited capacity. Substantial improvement in addressing the FRM information domains is possible by effectively modifying the existing solutions and purposefully adopting new solutions.

ST-LO (7 July -19 July 1944)

DTIC Science & Technology

1946-08-21

artillery, but the main troubles came in operating heavy vehicles through muddy fields, slippery trails, and hedgerows. The II9th Infantry felt...Army’s attack began, continued to balk aerial reconnaissance and to make the roads slippery and muddy. The axis of advance of the attacking task...JuLTI J ! VII - XIX CoIIps IOUNOIJIY AS OF 6 JULY .....lIT, If’ III IIIET!!!!. I o " ’" •’" .. flU ( \\ J Elm ’ "’/1’ v,",f «. ? --/ F:rm~91

Radioactive ion beam acceleration at MAFF

NASA Astrophysics Data System (ADS)

Pasini, M.; Kester, O.; Habs, D.; Groß, M.; Sieber, T.; Maier, H. J.; Assmann, W.; Krüken, R.; Faestermann, T.; Schempp, A.; Ratzinger, U.; Safvan, C. P.

2004-12-01

In April 2003, the German safety commission has given the final approval for the oper- ation of the high flux reactor FRM-II. This is an important step towards the development and installation of the Munich accelerator for fission fragments (MAFF), which will deliver highest intensities of neutron rich fission fragments. The acceleration chain of MAFF [1] consists of a charge breeder, which will deliver the ions with a mass to charge ratio of A/q ⩽ 6.3 irrespective of the mass range, and with a repetition rate of maximum 50 Hz. The LINAC operating at 10% duty cycle is composed of a 101.28 IH-RFQ, which will boost up the energy from 2.5 up to 300 keV/u, three IH-tanks that will deliver an energy of 5.4 MeV/u and 2 seven gap IH-resonators that are used to vary the final energy up to a maximum of 5.9 MeV/u. Currently beam dynamics revisions are in progress especially in the low energy section, since the experimental program has requested specific time structures of the beam for TOF experiments. The status of the beam dynamics studies as well as the status of the single components of the accelerator will be presented in this paper.

A CAMAC based real-time noise analysis system for nuclear reactors

NASA Astrophysics Data System (ADS)

Ciftcioglu, Özer

1987-05-01

A CAMAC based real-time noise analysis system was designed for the TRIGA MARK II nuclear reactor at the Institute for Nuclear Energy, Istanbul. The input analog signals obtained from the radiation detectors are introduced to the system through CAMAC interface. The signals converted into digital form are processed by a PDP-11 computer. The fast data processing based on auto/cross power spectral density computations is carried out by means of assembly written FFT algorithms in real-time and the spectra obtained are displayed on a CAMAC driven display system as an additional monitoring device. The system has the advantage of being software programmable and controlled by a CAMAC system so that it is operated under program control for reactor surveillance, anomaly detection and diagnosis. The system can also be used for the identification of nonstationary operational characteristics of the reactor in long term by comparing the noise power spectra with the corresponding reference noise patterns prepared in advance.

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

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

Erika N. Bailey

2011-10-10

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

Behavior of fiber reinforced mortar joints in masonry walls subjected to in-plane shear and out-of-plane bending

NASA Astrophysics Data System (ADS)

Armwood, Catherine K.

In this project, 26 fiber-reinforced mortar (FRM) mixtures are evaluated for their workability and strength characteristics. The specimens tested include two control mixtures and 24 FRMs. The mixtures were made of two types of binders; Type N Portland cement lime (Type N-PCL) and Natural Hydrated Lime 5 (NHL5); and 6 fiber types (5 synthetic fibers and one organic). When tested in flexure, the results indicate that majority of the synthetic fiber mixtures enhanced the performance of the mortar and the nano-nylon and horse hair fibers were the least effective in improving the mortar's modulus of rupture, ductility, and energy absorption. Four FRMs that improved the mortar's mechanical properties most during the flexural strength test were then used to conduct additional experiments. The FRM's compressive strength, as well as flexural and shear bond strength with clay and concrete masonry units were determined. Those four mixtures included Type N-PCL as the binder and 4 synthetic fibers. They were evaluated at a standard laboratory flow rate of 110% +/- 5% and a practical field flow rate of 130% +/- 5%. Results indicate that the use of fibers decreases the compressive strength of the mortar most of the time. However, the bond strength test results were promising: 81% of the FRM mixtures increased the flexural bond strength of the prism. The mixtures at 110 +/- 5% flow rate bonded better with concrete bricks and those ate 130+/-5% flow rate bonded better with clay bricks. The results of the shear bond strength show 50% of the FRM mixtures improved the shear bond strength. The FRM mixtures at 110+/-5% flow rate bonded with clay units provided the most improvement in shear bond strength compared to control specimen results. Along with detailed discussions and derived conclusions of these experiments, this dissertation includes recommendations for the most feasible FRM for different applications.

NEUTRONIC REACTOR

DOEpatents

Wigner, E.P.

1958-04-22

A nuclear reactor for isotope production is described. This reactor is designed to provide a maximum thermal neutron flux in a region adjacent to the periphery of the reactor rather than in the center of the reactor. The core of the reactor is generally centrally located with respect tn a surrounding first reflector, constructed of beryllium. The beryllium reflector is surrounded by a second reflector, constructed of graphite, which, in tune, is surrounded by a conventional thermal shield. Water is circulated through the core and the reflector and functions both as a moderator and a coolant. In order to produce a greatsr maximum thermal neutron flux adjacent to the periphery of the reactor rather than in the core, the reactor is designed so tbat the ratio of neutron scattering cross section to neutron absorption cross section averaged over all of the materials in the reflector is approximately twice the ratio of neutron scattering cross section to neutron absorption cross section averaged over all of the material of the core of the reactor.

Control of reactor coolant flow path during reactor decay heat removal

DOEpatents

Hunsbedt, Anstein N.

1988-01-01

An improved reactor vessel auxiliary cooling system for a sodium cooled nuclear reactor is disclosed. The sodium cooled nuclear reactor is of the type having a reactor vessel liner separating the reactor hot pool on the upstream side of an intermediate heat exchanger and the reactor cold pool on the downstream side of the intermediate heat exchanger. The improvement includes a flow path across the reactor vessel liner flow gap which dissipates core heat across the reactor vessel and containment vessel responsive to a casualty including the loss of normal heat removal paths and associated shutdown of the main coolant liquid sodium pumps. In normal operation, the reactor vessel cold pool is inlet to the suction side of coolant liquid sodium pumps, these pumps being of the electromagnetic variety. The pumps discharge through the core into the reactor hot pool and then through an intermediate heat exchanger where the heat generated in the reactor core is discharged. Upon outlet from the heat exchanger, the sodium is returned to the reactor cold pool. The improvement includes placing a jet pump across the reactor vessel liner flow gap, pumping a small flow of liquid sodium from the lower pressure cold pool into the hot pool. The jet pump has a small high pressure driving stream diverted from the high pressure side of the reactor pumps. During normal operation, the jet pumps supplement the normal reactor pressure differential from the lower pressure cold pool to the hot pool. Upon the occurrence of a casualty involving loss of coolant pump pressure, and immediate cooling circuit is established by the back flow of sodium through the jet pumps from the reactor vessel hot pool to the reactor vessel cold pool. The cooling circuit includes flow into the reactor vessel liner flow gap immediate the reactor vessel wall and containment vessel where optimum and immediate discharge of residual reactor heat occurs.

Nanoscale zero-valent iron (nZVI) for the treatment of concentrated Cu(II) wastewater: a field demonstration.

PubMed

Li, Shaolin; Wang, Wei; Yan, Weile; Zhang, Wei-xian

2014-03-01

A field demonstration was conducted to assess the feasibility of nanoscale zero-valent iron (nZVI) for the treatment of wastewater containing high levels of Cu(II). Pilot tests were performed at a printed-circuit-board manufacturing plant, treating 250,000 L of wastewater containing 70 mg L(-1) Cu(II) with a total of 55 kg of nZVI. A completely mixed reactor of 1,600 L was operated continuously with flow rates ranging from 1000 to 2500 L h(-1). The average Cu(II) removal efficiency was greater than 96% with 0.20 g L(-1) nZVI and a hydraulic retention time of 100 min. The nZVI reactor achieved a remarkably high volumetric loading rate of 1876 g Cu per m(3) per day for Cu(II) removal, surpassing the loading rates of conventional technologies by more than one order of magnitude. The average removal capacity of nZVI for Cu(II) was 0.343 g Cu per gram of Fe. The Cu(II) removal efficiency can be reliably regulated by the solution Eh, which in turn is a function of nZVI input and hydraulic retention time. The ease of separation and recycling of nZVI contribute to process up-scalability and cost effectiveness. Cu(II) was reduced to metallic copper and cuprite (Cu2O). The end product is a valuable composite of iron and copper (∼20-25%), which can partially offset the treatment costs.

40 CFR 53.2 - General requirements for a reference method determination.

Code of Federal Regulations, 2010 CFR

2010-07-01

...) Manual methods—(1) Sulfur dioxide (SO 2 ) and lead. For measuring SO2 and lead, appendices A and G of part 50 of this chapter specify unique manual FRM for measuring these pollutants. Except as provided in § 53.16, other manual methods for SO2 and lead will not be considered for FRM determinations under this...

NEUTRONIC REACTOR

DOEpatents

Fermi, E.; Zinn, W.H.; Anderson, H.L.

1958-09-16

Means are presenied for increasing the reproduction ratio of a gaphite- moderated neutronic reactor by diminishing the neutron loss due to absorption or capture by gaseous impurities within the reactor. This means comprised of a fluid-tight casing or envelope completely enclosing the reactor and provided with a valve through which the casing, and thereby the reactor, may be evacuated of atmospheric air.

The biological function of consciousness

PubMed Central

Earl, Brian

2014-01-01

This research is an investigation of whether consciousness—one's ongoing experience—influences one's behavior and, if so, how. Analysis of the components, structure, properties, and temporal sequences of consciousness has established that, (1) contrary to one's intuitive understanding, consciousness does not have an active, executive role in determining behavior; (2) consciousness does have a biological function; and (3) consciousness is solely information in various forms. Consciousness is associated with a flexible response mechanism (FRM) for decision-making, planning, and generally responding in nonautomatic ways. The FRM generates responses by manipulating information and, to function effectively, its data input must be restricted to task-relevant information. The properties of consciousness correspond to the various input requirements of the FRM; and when important information is missing from consciousness, functions of the FRM are adversely affected; both of which indicate that consciousness is the input data to the FRM. Qualitative and quantitative information (shape, size, location, etc.) are incorporated into the input data by a qualia array of colors, sounds, and so on, which makes the input conscious. This view of the biological function of consciousness provides an explanation why we have experiences; why we have emotional and other feelings, and why their loss is associated with poor decision-making; why blindsight patients do not spontaneously initiate responses to events in their blind field; why counter-habitual actions are only possible when the intended action is in mind; and the reason for inattentional blindness. PMID:25140159

The U.S. RERTR program status and progress.

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

Travelli, A.

1998-01-21

The progress of the Reduced Enrichment Research and Test Reactor (RERTR) Program since its inception in 1978 is described. A brief summary of the results which the RERTR Program had achieved by the end of 1996 in collaboration with its many international partners is followed by a detailed review of the major events, findings, and activities of 1997. Significant progress has been made during the past year. In the area of U.S. acceptance of spent fuel from foreign research reactors, several shipments have taken place and additional are being planned. Intense fuel development activities are in progress, including procurement ofmore » equipment, screening of candidate materials, and production of microplates. Irradiation of the first series of microplates began in August 1997 in the Advanced Test Reactor, in Idaho. Progress has been made in the Russian RERTR program, which aims to develop and demonstrate within five years the technical means needed to convert Russian-supplied research reactors to LEU fuels. The study of an alternative LEU core for the FRM-II design has been extended to address, with favorable results, controversial performance issues which were raised at last year's meeting. Progress was also made on several aspects of producing molybdenum-99 from fission targets utilizing LEU instead of HEU. Various types of targets and processes are being pursued, with FDA approval of an LEU process projected to occur within two years. The feasibility of LEU Fuel conversion for three important DOE research reactors (BMRR, HFBR, and HFIR) has been evaluated by the RERTR program. In spite of the many momentous events which have occurred during the intervening years, and the excellent progress achieved, the most important challenges that the RERTR program faces today are not very different in type from those that were faced during the first RERTR meeting. Now, as then, the most important task is to develop new LEU fuels satisfying requirements which cannot be

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…

Measurement of the ROT effect in the neutron induced fission of 235U in the 0.3 eV resonance at a hot source of polarized neutrons

NASA Astrophysics Data System (ADS)

Kopatch, Yuri; Novitsky, Vadim; Ahmadov, Gadir; Gagarsky, Alexei; Berikov, Daniyar; Danilyan, Gevorg; Hutanu, Vladimir; Klenke, Jens; Masalovich, Sergey

2018-03-01

The TRI and ROT asymmetries in fission of heavy nuclei have been extensively studied during more than a decade. The effects were first discovered in the ternary fission in a series of experiments performed at the ILL reactor (Grenoble) by a collaboration of Russian and European institutes, and were carefully measured for a number of fissioning nuclei. Later on, the ROT effect has been observed in the emission of prompt gamma rays and neutrons in fission of 235U and 233U, although its value was an order of magnitude smaller than in the α-particle emission from ternary fission. All experiments performed so far are done with cold polarized neutrons, what assumes a mixture of several spin states, the weights of these states being not well known. The present paper describes the first attempt to get "clean" data by performing the measurement of gamma and neutron asymmetries in an isolated resonance of 235U at the POLI instrument of the FRM2 reactor in Garching.

CONVECTION REACTOR

DOEpatents

Hammond, R.P.; King, L.D.P.

1960-03-22

An homogeneous nuclear power reactor utilizing convection circulation of the liquid fuel is proposed. The reactor has an internal heat exchanger looated in the same pressure vessel as the critical assembly, thereby eliminating necessity for handling the hot liquid fuel outside the reactor pressure vessel during normal operation. The liquid fuel used in this reactor eliminates the necessity for extensive radiolytic gas rocombination apparatus, and the reactor is resiliently pressurized and, without any movable mechanical apparatus, automatically regulates itself to the condition of criticality during moderate variations in temperature snd pressure and shuts itself down as the pressure exceeds a predetermined safe operating value.

Spent caustic oxidation using electro-generated Fenton's reagent in a batch reactor.

PubMed

Rodriguez, Nicolas; Hansen, Henrik K; Nunez, Patricio; Guzman, Jaime

2008-07-01

This work shows the results of four Electro-Fenton laboratory tests to reduce the chemical oxygen demand (COD) in spent caustic solutions. The treatment consisted of (i) a pH reduction followed by (ii) an Electro-Fenton process, which was analyzed in this work. The Fenton's reagent was produced in a specially designed reactor, where the waste stream flowed through a labyrinth made by ferrous plates. These plates acted as sacrificial anodes-releasing Fe(2 +) cations to the solution, where H(2)O(2) was also added. The Electro-Fenton process was analyzed varying the ferrous ion concentration ([Fe(+ 2)]), the spent caustic's initial temperature and the initial pH. Close to 95% removal of COD (from 8800 mg L(- 1)) was achieved at a pH of 4, a temperature of 40 degrees C and 100 mg L(- 1) of Fe(+ 2) (applying 1 A). Two models were considered to simulate the behavior of the reactor considering (i) axial dispersion and (ii) kinetic rate, respectively. The model that was based on kinetics, proved to be the slightly closest fit to the experimental values.

BOILING REACTORS

DOEpatents

Untermyer, S.

1962-04-10

A boiling reactor having a reactivity which is reduced by an increase in the volume of vaporized coolant therein is described. In this system unvaporized liquid coolant is extracted from the reactor, heat is extracted therefrom, and it is returned to the reactor as sub-cooled liquid coolant. This reduces a portion of the coolant which includes vaporized coolant within the core assembly thereby enhancing the power output of the assembly and rendering the reactor substantially self-regulating. (AEC)

Light Water Reactor Sustainability Program Advanced Instrumentation, Information, and Control Systems Technologies Technical Program Plan for FY 2016

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

Hallbert, Bruce Perry; Thomas, Kenneth David

2015-10-01

Reliable instrumentation, information, and control (II&C) systems technologies are essential to ensuring safe and efficient operation of the U.S. light water reactor (LWR) fleet. These technologies affect every aspect of nuclear power plant (NPP) and balance-of-plant operations. In 1997, the National Research Council conducted a study concerning the challenges involved in modernization of digital instrumentation and control systems in NPPs. Their findings identified the need for new II&C technology integration.

Nuclear Reactor Physics

NASA Astrophysics Data System (ADS)

Stacey, Weston M.

2001-02-01

An authoritative textbook and up-to-date professional's guide to basic and advanced principles and practices Nuclear reactors now account for a significant portion of the electrical power generated worldwide. At the same time, the past few decades have seen an ever-increasing number of industrial, medical, military, and research applications for nuclear reactors. Nuclear reactor physics is the core discipline of nuclear engineering, and as the first comprehensive textbook and reference on basic and advanced nuclear reactor physics to appear in a quarter century, this book fills a large gap in the professional literature. Nuclear Reactor Physics is a textbook for students new to the subject, for others who need a basic understanding of how nuclear reactors work, as well as for those who are, or wish to become, specialists in nuclear reactor physics and reactor physics computations. It is also a valuable resource for engineers responsible for the operation of nuclear reactors. Dr. Weston Stacey begins with clear presentations of the basic physical principles, nuclear data, and computational methodology needed to understand both the static and dynamic behaviors of nuclear reactors. This is followed by in-depth discussions of advanced concepts, including extensive treatment of neutron transport computational methods. As an aid to comprehension and quick mastery of computational skills, he provides numerous examples illustrating step-by-step procedures for performing the calculations described and chapter-end problems. Nuclear Reactor Physics is a useful textbook and working reference. It is an excellent self-teaching guide for research scientists, engineers, and technicians involved in industrial, research, and military applications of nuclear reactors, as well as government regulators who wish to increase their understanding of nuclear reactors.

Light Water Reactor Sustainability Program: Digital Technology Business Case Methodology Guide

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

Thomas, Ken; Lawrie, Sean; Hart, Adam

The Department of Energy’s (DOE’s) Light Water Reactor Sustainability Program aims to develop and deploy technologies that will make the existing U.S. nuclear fleet more efficient and competitive. The program has developed a standard methodology for determining the impact of new technologies in order to assist nuclear power plant (NPP) operators in building sound business cases. The Advanced Instrumentation, Information, and Control (II&C) Systems Technologies Pathway is part of the DOE’s Light Water Reactor Sustainability (LWRS) Program. It conducts targeted research and development (R&D) to address aging and reliability concerns with the legacy instrumentation and control and related information systemsmore » of the U.S. operating light water reactor (LWR) fleet. This work involves two major goals: (1) to ensure that legacy analog II&C systems are not life-limiting issues for the LWR fleet and (2) to implement digital II&C technology in a manner that enables broad innovation and business improvement in the NPP operating model. Resolving long-term operational concerns with the II&C systems contributes to the long-term sustainability of the LWR fleet, which is vital to the nation’s energy and environmental security. The II&C Pathway is conducting a series of pilot projects that enable the development and deployment of new II&C technologies in existing nuclear plants. Through the LWRS program, individual utilities and plants are able to participate in these projects or otherwise leverage the results of projects conducted at demonstration plants. Performance advantages of the new pilot project technologies are widely acknowledged, but it has proven difficult for utilities to derive business cases for justifying investment in these new capabilities. Lack of a business case is often cited by utilities as a barrier to pursuing wide-scale application of digital technologies to nuclear plant work activities. The decision to move forward with funding usually

NUCLEAR REACTOR

DOEpatents

Treshow, M.

1961-09-01

A boiling-water nuclear reactor is described wherein control is effected by varying the moderator-to-fuel ratio in the reactor core. This is accomplished by providing control tubes containing a liquid control moderator in the reactor core and providing means for varying the amount of control moderatcr within the control tubes.

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

Cooking Matters for Adults Improves Food Resource Management Skills and Self-confidence Among Low-Income Participants.

PubMed

Pooler, Jennifer A; Morgan, Ruth E; Wong, Karen; Wilkin, Margaret K; Blitstein, Jonathan L

Determine the impact of Cooking Matters for Adults (CM) on food resource management (FRM) skills and self-confidence 6 months after course completion. Quasi-experimental design with nonequivalent comparison group and 6-month follow-up. Cooking Matters for Adults programs in CA, CO, ME, MA, MI, and OR. Participants in CM attending classes in April to July, 2016 (n = 332); comparison group (n = 336). Cooking Matters for Adults educated low-income adults to shop for and prepare healthy meals economically using hands-on meal preparation, facilitated discussion, and an interactive grocery store tour. Classes met for 2 hours, once a week for 6 weeks. Food resource management practices; FRM self-confidence (ie, in shopping for and preparing healthy foods on a budget); worrying that food might run out. Pearson's chi-square test and t tests identified measures associated with outcomes of interest and between-group differences. Repeated-measures linear mixed models with fixed and random effects were used to examine differences in outcomes between participants in CM and nonequivalent comparison group and to estimate the treatment effect of the program at 3 and 6 months after course completion. Six months after course completion, CM participants demonstrated improvements in all outcome measures of interest: Use of FRM practices improved (P = .002) as did FRM confidence (P < .001). Participants also worried less that food would run out before they had money to buy more (P = .03). This study demonstrated a positive impact of including FRM skills and confidence building in a nutrition education program, the effects of which could be seen for 6 months after participation in the program. Equipping low-income families with FRM skills allowed them to access healthier foods even during times of hardship. Copyright © 2017 Society for Nutrition Education and Behavior. All rights reserved.

Thermal neutron capture cross sections for 16,171,18O and 2H

NASA Astrophysics Data System (ADS)

Firestone, R. B.; Revay, Zs.

2016-04-01

Thermal neutron capture γ -ray spectra for 16,17,18O and 2H have been measured with guided cold neutron beams from the Forschungs-Neutronenquelle Heinz Maier-Leibnitz (FRM II) reactor and the Budapest Research Reactor (BRR) on natural and O,1817 enriched D2O targets. Complete neutron capture γ -ray decay schemes for the 16,17,18O(n ,γ ) reactions were measured. Absolute transition probabilities were determined for each reaction by a least-squares fit of the γ -ray intensities to the decay schemes after accounting for the contribution from internal conversion. The transition probability for the 870.76-keV γ ray from 16O(n ,γ ) was measured as Pγ(871 )=96.6 ±0.5 % and the thermal neutron cross section for this γ ray was determined as 0.164 ±0.003 mb by internal standardization with multiple targets containing oxygen and stoichiometric quantities of hydrogen, nitrogen, and carbon whose γ -ray cross sections were previously standardized. The γ -ray cross sections for the O,1817(n ,γ ) and 2H(n ,γ ) reactions were then determined relative to the 870.76-keV γ -ray cross section after accounting for the isotopic abundances in the targets. We determined the following total radiative thermal neutron cross sections for each isotope from the γ -ray cross sections and transition probabilities; σ0(16O )=0.170 ±0.003 mb; σ0(17O )=0.67 ±0.07 mb; σ0(18O )=0.141 ±0.006 mb; and σ0(2H )=0.489 ±0.006 mb.

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

NASA Astrophysics Data System (ADS)

Wright, Steven A.; Houts, Michael

2001-02-01

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

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

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

Van Lam Pham; Vinh Vinh Le; Ton Nghiem Huynh

2008-07-15

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

NEUTRONIC REACTOR SYSTEM

DOEpatents

Treshow, M.

1959-02-10

A reactor system incorporating a reactor of the heterogeneous boiling water type is described. The reactor is comprised essentially of a core submerged adwater in the lower half of a pressure vessel and two distribution rings connected to a source of water are disposed within the pressure vessel above the reactor core, the lower distribution ring being submerged adjacent to the uppcr end of the reactor core and the other distribution ring being located adjacent to the top of the pressure vessel. A feed-water control valve, responsive to the steam demand of the load, is provided in the feedwater line to the distribution rings and regulates the amount of feed water flowing to each distribution ring, the proportion of water flowing to the submerged distribution ring being proportional to the steam demand of the load. This invention provides an automatic means exterior to the reactor to control the reactivity of the reactor over relatively long periods of time without relying upon movement of control rods or of other moving parts within the reactor structure.

Automated power control system for reactor TRIGA PUSPATI

NASA Astrophysics Data System (ADS)

Ghazali, Anith Khairunnisa; Minhat, Mohd Sabri; Hassan, Mohd Khair

2017-01-01

Reactor TRIGA PUSPATI (RTP) Mark II type undergoes safe operation for more than 30 years and the only research reactor exists in Malaysia. The main safety feature of Instrumentation and Control (I&C) system design is such that any failure in the electronic, or its associated components, does not lead to an uncontrolled rate of reactivity. The existed controller using feedback approach to control the reactor power. This paper introduces proposed controllers such as Model Reference Adaptive Control (MRAC) and Proportional Integral Derivatives (PID) controller for the RTP simulation. In RTP, the most important considered parameter is the reactor power and act as nervous system. To design a controller for complex plant like RTP is quite difficult due to high cost and safety factors cause by the failure of the controller. Furthermore, to overcome these problems, a simulator can be used to replace functions the hardware and test could then be simulated using this simulator. In order to find the best controller, several controllers were proposed and the result will be analysed for study the performances of the controller. The output result will be used to find out the best RTP power controller using MATLAB/Simulink and gives result as close as the real RTP performances. Currently, the structures of RTP was design using MATLAB/Simulink tool that consist of fission chamber, controller, control rod position, height-to-worth of control rods and a RTP model. The controller will control the control rod position to make sure that the reactivity still under the limitation parameter. The results given from each controller will be analysed and validated through experiment data collected from RTP.

Reactor safety method

DOEpatents

Vachon, Lawrence J.

1980-03-11

This invention relates to safety means for preventing a gas cooled nuclear reactor from attaining criticality prior to start up in the event the reactor core is immersed in hydrogenous liquid. This is accomplished by coating the inside surface of the reactor coolant channels with a neutral absorbing material that will vaporize at the reactor's operating temperature.

Nuclear reactor neutron shielding

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

Speaker, Daniel P; Neeley, Gary W; Inman, James B

A nuclear reactor includes a reactor pressure vessel and a nuclear reactor core comprising fissile material disposed in a lower portion of the reactor pressure vessel. The lower portion of the reactor pressure vessel is disposed in a reactor cavity. An annular neutron stop is located at an elevation above the uppermost elevation of the nuclear reactor core. The annular neutron stop comprises neutron absorbing material filling an annular gap between the reactor pressure vessel and the wall of the reactor cavity. The annular neutron stop may comprise an outer neutron stop ring attached to the wall of the reactormore » cavity, and an inner neutron stop ring attached to the reactor pressure vessel. An excore instrument guide tube penetrates through the annular neutron stop, and a neutron plug comprising neutron absorbing material is disposed in the tube at the penetration through the neutron stop.« less

Spinning fluids reactor

DOEpatents

Miller, Jan D; Hupka, Jan; Aranowski, Robert

2012-11-20

A spinning fluids reactor, includes a reactor body (24) having a circular cross-section and a fluid contactor screen (26) within the reactor body (24). The fluid contactor screen (26) having a plurality of apertures and a circular cross-section concentric with the reactor body (24) for a length thus forming an inner volume (28) bound by the fluid contactor screen (26) and an outer volume (30) bound by the reactor body (24) and the fluid contactor screen (26). A primary inlet (20) can be operatively connected to the reactor body (24) and can be configured to produce flow-through first spinning flow of a first fluid within the inner volume (28). A secondary inlet (22) can similarly be operatively connected to the reactor body (24) and can be configured to produce a second flow of a second fluid within the outer volume (30) which is optionally spinning.

Full-length U-xPu-10Zr (x = 0, 8, 19 wt.%) fast reactor fuel test in FFTF

NASA Astrophysics Data System (ADS)

Porter, D. L.; Tsai, Hanchung

2012-08-01

The Integral Fast Reactor-1 (IFR-1) experiment performed in the Fast Flux Test Facility (FFTF) was the only U-Pu-10Zr (Pu-0, 8 and 19 wt.%) metallic fast reactor test with commercial-length (91.4-cm active fuel-column length) conducted to date. With few remaining test reactors, there is little opportunity for performing another test with a long active fuel column. The assembly was irradiated to the goal burnup of 10 at.%. The beginning-of-life (BOL) peak cladding temperature of the hottest pin was 608 °C, cooling to 522 °C at end-of-life (EOL). Selected fuel pins were examined non-destructively using neutron radiography, precision axial gamma scanning, and both laser and spiral contact cladding profilometry. Destructive exams included plenum gas pressure, volume, and gas composition determinations on a number of pins followed by optical metallography, electron probe microanalysis (EPMA), and alpha and beta-gamma autoradiography on a single U-19Pu-10Zr pin. The post-irradiation examinations (PIEs) showed very few differences compared to the short-pin (34.3-cm fuel column) testing performed on fuels of similar composition in Experimental Breeder Reactor-II (EBR-II). The fuel column grew axially slightly less than observed in the short pins, but with the same pattern of decreasing growth with increasing Pu content. There was a difference in the fuel-cladding chemical interaction (FCCI) in that the maximum cladding penetration by interdiffusion with fuel/fission products did not occur at the top of the fuel column where the cladding temperature is highest, as observed in EBR-II tests. Instead, the more exaggerated fission-rate profile of the FFTF pins resulted in a peak FCCI at ˜0.7 X/L axial location along the fuel column. This resulted from a higher production of rare-earth fission products at this location and a higher ΔT between fuel center and cladding than at core center, together providing more rare earths at the cladding and more FCCI. This behavior could

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

2016-11-01

Uncertainty and sensitivity analysis is an indispensable element of any substantial attempt in reactor simulation validation. The quantification of uncertainties in nuclear engineering has grown more important and the IAEA Coordinated Research Program (CRP) on High-Temperature Gas Cooled Reactor (HTGR) initiated in 2012 aims to investigate the various uncertainty quantification methodologies for this type of reactors. The first phase of the CRP is dedicated to the estimation of cell and lattice model uncertainties due to the neutron cross sections co-variances. Phase II is oriented towards the investigation of propagated uncertainties from the lattice to the coupled neutronics/thermal hydraulics core calculations.more » Nominal results for the prismatic single block (Ex.I-2a) and super cell models (Ex.I-2c) have been obtained using the SCALE 6.1.3 two-dimensional lattice code NEWT coupled to the TRITON sequence for cross section generation. In this work, the TRITON/NEWT-flux-weighted cross sections obtained for Ex.I-2a and various models of Ex.I-2c is utilized to perform a sensitivity analysis of the MHTGR-350 core power densities and eigenvalues. The core solutions are obtained with the INL coupled code PHISICS/RELAP5-3D, utilizing a fixed-temperature feedback for Ex. II-1a.. It is observed that the core power density does not vary significantly in shape, but the magnitude of these variations increases as the moderator-to-fuel ratio increases in the super cell lattice models.« less

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.

Considerations of Alloy 617 Application in the Gen IV Nuclear Reactor Systems - Part II: Metallurgical Property Challenges

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

Ren, Weiju

2010-01-01

Alloy 617 is currently considered as a leading candidate material for high temperature components in the Gen IV Nuclear Reactor Systems. Because of the unprecedented severe working conditions beyond its commercial service experience required by the Gen IV systems, the alloy faces various challenges in both mechanical and metallurgical properties. Following a previous paper discussing the mechanical property challenges, this paper is focused on the challenges and issues in metallurgical properties of the alloy for the intended nuclear application. Considerations are given in details about its metallurgical stability and aging evolution, aging effects on mechanical properties, potential Co hazard, andmore » internal oxidation. Some research and development activities are suggested with discussions on viability to satisfy the Gen IV Nuclear Reactor System needs.« less

In-situ material-motion diagnostics and fuel radiography in experimental reactors

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

DeVolpi, A.

1982-01-01

Material-motion monitoring has become a routine part of in-pile transient reactor-safety experiments. Diagnostic systems, such as the fast-neutron hodoscope, were developed for the purpose of providing direct time-resolved data on pre-failure fuel motion, cladding-breach time and location, and post-failure fuel relocation. Hodoscopes for this purpose have been installed at TREAT and CABRI; other types of imaging systems that have been tested are a coded-aperture at ACRR and a pinhole at TREAT. Diagnostic systems that use penetrating radiation emitted from the test section can non-invasively monitor fuel without damage to the measuring instrument during the radiographic images of test sections installedmore » in the reator. Studies have been made of applications of hodoscopes to other experimental reactors, including PBF, FARET, STF, ETR, EBR-II, SAREF-STF, and DMT.« less

Work Domain Analysis of a Predecessor Sodium-cooled Reactor as Baseline for AdvSMR Operational Concepts

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

Ronald Farris; David Gertman; Jacques Hugo

This report presents the results of the Work Domain Analysis for the Experimental Breeder Reactor (EBR-II). This is part of the phase of the research designed to incorporate Cognitive Work Analysis in the development of a framework for the formalization of an Operational Concept (OpsCon) for Advanced Small Modular Reactors (AdvSMRs). For a new AdvSMR design, information obtained through Cognitive Work Analysis, combined with human performance criteria, can and should be used in during the operational phase of a plant to assess the crew performance aspects associated with identified AdvSMR operational concepts. The main objective of this phase was tomore » develop an analytical and descriptive framework that will help systems and human factors engineers to understand the design and operational requirements of the emerging generation of small, advanced, multi-modular reactors. Using EBR-II as a predecessor to emerging sodium-cooled reactor designs required the application of a method suitable to the structured and systematic analysis of the plant to assist in identifying key features of the work associated with it and to clarify the operational and other constraints. The analysis included the identification and description of operating scenarios that were considered characteristic of this type of nuclear power plant. This is an invaluable aspect of Operational Concept development since it typically reveals aspects of future plant configurations that will have an impact on operations. These include, for example, the effect of core design, different coolants, reactor-to-power conversion unit ratios, modular plant layout, modular versus central control rooms, plant siting, and many more. Multi-modular plants in particular are expected to have a significant impact on overall OpsCon in general, and human performance in particular. To support unconventional modes of operation, the modern control room of a multi-module plant would typically require advanced HSIs that

Mackinawite (FeS) Reduces Mercury(II) under Sulfidic Conditions

PubMed Central

2015-01-01

Mercury (Hg) is a toxicant of global concern that accumulates in organisms as methyl Hg. The production of methyl Hg by anaerobic bacteria may be limited in anoxic sediments by the sequestration of divalent Hg [Hg(II)] into a solid phase or by the formation of elemental Hg [Hg(0)]. We tested the hypothesis that nanocrystalline mackinawite (tetragonal FeS), which is abundant in sediments where Hg is methylated, both sorbs and reduces Hg(II). Mackinawite suspensions were equilibrated with dissolved Hg(II) in batch reactors. Examination of the solid phase using Hg LIII-edge extended X-ray absorption fine structure (EXAFS) spectroscopy showed that Hg(II) was indeed reduced in FeS suspensions. Measurement of purgeable Hg using cold vapor atomic fluorescence spectrometry (CVAFS) from FeS suspensions and control solutions corroborated the production of Hg(0) that was observed spectroscopically. However, a fraction of the Hg(II) initially added to the suspensions remained in the divalent state, likely in the form of β-HgS-like clusters associated with the FeS surface or as a mixture of β-HgS and surface-associated species. Complexation by dissolved S(-II) in anoxic sediments hinders Hg(0) formation, but, by contrast, Hg(II)–S(-II) species are reduced in the presence of mackinawite, producing Hg(0) after only 1 h of reaction time. The results of our work support the idea that Hg(0) accounts for a significant fraction of the total Hg in wetland and estuarine sediments. PMID:25180562

Influence of fast alpha diffusion and thermal alpha buildup on tokamak reactor performance

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

Uckan, N.A.; Tolliver, J.S.; Houlberg, W.A.

1987-11-01

The effect of fast alpha diffusion and thermal alpha accumulation on the confinement capability of a candidate Engineering Test Reactor (ETR) plasma (Tokamak Ignition/Burn Experimental Reactor (TIBER-II)) in achieving ignition and steady-state driven operation has been assessed using both global and 1-1/2-D transport models. Estimates are made of the threshold for radial diffusion of fast alphas and thermal alpha buildup. It is shown that a relatively low level of radial transport, when combined with large gradients in the fast alpha density, leads to a significant radial flow with a deleterious effect on plasma performance. Similarly, modest levels of thermal alphamore » concentration significantly influence the ignition and steady-state burn capability. 23 refs., 9 figs., 4 tabs.« less

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

PubMed

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

2017-12-01

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

Request for Naval Reactors Comment on Proposed Prometheus Space Flight Nuclear Reactor High Tier Reactor Safety Requirements and for Naval Reactors Approval to Transmit These Requirements to JPL

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

D. Kokkinos

2005-04-28

The purpose of this letter is to request Naval Reactors comments on the nuclear reactor high tier requirements for the PROMETHEUS space flight reactor design, pre-launch operations, launch, ascent, operation, and disposal, and to request Naval Reactors approval to transmit these requirements to Jet Propulsion Laboratory to ensure consistency between the reactor safety requirements and the spacecraft safety requirements. The proposed PROMETHEUS nuclear reactor high tier safety requirements are consistent with the long standing safety culture of the Naval Reactors Program and its commitment to protecting the health and safety of the public and the environment. In addition, the philosophymore » on which these requirements are based is consistent with the Nuclear Safety Policy Working Group recommendations on space nuclear propulsion safety (Reference 1), DOE Nuclear Safety Criteria and Specifications for Space Nuclear Reactors (Reference 2), the Nuclear Space Power Safety and Facility Guidelines Study of the Applied Physics Laboratory.« less

Thermos reactors

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

Labrousse, M.; Lerouge, B.; Dupuy, G.

1978-04-01

THERMOS is a water reactor designed to provide hot water up to 120/sup 0/C for district heating or for desalination applications. It is a 100-MW reactor based on proven technology: oxide fuel plate elements, integrated primary circuit, and reactor vessel located in the bottom of a pool. As in swimming pool reactors, the pool is used for biological shielding, emergency core cooling, and fission product filtering (in case of an accident). Before economics, safety is the main characteristic of the concept: no fuel failure admitted, core under water in any accidental configuration, inspection of every ''nuclear'' component, and double-wall containment.

NEUTRONIC REACTOR

DOEpatents

Daniels, F.

1959-10-27

A reactor in which at least a portion of the moderator is in the form of movable refractory balls is described. In addition to their moderating capacity, these balls may serve as carriers for fissionable material or fertile material, or may serve in a coolant capacity to remove heat from the reactor. A pneumatic system is used to circulate the balls through the reactor.

Peak broadening and peak shift pole figures investigations by STRESS-SPEC diffractometer at FRM II

NASA Astrophysics Data System (ADS)

Gan, W. M.; Randau, C.; Hofmann, M.; Brokmeier, H. G.; Mueller, M.; Schreyer, A.

2012-02-01

This paper studied for the first time peak intensity, peak position and FHWM pole figures with one time measurement at the neutron diffractometer STRESS-SPEC via in-situ tensile deformation on austenitic steel. Fibre distribution with its evolution from central tensile direction to normal direction of these three kinds of pole figures was obtained. Variation of peak position and FWHM can be correlated to the reorientation of the texture component.

The RERTR Program : a status report.

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

Travelli, A.

1998-10-19

This paper describes the progress achieved by the Reduced Enrichment for Research and Test Reactors (RERTR) Program in collaboration with its many international partners since its inception in 1978. A brief summary of the results that the program had attained by the end of 1997 is followed by a detailed review of the major events, findings, and activities that took place in 1998. The past year was characterized by exceptionally important accomplishments and events for the RERTR program. Four additional shipments of spent fuel from foreign research reactors were accepted by the U.S. Altogether, 2,231 spent fuel assemblies from foreignmore » research reactors have been received by the U.S. under the acceptance policy. Fuel development activities began to yield solid results. Irradiations of the first two batches of microplates were completed. Preliminary postirradiation examinations of these microplates indicate excellent irradiation behavior of some of the fuel materials that were tested. These materials hold the promise of achieving the pro am goal of developing LEU research reactor fuels with uranium density in the 8-9 g /cm{sup 3} range. Progress was made in the Russian RERTR program, which aims to develop and demonstrate the technical means needed to convert Russian-supplied research reactors to LEU fuels. Feasibility studies for converting to LEU fuel four Russian-designed research reactors (IR-8 in Russia, Budapest research reactor in Hungary, MARIA in Poland, and WWR-SM in Uzbekistan) were completed. A new program activity began to study the feasibility of converting three Russian plutonium production reactors to the use of low-enriched U0{sub 2}-Al dispersion fuel, so that they can continue to produce heat and electricity without producing significant amounts of plutonium. The study of an alternative LEU core for the FRM-II design has been extended to address, with favorable results, the transient performance of the core under hypothetical accident

On the factors influencing the performance of solar reactors for water disinfection with photosensitized singlet oxygen.

PubMed

Manjón, Francisco; Villén, Laura; García-Fresnadillo, David; Orellana, Guillermo

2008-01-01

Two solar reactors based on compound parabolic collectors (CPCs) were optimized for water disinfection by photosensitized singlet oxygen (1O2) production in the heterogeneous phase. Sensitizing materials containing Ru(II) complexes immobilized on porous silicone were produced, photochemically characterized, and successfully tested for the inactivation of up to 10(4) CFU mL(-1) of waterborne Escherichia coli (gram-negative) or Enterococcus faecalis (gram-positive) bacteria. The main factors determining the performance of the solar reactors are the type of photosensitizing material, the sensitizer loading, the CPC collector geometry (fin- vs coaxial-type), the fluid rheology, and the balance between concurrent photothermal--photolytic and 1O2 effects on the microorganisms' inactivation. In this way, at the 40 degrees N latitude of Spain, water can be disinfected on a sunny day (0.6-0.8 MJ m(-2) L(-1) accumulated solar radiation dose in the 360-700 nm range, typically 5-6 h of sunlight) with a fin-type reactor containing 0.6 m2 of photosensitizing material saturated with tris(4,7-diphenyl-1,10-phenanthroline)ruthenium(II) (ca. 2.0 g m(-2)). The optimum rheological conditions require laminar-to-transitional water flow in both prototypes. The fin-type system showed better inactivation efficiency than the coaxial reactor due to a more important photolytic contribution. The durability of the sensitizing materials was tested and the operational lifetime of the photocatalyst is at least three months without any reduction in the bacteria inactivation efficiency. Solar water disinfection with 1O2-generating films is demonstrated to be an effective technique for use in isolated regions of developing countries with high yearly average sunshine.

Control Means for Reactor

DOEpatents

Manley, J. H.

1961-06-27

An apparatus for controlling a nuclear reactor includes a tank just below the reactor, tubes extending from the tank into the reactor, and a thermally expansible liquid neutron absorbent material in the tank. The liquid in the tank is exposed to a beam of neutrons from the reactor which heats the liquid causing it to expand into the reactor when the neutron flux in the reactor rises above a predetermincd danger point. Boron triamine may be used for this purpose.

REACTOR PHYSICS MODELING OF SPENT RESEARCH REACTOR FUEL FOR TECHNICAL NUCLEAR FORENSICS

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

Nichols, T.; Beals, D.; Sternat, M.

2011-07-18

Technical nuclear forensics (TNF) refers to the collection, analysis and evaluation of pre- and post-detonation radiological or nuclear materials, devices, and/or debris. TNF is an integral component, complementing traditional forensics and investigative work, to help enable the attribution of discovered radiological or nuclear material. Research is needed to improve the capabilities of TNF. One research area of interest is determining the isotopic signatures of research reactors. Research reactors are a potential source of both radiological and nuclear material. Research reactors are often the least safeguarded type of reactor; they vary greatly in size, fuel type, enrichment, power, and burn-up. Manymore » research reactors are fueled with highly-enriched uranium (HEU), up to {approx}93% {sup 235}U, which could potentially be used as weapons material. All of them have significant amounts of radiological material with which a radioactive dispersal device (RDD) could be built. Therefore, the ability to attribute if material originated from or was produced in a specific research reactor is an important tool in providing for the security of the United States. Currently there are approximately 237 operating research reactors worldwide, another 12 are in temporary shutdown and 224 research reactors are reported as shut down. Little is currently known about the isotopic signatures of spent research reactor fuel. An effort is underway at Savannah River National Laboratory (SRNL) to analyze spent research reactor fuel to determine these signatures. Computer models, using reactor physics codes, are being compared to the measured analytes in the spent fuel. This allows for improving the reactor physics codes in modeling research reactors for the purpose of nuclear forensics. Currently the Oak Ridge Research reactor (ORR) is being modeled and fuel samples are being analyzed for comparison. Samples of an ORR spent fuel assembly were taken by SRNL for analytical and

Planning and supervision of reactor defueling using discrete event techniques

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

Garcia, H.E.; Imel, G.R.; Houshyar, A.

1995-12-31

New fuel handling and conditioning activities for the defueling of the Experimental Breeder Reactor II are being performed at Argonne National Laboratory. Research is being conducted to investigate the use of discrete event simulation, analysis, and optimization techniques to plan, supervise, and perform these activities in such a way that productivity can be improved. The central idea is to characterize this defueling operation as a collection of interconnected serving cells, and then apply operational research techniques to identify appropriate planning schedules for given scenarios. In addition, a supervisory system is being developed to provide personnel with on-line information on themore » progress of fueling tasks and to suggest courses of action to accommodate changing operational conditions. This paper provides an introduction to the research in progress at ANL. In particular, it briefly describes the fuel handling configuration for reactor defueling at ANL, presenting the flow of material from the reactor grid to the interim storage location, and the expected contributions of this work. As an example of the studies being conducted for planning and supervision of fuel handling activities at ANL, an application of discrete event simulation techniques to evaluate different fuel cask transfer strategies is given at the end of the paper.« less

NUCLEAR REACTOR

DOEpatents

Grebe, J.J.

1959-07-14

High temperature reactors which are uniquely adapted to serve as the heat source for nuclear pcwered rockets are described. The reactor is comprised essentially of an outer tubular heat resistant casing which provides the main coolant passageway to and away from the reactor core within the casing and in which the working fluid is preferably hydrogen or helium gas which is permitted to vaporize from a liquid storage tank. The reactor core has a generally spherical shape formed entirely of an active material comprised of fissile material and a moderator material which serves as a diluent. The active material is fabricated as a gas permeable porous material and is interlaced in a random manner with very small inter-connecting bores or capillary tubes through which the coolant gas may flow. The entire reactor is divided into successive sections along the direction of the temperature gradient or coolant flow, each section utilizing materials of construction which are most advantageous from a nuclear standpoint and which at the same time can withstand the operating temperature of that particular zone. This design results in a nuclear reactor characterized simultaneously by a minimum critiral size and mass and by the ability to heat a working fluid to an extremely high temperature.

The diversity and unit of reactor noise theory

NASA Astrophysics Data System (ADS)

Kuang, Zhifeng

The study of reactor noise theory concerns questions about cause and effect relationships, and utilisation of random noise in nuclear reactor systems. The diversity of reactor noise theory arises from the variety of noise sources, the various mathematical treatments applied and various practical purposes. The neutron noise in zero- energy systems arises from the fluctuations in the number of neutrons per fission, the time between nuclear events, and the type of reactions. It can be used to evaluate system parameters. The mathematical treatment is based on the master equation of stochastic branching processes. The noise in power reactor systems is given rise by random processes of technological origin such as vibration of mechanical parts, boiling of the coolant, fluctuations of temperature and pressure. It can be used to monitor reactor behaviour with the possibility of detecting malfunctions at an early stage. The mathematical treatment is based on the Langevin equation. The unity of reactor noise theory arises from the fact that useful information from noise is embedded in the second moments of random variables, which lends the possibility of building up a unified mathematical description and analysis of the various reactor noise sources. Exploring such possibilities is the main subject among the three major topics reported in this thesis. The first subject is within the zero power noise in steady media, and we reported on the extension of the existing theory to more general cases. In Paper I, by use of the master equation approach, we have derived the most general Feynman- and Rossi-alpha formulae so far by taking the full joint statistics of the prompt and all the six groups of delayed neutron precursors, and a multiple emission source into account. The involved problems are solved with a combination of effective analytical techniques and symbolic algebra codes (Mathematica). Paper II gives a numerical evaluation of these formulae. An assessment of the

NEUTRONIC REACTOR

DOEpatents

Fraas, A.P.; Mills, C.B.

1961-11-21

A neutronic reactor in which neutron moderation is achieved primarily in its reflector is described. The reactor structure consists of a cylindrical central "island" of moderator and a spherical moderating reflector spaced therefrom, thereby providing an annular space. An essentially unmoderated liquid fuel is continuously passed through the annular space and undergoes fission while contained therein. The reactor, because of its small size, is particularly adapted for propulsion uses, including the propulsion of aircraft. (AEC)

Effect of Fe (II) in low-nitrogen sewage on the reactor performance and microbial community of an ANAMMOX biofilter.

PubMed

Zhang, Xiaojing; Zhou, Yue; Zhao, Siyu; Zhang, Rongrong; Peng, Zhaoxue; Zhai, Hanfei; Zhang, Hongzhong

2018-06-01

In this study, the effect of Fe (II) on Anaerobic Ammonium Oxidation (ANAMMOX) process was investigated by step-wise increasing the Fe (II) in influent from 1 to 50 mg L -1 . The nitrogen removal, biofilm property and the microbial community were analyzed in each phase. Results showed that, the anaerobic ammonia-oxidizing bacteria (AAOB) bioactivity and the nitrogen removal of ANAMMOX system were slightly improved to 0.58 from the initial 0.51 kg m -3 d -1 by Fe (II) in 1-5 mg L -1 . The nitrogen removal was suppressed and could recover to the initial level during the same period under 10-20 mg L -1 Fe (II), while it did not recover to the initial level under 30 mg L -1 Fe (II) and showed no recovery performance under 50 mg L -1 Fe (II). The irreversible suppression threshold of Fe (II) was calculated as 50 mg L -1 . The iron content in ANAMMOX biofilm presented linear correlation with the influent Fe (II) in 1-20 mg L -1 , which then tended to be stable when Fe (II) was higher. Dehydrogenase activity (DHA) showed similar and faster response to Fe (II) than the microbial activity, and it was an effective pre-indicator for the nitrogen removal performance in the ANAMMOX system suffered Fe (II). The Fe (II) feeding firstly led to the relative abundance of AAOB decreased to 11.04% from the initial 35.46%, and finally picked up to 19.39% after the long-term acclimatization. Copyright © 2018 Elsevier Ltd. All rights reserved.

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

NASA Astrophysics Data System (ADS)

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

2016-07-01

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

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

DOE PAGES

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

2016-07-05

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

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

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

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

2016-07-05

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

Isotopic evidence for nitrous oxide production pathways in a partial nitritation-anammox reactor.

PubMed

Harris, Eliza; Joss, Adriano; Emmenegger, Lukas; Kipf, Marco; Wolf, Benjamin; Mohn, Joachim; Wunderlin, Pascal

2015-10-15

Nitrous oxide (N2O) production pathways in a single stage, continuously fed partial nitritation-anammox reactor were investigated using online isotopic analysis of offgas N2O with quantum cascade laser absorption spectroscopy (QCLAS). N2O emissions increased when reactor operating conditions were not optimal, for example, high dissolved oxygen concentration. SP measurements indicated that the increase in N2O was due to enhanced nitrifier denitrification, generally related to nitrite build-up in the reactor. The results of this study confirm that process control via online N2O monitoring is an ideal method to detect imbalances in reactor operation and regulate aeration, to ensure optimal reactor conditions and minimise N2O emissions. Under normal operating conditions, the N2O isotopic site preference (SP) was much higher than expected - up to 40‰ - which could not be explained within the current understanding of N2O production pathways. Various targeted experiments were conducted to investigate the characteristics of N2O formation in the reactor. The high SP measurements during both normal operating and experimental conditions could potentially be explained by a number of hypotheses: i) unexpectedly strong heterotrophic N2O reduction, ii) unknown inorganic or anammox-associated N2O production pathway, iii) previous underestimation of SP fractionation during N2O production from NH2OH, or strong variations in SP from this pathway depending on reactor conditions. The second hypothesis - an unknown or incompletely characterised production pathway - was most consistent with results, however the other possibilities cannot be discounted. Further experiments are needed to distinguish between these hypotheses and fully resolve N2O production pathways in PN-anammox systems. Copyright © 2015 Elsevier Ltd. All rights reserved.

Classification of Reactor Facility Operational State Using SPRT Methods with Radiation Sensor Networks

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

Ramirez Aviles, Camila A.; Rao, Nageswara S.

We consider the problem of inferring the operational state of a reactor facility by using measurements from a radiation sensor network, which is deployed around the facility’s ventilation stack. The radiation emissions from the stack decay with distance, and the corresponding measurements are inherently random with parameters determined by radiation intensity levels at the sensor locations. We fuse measurements from network sensors to estimate the intensity at the stack, and use this estimate in a one-sided Sequential Probability Ratio Test (SPRT) to infer the on/off state of the reactor facility. We demonstrate the superior performance of this method over conventionalmore » majority vote fusers and individual sensors using (i) test measurements from a network of NaI sensors, and (ii) emulated measurements using radioactive effluents collected at a reactor facility stack. We analytically quantify the performance improvements of individual sensors and their networks with adaptive thresholds over those with fixed ones, by using the packing number of the radiation intensity space.« less

Reactor performances and microbial communities of biogas reactors: effects of inoculum sources.

PubMed

Han, Sheng; Liu, Yafeng; Zhang, Shicheng; Luo, Gang

2016-01-01

Anaerobic digestion is a very complex process that is mediated by various microorganisms, and the understanding of the microbial community assembly and its corresponding function is critical in order to better control the anaerobic process. The present study investigated the effect of different inocula on the microbial community assembly in biogas reactors treating cellulose with various inocula, and three parallel biogas reactors with the same inoculum were also operated in order to reveal the reproducibility of both microbial communities and functions of the biogas reactors. The results showed that the biogas production, volatile fatty acid (VFA) concentrations, and pH were different for the biogas reactors with different inocula, and different steady-state microbial community patterns were also obtained in different biogas reactors as reflected by Bray-Curtis similarity matrices and taxonomic classification. It indicated that inoculum played an important role in shaping the microbial communities of biogas reactor in the present study, and the microbial community assembly in biogas reactor did not follow the niche-based ecology theory. Furthermore, it was found that the microbial communities and reactor performances of parallel biogas reactors with the same inoculum were different, which could be explained by the neutral-based ecology theory and stochastic factors should played important roles in the microbial community assembly in the biogas reactors. The Bray-Curtis similarity matrices analysis suggested that inoculum affected more on the microbial community assembly compared to stochastic factors, since the samples with different inocula had lower similarity (10-20 %) compared to the samples from the parallel biogas reactors (30 %).

NUCLEAR REACTOR

DOEpatents

Miller, H.I.; Smith, R.C.

1958-01-21

This patent relates to nuclear reactors of the type which use a liquid fuel, such as a solution of uranyl sulfate in ordinary water which acts as the moderator. The reactor is comprised of a spherical vessel having a diameter of about 12 inches substantially surrounded by a reflector of beryllium oxide. Conventionnl control rods and safety rods are operated in slots in the reflector outside the vessel to control the operation of the reactor. An additional means for increasing the safety factor of the reactor by raising the ratio of delayed neutrons to prompt neutrons, is provided and consists of a soluble sulfate salt of beryllium dissolved in the liquid fuel in the proper proportion to obtain the result desired.

The application of probabilistic fracture analysis to residual life evaluation of embrittled reactor vessels

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

Dickson, T.L.; Simonen, F.A.

1992-05-01

Probabilistic fracture mechanics analysis is a major element of comprehensive probabilistic methodology on which current NRC regulatory requirements for pressurized water reactor vessel integrity evaluation are based. Computer codes such as OCA-P and VISA-II perform probabilistic fracture analyses to estimate the increase in vessel failure probability that occurs as the vessel material accumulates radiation damage over the operating life of the vessel. The results of such analyses, when compared with limits of acceptable failure probabilities, provide an estimation of the residual life of a vessel. Such codes can be applied to evaluate the potential benefits of plant-specific mitigating actions designedmore » to reduce the probability of failure of a reactor vessel. 10 refs.« less

The application of probabilistic fracture analysis to residual life evaluation of embrittled reactor vessels

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

Dickson, T.L.; Simonen, F.A.

1992-01-01

Probabilistic fracture mechanics analysis is a major element of comprehensive probabilistic methodology on which current NRC regulatory requirements for pressurized water reactor vessel integrity evaluation are based. Computer codes such as OCA-P and VISA-II perform probabilistic fracture analyses to estimate the increase in vessel failure probability that occurs as the vessel material accumulates radiation damage over the operating life of the vessel. The results of such analyses, when compared with limits of acceptable failure probabilities, provide an estimation of the residual life of a vessel. Such codes can be applied to evaluate the potential benefits of plant-specific mitigating actions designedmore » to reduce the probability of failure of a reactor vessel. 10 refs.« less

Reactor water cleanup system

DOEpatents

Gluntz, Douglas M.; Taft, William E.

1994-01-01

A reactor water cleanup system includes a reactor pressure vessel containing a reactor core submerged in reactor water. First and second parallel cleanup trains are provided for extracting portions of the reactor water from the pressure vessel, cleaning the extracted water, and returning the cleaned water to the pressure vessel. Each of the cleanup trains includes a heat exchanger for cooling the reactor water, and a cleaner for cleaning the cooled reactor water. A return line is disposed between the cleaner and the pressure vessel for channeling the cleaned water thereto in a first mode of operation. A portion of the cooled water is bypassed around the cleaner during a second mode of operation and returned through the pressure vessel for shutdown cooling.

Reactor water cleanup system

DOEpatents

Gluntz, D.M.; Taft, W.E.

1994-12-20

A reactor water cleanup system includes a reactor pressure vessel containing a reactor core submerged in reactor water. First and second parallel cleanup trains are provided for extracting portions of the reactor water from the pressure vessel, cleaning the extracted water, and returning the cleaned water to the pressure vessel. Each of the cleanup trains includes a heat exchanger for cooling the reactor water, and a cleaner for cleaning the cooled reactor water. A return line is disposed between the cleaner and the pressure vessel for channeling the cleaned water thereto in a first mode of operation. A portion of the cooled water is bypassed around the cleaner during a second mode of operation and returned through the pressure vessel for shutdown cooling. 1 figure.

System and method for the analysis of one or more compounds and/or species produced by a solution-based nuclear reactor

DOEpatents

Policke, Timothy A; Nygaard, Eric T

2014-05-06

The present invention relates generally to both a system and method for determining the composition of an off-gas from a solution nuclear reactor (e.g., an Aqueous Homogeneous Reactor (AHR)) and the composition of the fissioning solution from those measurements. In one embodiment, the present invention utilizes at least one quadrupole mass spectrometer (QMS) in a system and/or method designed to determine at least one or more of: (i) the rate of production of at least one gas and/or gas species from a nuclear reactor; (ii) the effect on pH by one or more nitrogen species; (iii) the rate of production of one or more fission gases; and/or (iv) the effect on pH of at least one gas and/or gas species other than one or more nitrogen species from a nuclear reactor.

97. ARAIII. ML1 reactor has been moved into GCRE reactor ...

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

97. ARA-III. ML-1 reactor has been moved into GCRE reactor building (ARA-608) for examination of corrosion on its underside and repair. May 24, 1963. Ineel photo no. 63-3485. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

NEUTRONIC REACTOR SHIELDING

DOEpatents

Borst, L.B.

1961-07-11

A special hydrogenous concrete shielding for reactors is described. In addition to Portland cement and water, the concrete essentially comprises 30 to 60% by weight barytes aggregate for enhanced attenuation of fast neutrons. The biological shields of AEC's Oak Ridge Graphite Reactor and Materials Testing Reactor are particular embodiments.

REACTOR

DOEpatents

Roman, W.G.

1961-06-27

A pressurized water reactor in which automatic control is achieved by varying the average density of the liquid moderator-cooiant is patented. Density is controlled by the temperature and power level of the reactor ftself. This control can be effected by the use of either plate, pellet, or tubular fuel elements. The fuel elements are disposed between upper and lower coolant plenum chambers and are designed to permit unrestricted coolant flow. The control chamber has an inlet opening communicating with the lower coolant plenum chamber and a restricted vapor vent communicating with the upper coolant plenum chamber. Thus, a variation in temperature of the fuel elements will cause a variation in the average moderator density in the chamber which directly affects the power level of the reactor.

Pressurized fluidized bed reactor

DOEpatents

Isaksson, J.

1996-03-19

A pressurized fluid bed reactor power plant includes a fluidized bed reactor contained within a pressure vessel with a pressurized gas volume between the reactor and the vessel. A first conduit supplies primary gas from the gas volume to the reactor, passing outside the pressure vessel and then returning through the pressure vessel to the reactor, and pressurized gas is supplied from a compressor through a second conduit to the gas volume. A third conduit, comprising a hot gas discharge, carries gases from the reactor, through a filter, and ultimately to a turbine. During normal operation of the plant, pressurized gas is withdrawn from the gas volume through the first conduit and introduced into the reactor at a substantially continuously controlled rate as the primary gas to the reactor. In response to an operational disturbance of the plant, the flow of gas in the first, second, and third conduits is terminated, and thereafter the pressure in the gas volume and in the reactor is substantially simultaneously reduced by opening pressure relief valves in the first and third conduits, and optionally by passing air directly from the second conduit to the turbine. 1 fig.

Pressurized fluidized bed reactor

DOEpatents

Isaksson, Juhani

1996-01-01

A pressurized fluid bed reactor power plant includes a fluidized bed reactor contained within a pressure vessel with a pressurized gas volume between the reactor and the vessel. A first conduit supplies primary gas from the gas volume to the reactor, passing outside the pressure vessel and then returning through the pressure vessel to the reactor, and pressurized gas is supplied from a compressor through a second conduit to the gas volume. A third conduit, comprising a hot gas discharge, carries gases from the reactor, through a filter, and ultimately to a turbine. During normal operation of the plant, pressurized gas is withdrawn from the gas volume through the first conduit and introduced into the reactor at a substantially continuously controlled rate as the primary gas to the reactor. In response to an operational disturbance of the plant, the flow of gas in the first, second, and third conduits is terminated, and thereafter the pressure in the gas volume and in the reactor is substantially simultaneously reduced by opening pressure relief valves in the first and third conduits, and optionally by passing air directly from the second conduit to the turbine.

Neutron fluxes in test reactors

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

Youinou, Gilles Jean-Michel

Communicate the fact that high-power water-cooled test reactors such as the Advanced Test Reactor (ATR), the High Flux Isotope Reactor (HFIR) or the Jules Horowitz Reactor (JHR) cannot provide fast flux levels as high as sodium-cooled fast test reactors. The memo first presents some basics physics considerations about neutron fluxes in test reactors and then uses ATR, HFIR and JHR as an illustration of the performance of modern high-power water-cooled test reactors.

151. ARAIII Reactor building (ARA608) Details of reactor pit and ...

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

151. ARA-III Reactor building (ARA-608) Details of reactor pit and instrument plan. Aerojet-general 880-area/GCRE-608-T-19. Date: November 1958. Ineel index code no. 063-0608-25-013-102678. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

NEUTRONIC REACTOR MANIPULATING DEVICE

DOEpatents

Ohlinger, L.A.

1962-08-01

A cable connecting a control rod in a reactor with a motor outside the reactor for moving the rod, and a helical conduit in the reactor wall, through which the cable passes are described. The helical shape of the conduit prevents the escape of certain harmful radiations from the reactor. (AEC)

Guideline for Performing Systematic Approach to Evaluate and Qualify Legacy Documents that Support Advanced Reactor Technology Activity

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

Honma, George

The establishment of a systematic process for the evaluation of historic technology information for use in advanced reactor licensing is described. Efforts are underway to recover and preserve Experimental Breeder Reactor II and Fast Flux Test Facility historical data. These efforts have generally emphasized preserving information from data-acquisition systems and hard-copy reports and entering it into modern electronic formats suitable for data retrieval and examination. The guidance contained in this document has been developed to facilitate consistent and systematic evaluation processes relating to quality attributes of historic technical information (with focus on sodium-cooled fast reactor (SFR) technology) that will bemore » used to eventually support licensing of advanced reactor designs. The historical information may include, but is not limited to, design documents for SFRs, research-and-development (R&D) data and associated documents, test plans and associated protocols, operations and test data, international research data, technical reports, and information associated with past U.S. Nuclear Regulatory Commission (NRC) reviews of SFR designs. The evaluation process is prescribed in terms of SFR technology, but the process can be used to evaluate historical information for any type of advanced reactor technology. An appendix provides a discussion of typical issues that should be considered when evaluating and qualifying historical information for advanced reactor technology fuel and source terms, based on current light water reactor (LWR) requirements and recent experience gained from Next Generation Nuclear Plant (NGNP).« less

A high performance neutron powder diffractometer at 3 MW Triga Mark-II research reactor in Bangladesh

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

Kamal, I., E-mail: imtiaz-kamal26@yahoo.com; Yunus, S. M., E-mail: yunussm11@yahoo.com; Datta, T. K., E-mail: tk-datta4@yahoo.com

2016-07-12

A high performance neutron diffractometer called Savar Neutron Diffractometer (SAND) was built and installed at radial beam port-2 of TRIGA Mark II research reactor at AERE, Savar, Dhaka, Bangladesh. Structural studies of materials are being done by this technique to characterize materials crystallograpohically and magnetically. The micro-structural information obtainable by neutron scattering method is very essential for determining its technological applications. This technique is unique for understanding the magnetic behavior in magnetic materials. Ceramic, steel, electronic and electric industries can be benefited from this facility for improving their products and fabrication process. This instrument consists of a Popovicimonochromator with amore » large linear position sensitive detector array. The monochromator consists of nine blades of perfect single crystal of silicon with 6 mm thickness each. The monochromator design was optimized to provide maximum flux on 3 mm diameter cylindrical sample with a relatively flat angular dependence of resolution. Five different wave lengths can be selected by orienting the crystal at various angles. A sapphire filter was used before the primary collimator to minimize the first neutron. The detector assembly is composed of 15 linear position sensitive proportional counters placed at either 1.1 m or 1.6 m from the sample position and enclosed in a air pad supported high density polythene shield. Position sensing is obtained by charge division using 1-wide NIM position encoding modules (PEM). The PEMs communicate with the host computer via USB. The detector when placed at 1.1 m, subtends 30° (2θ) at each step and covers 120° in 4 steps. When the detector is placed at 1.6 m it subtends 20° at each step and covers 120° in 6 steps. The instrument supports both low and high temperature sample environment. The instrument supports both low and high temperature sample environment. The diffractometer is a state-of-the art

A high performance neutron powder diffractometer at 3 MW Triga Mark-II research reactor in Bangladesh

NASA Astrophysics Data System (ADS)

Kamal, I.; Yunus, S. M.; Datta, T. K.; Zakaria, A. K. M.; Das, A. K.; Aktar, S.; Hossain, S.; Berliner, R.; Yelon, W. B.

2016-07-01

A high performance neutron diffractometer called Savar Neutron Diffractometer (SAND) was built and installed at radial beam port-2 of TRIGA Mark II research reactor at AERE, Savar, Dhaka, Bangladesh. Structural studies of materials are being done by this technique to characterize materials crystallograpohically and magnetically. The micro-structural information obtainable by neutron scattering method is very essential for determining its technological applications. This technique is unique for understanding the magnetic behavior in magnetic materials. Ceramic, steel, electronic and electric industries can be benefited from this facility for improving their products and fabrication process. This instrument consists of a Popovicimonochromator with a large linear position sensitive detector array. The monochromator consists of nine blades of perfect single crystal of silicon with 6mm thickness each. The monochromator design was optimized to provide maximum flux on 3mm diameter cylindrical sample with a relatively flat angular dependence of resolution. Five different wave lengths can be selected by orienting the crystal at various angles. A sapphire filter was used before the primary collimator to minimize the first neutron. The detector assembly is composed of 15 linear position sensitive proportional counters placed at either 1.1 m or 1.6 m from the sample position and enclosed in a air pad supported high density polythene shield. Position sensing is obtained by charge division using 1-wide NIM position encoding modules (PEM). The PEMs communicate with the host computer via USB. The detector when placed at 1.1 m, subtends 30˚ (2θ) at each step and covers 120˚ in 4 steps. When the detector is placed at 1.6 m it subtends 20˚ at each step and covers 120˚ in 6 steps. The instrument supports both low and high temperature sample environment. The instrument supports both low and high temperature sample environment. The diffractometer is a state-of-the art technology

A Comparison of Four Gravimetric Fine Particle Sampling Methods.

PubMed

Yanosky, Jeff D; Maclntosh, David L

2001-06-01

A study was conducted to compare four gravimetric methods of measuring fine particle (PM 2.5 ) concentrations in air: the BGI, Inc. PQ200 Federal Reference Method PM 2.5 (FRM) sampler; the Harvard-Marple Impactor (HI); the BGI, Inc. GK2.05 KTL Respirable/Thoracic Cyclone (KTL); and the AirMetrics MiniVol (MiniVol). Pairs of FRM, HI, and KTL samplers and one MiniVol sampler were collocated and 24-hr integrated PM 2.5 samples were collected on 21 days from January 6 through April 9, 2000. The mean and standard deviation of PM 2.5 levels from the FRM samplers were 13.6 and 6.8 μg/m 3 , respectively. Significant systematic bias was found between mean concentrations from the FRM and the MiniVol (1.14 μg/m 3 , p = 0.0007), the HI and the MiniVol (0.85 μg/m 3 , p = 0.0048), and the KTL and the MiniVol (1.23 μg/m 3 , p = 0.0078) according to paired t test analyses. Linear regression on all pairwise combinations of the sampler types was used to evaluate measurements made by the samplers. None of the regression intercepts was significantly different from 0, and only two of the regression slopes were significantly different from 1, that for the FRM and the MiniVol [β 1 = 0.91, 95% CI (0.83-0.99)] and that for the KTL and the MiniVol [ = 0.88, 95% CI (0.78-0.98)]. Regression R 2 terms were 0.96 or greater between all pairs of samplers, and regression root mean square error terms (RMSE) were 1.65 μg/m 3 or less. These results suggest that the MiniVol will underestimate measurements made by the FRM, the HI, and the KTL by an amount proportional to PM 2.5 concentration. Nonetheless, these results indicate that all of the sampler types are comparable if ~10% variation on the mean levels and on individual measurement levels is considered acceptable and the actual concentration is within the range of this study (5-35 μg/m 3 ).

A comparison of four gravimetric fine particle sampling methods.

PubMed

Yanosky, J D; MacIntosh, D L

2001-06-01

A study was conducted to compare four gravimetric methods of measuring fine particle (PM2.5) concentrations in air: the BGI, Inc. PQ200 Federal Reference Method PM2.5 (FRM) sampler; the Harvard-Marple Impactor (HI); the BGI, Inc. GK2.05 KTL Respirable/Thoracic Cyclone (KTL); and the AirMetrics MiniVol (MiniVol). Pairs of FRM, HI, and KTL samplers and one MiniVol sampler were collocated and 24-hr integrated PM2.5 samples were collected on 21 days from January 6 through April 9, 2000. The mean and standard deviation of PM2.5 levels from the FRM samplers were 13.6 and 6.8 microg/m3, respectively. Significant systematic bias was found between mean concentrations from the FRM and the MiniVol (1.14 microg/m3, p = 0.0007), the HI and the MiniVol (0.85 microg/m3, p = 0.0048), and the KTL and the MiniVol (1.23 microg/m3, p = 0.0078) according to paired t test analyses. Linear regression on all pairwise combinations of the sampler types was used to evaluate measurements made by the samplers. None of the regression intercepts was significantly different from 0, and only two of the regression slopes were significantly different from 1, that for the FRM and the MiniVol [beta1 = 0.91, 95% CI (0.83-0.99)] and that for the KTL and the MiniVol [beta1 = 0.88, 95% CI (0.78-0.98)]. Regression R2 terms were 0.96 or greater between all pairs of samplers, and regression root mean square error terms (RMSE) were 1.65 microg/m3 or less. These results suggest that the MiniVol will underestimate measurements made by the FRM, the HI, and the KTL by an amount proportional to PM2.5 concentration. Nonetheless, these results indicate that all of the sampler types are comparable if approximately 10% variation on the mean levels and on individual measurement levels is considered acceptable and the actual concentration is within the range of this study (5-35 microg/m3).

Characteristics and verification of a car-borne survey system for dose rates in air: KURAMA-II.

PubMed

Tsuda, S; Yoshida, T; Tsutsumi, M; Saito, K

2015-01-01

The car-borne survey system KURAMA-II, developed by the Kyoto University Research Reactor Institute, has been used for air dose rate mapping after the Fukushima Dai-ichi Nuclear Power Plant accident. KURAMA-II consists of a CsI(Tl) scintillation detector, a GPS device, and a control device for data processing. The dose rates monitored by KURAMA-II are based on the G(E) function (spectrum-dose conversion operator), which can precisely calculate dose rates from measured pulse-height distribution even if the energy spectrum changes significantly. The characteristics of KURAMA-II have been investigated with particular consideration to the reliability of the calculated G(E) function, dose rate dependence, statistical fluctuation, angular dependence, and energy dependence. The results indicate that 100 units of KURAMA-II systems have acceptable quality for mass monitoring of dose rates in the environment. Copyright © 2014 Elsevier Ltd. All rights reserved.

Direct Contact Heat Exchange Interfacial Phenomena for Liquid Metal Reactors: Part II - Void Fraction

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

Abdulla, S.; Liu, X.; Anderson, M.H.

One concept being considered for steam generation in innovative nuclear reactor applications, involves water coming into direct contact with a circulating molten metal. The vigorous agitation of the two fluids, the direct liquid-liquid contact and the consequent large interfacial area can give rise to large heat transfer coefficients and rapid steam generation. For an optimum design of such direct contact heat exchange and vaporization systems, detailed knowledge is necessary of the various flow regimes, interfacial transport phenomena, heat transfer and operational stability. In order to investigate the interfacial transport phenomena, heat transfer and operational stability of direct liquid-liquid contact, amore » series of experiments are being performed in a 1-d test facility at Argonne National Laboratory and a 2-d experimental facility at UW-Madison. Each of the experimental facilities primarily consist of a liquid-metal melt chamber, heated test section (10 cm diameter tube for 1-d facility and 10 cm 50 cm rectangle for 2-d facility), water injection system and steam suppression tank. This paper is part II which, primarily addresses results and analysis of a set of preliminary experiments and void fraction measurements conducted in the 2-d facility at UW-Madison, part I deals with the heat transfer in the 1-d test facility at Argonne National Laboratory. A real-time high energy X-ray imaging system was developed and utilized to visualize the multiphase flow and measure line-average local void fractions, time-dependent void fraction distribution as well as estimates of the vapor bubble sizes and velocities. These measurements allowed us to determine the volumetric heat transfer coefficient and gain insight into the local heat transfer mechanisms. In this study, the images were captured at frame rates of 100 fps with spatial resolution of about 7 mm with a full-field view of a 15 cm square and five different positions along the test section height. The full

Upgrade of Irradiation Test Capability of the Experimental Fast Reactor Joyo

NASA Astrophysics Data System (ADS)

Sekine, Takashi; Aoyama, Takafumi; Suzuki, Soju; Yamashita, Yoshioki

2003-06-01

The JOYO MK-II core was operated from 1983 to 2000 as fast neutron irradiation bed. In order to meet various requirements for irradiation tests for development of FBRs, the JOYO upgrading project named MK-III program was initiated. The irradiation capability in the MK-III core will be about four times larger than that of the MK-II core. Advanced irradiation test subassemblies such as capsule type subassembly and on-line instrumentation rig are planned. As an innovative reactor safety system, the irradiation test of Self-Actuated Shutdown System (SASS) will be conducted. In order to improve the accuracy of neutron fluence, the core management code system was upgraded, and the Monte Carlo code and Helium Accumulation Fluence Monitor (HAFM) were applied. The MK-III core is planned to achieve initial criticality in July 2003.

155. ARAIII Reactor building (ARA608) Details of reactor pit showing ...

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

155. ARA-III Reactor building (ARA-608) Details of reactor pit showing tray supports and fuel element storage rack. Aerojet-general 880-area/GCRE-608-MS-2. Date: November 1958. Ineel index code no. 063-0608-40-013-102625. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

FAST NEUTRON REACTOR

DOEpatents

Soodak, H.; Wigner, E.P.

1961-07-25

A reactor comprising fissionable material in concentration sufficiently high so that the average neutron enengy within the reactor is at least 25,000 ev is described. A natural uranium blanket surrounds the reactor, and a moderating reflector surrounds the blanket. The blanket is thick enough to substantially eliminate flow of neutrons from the reflector.

NEUTRONIC REACTOR

DOEpatents

Hurwitz, H. Jr.; Brooks, H.; Mannal, C.; Payne, J.H.; Luebke, E.A.

1959-03-24

A reactor of the heterogeneous, liquid cooled type is described. This reactor is comprised of a central region of a plurality of vertically disposed elongated tubes surrounded by a region of moderator material. The central region is comprised of a central core surrounded by a reflector region which is surrounded by a fast neutron absorber region, which in turn is surrounded by a slow neutron absorber region. Liquid sodium is used as the primary coolant and circulates through the core which contains the fuel elements. Control of the reactor is accomplished by varying the ability of the reflector region to reflect neutrons back into the core of the reactor. For this purpose the reflector is comprised of moderator and control elements having varying effects on reactivity, the control elements being arranged and actuated by groups to give regulation, shim, and safety control.

F Reactor Inspection

ScienceCinema

Grindstaff, Keith; Hathaway, Boyd; Wilson, Mike

2018-01-16

Workers from Mission Support Alliance, LLC., removed the welds around the steel door of the F Reactor before stepping inside the reactor to complete its periodic inspection. This is the first time the Department of Energy (DOE) has had the reactor open since 2008. The F Reactor is one of nine reactors along the Columbia River at the Department's Hanford Site in southeastern Washington State, where environmental cleanup has been ongoing since 1989. As part of the Tri-Party Agreement, the Department completes surveillance and maintenance activities of cocooned reactors periodically to evaluate the structural integrity of the safe storage enclosure and to ensure confinement of any remaining hazardous materials. "This entry marks a transition of sorts because the Hanford Long-Term Stewardship Program, for the first time, was responsible for conducting the entry and surveillance and maintenance activities," said Keith Grindstaff, Energy Department Long-Term Stewardship Program Manager. "As the River Corridor cleanup work is completed and transitioned to long-term stewardship, our program will manage any on-going requirements."

F Reactor Inspection

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

Grindstaff, Keith; Hathaway, Boyd; Wilson, Mike

2014-10-29

Workers from Mission Support Alliance, LLC., removed the welds around the steel door of the F Reactor before stepping inside the reactor to complete its periodic inspection. This is the first time the Department of Energy (DOE) has had the reactor open since 2008. The F Reactor is one of nine reactors along the Columbia River at the Department's Hanford Site in southeastern Washington State, where environmental cleanup has been ongoing since 1989. As part of the Tri-Party Agreement, the Department completes surveillance and maintenance activities of cocooned reactors periodically to evaluate the structural integrity of the safe storage enclosuremore » and to ensure confinement of any remaining hazardous materials. "This entry marks a transition of sorts because the Hanford Long-Term Stewardship Program, for the first time, was responsible for conducting the entry and surveillance and maintenance activities," said Keith Grindstaff, Energy Department Long-Term Stewardship Program Manager. "As the River Corridor cleanup work is completed and transitioned to long-term stewardship, our program will manage any on-going requirements."« less

ASME Material Challenges for Advanced Reactor Concepts

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

Piyush Sabharwall; Ali Siahpush

2013-07-01

This study presents the material Challenges associated with Advanced Reactor Concept (ARC) such as the Advanced High Temperature Reactor (AHTR). ACR are the next generation concepts focusing on power production and providing thermal energy for industrial applications. The efficient transfer of energy for industrial applications depends on the ability to incorporate cost-effective heat exchangers between the nuclear heat transport system and industrial process heat transport system. The heat exchanger required for AHTR is subjected to a unique set of conditions that bring with them several design challenges not encountered in standard heat exchangers. The corrosive molten salts, especially at highermore » temperatures, require materials throughout the system to avoid corrosion, and adverse high-temperature effects such as creep. Given the very high steam generator pressure of the supercritical steam cycle, it is anticipated that water tube and molten salt shell steam generators heat exchanger will be used. In this paper, the ASME Section III and the American Society of Mechanical Engineers (ASME) Section VIII requirements (acceptance criteria) are discussed. Also, the ASME material acceptance criteria (ASME Section II, Part D) for high temperature environment are presented. Finally, lack of ASME acceptance criteria for thermal design and analysis are discussed.« less

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

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

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

2016-07-07

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

Evaluation of the performance of high temperature conversion reactors for compound-specific oxygen stable isotope analysis.

PubMed

Hitzfeld, Kristina L; Gehre, Matthias; Richnow, Hans-Hermann

2017-05-01

In this study conversion conditions for oxygen gas chromatography high temperature conversion (HTC) isotope ratio mass spectrometry (IRMS) are characterised using qualitative mass spectrometry (IonTrap). It is shown that physical and chemical properties of a given reactor design impact HTC and thus the ability to accurately measure oxygen isotope ratios. Commercially available and custom-built tube-in-tube reactors were used to elucidate (i) by-product formation (carbon dioxide, water, small organic molecules), (ii) 2nd sources of oxygen (leakage, metal oxides, ceramic material), and (iii) required reactor conditions (conditioning, reduction, stability). The suitability of the available HTC approach for compound-specific isotope analysis of oxygen in volatile organic molecules like methyl tert-butyl ether is assessed. Main problems impeding accurate analysis are non-quantitative HTC and significant carbon dioxide by-product formation. An evaluation strategy combining mass spectrometric analysis of HTC products and IRMS 18 O/ 16 O monitoring for future method development is proposed.

International academic program in technologies of light-water nuclear reactors. Phases of development and implementation

NASA Astrophysics Data System (ADS)

Geraskin, N. I.; Glebov, V. B.

2017-01-01

The results of implementation of European educational projects CORONA and CORONA II dedicated to preserving and further developing nuclear knowledge and competencies in the area of technologies of light-water nuclear reactors are analyzed. Present article addresses issues of design and implementation of the program for specialized training in the branch of technologies of light-water nuclear reactors. The systematic approach has been used to construct the program for students of nuclear specialties, which corresponding to IAEA standards and commonly accepted nuclear principles recognized in the European Union. Possibilities of further development of the international cooperation between countries and educational institutions are analyzed. Special attention is paid to e-learning/distance training, nuclear knowledge preservation and interaction with European Nuclear Education Network.

REACTOR

DOEpatents

Szilard, L.

1963-09-10

A breeder reactor is described, including a mass of fissionable material that is less than critical with respect to unmoderated neutrons and greater than critical with respect to neutrons of average energies substantially greater than thermal, a coolant selected from sodium or sodium--potassium alloys, a control liquid selected from lead or lead--bismuth alloys, and means for varying the quantity of control liquid in the reactor. (AEC)

Improvement of the trace metal composition of medium for nitrite-dependent anaerobic methane oxidation bacteria: Iron (II) and copper (II) make a difference.

PubMed

He, Zhanfei; Geng, Sha; Pan, Yawei; Cai, Chaoyang; Wang, Jiaqi; Wang, Liqiao; Liu, Shuai; Zheng, Ping; Xu, Xinhua; Hu, Baolan

2015-11-15

Nitrite-dependent anaerobic methane oxidation (n-damo) is a potential bioprocess for treating nitrogen-containing wastewater. This process uses methane, an inexpensive and nontoxic end-product of anaerobic digestion, as an external electron donor. However, the low turnover rate and slow growth rate of n-damo functional bacteria limit the practical application of this process. In the present study, the short- and long-term effects of variations in trace metal concentrations on n-damo bacteria were investigated, and the concentrations of trace metal elements of medium were improved. The results were subsequently verified by a group of long-term inoculations (90 days) and were applied in a sequencing batch reactor (SBR) (84 days). The results indicated that iron (Fe(II)) and copper (Cu(II)) (20 and 10 μmol L(-1), respectively) significantly stimulated the activity and the growth of n-damo bacteria, whereas other trace metal elements, including zinc (Zn), molybdenum (Mo), cobalt (Co), manganese (Mn), and nickel (Ni), had no significant effect on n-damo bacteria in the tested concentration ranges. Interestingly, fluorescence in situ hybridization (FISH) showed that a large number of dense, large aggregates (10-50 μm) of n-damo bacteria were formed by cell adhesion in the SBR reactor after using the improved medium, and to our knowledge this is the first discovery of large aggregates of n-damo bacteria. Copyright © 2015 Elsevier Ltd. All rights reserved.

Green Silver Nanoparticles Based Dual Sensor for Toxic Hg (II) Ions.

PubMed

Sebastian, Maria; Aravind, Archana; Mathew, Beena

2018-06-11

The present study focuses on the utilization of green silver nanoparticles as they are more preferred for sensing applications due to their environment friendly nature. We have examined the optical and electrochemical sensing behavior of silver nanoparticles from Agaricus Bispores (AgNP-AB) towards Hg(II) ions. AgNP-AB was prepared by microwave reactor. The synthesized AgNPs have been used for the sensing of Hg(II) ions without the use of modifiers or further sophisticated instrumentation. The synthesized nanoparticles were successfully characterized by different techniques. AgNP-AB leads to aggregation with addition of Hg(II) ions in aqueous medium and developed a color change from brown to black which leads to the formation of AgNP-AB-Hg(II) complex. Moreover, the metal sensing ability of AgNPs has been explored using electrochemical studies. AgNP-AB modified platinum electrode (AgNP-AB/PE) was developed for the fast sensing of toxic Hg(II) ions. The sensor exhibits good limit of detection at 2.1x10-6M. The sensitivity of AgNP-AB/PE towards Hg(II) ion was analyzed with various metal ions. The sensing skill of developed system was successfully checked with real water sample from Vembanade Lake, Kumarakom, Kerala. The silver nanoparticles from Agaricus Bispoes are highly versatile and promising for different environmental applications. © 2018 IOP Publishing Ltd.

Green synthesis of isopropyl myristate in novel single phase medium Part II: Packed bed reactor (PBR) studies.

PubMed

Vadgama, Rajeshkumar N; Odaneth, Annamma A; Lali, Arvind M

2015-12-01

Isopropyl myristate is a useful functional molecule responding to the requirements of numerous fields of application in cosmetic, pharmaceutical and food industry. In the present work, lipase-catalyzed production of isopropyl myristate by esterification of myristic acid with isopropyl alcohol (molar ratio of 1:15) in the homogenous reaction medium was performed on a bench-scale packed bed reactors, in order to obtain suitable reaction performance data for upscaling. An immobilized lipase B from Candida antartica was used as the biocatalyst based on our previous study. The process intensification resulted in a clean and green synthesis process comprising a series of packed bed reactors of immobilized enzyme and water dehydrant. In addition, use of the single phase reaction system facilitates efficient recovery of the product with no effluent generated and recyclability of unreacted substrates. The single phase reaction system coupled with a continuous operating bioreactor ensures a stable operational life for the enzyme.

Pebble bed modular reactor safeguards: developing new approaches and implementing safeguards by design

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

Beyer, Brian David; Beddingfield, David H; Durst, Philip

2010-01-01

The design of the Pebble Bed Modular Reactor (PBMR) does not fit or seem appropriate to the IAEA safeguards approach under the categories of light water reactor (LWR), on-load refueled reactor (OLR, i.e. CANDU), or Other (prismatic HTGR) because the fuel is in a bulk form, rather than discrete items. Because the nuclear fuel is a collection of nuclear material inserted in tennis-ball sized spheres containing structural and moderating material and a PBMR core will contain a bulk load on the order of 500,000 spheres, it could be classified as a 'Bulk-Fuel Reactor.' Hence, the IAEA should develop unique safeguardsmore » criteria. In a multi-lab DOE study, it was found that an optimized blend of: (i) developing techniques to verify the plutonium content in spent fuel pebbles, (ii) improving burn-up computer codes for PBMR spent fuel to provide better understanding of the core and spent fuel makeup, and (iii) utilizing bulk verification techniques for PBMR spent fuel storage bins should be combined with the historic IAEA and South African approaches of containment and surveillance to verify and maintain continuity of knowledge of PBMR fuel. For all of these techniques to work the design of the reactor will need to accommodate safeguards and material accountancy measures to a far greater extent than has thus far been the case. The implementation of Safeguards-by-Design as the PBMR design progresses provides an approach to meets these safeguards and accountancy needs.« less

Modeling integrated fixed-film activated sludge and moving-bed biofilm reactor systems II: evaluation.

PubMed

Boltz, Joshua P; Johnson, Bruce R; Daigger, Glen T; Sandino, Julian; Elenter, Deborah

2009-06-01

A steady-state model presented by Boltz, Johnson, Daigger, and Sandino (2009) describing integrated fixed-film activated sludge (IFAS) and moving-bed biofilm reactor (MBBR) systems has been demonstrated to simulate, with reasonable accuracy, four wastewater treatment configurations with published operational data. Conditions simulated include combined carbon oxidation and nitrification (both IFAS and MBBR), tertiary nitrification MBBR, and post denitrification IFAS with methanol addition as the external carbon source. Simulation results illustrate that the IFAS/MBBR model is sufficiently accurate for describing ammonia-nitrogen reduction, nitrate/nitrite-nitrogen reduction and production, biofilm and suspended biomass distribution, and sludge production.

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

Hybrid adsorptive membrane reactor

NASA Technical Reports Server (NTRS)

Tsotsis, Theodore T. (Inventor); Sahimi, Muhammad (Inventor); Fayyaz-Najafi, Babak (Inventor); Harale, Aadesh (Inventor); Park, Byoung-Gi (Inventor); Liu, Paul K. T. (Inventor)

2011-01-01

A hybrid adsorbent-membrane reactor in which the chemical reaction, membrane separation, and product adsorption are coupled. Also disclosed are a dual-reactor apparatus and a process using the reactor or the apparatus.

Hybrid adsorptive membrane reactor

DOEpatents

Tsotsis, Theodore T [Huntington Beach, CA; Sahimi, Muhammad [Altadena, CA; Fayyaz-Najafi, Babak [Richmond, CA; Harale, Aadesh [Los Angeles, CA; Park, Byoung-Gi [Yeosu, KR; Liu, Paul K. T. [Lafayette Hill, PA

2011-03-01

A hybrid adsorbent-membrane reactor in which the chemical reaction, membrane separation, and product adsorption are coupled. Also disclosed are a dual-reactor apparatus and a process using the reactor or the apparatus.

Syngas Production By Thermochemical Conversion Of H2o And Co2 Mixtures Using A Novel Reactor Design

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

Pearlman, Howard; Chen, Chien-Hua

The Department of Energy awarded Advanced Cooling Technologies, Inc. (ACT) an SBIR Phase II contract (#DE-SC0004729) to develop a high-temperature solar thermochemical reactor for syngas production using water and/or carbon dioxide as feedstocks. The technology aims to provide a renewable and sustainable alternative to fossil fuels, promote energy independence and mitigate adverse issues associated with climate change by essentially recycling carbon from carbon dioxide emitted by the combustion of hydrocarbon fuels. To commercialize the technology and drive down the cost of solar fuels, new advances are needed in materials development and reactor design, both of which are integral elements inmore » this program.« less

NUCLEAR REACTOR

DOEpatents

Moore, R.V.; Bowen, J.H.; Dent, K.H.

1958-12-01

A heterogeneous, natural uranium fueled, solid moderated, gas cooled reactor is described, in which the fuel elements are in the form of elongated rods and are dlsposed within vertical coolant channels ln the moderator symmetrically arranged as a regular lattice in groups. This reactor employs control rods which operate in vertical channels in the moderator so that each control rod is centered in one of the fuel element groups. The reactor is enclosed in a pressure vessel which ls provided with access holes at the top to facilitate loading and unloadlng of the fuel elements, control rods and control rod driving devices.

Cross-flow electrochemical reactor cells, cross-flow reactors, and use of cross-flow reactors for oxidation reactions

DOEpatents

Balachandran, Uthamalingam; Poeppel, Roger B.; Kleefisch, Mark S.; Kobylinski, Thaddeus P.; Udovich, Carl A.

1994-01-01

This invention discloses cross-flow electrochemical reactor cells containing oxygen permeable materials which have both electron conductivity and oxygen ion conductivity, cross-flow reactors, and electrochemical processes using cross-flow reactor cells having oxygen permeable monolithic cores to control and facilitate transport of oxygen from an oxygen-containing gas stream to oxidation reactions of organic compounds in another gas stream. These cross-flow electrochemical reactors comprise a hollow ceramic blade positioned across a gas stream flow or a stack of crossed hollow ceramic blades containing a channel or channels for flow of gas streams. Each channel has at least one channel wall disposed between a channel and a portion of an outer surface of the ceramic blade, or a common wall with adjacent blades in a stack comprising a gas-impervious mixed metal oxide material of a perovskite structure having electron conductivity and oxygen ion conductivity. The invention includes reactors comprising first and second zones seprated by gas-impervious mixed metal oxide material material having electron conductivity and oxygen ion conductivity. Prefered gas-impervious materials comprise at least one mixed metal oxide having a perovskite structure or perovskite-like structure. The invention includes, also, oxidation processes controlled by using these electrochemical reactors, and these reactions do not require an external source of electrical potential or any external electric circuit for oxidation to proceed.

NEUTRONIC REACTOR

DOEpatents

Ohlinger, L.A.; Wigner, E.P.; Weinberg, A.M.; Young, G.J.

1958-09-01

This patent relates to neutronic reactors of the heterogeneous water cooled type, and in particular to a fuel element charging and discharging means therefor. In the embodiment illustrated the reactor contains horizontal, parallel coolant tubes in which the fuel elements are disposed. A loading cart containing a magnzine for holding a plurality of fuel elements operates along the face of the reactor at the inlet ends of the coolant tubes. The loading cart is equipped with a ram device for feeding fuel elements from the magazine through the inlot ends of the coolant tubes. Operating along the face adjacent the discharge ends of the tubes there is provided another cart means adapted to receive irradiated fuel elements as they are forced out of the discharge ends of the coolant tubes by the incoming new fuel elements. This cart is equipped with a tank coataining a coolant, such as water, into which the fuel elements fall, and a hydraulically operated plunger to hold the end of the fuel element being discharged. This inveation provides an apparatus whereby the fuel elements may be loaded into the reactor, irradiated therein, and unloaded from the reactor without stopping the fiow of the coolant and without danger to the operating personnel.

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.

Looking Southwest at Reactor Box Furnaces With Reactor Boxes and ...

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

Looking Southwest at Reactor Box Furnaces With Reactor Boxes and Repossessed Uranium in Recycle Recovery Building - Hematite Fuel Fabrication Facility, Recycle Recovery Building, 3300 State Road P, Festus, Jefferson County, MO

NEUTRONIC REACTOR

DOEpatents

Metcalf, H.E.; Johnson, H.W.

1961-04-01

BS>A nuclear reactor incorporating fuel rods passing through a moderator and including tubes of a material of higher Thermal conductivity than the fuel in contact with the fuel is described. The tubes extend beyond the active portion of the reactor into contant with a fiuld coolant.

A Virtual Reality Framework to Optimize Design, Operation and Refueling of GEN-IV Reactors.

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

Rizwan-uddin; Nick Karancevic; Stefano Markidis

2008-04-23

many GEN-IV candidate designs are currently under investigation. Technical issues related to material, safety and economics are being addressed at research laboratories, industry and in academia. After safety, economic feasibility is likely to be the most important crterion in the success of GEN-IV design(s). Lessons learned from the designers and operators of GEN-II (and GEN-III) reactors must play a vital role in achieving both safety and economic feasibility goals.

NEUTRONIC REACTOR POWER PLANT

DOEpatents

Metcalf, H.E.

1962-12-25

This patent relates to a nuclear reactor power plant incorporating an air-cooled, beryllium oxide-moderated, pebble bed reactor. According to the invention means are provided for circulating a flow of air through tubes in the reactor to a turbine and for directing a sidestream of the circu1ating air through the pebble bed to remove fission products therefrom as well as assist in cooling the reactor. (AEC)

REACTOR SHIELD

DOEpatents

Wigner, E.P.; Ohlinger, L.E.; Young, G.J.; Weinberg, A.M.

1959-02-17

Radiation shield construction is described for a nuclear reactor. The shield is comprised of a plurality of steel plates arranged in parallel spaced relationship within a peripheral shell. Reactor coolant inlet tubes extend at right angles through the plates and baffles are arranged between the plates at right angles thereto and extend between the tubes to create a series of zigzag channels between the plates for the circulation of coolant fluid through the shield. The shield may be divided into two main sections; an inner section adjacent the reactor container and an outer section spaced therefrom. Coolant through the first section may be circulated at a faster rate than coolant circulated through the outer section since the area closest to the reactor container is at a higher temperature and is more radioactive. The two sections may have separate cooling systems to prevent the coolant in the outer section from mixing with the more contaminated coolant in the inner section.

Fabrication and Testing of a Modular Micro-Pocket Fission Detector Instrumentation System for Test Nuclear Reactors

NASA Astrophysics Data System (ADS)

Reichenberger, Michael A.; Nichols, Daniel M.; Stevenson, Sarah R.; Swope, Tanner M.; Hilger, Caden W.; Roberts, Jeremy A.; Unruh, Troy C.; McGregor, Douglas S.

2018-01-01

Advancements in nuclear reactor core modeling and computational capability have encouraged further development of in-core neutron sensors. Measurement of the neutron-flux distribution within the reactor core provides a more complete understanding of the operating conditions in the reactor than typical ex-core sensors. Micro-Pocket Fission Detectors have been developed and tested previously but have been limited to single-node operation and have utilized highly specialized designs. The development of a widely deployable, multi-node Micro-Pocket Fission Detector assembly will enhance nuclear research capabilities. A modular, four-node Micro-Pocket Fission Detector array was designed, fabricated, and tested at Kansas State University. The array was constructed from materials that do not significantly perturb the neutron flux in the reactor core. All four sensor nodes were equally spaced axially in the array to span the fuel-region of the reactor core. The array was filled with neon gas, serving as an ionization medium in the small cavities of the Micro-Pocket Fission Detectors. The modular design of the instrument facilitates the testing and deployment of numerous sensor arrays. The unified design drastically improved device ruggedness and simplified construction from previous designs. Five 8-mm penetrations in the upper grid plate of the Kansas State University TRIGA Mk. II research nuclear reactor were utilized to deploy the array between fuel elements in the core. The Micro-Pocket Fission Detector array was coupled to an electronic support system which has been specially developed to support pulse-mode operation. The Micro-Pocket Fission Detector array composed of four sensors was used to monitor local neutron flux at a constant reactor power of 100 kWth at different axial locations simultaneously. The array was positioned at five different radial locations within the core to emulate the deployment of multiple arrays and develop a 2-dimensional measurement of

THE EXPERIENCE IN THE UNITED STATES WITH REACTOR OPERATION AND REACTOR SAFEGUARDS

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

McCullough, C.R.

1958-10-31

Reactors are operating or planned at locations in the United States in cities, near cities, and at remote locations. There is a general pattern that the higher power reactors are not in, but fairly uear cities, and the testing reactors for more hazardous experiments are at remote locations. A great deal has been done on the theoretical and experimental study of importunt features of reactor design. The metal-water reaction is still a theoretical possibility but tests of fuel element burnout under conditions approaching reactor operation gave no reaction. It appears that nucleate boiling does not necessarily result in steam blanketingmore » and fuel melting. Much attention is being given to the calculation of core kinetics but it is being found that temperature, power, and void coefficients cannot be calculated with accuracy and experiments are required. Some surprises are found giving positive localized void coefficients. Possible oscillatory behavior of reactors is being given careful study. No dangerous oscillations have been found in operating reactors but osciliations hare appeared in experimeats. The design of control and safety systems varies wvith different constructors. The relation of control to the kinetic behavior of the reactor is being studied. The importance of sensing element locations in order to know actual local reactor power level is being recognized. The time constants of instrumentation as related to reactor kinetics are being studied. Pressure vessels for reactors are being designed and manufactured. Many of these are beyond any previous experience. The stress problem is being given careful study. The effect of radiation is being studied experimentally. The stress problems of piping and pressure vessels is a difficult design problem being met successfully in reactor plants. The proper organization and procedure for operation of reactors is being evolved for resourch, testing, and power reactors. The importance of written standards and

Advanced Test Reactor Tour

ScienceCinema

Miley, Don

2017-12-21

The Advanced Test Reactor at Idaho National Laboratory is the foremost nuclear materials test reactor in the world. This virtual tour describes the reactor, how experiments are conducted, and how spent nuclear fuel is handled and stored.

PBF Reactor Building (PER620). Camera faces north into highbay/reactor pit ...

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

PBF Reactor Building (PER-620). Camera faces north into high-bay/reactor pit area. Inside from for reactor enclosure is in place. Photographer: John Capek. Date: March 15, 1967. INEEL negative no. 67-1769 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

Nuclear reactor control column

DOEpatents

Bachovchin, Dennis M.

1982-01-01

The nuclear reactor control column comprises a column disposed within the nuclear reactor core having a variable cross-section hollow channel and containing balls whose vertical location is determined by the flow of the reactor coolant through the column. The control column is divided into three basic sections wherein each of the sections has a different cross-sectional area. The uppermost section of the control column has the greatest cross-sectional area, the intermediate section of the control column has the smallest cross-sectional area, and the lowermost section of the control column has the intermediate cross-sectional area. In this manner, the area of the uppermost section can be established such that when the reactor coolant is flowing under normal conditions therethrough, the absorber balls will be lifted and suspended in a fluidized bed manner in the upper section. However, when the reactor coolant flow falls below a predetermined value, the absorber balls will fall through the intermediate section and into the lowermost section, thereby reducing the reactivity of the reactor core and shutting down the reactor.

Nuclear reactor overflow line

DOEpatents

Severson, Wayne J.

1976-01-01

The overflow line for the reactor vessel of a liquid-metal-cooled nuclear reactor includes means for establishing and maintaining a continuous bleed flow of coolant amounting to 5 to 10% of the total coolant flow through the overflow line to prevent thermal shock to the overflow line when the reactor is restarted following a trip. Preferably a tube is disposed concentrically just inside the overflow line extending from a point just inside the reactor vessel to an overflow tank and a suction line is provided opening into the body of liquid metal in the reactor vessel and into the annulus between the overflow line and the inner tube.

Use of freeze-casting in advanced burner reactor fuel design

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

Lang, A. L.; Yablinsky, C. A.; Allen, T. R.

2012-07-01

This paper will detail the modeling of a fast reactor with fuel pins created using a freeze-casting process. Freeze-casting is a method of creating an inert scaffold within a fuel pin. The scaffold is created using a directional solidification process and results in open porosity for emplacement of fuel, with pores ranging in size from 300 microns to 500 microns in diameter. These pores allow multiple fuel types and enrichments to be loaded into one fuel pin. Also, each pore could be filled with varying amounts of fuel to allow for the specific volume of fission gases created by thatmore » fuel type. Currently fast reactors, including advanced burner reactors (ABR's), are not economically feasible due to the high cost of operating the reactors and of reprocessing the fuel. However, if the fuel could be very precisely placed, such as within a freeze-cast scaffold, this could increase fuel performance and result in a valid design with a much lower cost per megawatt. In addition to competitive costs, freeze-cast fuel would also allow for selective breeding or burning of actinides within specific locations in fast reactors. For example, fast flux peak locations could be utilized on a minute scale to target specific actinides for transmutation. Freeze-cast fuel is extremely flexible and has great potential in a variety of applications. This paper performs initial modeling of freeze-cast fuel, with the generic fast reactor parameters for this model based on EBR-II. The core has an assumed power of 62.5 MWt. The neutronics code used was Monte Carlo N-Particle (MCNP5) transport code. Uniform pore sizes were used in increments of 100 microns. Two different freeze-cast scaffold materials were used: ceramic (MgO-ZrO{sub 2}) and steel (SS316L). Separate models were needed for each material because the freeze-cast ceramic and metal scaffolds have different structural characteristics and overall porosities. Basic criticality results were compiled for the various models

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.

Methanation assembly using multiple reactors

DOEpatents

Jahnke, Fred C.; Parab, Sanjay C.

2007-07-24

A methanation assembly for use with a water supply and a gas supply containing gas to be methanated in which a reactor assembly has a plurality of methanation reactors each for methanating gas input to the assembly and a gas delivery and cooling assembly adapted to deliver gas from the gas supply to each of said methanation reactors and to combine water from the water supply with the output of each methanation reactor being conveyed to a next methanation reactor and carry the mixture to such next methanation reactor.

NEUTRONIC REACTOR SYSTEM

DOEpatents

Goett, J.J.

1961-01-24

A system is described which includes a neutronic reactor containing a dispersion of fissionable material in a liquid moderator as fuel and a conveyor to which a portion of the dispersion may be passed and wherein the self heat of the slurry evaporates the moderator. Means are provided for condensing the liquid moderator and returning it to the reactor and for conveying the dried fissionable material away from the reactor.

Neutronic Reactor III

NASA Astrophysics Data System (ADS)

Fermi, Enrico; Zinn, Walter H.; Anderson, Herbert L.

An improvement of the reactors described in the previous Patents, aimed at increasing the reproduction factor, is reported here, such improvement being obtained by diminishing the neutron loss due to impurities within the reactor. This is achieved by encasing the reactor in a rubberized balloon cloth housing (or something like this) in order to eliminate the atmospheric air therefrom, thus eliminating both the effect of the danger coefficient of nitrogen (70% of the atmospheric air) and that of the argon present in the air, which can become radioactive. Since the removal of the air from the reactor may result in structural problems, caused by the forces brought into play by that evacuation, the reactor is then filled with a non-reactive (from a chemical and nuclear standpoint) gas such as helium or carbon dioxide. It is interesting to point out that the authors consider also the possibility to control (a little) the reproduction ratio of the reactor by varying the air content of it. Just a rapid mention of the main idea of the present Patent (i.e. the encasing of the pile in a balloon cloth) appeared in [Fermi (1942f)], but no detailed description of the system considered here is reported in any other published paper.

Adsorption of Pb(II), Cu(II), Cd(II), Zn(II), Ni(II), Fe(II), and As(V) on bacterially produced metal sulfides.

PubMed

Jong, Tony; Parry, David L

2004-07-01

The adsorption of Pb(II), Cu(II), Cd(II), Zn(II), Ni(II), Fe(II) and As(V) onto bacterially produced metal sulfide (BPMS) material was investigated using a batch equilibrium method. It was found that the sulfide material had adsorptive properties comparable with those of other adsorbents with respect to the specific uptake of a range of metals and, the levels to which dissolved metal concentrations in solution can be reduced. The percentage of adsorption increased with increasing pH and adsorbent dose, but decreased with increasing initial dissolved metal concentration. The pH of the solution was the most important parameter controlling adsorption of Cd(II), Cu(II), Fe(II), Ni(II), Pb(II), Zn(II), and As(V) by BPMS. The adsorption data were successfully modeled using the Langmuir adsorption isotherm. Desorption experiments showed that the reversibility of adsorption was low, suggesting high-affinity adsorption governed by chemisorption. The mechanism of adsorption for the divalent metals was thought to be the formation of strong, inner-sphere complexes involving surface hydroxyl groups. However, the mechanism for the adsorption of As(V) by BPMS appears to be distinct from that of surface hydroxyl exchange. These results have important implications to the management of metal sulfide sludge produced by bacterial sulfate reduction.

Period meter for reactors

DOEpatents

Rusch, Gordon K.

1976-01-06

An improved log N amplifier type nuclear reactor period meter with reduced probability for noise-induced scrams is provided. With the reactor at low power levels a sampling circuit is provided to determine the reactor period by measuring the finite change in the amplitude of the log N amplifier output signal for a predetermined time period, while at high power levels, differentiation of the log N amplifier output signal provides an additional measure of the reactor period.

Comparing the new generation accelerator driven subcritical reactor system (ADS) to traditional critical reactors

NASA Astrophysics Data System (ADS)

Kemah, Elif; Akkaya, Recep; Tokgöz, Seyit Rıza

2017-02-01

In recent years, the accelerator driven subcritical reactors have taken great interest worldwide. The Accelerator Driven System (ADS) has been used to produce neutron in subcritical state by the external proton beam source. These reactors, which are hybrid systems, are important in production of clean and safe energy and conversion of radioactive waste. The ADS with the selection of reliability and robust target materials have been the new generation of fission reactors. In addition, in the ADS Reactors the problems of long-lived radioactive fission products and waste actinides encountered in the fission process of the reactor during incineration can be solved, and ADS has come to the forefront of thorium as fuel for the reactors.

Advantages of liquid fluoride thorium reactor in comparison with light water reactor

NASA Astrophysics Data System (ADS)

Bahri, Che Nor Aniza Che Zainul; Majid, Amran Ab.; Al-Areqi, Wadeeah M.

2015-04-01

Liquid Fluoride Thorium Reactor (LFTR) is an innovative design for the thermal breeder reactor that has important potential benefits over the traditional reactor design. LFTR is fluoride based liquid fuel, that use the thorium dissolved in salt mixture of lithium fluoride and beryllium fluoride. Therefore, LFTR technology is fundamentally different from the solid fuel technology currently in use. Although the traditional nuclear reactor technology has been proven, it has perceptual problems with safety and nuclear waste products. The aim of this paper is to discuss the potential advantages of LFTR in three aspects such as safety, fuel efficiency and nuclear waste as an alternative energy generator in the future. Comparisons between LFTR and Light Water Reactor (LWR), on general principles of fuel cycle, resource availability, radiotoxicity and nuclear weapon proliferation shall be elaborated.

PBF Reactor Building (PER620). After lowering reactor vessel onto blocks, ...

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

PBF Reactor Building (PER-620). After lowering reactor vessel onto blocks, it is rolled on logs into PBF. Metal framework under vessel is handling device. Various penetrations in reactor bottom were for instrumentation, poison injection, drains. Large one, below center "manhole" was for primary coolant. Photographer: Larry Page. Date: February 13, 1970. INEEL negative no. 70-736 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

NEUTRONIC REACTOR CONSTRUCTION AND OPERATION

DOEpatents

West, J.M.; Weills, J.T.

1960-03-15

A method is given for operating a nuclear reactor having a negative coefficient of reactivity to compensate for the change in reactor reactivity due to the burn-up of the xenon peak following start-up of the reactor. When it is desired to start up the reactor within less than 72 hours after shutdown, the temperature of the reactor is lowered prior to start-up, and then gradually raised after start-up.

Anaerobic biodegradation of diesel fuel-contaminated wastewater in a fluidized bed reactor.

PubMed

Cuenca, M Alvarez; Vezuli, J; Lohi, A; Upreti, S R

2006-06-01

Diesel fuel spills have a major impact on the quality of groundwater. In this work, the performance of an Anaerobic Fluidized Bed Reactor (AFBR) treating synthetic wastewater is experimentally evaluated. The wastewater comprises tap water containing 100, 200 and 300 mg/L of diesel fuel and nutrients. Granular, inert, activated carbon particles are employed to provide support for biomass inside the reactor where diesel fuel is the sole source of carbon for anaerobic microorganisms. For different rates of organic loading, the AFBR performance is evaluated in terms of the removal of diesel fuel as well as chemical oxygen demand (COD) from wastewater. For the aforementioned diesel fuel concentrations and a wastewater flow rate of 1,200 L/day, the COD removal ranges between 61.9 and 84.1%. The concentration of diesel fuel in the effluent is less than 50 mg/L, and meets the Level II groundwater standards of the MUST guidelines of Alberta.

Reactor operation environmental information document

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

Haselow, J.S.; Price, V.; Stephenson, D.E.

1989-12-01

The Savannah River Site (SRS) produces nuclear materials, primarily plutonium and tritium, to meet the requirements of the Department of Defense. These products have been formed in nuclear reactors that were built during 1950--1955 at the SRS. K, L, and P reactors are three of five reactors that have been used in the past to produce the nuclear materials. All three of these reactors discontinued operation in 1988. Currently, intense efforts are being extended to prepare these three reactors for restart in a manner that protects human health and the environment. To document that restarting the reactors will have minimalmore » impacts to human health and the environment, a three-volume Reactor Operations Environmental Impact Document has been prepared. The document focuses on the impacts of restarting the K, L, and P reactors on both the SRS and surrounding areas. This volume discusses the geology, seismology, and subsurface hydrology. 195 refs., 101 figs., 16 tabs.« less

Reactor shutdown experience

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

Cletcher, J.W.

1995-10-01

This is a regular report of summary statistics relating to recent reactor shutdown experience. The information includes both number of events and rates of occurence. It was compiled from data about operating events that were entered into the SCSS data system by the Nuclear Operations Analysis Center at the Oak ridge National Laboratory and covers the six mont period of July 1 to December 31, 1994. Cumulative information, starting from May 1, 1994, is also reported. Updates on shutdown events included in earlier reports is excluded. Information on shutdowns as a function of reactor power at the time of themore » shutdown for both BWR and PWR reactors is given. Data is also discerned by shutdown type and reactor age.« less

THERMAL NEUTRONIC REACTOR

DOEpatents

Spinrad, B.I.

1960-01-12

A novel thermal reactor was designed in which a first reflector formed from a high atomic weight, nonmoderating material is disposed immediately adjacent to the reactor core. A second reflector composed of a moderating material is disposed outwardly of the first reflector. The advantage of this novel reflector arrangement is that the first reflector provides a high slow neutron flux in the second reflector, where irradiation experiments may be conducted with a small effect on reactor reactivity.

Bioconversion reactor

DOEpatents

McCarty, Perry L.; Bachmann, Andre

1992-01-01

A bioconversion reactor for the anaerobic fermentation of organic material. The bioconversion reactor comprises a shell enclosing a predetermined volume, an inlet port through which a liquid stream containing organic materials enters the shell, and an outlet port through which the stream exits the shell. A series of vertical and spaced-apart baffles are positioned within the shell to force the stream to flow under and over them as it passes from the inlet to the outlet port. The baffles present a barrier to the microorganisms within the shell causing them to rise and fall within the reactor but to move horizontally at a very slow rate. Treatment detention times of one day or less are possible.

Advantages of liquid fluoride thorium reactor in comparison with light water reactor

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

Bahri, Che Nor Aniza Che Zainul, E-mail: anizazainul@gmail.com; Majid, Amran Ab.; Al-Areqi, Wadeeah M.

2015-04-29

Liquid Fluoride Thorium Reactor (LFTR) is an innovative design for the thermal breeder reactor that has important potential benefits over the traditional reactor design. LFTR is fluoride based liquid fuel, that use the thorium dissolved in salt mixture of lithium fluoride and beryllium fluoride. Therefore, LFTR technology is fundamentally different from the solid fuel technology currently in use. Although the traditional nuclear reactor technology has been proven, it has perceptual problems with safety and nuclear waste products. The aim of this paper is to discuss the potential advantages of LFTR in three aspects such as safety, fuel efficiency and nuclearmore » waste as an alternative energy generator in the future. Comparisons between LFTR and Light Water Reactor (LWR), on general principles of fuel cycle, resource availability, radiotoxicity and nuclear weapon proliferation shall be elaborated.« less

Hybrid plasmachemical reactor

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

Lelevkin, V. M., E-mail: lelevkin44@mail.ru; Smirnova, Yu. G.; Tokarev, A. V.

2015-04-15

A hybrid plasmachemical reactor on the basis of a dielectric barrier discharge in a transformer is developed. The characteristics of the reactor as functions of the dielectric barrier discharge parameters are determined.

New techniques for modeling the reliability of reactor pressure vessels

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

Johnson, K.I.; Simonen, F.A.; Liebetrau, A.M.

1986-01-01

In recent years several probabilistic fracture mechanics codes, including the VISA code, have been developed to predict the reliability of reactor pressure vessels. This paper describes several new modeling techniques used in a second generation of the VISA code entitled VISA-II. Results are presented that show the sensitivity of vessel reliability predictions to such factors as inservice inspection to detect flaws, random positioning of flaws within the vessel wall thickness, and fluence distributions that vary throughout the vessel. The algorithms used to implement these modeling techniques are also described. Other new options in VISA-II are also described in this paper.more » The effect of vessel cladding has been included in the heat transfer, stress, and fracture mechanics solutions in VISA-II. The algorithms for simulating flaws has been changed to consider an entire vessel rather than a single flaw in a single weld. The flaw distribution was changed to include the distribution of both flaw depth and length. A menu of several alternate equations has been included to predict the shift in RT/sub NDT/. For flaws that arrest and later re-initiate, an option was also included to allow correlating the current arrest toughness with subsequent initiation toughnesses.« less

New techniques for modeling the reliability of reactor pressure vessels

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

Johnson, K.I.; Simonen, F.A.; Liebetrau, A.M.

1985-12-01

In recent years several probabilistic fracture mechanics codes, including the VISA code, have been developed to predict the reliability of reactor pressure vessels. This paper describes new modeling techniques used in a second generation of the VISA code entitled VISA-II. Results are presented that show the sensitivity of vessel reliability predictions to such factors as inservice inspection to detect flaws, random positioning of flaws within the vessel walls thickness, and fluence distributions that vary through-out the vessel. The algorithms used to implement these modeling techniques are also described. Other new options in VISA-II are also described in this paper. Themore » effect of vessel cladding has been included in the heat transfer, stress, and fracture mechanics solutions in VISA-II. The algorithm for simulating flaws has been changed to consider an entire vessel rather than a single flaw in a single weld. The flaw distribution was changed to include the distribution of both flaw depth and length. A menu of several alternate equations has been included to predict the shift in RTNDT. For flaws that arrest and later re-initiate, an option was also included to allow correlating the current arrest thoughness with subsequent initiation toughnesses. 21 refs.« less

HOMOGENEOUS NUCLEAR POWER REACTOR

DOEpatents

King, L.D.P.

1959-09-01

A homogeneous nuclear power reactor utilizing forced circulation of the liquid fuel is described. The reactor does not require fuel handling outside of the reactor vessel during any normal operation including complete shutdown to room temperature, the reactor being selfregulating under extreme operating conditions and controlled by the thermal expansion of the liquid fuel. The liquid fuel utilized is a uranium, phosphoric acid, and water solution which requires no gus exhaust system or independent gas recombining system, thereby eliminating the handling of radioiytic gas.

Update on reactors and reactor instruments in Asia

NASA Astrophysics Data System (ADS)

Rao, K. R.

1991-10-01

The 1980s have seen the commissioning of several medium flux (∼10 14 neutrons/cm 2s) research reactors in Asia. The reactors are based on indigenous design and development in India and China. At Dhruva reactor (India), a variety of neutron spectrometers have been established that have provided useful data related to the structure of high- Tc materials, phonon density of states, magnetic moment distributions and micellar aggregation during the last couple of years. Polarised neutron analysis, neutron interferometry and neutron spin echo methods are some of the new techniques under development. The spectrometers and associated automaton, detectors and neutron guides have all been indigenously developed. This paper summarises the developments and on-going activities in Bangladesh, China, India, Indonesia, Korea, Malaysia, the Philippines and Thailand.

Solvent refined coal reactor quench system

DOEpatents

Thorogood, Robert M.

1983-01-01

There is described an improved SRC reactor quench system using a condensed product which is recycled to the reactor and provides cooling by evaporation. In the process, the second and subsequent reactors of a series of reactors are cooled by the addition of a light oil fraction which provides cooling by evaporation in the reactor. The vaporized quench liquid is recondensed from the reactor outlet vapor stream.

Method to Reduce Long-lived Fission Products by Nuclear Transmutations with Fast Spectrum Reactors.

PubMed

Chiba, Satoshi; Wakabayashi, Toshio; Tachi, Yoshiaki; Takaki, Naoyuki; Terashima, Atsunori; Okumura, Shin; Yoshida, Tadashi

2017-10-24

Transmutation of long-lived fission products (LLFPs: 79 Se, 93 Zr, 99 Tc, 107 Pd, 129 I, and 135 Cs) into short-lived or non-radioactive nuclides by fast neutron spectrum reactors without isotope separation has been proposed as a solution to the problem of radioactive wastes disposal. Despite investigation of many methods, such transmutation remains technologically difficult. To establish an effective and efficient transmutation system, we propose a novel neutron moderator material, yttrium deuteride (YD 2 ), to soften the neutron spectrum leaking from the reactor core. Neutron energy spectra and effective half-lives of LLFPs, transmutation rates, and support ratios were evaluated with the continuous-energy Monte Carlo code MVP-II/MVP-BURN and the JENDL-4.0 cross section library. With the YD 2 moderator in the radial blanket and shield regions, effective half-lives drastically decreased from 106 to 102 years and the support ratios reached 1.0 for all six LLFPs. This successful development and implementation of a transmutation system for LLFPs without isotope separation contributes to a the ability of fast spectrum reactors to reduce radioactive waste by consuming their own LLFPs.

Improved sample manipulation at the STRESS-SPEC neutron diffractometer using an industrial 6-axis robot for texture and strain analyses

NASA Astrophysics Data System (ADS)

Randau, C.; Brokmeier, H. G.; Gan, W. M.; Hofmann, M.; Voeller, M.; Tekouo, W.; Al-hamdany, N.; Seidl, G.; Schreyer, A.

2015-09-01

The materials science neutron diffractometer STRESS-SPEC located at FRM II is a dedicated instrument for strain and pole figure measurements. Both methods make complementary demands on sample handling. On one hand pole figure measurements need a high degree of freedom to orient small samples and on the other hand in strain investigations it is often necessary to handle large and heavy components. Therefore a robot based sample positioning system was developed, which has the capability to provide both possibilities. Based on this new robot system further developments like a full automated sample changer system for texture measurements were accomplished. Moreover this system opens the door for combined strain and texture analysis at STRESS-SPEC.

Nonstationary decision model for flood risk decision scaling

NASA Astrophysics Data System (ADS)

Spence, Caitlin M.; Brown, Casey M.

2016-11-01

Hydroclimatic stationarity is increasingly questioned as a default assumption in flood risk management (FRM), but successor methods are not yet established. Some potential successors depend on estimates of future flood quantiles, but methods for estimating future design storms are subject to high levels of uncertainty. Here we apply a Nonstationary Decision Model (NDM) to flood risk planning within the decision scaling framework. The NDM combines a nonstationary probability distribution of annual peak flow with optimal selection of flood management alternatives using robustness measures. The NDM incorporates structural and nonstructural FRM interventions and valuation of flows supporting ecosystem services to calculate expected cost of a given FRM strategy. A search for the minimum-cost strategy under incrementally varied representative scenarios extending across the plausible range of flood trend and value of the natural flow regime discovers candidate FRM strategies that are evaluated and compared through a decision scaling analysis (DSA). The DSA selects a management strategy that is optimal or close to optimal across the broadest range of scenarios or across the set of scenarios deemed most likely to occur according to estimates of future flood hazard. We illustrate the decision framework using a stylized example flood management decision based on the Iowa City flood management system, which has experienced recent unprecedented high flow episodes. The DSA indicates a preference for combining infrastructural and nonstructural adaptation measures to manage flood risk and makes clear that options-based approaches cannot be assumed to be "no" or "low regret."

Solvent refined coal reactor quench system

DOEpatents

Thorogood, R.M.

1983-11-08

There is described an improved SRC reactor quench system using a condensed product which is recycled to the reactor and provides cooling by evaporation. In the process, the second and subsequent reactors of a series of reactors are cooled by the addition of a light oil fraction which provides cooling by evaporation in the reactor. The vaporized quench liquid is recondensed from the reactor outlet vapor stream. 1 fig.

Intense positron beam as a source for production of electron-positron plasma

NASA Astrophysics Data System (ADS)

Stoneking, M. R.; Horn-Stanja, J.; Stenson, E. V.; Pedersen, T. Sunn; Saitoh, H.; Hergenhahn, U.; Niemann, H.; Paschkowski, N.; Hugenschmidt, C.; Piochacz, C.

2016-10-01

We aim to produce magnetically confined, short Debye length electron-positron plasma and test predicted properties for such systems. A first challenge is obtaining large numbers of positrons; a table-top experiment (system size 5 cm) with a temperature less than 5 eV requires about 1010 positrons to have more than 10 Debye lengths in the system. The NEPOMUC facility at the FRM II research reactor in Germany is one of the world's most intense positron sources. We report on characterization (using a retarding field energy analyzer with magnetic field gradient) of the NEPOMUC beam as delivered to the open beam port at various beam energies and in both the re-moderated and primary beam configurations in order to design optimal trapping (and accumulation) schemes for production of electron-positron plasma. The intensity of the re-moderated (primary) beam is in the range 2 -3 x 107 /s (1 - 5 x 108 /s). The re-moderated beam is currently the most promising for direct injection and confinement experiments; it has a parallel energy spread of 15 - 35% and the transverse energy spread is 6 - 15% of the parallel energy. We report on the implications for injection and trapping in a dipole magnetic field as well as plans for beam development, in situ re-moderation, and accumulation. We also report results demonstrating a difference in phosphor luminescent response to low energy positrons versus electrons.

REACTOR FUEL SCAVENGING MEANS

DOEpatents

Coffinberry, A.S.

1962-04-10

A process for removing fission products from reactor liquid fuel without interfering with the reactor's normal operation or causing a significant change in its fuel composition is described. The process consists of mixing a liquid scavenger alloy composed of about 44 at.% plutoniunm, 33 at.% lanthanum, and 23 at.% nickel or cobalt with a plutonium alloy reactor fuel containing about 3 at.% lanthanum; removing a portion of the fuel and scavenger alloy from the reactor core and replacing it with an equal amount of the fresh scavenger alloy; transferring the portion to a quiescent zone where the scavenger and the plutonium fuel form two distinct liquid layers with the fission products being dissolved in the lanthanum-rich scavenger layer; and the clean plutonium-rich fuel layer being returned to the reactor core. (AEC)

Reactor-Produced Medical Radionuclides

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

Mirzadeh, Saed; Mausner, Leonard; Garland, Marc A

2011-01-01

The therapeutic use of radionuclides in nuclear medicine, oncology and cardiology is the most rapidly growing use of medical radionuclides. Since most therapeutic radionuclides are neutron rich and decay by beta emission, they are reactor-produced. This chapter deals mainly with production approaches with neutrons. Neutron interactions with matter, neutron transmission and activation rates, and neutron spectra of nuclear reactors are discussed in some detail. Further, a short discussion of the neutron-energy dependence of cross sections, reaction rates in thermal reactors, cross section measurements and flux monitoring, and general equations governing the reactor production of radionuclides are presented. Finally, the chaptermore » is concluded by providing a number of examples encompassing the various possible reaction routes for production of a number of medical radionuclides in a reactor.« less

NEUTRONIC REACTOR

DOEpatents

Metcalf, H.E.

1957-10-01

A reactor of the type which preferably uses plutonium as the fuel and a liquid moderator, preferably ordinary water, and which produces steam within the reactor core due to the heat of the chain reaction is described. In the reactor shown the fuel elements are essentially in the form of trays and are ventically stacked in spaced relationship. The water moderator is continuously supplied to the trays to maintain a constant level on the upper surfaces of the fuel element as it is continually evaporated by the heat. The steam passes out through the spaces between the fuel elements and is drawn off at the top of the core. The fuel elements are clad in aluminum to prevent deterioration thereof with consequent contamimation of the water.

REACTOR CONTROL

DOEpatents

Fortescue, P.; Nicoll, D.

1962-04-24

A control system employed with a high pressure gas cooled reactor in which a control rod is positioned for upward and downward movement into the neutron field from a position beneath the reactor is described. The control rod is positioned by a coupled piston cylinder releasably coupled to a power drive means and the pressurized coolant is directed against the lower side of the piston. The coolant pressure is offset by a higher fiuid pressure applied to the upper surface of the piston and means are provided for releasing the higher pressure on the upper side of the piston so that the pressure of the coolant drives the piston upwardly, forcing the coupled control rod into the ncutron field of the reactor. (AEC)

NUCLEAR REACTOR CONTROL SYSTEM

DOEpatents

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

1959-11-01

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

Reactor vessel support system. [LMFBR

DOEpatents

Golden, M.P.; Holley, J.C.

1980-05-09

A reactor vessel support system includes a support ring at the reactor top supported through a box ring on a ledge of the reactor containment. The box ring includes an annular space in the center of its cross-section to reduce heat flow and is keyed to the support ledge to transmit seismic forces from the reactor vessel to the containment structure. A coolant channel is provided at the outside circumference of the support ring to supply coolant gas through the keyways to channels between the reactor vessel and support ledge into the containment space.

Attrition reactor system

DOEpatents

Scott, Charles D.; Davison, Brian H.

1993-01-01

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

Effect of Cu(II), Cd(II) and Zn(II) on Pb(II) biosorption by algae Gelidium-derived materials.

PubMed

Vilar, Vítor J P; Botelho, Cidália M S; Boaventura, Rui A R

2008-06-15

Biosorption of Pb(II), Cu(II), Cd(II) and Zn(II) from binary metal solutions onto the algae Gelidium sesquipedale, an algal industrial waste and a waste-based composite material was investigated at pH 5.3, in a batch system. Binary Pb(II)/Cu(II), Pb(II)/Cd(II) and Pb(II)/Zn(II) solutions have been tested. For the same equilibrium concentrations of both metal ions (1 mmol l(-1)), approximately 66, 85 and 86% of the total uptake capacity of the biosorbents is taken by lead ions in the systems Pb(II)/Cu(II), Pb(II)/Cd(II) and Pb(II)/Zn(II), respectively. Two-metal results were fitted to a discrete and a continuous model, showing the inhibition of the primary metal biosorption by the co-cation. The model parameters suggest that Cd(II) and Zn(II) have the same decreasing effect on the Pb(II) uptake capacity. The uptake of Pb(II) was highly sensitive to the presence of Cu(II). From the discrete model it was possible to obtain the Langmuir affinity constant for Pb(II) biosorption. The presence of the co-cations decreases the apparent affinity of Pb(II). The experimental results were successfully fitted by the continuous model, at different pH values, for each biosorbent. The following sequence for the equilibrium affinity constants was found: Pb>Cu>Cd approximately Zn.

Microfluidic electrochemical reactors

DOEpatents

Nuzzo, Ralph G [Champaign, IL; Mitrovski, Svetlana M [Urbana, IL

2011-03-22

A microfluidic electrochemical reactor includes an electrode and one or more microfluidic channels on the electrode, where the microfluidic channels are covered with a membrane containing a gas permeable polymer. The distance between the electrode and the membrane is less than 500 micrometers. The microfluidic electrochemical reactor can provide for increased reaction rates in electrochemical reactions using a gaseous reactant, as compared to conventional electrochemical cells. Microfluidic electrochemical reactors can be incorporated into devices for applications such as fuel cells, electrochemical analysis, microfluidic actuation, pH gradient formation.

REACTOR CONTROL SYSTEM

DOEpatents

MacNeill, J.H.; Estabrook, J.Y.

1960-05-10

A reactor control system including a continuous tape passing through a first coolant passageway, over idler rollers, back through another parallel passageway, and over motor-driven rollers is described. Discrete portions of fuel or poison are carried on two opposed active sections of the tape. Driving the tape in forward or reverse directions causes both active sections to be simultaneously inserted or withdrawn uniformly, tending to maintain a more uniform flux within the reactor. The system is particularly useful in mobile reactors, where reduced inertial resistance to control rod movement is important.

NEUTRONIC REACTOR CONTROL

DOEpatents

Metcalf, H.E.

1958-10-14

Methods of controlling reactors are presented. Specifically, a plurality of neutron absorber members are adjustably disposed in the reactor core at different distances from the center thereof. The absorber members extend into the core from opposite faces thereof and are operated by motive means coupled in a manner to simultaneously withdraw at least one of the absorber members while inserting one of the other absorber members. This feature effects fine control of the neutron reproduction ratio by varying the total volume of the reactor effective in developing the neutronic reaction.

LWR pressure vessel surveillance dosimetry improvement program: LWR power reactor surveillance physics-dosimetry data base compendium

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

McElroy, W.N.

1985-08-01

This NRC physics-dosimetry compendium is a collation of information and data developed from available research and commercial light water reactor vessel surveillance program (RVSP) documents and related surveillance capsule reports. The data represents the results of the HEDL least-squares FERRET-SAND II Code re-evaluation of exposure units and values for 47 PWR and BWR surveillance capsules for W, B and W, CE, and GE power plants. Using a consistent set of auxiliary data and dosimetry-adjusted reactor physics results, the revised fluence values for E > 1 MeV averaged 25% higher than the originally reported values. The range of fluence values (new/old)more » was from a low of 0.80 to a high of 2.38. These HEDL-derived FERRET-SAND II exposure parameter values are being used for NRC-supported HEDL and other PWR and BWR trend curve data development and testing studies. These studies are providing results to support Revision 2 of Regulatory Guide 1.99. As stated by Randall (Ra84), the Guide is being updated to reflect recent studies of the physical basis for neutron radiation damage and efforts to correlate damage to chemical composition and fluence.« less

Status of French reactors

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

Ballagny, A.

1997-08-01

The status of French reactors is reviewed. The ORPHEE and RHF reactors can not be operated with a LEU fuel which would be limited to 4.8 g U/cm{sup 3}. The OSIRIS reactor has already been converted to LEU. It will use U{sub 3}Si{sub 2} as soon as its present stock of UO{sub 2} fuel is used up, at the end of 1994. The decision to close down the SILOE reactor in the near future is not propitious for the start of a conversion process. The REX 2000 reactor, which is expected to be commissioned in 2005, will use LEU (exceptmore » if the fast neutrons core option is selected). Concerning the end of the HEU fuel cycle, the best option is reprocessing followed by conversion of the reprocessed uranium to LEU.« less

Startup of reactors for anoxic ammonium oxidation: experiences from the first full-scale anammox reactor in Rotterdam.

PubMed

van der Star, Wouter R L; Abma, Wiebe R; Blommers, Dennis; Mulder, Jan-Willem; Tokutomi, Takaaki; Strous, Marc; Picioreanu, Cristian; van Loosdrecht, Mark C M

2007-10-01

The first full-scale anammox reactor in the world was started in Rotterdam (NL). The reactor was scaled-up directly from laboratory-scale to full-scale and treats up to 750 kg-N/d. In the initial phase of the startup, anammox conversions could not be identified by traditional methods, but quantitative PCR proved to be a reliable indicator for growth of the anammox population, indicating an anammox doubling time of 10-12 days. The experience gained during this first startup in combination with the availability of seed sludge from this reactor, will lead to a faster startup of anammox reactors in the future. The anammox reactor type employed in Rotterdam was compared to other reactor types for the anammox process. Reactors with a high specific surface area like the granular sludge reactor employed in Rotterdam provide the highest volumetric loading rates. Mass transfer of nitrite into the biofilm is limiting the conversion of those reactor types that have a lower specific surface area. Now the first full-scale commercial anammox reactor is in operation, a consistent and descriptive nomenclature is suggested for reactors in which the anammox process is employed.

Free metal ion depletion by "Good's" buffers. III. N-(2-acetamido)iminodiacetic acid, 2:1 complexes with zinc(II), cobalt(II), nickel(II), and copper(II); amide deprotonation by Zn(II), Co(II), and Cu(II).

PubMed

Lance, E A; Rhodes, C W; Nakon, R

1983-09-01

Potentiometric, visible, infrared, electron spin, and nuclear magnetic resonance studies of the complexation of N-(2-acetamido)iminodiacetic acid (H2ADA) by Ca(II), Mg(II), Mn(II), Zn(II), Co(II), Ni(II), and Cu(II) are reported. Ca(II) and Mg(II) were found not to form 2:1 ADA2- to M(II) complexes, while Mn(II), Cu(II), Ni(II), Zn(II), and Co(II) did form 2:1 metal chelates at or below physiological pH values. Co(II) and Zn(II), but not Cu(II), were found to induce stepwise deprotonation of the amide groups to form [M(H-1ADA)4-(2)]. Formation (affinity) constants for the various metal complexes are reported, and the probable structures of the various metal chelates in solution are discussed on the basis of various spectral data.

Dinuclear complexes containing linear M-F-M [M = Mn(II), Fe(II), Co(II), Ni(II), Cu(II), Zn(II), Cd(II)] bridges: trends in structures, antiferromagnetic superexchange interactions, and spectroscopic properties.

PubMed

Reger, Daniel L; Pascui, Andrea E; Smith, Mark D; Jezierska, Julia; Ozarowski, Andrew

2012-11-05

The reaction of M(BF(4))(2)·xH(2)O, where M is Fe(II), Co(II), Ni(II), Cu(II), Zn(II), and Cd(II), with the new ditopic ligand m-bis[bis(3,5-dimethyl-1-pyrazolyl)methyl]benzene (L(m)*) leads to the formation of monofluoride-bridged dinuclear metallacycles of the formula [M(2)(μ-F)(μ-L(m)*)(2)](BF(4))(3). The analogous manganese(II) species, [Mn(2)(μ-F)(μ-L(m)*)(2)](ClO(4))(3), was isolated starting with Mn(ClO(4))(2)·6H(2)O using NaBF(4) as the source of the bridging fluoride. In all of these complexes, the geometry around the metal centers is trigonal bipyramidal, and the fluoride bridges are linear. The (1)H, (13)C, and (19)F NMR spectra of the zinc(II) and cadmium(II) compounds and the (113)Cd NMR of the cadmium(II) compound indicate that the metallacycles retain their structure in acetonitrile and acetone solution. The compounds with M = Mn(II), Fe(II), Co(II), Ni(II), and Cu(II) are antiferromagnetically coupled, although the magnitude of the coupling increases dramatically with the metal as one moves to the right across the periodic table: Mn(II) (-6.7 cm(-1)) < Fe(II) (-16.3 cm(-1)) < Co(II) (-24.1 cm(-1)) < Ni(II) (-39.0 cm(-1)) ≪ Cu(II) (-322 cm(-1)). High-field EPR spectra of the copper(II) complexes were interpreted using the coupled-spin Hamiltonian with g(x) = 2.150, g(y) = 2.329, g(z) = 2.010, D = 0.173 cm(-1), and E = 0.089 cm(-1). Interpretation of the EPR spectra of the iron(II) and manganese(II) complexes required the spin Hamiltonian using the noncoupled spin operators of two metal ions. The values g(x) = 2.26, g(y) = 2.29, g(z) = 1.99, J = -16.0 cm(-1), D(1) = -9.89 cm(-1), and D(12) = -0.065 cm(-1) were obtained for the iron(II) complex and g(x) = g(y) = g(z) = 2.00, D(1) = -0.3254 cm(-1), E(1) = -0.0153, J = -6.7 cm(-1), and D(12) = 0.0302 cm(-1) were found for the manganese(II) complex. Density functional theory (DFT) calculations of the exchange integrals and the zero-field splitting on manganese(II) and iron(II) ions were performed

NUCLEAR REACTOR

DOEpatents

Breden, C.R.; Dietrich, J.R.

1961-06-20

A water-soluble non-volatile poison may be introduced into a reactor to nullify excess reactivity. The poison is removed by passing a side stream of the water containing the soluble poison to an evaporation chamber. The vapor phase is returned to the reactor to decrease the concentration of soluble poison and the liquid phase is returned to increase the concentration of soluble poison.

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.

Attrition reactor system

DOEpatents

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

1993-09-28

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

REACTOR PHYSICS CONSTANTS

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

None

1963-07-01

This second edition is based on data available on March 15, 1961. Sections on constants necessary for the interpretation of experimental data and on digital computer programs for reactor design and reactor physics have been added. 1344 references. (D.C.W.)

High Temperature Reactor (HTR) Deep Burn Core and Fuel Analysis: Design Selection for the Prismatic Block Reactor

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

Francesco Venneri; Chang-Keun Jo; Jae-Man Noh

2010-09-01

The Deep Burn (DB) Project is a U.S. Department of Energy sponsored feasibility study of Transuranic Management using high burnup fuel in the high temperature helium cooled reactor (HTR). The DB Project consists of seven tasks: project management, core and fuel analysis, spent fuel management, fuel cycle integration, TRU fuel modeling, TRU fuel qualification, and HTR fuel recycle. In the Phase II of the Project, we conducted nuclear analysis of TRU destruction/utilization in the HTR prismatic block design (Task 2.1), deep burn fuel/TRISO microanalysis (Task 2.3), and synergy with fast reactors (Task 4.2). The Task 2.1 covers the core physicsmore » design, thermo-hydraulic CFD analysis, and the thermofluid and safety analysis (low pressure conduction cooling, LPCC) of the HTR prismatic block design. The Task 2.3 covers the analysis of the structural behavior of TRISO fuel containing TRU at very high burnup level, i.e. exceeding 50% of FIMA. The Task 4.2 includes the self-cleaning HTR based on recycle of HTR-generated TRU in the same HTR. Chapter IV contains the design and analysis results of the 600MWth DB-HTR core physics with the cycle length, the average discharged burnup, heavy metal and plutonium consumptions, radial and axial power distributions, temperature reactivity coefficients. Also, it contains the analysis results of the 450MWth DB-HTR core physics and the analysis of the decay heat of a TRU loaded DB-HTR core. The evaluation of the hot spot fuel temperature of the fuel block in the DB-HTR (Deep-Burn High Temperature Reactor) core under full operating power conditions are described in Chapter V. The investigated designs are the 600MWth and 460MWth DB-HTRs. In Chapter VI, the thermo-fluid and safety of the 600MWth DB-HTRs has been analyzed to investigate a thermal-fluid design performance at the steady state and a passive safety performance during an LPCC event. Chapter VII describes the analysis results of the TRISO fuel microanalysis of the 600MWth

Method of Operating a Neutronic Reactor

NASA Astrophysics Data System (ADS)

Fermi, Enrico; Szilard, Leo

reactor, in order to increase the multiplication factor K, are considered as well. A description of the possible uses of nuclear reactors, other than as power supplies, including the production of collimated beams of fast neutrons, the production of plutonium (a fissionable material usable in other reactors) or several other radioactive isotopes (for possible utilization in medicine) is as well given. As it results clear, no published reference article behind the present Patent exists. Some partial results may be found in several papers2 of Volume II of [Fermi (1962)] (see, for example, [Fermi (1952)]), but here very many technical data and some information of historic interest (mainly on the experiments performed in order to obtain the data reported) are given. The most "relevant" change of Patent No. 2,798,847 with respect to the original Patent No. 2,708,656 is the replacement of the 8 claims of the previous one by the following only one claim, which well summarizes the work done: "A method of operating a neutronic reactor including an active portion having a neutron reproduction ratio substantially in excess of unity in the absence of high neutron absorbing bodies, said method comprising the steps of inserting in the active portion a shim member consisting essentially of a high neutron absorbing body in an amount to reduce the neutron reproduction ratio to a value slightly higher than unit to prevent a dangerous reactivity level, controlling the reaction by moving a control member consisting essentially of a second high neutron absorbing body inwardly and outwardly in response to variations in neutron density, to maintain the neutron reproduction ratio substantially at unity, and withdrawing successive portions of the shim member to the extent necessary to enable the reactor to be controlled by movement of the control member after the neutron reproduction value has been lowered to the point where the outward movement of the control member is insufficient to maintain

Low Absorption Vitreous Carbon Reactors for Operando XAS: A Case Study on Cu/Zeolites for Selective Catalytic Reduction of NOx by NH3

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

Kispersky, Vincent F.; Kropf, A. Jeremy; Ribeiro, Fabio H.

2012-01-01

We describe the use of vitreous carbon as an improved reactor material for an operando X-ray absorption spectroscopy (XAS) plug-flow reactor. These tubes significantly broaden the operating range for operando experiments. Using selective catalytic reduction (SCR) of NO x by NH₃ on Cu/Zeolites (SSZ-13, SAPO-34 and ZSM-5) as an example reaction, we illustrate the high-quality XAS data achievable with these reactors. The operando experiments showed that in Standard SCR conditions of 300 ppm NO, 300 ppm NH₃, 5% O₂, 5% H₂O, 5% CO₂ and balance He at 200 °C, the Cu was a mixture of Cu(I) and Cu(II) oxidation states.more » XANES and EXAFS fitting found the percent of Cu(I) to be 15%, 45% and 65% for SSZ-13, SAPO-34 and ZSM-5, respectively. For Standard SCR, the catalytic rates per mole of Cu for Cu/SSZ-13 and Cu/SAPO-34 were about one third of the rate per mole of Cu on Cu/ZSM-5. Based on the apparent lack of correlation of rate with the presence of Cu(I), we propose that the reaction occurs via a redox cycle of Cu(I) and Cu(II). Cu(I) was not found in in situSCR experiments on Cu/Zeolites under the same conditions, demonstrating a possible pitfall of in situ measurements. A Cu/SiO₂ catalyst, reduced in H₂ at 300 °C, was also used to demonstrate the reactor's operando capabilities using a bending magnet beamline. Analysis of the EXAFS data showed the Cu/SiO₂ catalyst to be in a partially reduced Cu metal–Cu(I) state. In addition to improvements in data quality, the reactors are superior in temperature, stability, strength and ease of use compared to previously proposed borosilicate glass, polyimide tubing, beryllium and capillary reactors. The solid carbon tubes are non-porous, machinable, can be operated at high pressure (tested at 25 bar), are inert, have high material purity and high X-ray transmittance.« less

Improved vortex reactor system

DOEpatents

Diebold, James P.; Scahill, John W.

1995-01-01

An improved vortex reactor system for affecting fast pyrolysis of biomass and Refuse Derived Fuel (RDF) feed materials comprising: a vortex reactor having its axis vertically disposed in relation to a jet of a horizontally disposed steam ejector that impels feed materials from a feeder and solids from a recycle loop along with a motive gas into a top part of said reactor.

Reactor monitoring using antineutrino detectors

NASA Astrophysics Data System (ADS)

Bowden, N. S.

2011-08-01

Nuclear reactors have served as the antineutrino source for many fundamental physics experiments. The techniques developed by these experiments make it possible to use these weakly interacting particles for a practical purpose. The large flux of antineutrinos that leaves a reactor carries information about two quantities of interest for safeguards: the reactor power and fissile inventory. Measurements made with antineutrino detectors could therefore offer an alternative means for verifying the power history and fissile inventory of a reactor as part of International Atomic Energy Agency (IAEA) and/or other reactor safeguards regimes. Several efforts to develop this monitoring technique are underway worldwide.

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

Nondestructive assay of EBR-II blanket elements using resonance transmission analysis

NASA Astrophysics Data System (ADS)

Klann, Raymond Todd

1998-10-01

Resonance transmission analysis utilizing a filtered reactor beam was examined as a means of determining the 239Pu content in Experimental Breeder Reactor - II depleted uranium blanket elements. The technique uses cadmium and gadolinium filters along with a 239Pu fission chamber to isolate the 0.3 eV resonance in 239Pu. In the energy range of this resonance (0.1 eV to 0.5 eV), the total microscopic cross-section of 239Pu is significantly greater than the cross- sections of 238U and 235U. This large difference allows small changes in the 239Pu content of a sample to result in large changes in the mass signal response. Tests with small stacks of depleted uranium and 239Pu foils indicate a significant change in response based on the 239Pu content of the foil stack. In addition, the tests indicate good agreement between the measured and predicted values of 239Pu up to approximately two weight percent.

Binding Selectivity of Methanobactin from Methylosinus trichosporium OB3b for Copper(I), Silver(I), Zinc(II), Nickel(II), Cobalt(II), Manganese(II), Lead(II), and Iron(II).

PubMed

McCabe, Jacob W; Vangala, Rajpal; Angel, Laurence A

2017-12-01

Methanobactin (Mb) from Methylosinus trichosporium OB3b is a member of a class of metal binding peptides identified in methanotrophic bacteria. Mb will selectively bind and reduce Cu(II) to Cu(I), and is thought to mediate the acquisition of the copper cofactor for the enzyme methane monooxygenase. These copper chelating properties of Mb make it potentially useful as a chelating agent for treatment of diseases where copper plays a role including Wilson's disease, cancers, and neurodegenerative diseases. Utilizing traveling wave ion mobility-mass spectrometry (TWIMS), the competition for the Mb copper binding site from Ag(I), Pb(II), Co(II), Fe(II), Mn(II), Ni(II), and Zn(II) has been determined by a series of metal ion titrations, pH titrations, and metal ion displacement titrations. The TWIMS analyses allowed for the explicit identification and quantification of all the individual Mb species present during the titrations and measured their collision cross-sections and collision-induced dissociation patterns. The results showed Ag(I) and Ni(II) could irreversibly bind to Mb and not be effectively displaced by Cu(I), whereas Ag(I) could also partially displace Cu(I) from the Mb complex. At pH ≈ 6.5, the Mb binding selectivity follows the order Ag(I)≈Cu(I)>Ni(II)≈Zn(II)>Co(II)>Mn(II)≈Pb(II)>Fe(II), and at pH 7.5 to 10.4 the order is Ag(I)>Cu(I)>Ni(II)>Co(II)>Zn(II)>Mn(II)≈Pb(II)>Fe(II). Breakdown curves of the disulfide reduced Cu(I) and Ag(I) complexes showed a correlation existed between their relative stability and their compact folded structure indicated by their CCS. Fluorescence spectroscopy, which allowed the determination of the binding constant, compared well with the TWIMS analyses, with the exception of the Ni(II) complex. Graphical abstract ᅟ.

Improved vortex reactor system

DOEpatents

Diebold, J.P.; Scahill, J.W.

1995-05-09

An improved vortex reactor system is described for affecting fast pyrolysis of biomass and Refuse Derived Fuel (RDF) feed materials comprising: a vortex reactor having its axis vertically disposed in relation to a jet of a horizontally disposed steam ejector that impels feed materials from a feeder and solids from a recycle loop along with a motive gas into a top part of said reactor. 12 figs.

High solids fermentation reactor

DOEpatents

Wyman, Charles E.; Grohmann, Karel; Himmel, Michael E.; Richard, Christopher J.

1993-03-02

A fermentation reactor and method for fermentation of materials having greater than about 10% solids. The reactor includes a rotatable shaft along the central axis, the shaft including rods extending outwardly to mix the materials. The reactor and method are useful for anaerobic digestion of municipal solid wastes to produce methane, for production of commodity chemicals from organic materials, and for microbial fermentation processes.

High solids fermentation reactor

DOEpatents

Wyman, Charles E.; Grohmann, Karel; Himmel, Michael E.; Richard, Christopher J.

1993-01-01

A fermentation reactor and method for fermentation of materials having greater than about 10% solids. The reactor includes a rotatable shaft along the central axis, the shaft including rods extending outwardly to mix the materials. The reactor and method are useful for anaerobic digestion of municipal solid wastes to produce methane, for production of commodity chemicals from organic materials, and for microbial fermentation processes.

Morphological evolution of copper nanoparticles: Microemulsion reactor system versus batch reactor system

NASA Astrophysics Data System (ADS)

Xia, Ming; Tang, Zengmin; Kim, Woo-Sik; Yu, Taekyung; Park, Bum Jun

2017-07-01

In the synthesis of nanoparticles, the reaction rate is important to determine the morphology of nanoparticles. We investigated morphology evolution of Cu nanoparticles in this two different reactors, microemulsion reactor and batch reactor. In comparison with the batch reactor system, the enhanced mass and heat transfers in the emulsion system likely led to the relatively short nucleation time and the highly homogeneous environment in the reaction mixture, resulting in suppressing one or two dimensional growth of the nanoparticles. We believe that this work can offer a good model system to quantitatively understand the crystal growth mechanism that depends strongly on the local monomer concentration, the efficiency of heat transfer, and the relative contribution of the counter ions (Br- and Cl-) as capping agents.

Polymerization Reactor Engineering.

ERIC Educational Resources Information Center

Skaates, J. Michael

1987-01-01

Describes a polymerization reactor engineering course offered at Michigan Technological University which focuses on the design and operation of industrial polymerization reactors to achieve a desired degree of polymerization and molecular weight distribution. Provides a list of the course topics and assigned readings. (TW)

NEUTRONIC REACTOR CONTROL

DOEpatents

Hurwitz, H. Jr.

1960-04-01

An apparatus is described for indicating the approach to prompt criticality of a neutronic reactor and comprises means for oscillating an absorber in the reactor, a detector for measuring neutron flux in the reactor, two channels into which the output of the detector can be directed, one of which includes a narrow band filter with band pass frequency equal to that of the oscillator, and means for indicating the ratio of the signal produced by the channel with the filter to the signal produced by the other channel, which constitutes an indication of the approach to prompt criticality.

Reactor Dosimetry State of the Art 2008

NASA Astrophysics Data System (ADS)

Voorbraak, Wim; Debarberis, Luigi; D'Hondt, Pierre; Wagemans, Jan

2009-08-01

. Williams, A. P. Ribaric and T. Schnauber. Agile high-fidelity MCNP model development techniques for rapid mechanical design iteration / J. A. Kulesza.Extension of Raptor-M3G to r-8-z geometry for use in reactor dosimetry applications / M. A. Hunter, G. Longoni and S. L. Anderson. In vessel exposure distributions evaluated with MCNP5 for Atucha II / J. M. Longhino, H. Blaumann and G. Zamonsky. Atucha I nuclear power plant azimutal ex-vessel flux profile evaluation / J. M. Longhino ... [et al.]. UFTR thermal column characterization and redesign for maximized thermal flux / C. Polit and A. Haghighat. Activation counter using liquid light-guide for dosimetry of neutron burst / M. Hayashi ... [et al.]. Control rod reactivity curves for the annular core research reactor / K. R. DePriest ... [et al.]. Specification of irradiation conditions in VVER-440 surveillance positions / V. Kochkin ... [et al.]. Simulations of Mg-Ar ionisation and TE-TE ionisation chambers with MCNPX in a straightforward gamma and beta irradiation field / S. Nievaart ... [et al.]. The change of austenitic stainless steel elements content in the inner parts of VVER-440 reactor during operation / V. Smutný, J. Hep and P. Novosad. Fast neutron environmental spectrometry using disk activation / G. Lövestam ... [et al.]. Optimization of the neutron activation detector location scheme for VVER-lOOO ex-vessel dosimetry / V. N. Bukanov ... [et al.]. Irradiation conditions for surveillance specimens located into plane containers installed in the WWER-lOOO reactor of unit 2 of the South-Ukrainian NPP / O. V. Grytsenko. V. N. Bukanov and S. M. Pugach. Conformity between LRO mock-ups and VVERS NPP RPV neutron flux attenuation / S. Belousov. Kr. Ilieva and D. Kirilova. FLUOLE: a new relevant experiment for PWR pressure vessel surveillance / D. Beretz ... [et al.]. Transport of neutrons and photons through the iron and water layers / M. J. Kost'ál ... [et al.]. Condition evaluation of spent nuclear fuel assemblies

Thermal-Hydraulic Analysis of an Experimental Reactor Cavity Cooling System with Air. Part I: Experiments; Part II: Separate Effects Tests and Modeling

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

Corradin, Michael; Anderson, M.; Muci, M.

This experimental study investigates the thermal hydraulic behavior and the heat removal performance for a scaled Reactor Cavity Cooling System (RCCS) with air. A quarter-scale RCCS facility was designed and built based on a full-scale General Atomics (GA) RCCS design concept for the Modular High Temperature Gas Reactor (MHTGR). The GA RCCS is a passive cooling system that draws in air to use as the cooling fluid to remove heat radiated from the reactor pressure vessel to the air-cooled riser tubes and discharged the heated air into the atmosphere. Scaling laws were used to preserve key aspects and to maintainmore » similarity. The scaled air RCCS facility at UW-Madison is a quarter-scale reduced length experiment housing six riser ducts that represent a 9.5° sector slice of the full-scale GA air RCCS concept. Radiant heaters were used to simulate the heat radiation from the reactor pressure vessel. The maximum power that can be achieved with the radiant heaters is 40 kW with a peak heat flux of 25 kW per meter squared. The quarter-scale RCCS was run under different heat loading cases and operated successfully. Instabilities were observed in some experiments in which one of the two exhaust ducts experienced a flow reversal for a period of time. The data and analysis presented show that the RCCS has promising potential to be a decay heat removal system during an accident scenario.« less

STUDIES OF FAST REACTOR FUEL ELEMENT BEHAVIOR UNDER TRANSIENT HEATING TO FAILURE. I. INITIAL EXPERIMENTS ON METALLIC SAMPLES IN THE ABSENCE OF COOLANT

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

Dickerman, C. E.; Sowa, E. S.; Okrent, D.

1961-08-01

Meltdown tests on single metallic unirradiated fuel elements in TREAT are described. The fuel elements (EBRII Mark I fuel pins, EBR-II fuel pins with retractory Nb or Ta cladding, and Fermi-I fuel pins) are tested in an inert atmosphere, with no coolant. The fuel elements are exposed to reactor power bursts of 200 msec to 25 sec duration, under conditions simulating fast reactor operations. For these tests, the type of power burst, the integrated power, the fuel enrichment, the maximum cladding temperature, and the effects of the test on the fuel element are recorded. ( T.F.H.)

Safety survey report EBR-II safety survey, ANL-west health protection, industrial safety and fire protection survey

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

Dunbar, K.A.

1972-01-10

A safety survey covering the disciplines of Reactor Safety, Nuclear Criticality Safety, Health Protection and Industrial Safety and Fire Protection was conducted at the ANL-West EBR-II FEF Complex during the period January 10-18, 1972. In addition, the entire ANL-West site was surveyed for Health Protection and Industrial Safety and Fire Protection. The survey was conducted by members of the AEC Chicago Operations Office, a member of RDT-HQ and a member of the RDT-ID site office. Eighteen recommendations resulted from the survey, eleven in the area of Industrial Safety and Fire Protection, five in the area of Reactor Safety and twomore » in the area of Nuclear Criticality Safety.« less

Healthy Choices for Every Body Adult Curriculum Improves Participants' Food Resource Management Skills and Food Safety Practices.

PubMed

Adedokun, Omolola A; Plonski, Paula; Jenkins-Howard, Brooke; Cotterill, Debra B; Vail, Ann

2018-06-01

To evaluate the impact of the University of Kentucky's Healthy Choices for Every Body (HCEB) adult nutrition education curriculum on participants' food resource management (FRM) skills and food safety practices. A quasi-experimental design was employed using propensity score matching to pair 8 intervention counties with 8 comparison counties. Independent-samples t tests and ANCOVA models compared gains in FRM skills and food safety practices between the intervention and comparison groups (n = 413 and 113, respectively). Propensity score matching analysis showed a statistical balance and similarities between the comparison and intervention groups. Food resource management and food safety gain scores were statistically significantly higher for the intervention group (P < .001), with large effect sizes (d = 0.9) for both variables. The group differences persisted even after controlling for race and age in the ANCOVA models. The HCEB curriculum was effective in improving the FRM skills and food safety practices of participants. Copyright © 2018 Society for Nutrition Education and Behavior. Published by Elsevier Inc. All rights reserved.

Binding Selectivity of Methanobactin from Methylosinus trichosporium OB3b for Copper(I), Silver(I), Zinc(II), Nickel(II), Cobalt(II), Manganese(II), Lead(II), and Iron(II)

NASA Astrophysics Data System (ADS)

McCabe, Jacob W.; Vangala, Rajpal; Angel, Laurence A.

2017-12-01

Methanobactin (Mb) from Methylosinus trichosporium OB3b is a member of a class of metal binding peptides identified in methanotrophic bacteria. Mb will selectively bind and reduce Cu(II) to Cu(I), and is thought to mediate the acquisition of the copper cofactor for the enzyme methane monooxygenase. These copper chelating properties of Mb make it potentially useful as a chelating agent for treatment of diseases where copper plays a role including Wilson's disease, cancers, and neurodegenerative diseases. Utilizing traveling wave ion mobility-mass spectrometry (TWIMS), the competition for the Mb copper binding site from Ag(I), Pb(II), Co(II), Fe(II), Mn(II), Ni(II), and Zn(II) has been determined by a series of metal ion titrations, pH titrations, and metal ion displacement titrations. The TWIMS analyses allowed for the explicit identification and quantification of all the individual Mb species present during the titrations and measured their collision cross-sections and collision-induced dissociation patterns. The results showed Ag(I) and Ni(II) could irreversibly bind to Mb and not be effectively displaced by Cu(I), whereas Ag(I) could also partially displace Cu(I) from the Mb complex. At pH ≈ 6.5, the Mb binding selectivity follows the order Ag(I)≈Cu(I)>Ni(II)≈Zn(II)>Co(II)>>Mn(II)≈Pb(II)>Fe(II), and at pH 7.5 to 10.4 the order is Ag(I)>Cu(I)>Ni(II)>Co(II)>Zn(II)>Mn(II)≈Pb(II)>Fe(II). Breakdown curves of the disulfide reduced Cu(I) and Ag(I) complexes showed a correlation existed between their relative stability and their compact folded structure indicated by their CCS. Fluorescence spectroscopy, which allowed the determination of the binding constant, compared well with the TWIMS analyses, with the exception of the Ni(II) complex. [Figure not available: see fulltext.

NUCLEAR REACTOR

DOEpatents

Christy, R.F.

1958-07-15

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

Nuclear reactor building

DOEpatents

Gou, Perng-Fei; Townsend, Harold E.; Barbanti, Giancarlo

1994-01-01

A reactor building for enclosing a nuclear reactor includes a containment vessel having a wetwell disposed therein. The wetwell includes inner and outer walls, a floor, and a roof defining a wetwell pool and a suppression chamber disposed thereabove. The wetwell and containment vessel define a drywell surrounding the reactor. A plurality of vents are disposed in the wetwell pool in flow communication with the drywell for channeling into the wetwell pool steam released in the drywell from the reactor during a LOCA for example, for condensing the steam. A shell is disposed inside the wetwell and extends into the wetwell pool to define a dry gap devoid of wetwell water and disposed in flow communication with the suppression chamber. In a preferred embodiment, the wetwell roof is in the form of a slab disposed on spaced apart support beams which define therebetween an auxiliary chamber. The dry gap, and additionally the auxiliary chamber, provide increased volume to the suppression chamber for improving pressure margin.

A simulator-based nuclear reactor emergency response training exercise.

PubMed

Waller, Edward; Bereznai, George; Shaw, John; Chaput, Joseph; Lafortune, Jean-Francois

Training offsite emergency response personnel basic awareness of onsite control room operations during nuclear power plant emergency conditions was the primary objective of a week-long workshop conducted on a CANDU® virtual nuclear reactor simulator available at the University of Ontario Institute of Technology, Oshawa, Canada. The workshop was designed to examine both normal and abnormal reactor operating conditions, and to observe the conditions in the control room that may have impact on the subsequent offsite emergency response. The workshop was attended by participants from a number of countries encompassing diverse job functions related to nuclear emergency response. Objectives of the workshop were to provide opportunities for participants to act in the roles of control room personnel under different reactor operating scenarios, providing a unique experience for participants to interact with the simulator in real-time, and providing increased awareness of control room operations during accident conditions. The ability to "pause" the simulator during exercises allowed the instructors to evaluate and critique the performance of participants, and to provide context with respect to potential offsite emergency actions. Feedback from the participants highlighted (i) advantages of observing and participating "hands-on" with operational exercises, (ii) their general unfamiliarity with control room operational procedures and arrangements prior to the workshop, (iii) awareness of the vast quantity of detailed control room procedures for both normal and transient conditions, and (iv) appreciation of the increased workload for the operators in the control room during a transient from normal operations. Based upon participant feedback, it was determined that the objectives of the training had been met, and that future workshops should be conducted.

FLOW SYSTEM FOR REACTOR

DOEpatents

Zinn, W.H.

1963-06-11

A reactor is designed with means for terminating the reaction when returning coolant is below a predetermined temperature. Coolant flowing from the reactor passes through a heat exchanger to a lower reservoir, and then circulates between the lower reservoir and an upper reservoir before being returned to the reactor. Means responsive to the temperature of the coolant in the return conduit terminate the chain reaction when the temperature reaches a predetermined minimum value. (AEC)

Early administration of tolvaptan preserves renal function in elderly patients with acute decompensated heart failure.

PubMed

Kimura, Kazuhiro; Momose, Tomoyasu; Hasegawa, Tomoya; Morita, Takehiro; Misawa, Takuo; Motoki, Hirohiko; Izawa, Atsushi; Ikeda, Uichi

2016-05-01

Loop diuretics used in the treatment of heart failure often induce renal impairment. This study was conducted in order to evaluate the renal protective effect of adding tolvaptan (TLV), compared to increasing the furosemide (FRM) dose, for the treatment of acute decompensated heart failure (ADHF) in a real-world elderly patient population. This randomized controlled trial enrolled 52 consecutive hospitalized patients (age 83.4±9.6 years) with ADHF. The patients were assigned alternately to either the TLV group (TLV plus conventional treatment, n=26) or the FRM group (increasing the dose of FRM, n=26). TLV was administered within 24h from admission. The incidence of worsening renal function (WRF) within 7 days from admission was significantly lower in the TLV group (26.9% vs. 57.7%, p=0.025). Furthermore, the rates of occurrence of persistent and late-onset (≥5 days from admission) WRF were significantly lower in the TLV group. Persistent and late-onset WRF were significantly associated with a higher incidence of cardiac death or readmission for worsening heart failure in the 90 days following discharge, compared to transient and early-onset WRF, respectively. Early administration of TLV, compared to increased FRM dosage, reduces the incidence of WRF in real-world elderly ADHF patients. In addition, it reduces the occurrence of 'worse' WRF-persistent and late-onset WRF-which are associated with increased rates of cardiac death or readmission for worsening heart failure in the 90 days after discharge. Copyright © 2015 Japanese College of Cardiology. Published by Elsevier Ltd. All rights reserved.

Benefits and Limitations of Real Options Analysis for the Practice of River Flood Risk Management

NASA Astrophysics Data System (ADS)

Kind, Jarl M.; Baayen, Jorn H.; Botzen, W. J. Wouter

2018-04-01

Decisions on long-lived flood risk management (FRM) investments are complex because the future is uncertain. Flexibility and robustness can be used to deal with future uncertainty. Real options analysis (ROA) provides a welfare-economics framework to design and evaluate robust and flexible FRM strategies under risk or uncertainty. Although its potential benefits are large, ROA is hardly used in todays' FRM practice. In this paper, we investigate benefits and limitations of a ROA, by applying it to a realistic FRM case study for an entire river branch. We illustrate how ROA identifies optimal short-term investments and values future options. We develop robust dike investment strategies and value the flexibility offered by additional room for the river measures. We benchmark the results of ROA against those of a standard cost-benefit analysis and show ROA's potential policy implications. The ROA for a realistic case requires a high level of geographical detail, a large ensemble of scenarios, and the inclusion of stakeholders' preferences. We found several limitations of applying the ROA. It is complex. In particular, relevant sources of uncertainty need to be recognized, quantified, integrated, and discretized in scenarios, requiring subjective choices and expert judgment. Decision trees have to be generated and stakeholders' preferences have to be translated into decision rules. On basis of this study, we give general recommendations to use high discharge scenarios for the design of measures with high fixed costs and few alternatives. Lower scenarios may be used when alternatives offer future flexibility.

SELF-REGULATING BOILING-WATER NUCLEAR REACTORS

DOEpatents

Ransohoff, J.A.; Plawchan, J.D.

1960-08-16

A boiling-water reactor was designed which comprises a pressure vessel containing a mass of water, a reactor core submerged within the water, a reflector tank disposed within the reactor, the reflector tank being open at the top to the interior of the pressure vessel, and a surge tank connected to the reflector tank. In operation the reflector level changes as a function of the pressure witoin the reactor so that the reactivity of the reactor is automatically controlled.

Nuclear reactor reflector

DOEpatents

Hopkins, Ronald J.; Land, John T.; Misvel, Michael C.

1994-01-01

A nuclear reactor reflector is disclosed that comprises a stack of reflector blocks with vertical water flow passages to cool the reflector. The interface between blocks is opposite support points for reactor fuel rods. Water flows between the reflector and the reactor barrel from passages in a bottom block. The top block contains a flange to limit this flow and the flange has a slot to receive an alignment pin that is welded to the barrel. The pin is held in the slot by two removable shims. Alignment bars extend the length of the stack in slots machined in each block when the stack is assembled.

Nuclear reactor reflector

DOEpatents

Hopkins, R.J.; Land, J.T.; Misvel, M.C.

1994-06-07

A nuclear reactor reflector is disclosed that comprises a stack of reflector blocks with vertical water flow passages to cool the reflector. The interface between blocks is opposite support points for reactor fuel rods. Water flows between the reflector and the reactor barrel from passages in a bottom block. The top block contains a flange to limit this flow and the flange has a slot to receive an alignment pin that is welded to the barrel. The pin is held in the slot by two removable shims. Alignment bars extend the length of the stack in slots machined in each block when the stack is assembled. 12 figs.

Imaging Fukushima Daiichi reactors with muons

NASA Astrophysics Data System (ADS)

Miyadera, Haruo; Borozdin, Konstantin N.; Greene, Steve J.; Lukić, Zarija; Masuda, Koji; Milner, Edward C.; Morris, Christopher L.; Perry, John O.

2013-05-01

A study of imaging the Fukushima Daiichi reactors with cosmic-ray muons to assess the damage to the reactors is presented. Muon scattering imaging has high sensitivity for detecting uranium fuel and debris even through thick concrete walls and a reactor pressure vessel. Technical demonstrations using a reactor mockup, detector radiation test at Fukushima Daiichi, and simulation studies have been carried out. These studies establish feasibility for the reactor imaging. A few months of measurement will reveal the spatial distribution of the reactor fuel. The muon scattering technique would be the best and probably the only way for Fukushima Daiichi to make this determination in the near future.

NEUTRONIC REACTORS

DOEpatents

Vernon, H.C.

1959-01-13

A neutronic reactor of the heterogeneous, fluid cooled tvpe is described. The reactor is comprised of a pressure vessel containing the moderator and a plurality of vertically disposed channels extending in spaced relationship through the moderator. Fissionable fuel material is placed within the channels in spaced relationship thereto to permit circulation of the coolant fluid. Separate means are provided for cooling the moderator and for circulating a fluid coolant thru the channel elements to cool the fuel material.

NEUTRONIC REACTORS

DOEpatents

Wigner, E.P.; Young, G.J.

1958-10-14

A method is presented for loading and unloading rod type fuel elements of a neutronic reactor of the heterogeneous, solld moderator, liquid cooled type. In the embodiment illustrated, the fuel rods are disposed in vertical coolant channels in the reactor core. The fuel rods are loaded and unloaded through the upper openings of the channels which are immersed in the coolant liquid, such as water. Unloading is accomplished by means of a coffer dam assembly having an outer sleeve which is placed in sealing relation around the upper opening. A radiation shield sleeve is disposed in and reciprocable through the coffer dam sleeve. A fuel rod engaging member operates through the axial bore in the radiation shield sleeve to withdraw the fuel rod from its position in the reactor coolant channel into the shield, the shield snd rod then being removed. Loading is accomplished in the reverse procedure.

POWER REACTOR

DOEpatents

Zinn, W.H.

1958-07-01

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

REACTOR-FLASH BOILER-FLYWHEEL POWER PLANT

DOEpatents

Loeb, E.

1961-01-17

A power generator in the form of a flywheel with four reactors positioned about its rim is described. The reactors are so positioned that steam, produced in the reactor, exists tangentially to the flywheel, giving it a rotation. The reactors are incompletely moderated without water. The water enters the flywheel at its axis, under sufficient pressure to force it through the reactors, where it is converted to steam. The fuel consists of parallel twisted ribbons assembled to approximate a cylinder.

NEUTRONIC REACTOR CHARGING AND DISCHARGING

DOEpatents

Zinn, W.H.

1959-07-14

A method and arrangement is presented for removing a fuel element from a neutronic reactor tube through which a liquid coolant is being circulaled. The fuel element is moved into a section of the tube beyond the reactor proper, and then the coolant in the tube between the fuel element and the reactor proper is frozen, so that the fuel element may be removed from the tube without loss of the coolant therein. The method is particularly useful in the case of a liquid metal- cooled reactor.

Non-equilibrium radiation nuclear reactor

NASA Technical Reports Server (NTRS)

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

1978-01-01

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

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.

Regulatory impact analysis and regulatory support document: Control of air pollution; determination of significance for nonroad sources and emission standards for new nonroad compression-ignition engines at or above 37 kilowatts (50 horsepower). Final report

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

Trimble, T.; North, D.R.; Green, K.A.H.

1994-05-27

The regulatory impact analysis and support document provides additional information in support of the Final Rulemaking (FRM). This FRM will regulate all new nonroad compression-ignition engines greater than or equal to 37 kilowatts (50 hp), except engines which propel or are used on marine vessels, aircraft engines, engines which propel locomotives, and engines regulated by the Mining, Safety, and Health Administration. The regulated engines are hereafter referred to as nonroad large CI engines. The goal of this regulation is to substantially reduce NOx emission and smoke from nonroad large CI engines beginning in the 1996 model year.

Breeder Reactors, Understanding the Atom Series.

ERIC Educational Resources Information Center

Mitchell, Walter, III; Turner, Stanley E.

The theory of breeder reactors in relationship to a discussion of fission is presented. Different kinds of reactors are characterized by the cooling fluids used, such as liquid metal, gas, and molten salt. The historical development of breeder reactors over the past twenty-five years includes specific examples of reactors. The location and a brief…

Imaging Fukushima Daiichi reactors with muons

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

Miyadera, Haruo; Borozdin, Konstantin N.; Greene, Steve J.

2013-05-15

A study of imaging the Fukushima Daiichi reactors with cosmic-ray muons to assess the damage to the reactors is presented. Muon scattering imaging has high sensitivity for detecting uranium fuel and debris even through thick concrete walls and a reactor pressure vessel. Technical demonstrations using a reactor mockup, detector radiation test at Fukushima Daiichi, and simulation studies have been carried out. These studies establish feasibility for the reactor imaging. A few months of measurement will reveal the spatial distribution of the reactor fuel. The muon scattering technique would be the best and probably the only way for Fukushima Daiichi tomore » make this determination in the near future.« less

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.

REACTOR FUEL ELEMENTS TESTING CONTAINER

DOEpatents

Whitham, G.K.; Smith, R.R.

1963-01-15

This patent shows a method for detecting leaks in jacketed fuel elements. The element is placed in a sealed tank within a nuclear reactor, and, while the reactor operates, the element is sparged with gas. The gas is then led outside the reactor and monitored for radioactive Xe or Kr. (AEC)

Nuclear reactor building

DOEpatents

Gou, P.F.; Townsend, H.E.; Barbanti, G.

1994-04-05

A reactor building for enclosing a nuclear reactor includes a containment vessel having a wetwell disposed therein. The wetwell includes inner and outer walls, a floor, and a roof defining a wetwell pool and a suppression chamber disposed there above. The wetwell and containment vessel define a drywell surrounding the reactor. A plurality of vents are disposed in the wetwell pool in flow communication with the drywell for channeling into the wetwell pool steam released in the drywell from the reactor during a LOCA for example, for condensing the steam. A shell is disposed inside the wetwell and extends into the wetwell pool to define a dry gap devoid of wetwell water and disposed in flow communication with the suppression chamber. In a preferred embodiment, the wetwell roof is in the form of a slab disposed on spaced apart support beams which define there between an auxiliary chamber. The dry gap, and additionally the auxiliary chamber, provide increased volume to the suppression chamber for improving pressure margin. 4 figures.

WATER BOILER REACTOR

DOEpatents

King, L.D.P.

1960-11-22

As its name implies, this reactor utilizes an aqueous solution of a fissionable element salt, and is also conventional in that it contains a heat exchanger cooling coil immersed in the fuel. Its novelty lies in the utilization of a cylindrical reactor vessel to provide a critical region having a large and constant interface with a supernatant vapor region, and the use of a hollow sleeve coolant member suspended from the cover assembly in coaxial relation with the reactor vessel. Cool water is circulated inside this hollow coolant member, and a gap between its outer wall and the reactor vessel is used to carry off radiolytic gases for recombination in an external catalyst chamber. The central passage of the coolant member defines a reflux condenser passage into which the externally recombined gases are returned and condensed. The large and constant interface between fuel solution and vapor region prevents the formation of large bubbles and minimizes the amount of fuel salt carried off by water vapor, thus making possible higher flux densities, specific powers and power densities.

COOLED NEUTRONIC REACTOR

DOEpatents

Binner, C.R.; Wilkie, C.B.

1958-03-18

This patent relates to a design for a reactor of the type in which a fluid coolant is flowed through the active portion of the reactor. This design provides for the cooling of the shielding material as well as the reactor core by the same fluid coolant. The core structure is a solid moderator having coolant channels in which are disposed the fuel elements in rod or slug form. The coolant fluid enters the chamber in the shield, in which the core is located, passes over the inner surface of said chamber, enters the core structure at the center, passes through the coolant channels over the fuel elements and out through exhaust ducts.

PBF Reactor Building (PER620). Reactor vessel arrives from gate city ...

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

PBF Reactor Building (PER-620). Reactor vessel arrives from gate city steel at door of PBF. On flatbed, it is too high to fit under door. Photographer: Larry Page. Date: February 13, 1970. INEEL negative no. 70-737 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

Shutdown system for a nuclear reactor

DOEpatents

Groh, E.F.; Olson, A.P.; Wade, D.C.; Robinson, B.W.

1984-06-05

An ultimate shutdown system is provided for termination of neutronic activity in a nuclear reactor. The shutdown system includes bead chains comprising spherical containers suspended on a flexible cable. The containers are comprised of mating hemispherical shells which provide a ruggedized enclosure for reactor poison material. The bead chains, normally suspended above the reactor core on storage spools, are released for downward travel upon command from an external reactor monitor. The chains are capable of horizontal movement, so as to flow around obstructions in the reactor during their downward motion. 8 figs.

Shutdown system for a nuclear reactor

DOEpatents

Groh, Edward F.; Olson, Arne P.; Wade, David C.; Robinson, Bryan W.

1984-01-01

An ultimate shutdown system is provided for termination of neutronic activity in a nuclear reactor. The shutdown system includes bead chains comprising spherical containers suspended on a flexible cable. The containers are comprised of mating hemispherical shells which provide a ruggedized enclosure for reactor poison material. The bead chains, normally suspended above the reactor core on storage spools, are released for downward travel upon command from an external reactor monitor. The chains are capable of horizontal movement, so as to flow around obstructions in the reactor during their downward motion.

Control rod drive for reactor shutdown

DOEpatents

McKeehan, Ernest R.; Shawver, Bruce M.; Schiro, Donald J.; Taft, William E.

1976-01-20

A means for rapidly shutting down or scramming a nuclear reactor, such as a liquid metal-cooled fast breeder reactor, and serves as a backup to the primary shutdown system. The control rod drive consists basically of an in-core assembly, a drive shaft and seal assembly, and a control drive mechanism. The control rod is driven into the core region of the reactor by gravity and hydraulic pressure forces supplied by the reactor coolant, thus assuring that common mode failures will not interfere with or prohibit scramming the reactor when necessary.

Catalytic reactor

DOEpatents

Aaron, Timothy Mark [East Amherst, NY; Shah, Minish Mahendra [East Amherst, NY; Jibb, Richard John [Amherst, NY

2009-03-10

A catalytic reactor is provided with one or more reaction zones each formed of set(s) of reaction tubes containing a catalyst to promote chemical reaction within a feed stream. The reaction tubes are of helical configuration and are arranged in a substantially coaxial relationship to form a coil-like structure. Heat exchangers and steam generators can be formed by similar tube arrangements. In such manner, the reaction zone(s) and hence, the reactor is compact and the pressure drop through components is minimized. The resultant compact form has improved heat transfer characteristics and is far easier to thermally insulate than prior art compact reactor designs. Various chemical reactions are contemplated within such coil-like structures such that as steam methane reforming followed by water-gas shift. The coil-like structures can be housed within annular chambers of a cylindrical housing that also provide flow paths for various heat exchange fluids to heat and cool components.

Reactor for making uniform capsules

NASA Technical Reports Server (NTRS)

Wang, Taylor G. (Inventor); Anikumar, Amrutur V. (Inventor); Lacik, Igor (Inventor)

1999-01-01

The present invention provides a novel reactor for making capsules with uniform membrane. The reactor includes a source for providing a continuous flow of a first liquid through the reactor; a source for delivering a steady stream of drops of a second liquid to the entrance of the reactor; a main tube portion having at least one loop, and an exit opening, where the exit opening is at a height substantially equal to the entrance. In addition, a method for using the novel reactor is provided. This method involves providing a continuous stream of a first liquid; introducing uniformly-sized drops of the second liquid into the stream of the first liquid; allowing the drops to react in the stream for a pre-determined period of time; and collecting the capsules.

Anaerobic reactor/high rate pond combined technology for sewage treatment in the Mediterranean area.

PubMed

El Hafiane, F; El Hamouri, B

2005-01-01

Two high-rate, anaerobic/aerobic units were used to treat the sewage of the Institut Agronomique st Vétérinaire Hassan II (Morocco) campus in a 1,100 m2-plant designed for 1,500 e.p. and receiving 63 m3 per day. The anaerobic pre-treatment consisted of a two-step up-flow anaerobic reactor (TSUAR) comprising two reactors and one external settler all in series. The aerobic line, or post-treatment, consisted of a high-rate algal pond (HRAP) and one maturation pond in series. The system totalized a hydraulic retention time (HRT) of 9 days. A gravel filter (GF) was constructed behind the TSUAR to trap low-density particles. The TSUAR removed 80% of COD and 90% of SS within 48 h. Solids retention time in the reactors averaged 32 d with a specific sludge production of 0.28 g SS g(-1) COD removed. Almost 93% of the sludge evacuated from the settler was stabilized. Specific biogas production from both reactors was 0.25m3 kg(-1) COD removed. Used in this configuration, the HRAP lost its BOD removal activity and increased its nutrients and pathogens removal capabilities (tertiary treatment). Results showed that 85% of total nitrogen and 48% of total phosphorus were removed by the HRAP. Land area requirement of this combination was less than 1 m2 per capita and filtered final effluent was of excellent quality (COD, 82 mg/l; TKN, 8.3 mg/l; total P, 2.7 mg/l, faecal coliforms, 2.4 10(3)/100 ml and zero helminths eggs).

Plasma-material Interactions in Current Tokamaks and their Implications for Next-step Fusion Reactors

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

Federici, G.; Skinner, C.H.; Brooks, J.N.

2001-01-10

The major increase in discharge duration and plasma energy in a next-step DT [deuterium-tritium] fusion reactor will give rise to important plasma-material effects that will critically influence its operation, safety, and performance. Erosion will increase to a scale of several centimeters from being barely measurable at a micron scale in today's tokamaks. Tritium co-deposited with carbon will strongly affect the operation of machines with carbon plasma-facing components. Controlling plasma wall interactions is critical to achieving high performance in present-day tokamaks and this is likely to continue to be the case in the approach to practical fusion reactors. Recognition of themore » important consequences of these phenomena has stimulated an internationally coordinated effort in the field of plasma-surface interactions supporting the Engineering Design Activities of the International Thermonuclear Experimental Reactor (ITER) project and significant progress has been made in better under standing these issues. This paper reviews the underlying physical processes and the existing experimental database of plasma-material interactions both in tokamaks and laboratory simulation facilities for conditions of direct relevance to next-step fusion reactors. Two main topical groups of interactions are considered: (i) erosion/redeposition from plasma sputtering and disruptions, including dust and flake generation, (ii) tritium retention and removal. The use of modeling tools to interpret the experimental results and make projections for conditions expected in future devices is explained. Outstanding technical issues and specific recommendations on potential R and D [Research and Development] avenues for their resolution are presented.« less

Thermionic switched self-actuating reactor shutdown system

DOEpatents

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

1989-01-01

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

NUCLEAR REACTOR FUEL SYSTEMS

DOEpatents

Thamer, B.J.; Bidwell, R.M.; Hammond, R.P.

1959-09-15

Homogeneous reactor fuel solutions are reported which provide automatic recombination of radiolytic gases and exhibit large thermal expansion characteristics, thereby providing stability at high temperatures and enabling reactor operation without the necessity of apparatus to recombine gases formed by the radiolytic dissociation of water in the fuel and without the necessity of liquid fuel handling outside the reactor vessel except for recovery processes. The fuels consist of phosphoric acid and water solutions of enriched uranium, wherein the uranium is in either the hexavalent or tetravalent state.

Computational analysis of the dose rates at JSI TRIGA reactor irradiation facilities.

PubMed

Ambrožič, K; Žerovnik, G; Snoj, L

2017-12-01

The JSI TRIGA Mark II, IJS research reactor is equipped with numerous irradiation positions, where samples can be irradiated by neutrons and γ-rays. Irradiation position selection is based on its properties, such as physical size and accessibility, as well as neutron and γ-ray spectra, flux and dose intensities. This paper presents an overview on the neutron and γ-ray fluxes, spectra and dose intensities calculations using Monte Carlo MCNP software and ENDF/B-VII.0 nuclear data libraries. The dose-rates are presented in terms of ambient dose equivalents, air kerma, and silicon dose equivalent. At full reactor power the neutron ambient dose equivalent ranges from 5.5×10 3 Svh -1 to 6×10 6 Svh -1 , silicon dose equivalent from 6×10 2 Gy/h si to 3×10 5 Gy/h si , and neutron air kerma from 4.3×10 3 Gyh -1 to 2×10 5 Gyh -1 . Ratio of fast (1MeVreactor power from 3.4×10 3 Svh -1 to 3.6×10 5 Svh -1 and γ air kerma range 3.1×10 3 Gyh -1 to 2.9×10 5 Gyh -1 . Copyright © 2017 Elsevier Ltd. All rights reserved.

Neutrino oscillation studies with reactors

PubMed Central

Vogel, P.; Wen, L.J.; Zhang, C.

2015-01-01

Nuclear reactors are one of the most intense, pure, controllable, cost-effective and well-understood sources of neutrinos. Reactors have played a major role in the study of neutrino oscillations, a phenomenon that indicates that neutrinos have mass and that neutrino flavours are quantum mechanical mixtures. Over the past several decades, reactors were used in the discovery of neutrinos, were crucial in solving the solar neutrino puzzle, and allowed the determination of the smallest mixing angle θ13. In the near future, reactors will help to determine the neutrino mass hierarchy and to solve the puzzling issue of sterile neutrinos. PMID:25913819

Neutrino oscillation studies with reactors

DOE PAGES

Vogel, P.; Wen, L.J.; Zhang, C.

2015-04-27

Nuclear reactors are one of the most intense, pure, controllable, cost-effective and well-understood sources of neutrinos. Reactors have played a major role in the study of neutrino oscillations, a phenomenon that indicates that neutrinos have mass and that neutrino flavours are quantum mechanical mixtures. Over the past several decades, reactors were used in the discovery of neutrinos, were crucial in solving the solar neutrino puzzle, and allowed the determination of the smallest mixing angle θ 13. In the near future, reactors will help to determine the neutrino mass hierarchy and to solve the puzzling issue of sterile neutrinos.

Neutrino oscillation studies with reactors.

PubMed

Vogel, P; Wen, L J; Zhang, C

2015-04-27

Nuclear reactors are one of the most intense, pure, controllable, cost-effective and well-understood sources of neutrinos. Reactors have played a major role in the study of neutrino oscillations, a phenomenon that indicates that neutrinos have mass and that neutrino flavours are quantum mechanical mixtures. Over the past several decades, reactors were used in the discovery of neutrinos, were crucial in solving the solar neutrino puzzle, and allowed the determination of the smallest mixing angle θ13. In the near future, reactors will help to determine the neutrino mass hierarchy and to solve the puzzling issue of sterile neutrinos.

When Do Commercial Reactors Permanently Shut Down?

EIA Publications

2011-01-01

For those wishing to obtain current data, the following resources are available: U.S. reactors, go to the Energy Information Administration's nuclear reactor shutdown list. (Note: As of April 30, 2010, the last U.S. reactor to permanently shut down was Big Rock Point in 1997.) Foreign Reactors, go to the Power Reactor Information System (PRIS) on the International Atomic Energy Agency's website.

Transmutation of actinides in power reactors.

PubMed

Bergelson, B R; Gerasimov, A S; Tikhomirov, G V

2005-01-01

Power reactors can be used for partial short-term transmutation of radwaste. This transmutation is beneficial in terms of subsequent storage conditions for spent fuel in long-term storage facilities. CANDU-type reactors can transmute the main minor actinides from two or three reactors of the VVER-1000 type. A VVER-1000-type reactor can operate in a self-service mode with transmutation of its own actinides.

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.

Self isolating high frequency saturable reactor

DOEpatents

Moore, James A.

1998-06-23

The present invention discloses a saturable reactor and a method for decoupling the interwinding capacitance from the frequency limitations of the reactor so that the equivalent electrical circuit of the saturable reactor comprises a variable inductor. The saturable reactor comprises a plurality of physically symmetrical magnetic cores with closed loop magnetic paths and a novel method of wiring a control winding and a RF winding. The present invention additionally discloses a matching network and method for matching the impedances of a RF generator to a load. The matching network comprises a matching transformer and a saturable reactor.

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

Moon base reactor system

NASA Technical Reports Server (NTRS)

Chavez, H.; Flores, J.; Nguyen, M.; Carsen, K.

1989-01-01

The objective of our reactor design is to supply a lunar-based research facility with 20 MW(e). The fundamental layout of this lunar-based system includes the reactor, power conversion devices, and a radiator. The additional aim of this reactor is a longevity of 12 to 15 years. The reactor is a liquid metal fast breeder that has a breeding ratio very close to 1.0. The geometry of the core is cylindrical. The metallic fuel rods are of beryllium oxide enriched with varying degrees of uranium, with a beryllium core reflector. The liquid metal coolant chosen was natural lithium. After the liquid metal coolant leaves the reactor, it goes directly into the power conversion devices. The power conversion devices are Stirling engines. The heated coolant acts as a hot reservoir to the device. It then enters the radiator to be cooled and reenters the Stirling engine acting as a cold reservoir. The engines' operating fluid is helium, a highly conductive gas. These Stirling engines are hermetically sealed. Although natural lithium produces a lower breeding ratio, it does have a larger temperature range than sodium. It is also corrosive to steel. This is why the container material must be carefully chosen. One option is to use an expensive alloy of cerbium and zirconium. The radiator must be made of a highly conductive material whose melting point temperature is not exceeded in the reactor and whose structural strength can withstand meteor showers.

A NEUTRONIC REACTOR

DOEpatents

Luebke, E.A.; Vandenberg, L.B.

1959-09-01

A nuclear reactor for producing thermoelectric power is described. The reactor core comprises a series of thermoelectric assemblies, each assembly including fissionable fuel as an active element to form a hot junction and a thermocouple. The assemblies are disposed parallel to each other to form spaces and means are included for Introducing an electrically conductive coolant between the assemblies to form cold junctions of the thermocouples. An electromotive force is developed across the entire series of the thermoelectric assemblies due to fission heat generated in the fuel causing a current to flow perpendicular to the flow of coolant and is distributed to a load outside of the reactor by means of bus bars electrically connected to the outermost thermoelectric assembly.

REACTOR CONTROL DEVICE

DOEpatents

Graham, R.H.

1962-09-01

A wholly mechanical compact control device is designed for automatically rendering the core of a fission reactor subcritical in response to core temperatures in excess of the design operating temperature limit. The control device comprises an expansible bellows interposed between the base of a channel in a reactor core and the inner end of a fuel cylinder therein which is normally resiliently urged inwardly. The bellows contains a working fluid which undergoes a liquid to vapor phase change at a temperature substantially equal to the design temperature limit. Hence, the bellows abruptiy expands at this limiting temperature to force the fuel cylinder outward and render the core subcritical. The control device is particularly applicable to aircraft propulsion reactor service. (AEC)

EMERGENCY SHUTDOWN FOR NUCLEAR REACTORS

DOEpatents

Paget, J.A.; Koutz, S.L.; Stone, R.S.; Stewart, H.B.

1963-12-24

An emergency shutdown or scram apparatus for use in a nuclear reactor that includes a neutron absorber suspended from a temperature responsive substance that is selected to fail at a preselected temperature in excess of the normal reactor operating temperature, whereby the neutron absorber is released and allowed to fall under gravity to a preselected position within the reactor core is presented. (AEC)

NUCLEAR REACTOR

DOEpatents

Young, G.

1963-01-01

This patent covers a power-producing nuclear reactor in which fuel rods of slightly enriched U are moderated by heavy water and cooled by liquid metal. The fuel rods arranged parallel to one another in a circle are contained in a large outer closed-end conduit that extends into a tank containing the heavy water. Liquid metal is introduced into the large conduit by a small inner conduit that extends within the circle of fuel rods to a point near the lower closed end of the outer conduit. (AEC) Production Reactors

Microchannel Reactors for ISRU Applications

NASA Astrophysics Data System (ADS)

Carranza, Susana; Makel, Darby B.; Blizman, Brandon; Ward, Benjamin J.

2005-02-01

Affordable planning and execution of prolonged manned space missions depend upon the utilization of local resources and the waste products which are formed in manned spacecraft and surface bases. Successful in-situ resources utilization (ISRU) will require component technologies which provide optimal size, weight, volume, and power efficiency. Microchannel reactors enable the efficient chemical processing of in situ resources. The reactors can be designed for the processes that generate the most benefit for each mission. For instance, propellants (methane) can be produced from carbon dioxide from the Mars atmosphere using the Sabatier reaction and ethylene can be produced from the partial oxidation of methane. A system that synthesizes ethylene could be the precursor for systems to synthesize ethanol and polyethylene. Ethanol can be used as a nutrient for Astrobiology experiments, as well as the production of nutrients for human crew (e.g. sugars). Polyethylene can be used in the construction of habitats, tools, and replacement parts. This paper will present recent developments in miniature chemical reactors using advanced Micro Electro Mechanical Systems (MEMS) and microchannel technology to support ISRU of Mars and lunar missions. Among other applications, the technology has been demonstrated for the Sabatier process and for the partial oxidation of methane. Microchannel reactors were developed based on ceramic substrates as well as metal substrates. In both types of reactors, multiple layers coated with catalytic material are bonded, forming a monolithic structure. Such reactors are readily scalable with the incorporation of extra layers. In addition, this reactor structure minimizes pressure drop and catalyst settling, which are common problems in conventional packed bed reactors.

Auxiliary reactor for a hydrocarbon reforming system

DOEpatents

Clawson, Lawrence G.; Dorson, Matthew H.; Mitchell, William L.; Nowicki, Brian J.; Bentley, Jeffrey M.; Davis, Robert; Rumsey, Jennifer W.

2006-01-17

An auxiliary reactor for use with a reformer reactor having at least one reaction zone, and including a burner for burning fuel and creating a heated auxiliary reactor gas stream, and heat exchanger for transferring heat from auxiliary reactor gas stream and heat transfer medium, preferably two-phase water, to reformer reaction zone. Auxiliary reactor may include first cylindrical wall defining a chamber for burning fuel and creating a heated auxiliary reactor gas stream, the chamber having an inlet end, an outlet end, a second cylindrical wall surrounding first wall and a second annular chamber there between. The reactor being configured so heated auxiliary reactor gas flows out the outlet end and into and through second annular chamber and conduit which is disposed in second annular chamber, the conduit adapted to carry heat transfer medium and being connectable to reformer reaction zone for additional heat exchange.

Propellant actuated nuclear reactor steam depressurization valve

DOEpatents

Ehrke, Alan C.; Knepp, John B.; Skoda, George I.

1992-01-01

A nuclear fission reactor combined with a propellant actuated depressurization and/or water injection valve is disclosed. The depressurization valve releases pressure from a water cooled, steam producing nuclear reactor when required to insure the safety of the reactor. Depressurization of the reactor pressure vessel enables gravity feeding of supplementary coolant water through the water injection valve to the reactor pressure vessel to prevent damage to the fuel core.

RADIATION FACILITY FOR NUCLEAR REACTORS

DOEpatents

Currier, E.L. Jr.; Nicklas, J.H.

1961-12-12

A radiation facility is designed for irradiating samples in close proximity to the core of a nuclear reactor. The facility comprises essentially a tubular member extending through the biological shield of the reactor and containing a manipulatable rod having the sample carrier at its inner end, the carrier being longitudinally movable from a position in close proximity to the reactor core to a position between the inner and outer faces of the shield. Shield plugs are provided within the tubular member to prevent direct radiation from the core emanating therethrough. In this device, samples may be inserted or removed during normal operation of the reactor without exposing personnel to direct radiation from the reactor core. A storage chamber is also provided within the radiation facility to contain an irradiated sample during the period of time required to reduce the radioactivity enough to permit removal of the sample for external handling. (AEC)

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.

Fast-acting nuclear reactor control device

DOEpatents

Kotlyar, Oleg M.; West, Phillip B.

1993-01-01

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

High Solid Fed-batch Butanol Fermentation with Simultaneous Product Recovery: Part II - Process Integration.

PubMed

Qureshi, Nasib; Klasson, K Thomas; Saha, Badal C; Liu, Siqing

2018-04-25

In these studies liquid hot water (LHW) pretreated and enzymatically hydrolyzed Sweet Sorghum Bagasse (SSB) hydrolyzates were fermented in a fed-batch reactor. As reported in the preceding paper, the culture was not able to ferment the hydrolyzate I in a batch process due to presence of high level of toxic chemicals, in particular acetic acid released from SSB during the hydrolytic process. To be able to ferment the hydrolyzate I obtained from 250 gL -1 SSB hydrolysis, a fed-batch reactor with in-situ butanol recovery was devised. The process was started with the hydrolyzate II and when good cell growth and vigorous fermentation were observed, the hydrolyzate I was slowly fed to the reactor. In this manner the culture was able to ferment all the sugars present in both the hydrolyzates to acetone butanol ethanol (ABE). In a control batch reactor in which ABE was produced from glucose, ABE productivity and yield of 0.42 gL -1 h -1 and 0.36 were obtained, respectively. In the fed-batch reactor fed with SSB hydrolyzates these productivity and yield values were 0.44 gL -1 h -1 and 0.45, respectively. ABE yield in the integrated system was high due to utilization of acetic acid to convert to ABE. In summary we were able to utilize both the hydrolyzates obtained from LHW pretreated and enzymatically hydrolyzed SSB (250 gL -1 ) and convert them to ABE. Complete fermentation was possible due to simultaneous recovery of ABE by vacuum. This article is protected by copyright. All rights reserved. © 2018 American Institute of Chemical Engineers.

The clinical utility of fibrin-related biomarkers in sepsis.

PubMed

Toh, Julien M H; Ken-Dror, Gie; Downey, Colin; Abrams, Simon T

2013-12-01

Sepsis is associated with systemic inflammatory responses and induction of intravascular fibrin formation. Our aim is to investigate whether three fibrin-related markers (FRM) reflect the extent of coagulation activation in vivo and evaluate their clinical usefulness in identifying as well as monitoring patients with sepsis. Fibrin-degradation products (FDP), D-dimer and soluble fibrin monomer assays were measured on plasma samples from patients in the ICU with sepsis (n = 37), systemic inflammatory response syndrome (SIRS) (n = 35) and healthy individuals (n = 15). The levels were correlated with each other and also with fibrinogen, prothrombin time, platelets and antithrombin III. Clinical correlation was also performed for the diagnosis of sepsis and longitudinal monitoring for survival or death.There was strong correlation between the three FRM (r = 0.38-0.93, P < 0.0001) with only fibrin monomer correlating significantly with prothrombin time, fibrinogen and platelet levels. Clinically, all three FRM could discriminate between patients with sepsis, SIRS and healthy individuals with FDP, and D-dimer showing statistical significance (P < 0.05). No FRM predicted outcome from a single measurement but FDP was significantly able to predict patient survival from serial samples [mean FDP (μg/ml) from 35.36 to 21.37 (first to third ICU-day), P < 0.05]. Fibrin monomer appears the most sensitive indicator of coagulation activation, whereas D-dimer and FDP levels can significantly differentiate ICU patients with sepsis from those without. In addition, FDP would be preferable for monitoring with its statistically significant time-dependent prediction of survival or death from sepsis.

REACTOR CONTROL

DOEpatents

Ruano, W.J.

1957-12-10

This patent relates to nuclear reactors of the type which utilize elongited rod type fuel elements immersed in a liquid moderator and shows a design whereby control of the chain reaction is obtained by varying the amount of moderator or reflector material. A central tank for containing liquid moderator and fuel elements immersed therein is disposed within a surrounding outer tank providing an annular space between the two tanks. This annular space is filled with liquid moderator which functions as a reflector to reflect neutrons back into the central reactor tank to increase the reproduction ratio. Means are provided for circulating and cooling the moderator material in both tanks and additional means are provided for controlling separately the volume of moderator in each tank, which latter means may be operated automatically by a neutron density monitoring device. The patent also shows an arrangement for controlling the chain reaction by injecting and varying an amount of poisoning material in the moderator used in the reflector portion of the reactor.

Delayed Gamma Measurements in Different Nuclear Research Reactors Bringing Out the Importance of the Delayed Contribution in Gamma Flux Calculations

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

Fourmentel, D.; Radulovic, V.; Barbot, L.

Neutron and gamma flux levels are key parameters in nuclear research reactors. In Material Testing Reactors, such as the future Jules Horowitz Reactor, under construction at the French Alternative Energies and Atomic Energy Commission (CEA Cadarache, France), the expected gamma flux levels are very high (nuclear heating is of the order of 20 W/g at 100 MWth). As gamma rays deposit their energy in the reactor structures and structural materials it is important to take them into account when designing irradiation devices. There are only a few sensors which allow measurements of the nuclear heating ; a recent development atmore » the CEA Cadarache allows measurements of the gamma flux using a miniature ionization chamber (MIC). The measured MIC response is often compared with calculation using modern Monte Carlo (MC) neutron and photon transport codes, such as TRIPOLI-4 and MCNP6. In these calculations only the production of prompt gamma rays in the reactor is usually modelled thus neglecting the delayed gamma rays. Hence calculations and measurements are usually in better accordance for the neutron flux than for the gamma flux. In this paper we study the contribution of delayed gamma rays to the total MIC signal in order to estimate the systematic error in gamma flux MC calculations. In order to experimentally determine the delayed gamma flux contributions to the MIC response, we performed gamma flux measurements with CEA developed MIC at three different research reactors: the OSIRIS reactor (MTR - 70 MWth at CEA Saclay, France), the TRIGA MARK II reactor (TRIGA - 250 kWth at the Jozef Stefan Institute, Slovenia) and the MARIA reactor (MTR - 30 MWth at the National Center for Nuclear Research, Poland). In order to experimentally assess the delayed gamma flux contribution to the total gamma flux, several reactor shut down (scram) experiments were performed specifically for the purpose of the measurements. Results show that on average about 30 % of the MIC signal is

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

Research Program of a Super Fast Reactor

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

Oka, Yoshiaki; Ishiwatari, Yuki; Liu, Jie

2006-07-01

Research program of a supercritical-pressure light water cooled fast reactor (Super Fast Reactor) is funded by MEXT (Ministry of Education, Culture, Sports, Science and Technology) in December 2005 as one of the research programs of Japanese NERI (Nuclear Energy Research Initiative). It consists of three programs. (1) development of Super Fast Reactor concept; (2) thermal-hydraulic experiments; (3) material developments. The purpose of the concept development is to pursue the advantage of high power density of fast reactor over thermal reactors to achieve economic competitiveness of fast reactor for its deployment without waiting for exhausting uranium resources. Design goal is notmore » breeding, but maximizing reactor power by using plutonium from spent LWR fuel. MOX will be the fuel of the Super Fast Reactor. Thermal-hydraulic experiments will be conducted with HCFC22 (Hydro chlorofluorocarbons) heat transfer loop of Kyushu University and supercritical water loop at JAEA. Heat transfer data including effect of grid spacers will be taken. The critical flow and condensation of supercritical fluid will be studied. The materials research includes the development and testing of austenitic stainless steel cladding from the experience of PNC1520 for LMFBR. Material for thermal insulation will be tested. SCWR (Supercritical-Water Cooled Reactor) of GIF (Generation-4 International Forum) includes both thermal and fast reactors. The research of the Super Fast Reactor will enhance SCWR research and the data base. The research period will be until March 2010. (authors)« less

Tritium resources available for fusion reactors

NASA Astrophysics Data System (ADS)

Kovari, M.; Coleman, M.; Cristescu, I.; Smith, R.

2018-02-01

The tritium required for ITER will be supplied from the CANDU production in Ontario, but while Ontario may be able to supply 8 kg for a DEMO fusion reactor in the mid-2050s, it will not be able to provide 10 kg at any realistic starting time. The tritium required to start DEMO will depend on advances in plasma fuelling efficiency, burnup fraction, and tritium processing technology. It is in theory possible to start up a fusion reactor with little or no tritium, but at an estimated cost of 2 billion per kilogram of tritium saved, it is not economically sensible. Some heavy water reactor tritium production scenarios with varying degrees of optimism are presented, with the assumption that only Canada, the Republic of Korea, and Romania make tritium available to the fusion community. Results for the tritium available for DEMO in 2055 range from zero to 30 kg. CANDU and similar heavy water reactors could in theory generate additional tritium in a number of ways: (a) adjuster rods containing lithium could be used, giving 0.13 kg per year per reactor; (b) a fuel bundle with a burnable absorber has been designed for CANDU reactors, which might be adapted for tritium production; (c) tritium production could be increased by 0.05 kg per year per reactor by doping the moderator with lithium-6. If a fusion reactor is started up around 2055, governments in Canada, Argentina, China, India, South Korea and Romania will have the opportunity in the years leading up to that to take appropriate steps: (a) build, refurbish or upgrade tritium extraction facilities; (b) extend the lives of heavy water reactors, or build new ones; (c) reduce tritium sales; (d) boost tritium production in the remaining heavy water reactors. All of the alternative production methods considered have serious economic and regulatory drawbacks, and the risk of diversion of tritium or lithium-6 would also be a major concern. There are likely to be serious problems with supplying tritium for future

Manganese(II), iron(II), cobalt(II), and copper(II) complexes of an extended inherently chiral tris-bipyridyl cage.

PubMed

Perkins, David F; Lindoy, Leonard F; McAuley, Alexander; Meehan, George V; Turner, Peter

2006-01-17

Manganese(II), iron(II), cobalt(II), and copper(II) derivatives of two inherently chiral, Tris(bipyridyl) cages (L and L') of type [ML]-(PF(6))(2)(solvent)(n) and [FeL'](ClO(4))(2) are reported, where L is the hexa-tertiary butyl-substituted derivative of L'. These products were obtained by using the free cage and metal template procedures; the latter involved the reductive amination of the respective Tris-dialdehyde precursor complexes of iron(II), cobalt(II), or nickel(II). Electrochemical, EPR, and NMR studies have been used to probe the nature of the individual complexes. X-ray structures of the manganese(II), iron(II), and copper(II) complexes of L and the iron(II) complex of L' are presented; these are compared with the previously reported structures of the corresponding nickel(II) complex and metal-free cage (L). In each complex the metal cation occupies the cage's central cavity and is coordinated to six nitrogens from the three bipyridyl groups. The cations [MnL](2+) and [FeL](2+) are isostructural but both exhibit a different arrangement of the bound cage to that observed in the corresponding nickel(II) and copper(II) complexes. The latter have an exo-exo arrangement of the bridgehead nitrogen lone pairs, with the metal inducing a triple helical twist that extends approximately 22 A along the axial length of each complex. In contrast, [MnL](2+) and [FeL](2+) have their terminal nitrogen lone pairs directed endo, causing a significant change in the configuration of the bound ligand. In [FeL'](2+), the cage has both bridgehead nitrogen lone pairs orientated exo. Semiempirical calculations indicate that the observed endo-endo and exo-exo arrangements are of comparable energy.

CONTROL FOR NEUTRONIC REACTOR

DOEpatents

Lichtenberger, H.V.; Cameron, R.A.

1959-03-31

S>A control rod operating device in a nuclear reactor of the type in which the control rod is gradually withdrawn from the reactor to a position desired during stable operation is described. The apparatus is comprised essentially of a stop member movable in the direction of withdrawal of the control rod, a follower on the control rod engageable with the stop and means urging the follower against the stop in the direction of withdrawal. A means responsive to disengagement of the follower from the stop is provided for actuating the control rod to return to the reactor shut-down position.

Spherical torus fusion reactor

DOEpatents

Martin Peng, Y.K.M.

1985-10-03

The object of this invention is to provide a compact torus fusion reactor with dramatic simplification of plasma confinement design. Another object of this invention is to provide a compact torus fusion reactor with low magnetic field and small aspect ratio stable plasma confinement. In accordance with the principles of this invention there is provided a compact toroidal-type plasma confinement fusion reactor in which only the indispensable components inboard of a tokamak type of plasma confinement region, mainly a current conducting medium which carries electrical current for producing a toroidal magnet confinement field about the toroidal plasma region, are retained.

Self-actuating reactor shutdown system

DOEpatents

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

1988-01-01

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

Assessment of Sensor Technologies for Advanced Reactors

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

Korsah, Kofi; Kisner, R. A.; Britton Jr., C. L.

This paper provides an assessment of sensor technologies and a determination of measurement needs for advanced reactors (AdvRx). It is a summary of a study performed to provide the technical basis for identifying and prioritizing research targets within the instrumentation and control (I&C) Technology Area under the Department of Energy’s (DOE’s) Advanced Reactor Technology (ART) program. The study covered two broad reactor technology categories: High Temperature Reactors and Fast Reactors. The scope of “High temperature reactors” included Gen IV reactors whose coolant exit temperatures exceed ≈650 °C and are moderated (as opposed to fast reactors). To bound the scope formore » fast reactors, this report reviewed relevant operating experience from US-operated Sodium Fast Reactor (SFR) and relevant test experience from the Fast Flux Test Facility (FFTF). For high temperature reactors the study showed that in many cases instrumentation have performed reasonably well in research and demonstration reactors. However, even in cases where the technology is “mature” (such as thermocouples), HTGRs can benefit from improved technologies. Current HTGR instrumentation is generally based on decades-old technology and adapting newer technologies could provide significant advantages. For sodium fast reactors, the study found that several key research needs arise around (1) radiation-tolerant sensor design for in-vessel or in-core applications, where possible non-invasive sensing approaches for key parameters that minimize the need to deploy sensors in-vessel, (2) approaches to exfiltrating data from in-vessel sensors while minimizing penetrations, (3) calibration of sensors in-situ, and (4) optimizing sensor placements to maximize the information content while minimizing the number of sensors needed.« less

Fast quench reactor and method

DOEpatents

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

1998-05-12

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

Competitive adsorption of copper(II), cadmium(II), lead(II) and zinc(II) onto basic oxygen furnace slag.

PubMed

Xue, Yongjie; Hou, Haobo; Zhu, Shujing

2009-02-15

Polluted and contaminated water can often contain more than one heavy metal species. It is possible that the behavior of a particular metal species in a solution system will be affected by the presence of other metals. In this study, we have investigated the adsorption of Cd(II), Cu(II), Pb(II), and Zn(II) onto basic oxygen furnace slag (BOF slag) in single- and multi-element solution systems as a function of pH and concentration, in a background solution of 0.01M NaNO(3). In adsorption edge experiments, the pH was varied from 2.0 to 13.0 with total metal concentration 0.84mM in the single element system and 0.21mM each of Cd(II), Cu(II), Pb(II), and Zn(II) in the multi-element system. The value of pH(50) (the pH at which 50% adsorption occurs) was found to follow the sequence Zn>Cu>Pb>Cd in single-element systems, but Pb>Cu>Zn>Cd in the multi-element system. Adsorption isotherms at pH 6.0 in the multi-element systems showed that there is competition among various metals for adsorption sites on BOF slag. The adsorption and potentiometric titrations data for various slag-metal systems were modeled using an extended constant-capacitance surface complexation model that assumed an ion-exchange process below pH 6.5 and the formation of inner-sphere surface complexes at higher pH. Inner-sphere complexation was more dominant for the Cu(II), Pb(II) and Zn(II) systems.

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.

Space Nuclear Reactor Engineering

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

Poston, David Irvin

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

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.

Fossil fuel furnace reactor

DOEpatents

Parkinson, William J.

1987-01-01

A fossil fuel furnace reactor is provided for simulating a continuous processing plant with a batch reactor. An internal reaction vessel contains a batch of shale oil, with the vessel having a relatively thin wall thickness for a heat transfer rate effective to simulate a process temperature history in the selected continuous processing plant. A heater jacket is disposed about the reactor vessel and defines a number of independent controllable temperature zones axially spaced along the reaction vessel. Each temperature zone can be energized to simulate a time-temperature history of process material through the continuous plant. A pressure vessel contains both the heater jacket and the reaction vessel at an operating pressure functionally selected to simulate the continuous processing plant. The process yield from the oil shale may be used as feedback information to software simulating operation of the continuous plant to provide operating parameters, i.e., temperature profiles, ambient atmosphere, operating pressure, material feed rates, etc., for simulation in the batch reactor.

The new neutron radiography/tomography/imaging station DINGO at OPAL

NASA Astrophysics Data System (ADS)

Garbe, U.; Randall, T.; Hughes, C.

2011-09-01

A new neutron imaging instrument will be built to support the area of neutron imaging research (neutron radiography/tomography) at ANSTO. The instrument will be designed for an international user community and for routine quality control for defence, industrial, mining, space and aircraft applications. It will also be a useful tool for assessing oil and water flow in sedimentary rock reservoirs (like the North West Shelf), assessing water damage in aircraft components, and the study of hydrogen distribution and cracking in steel. The instrument is planned to be completed by the end of June 2013 and is currently in the design stage. The usable neutron flux is mainly determined by the neutron source, but it also depends on the instrument position and the resolution. The designated instrument position for DINGO is the beam port HB-2 in the reactor hall. The estimated flux for an L/ D of approximately 250 at HB-2 is calculated by Mcstas simulation in a range of 4.75×10 7 n/cm 2 s, which is in the same range of other facilities like ANSTARES (FRM II; Schillinger et al., 2004 [1]) or BT2 (NIST; Hussey et al., 2005 [2]). A special feature of DINGO is the in-pile collimator place in front of the main shutter at HB-2. The collimator offers two pinholes with a possible L/ D of 250 and 1000. A secondary collimator will separate the two beams and block one. The whole instrument will operate in two different positions, one for high resolution and the other for high speed.

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

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

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

1963-01-01

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

REACTOR UNLOADING MEANS

DOEpatents

Cooper, C.M.

1957-08-20

A means for remotely unloading irradiated fuel slugs from a neutronic reactor core and conveying them to a remote storage tank is reported. The means shown is specifically adapted for use with a reactor core wherein the fuel slugs are slidably held in end to end abutting relationship in the horizontal coolant flow tubes, the slugs being spaced from tae internal walls of the tubes to permit continuous circulation of coolant water therethrough. A remotely operated plunger at the charging ends of the tubes is used to push the slugs through the tubes and out the discharge ends into a special slug valve which transfers the slug to a conveying tube leading into a storage tank. Water under pressure is forced through the conveying tube to circulate around the slug to cool it and also to force the slug through the conveving tube into the storage tank. The slug valve and conveying tube are shielded to prevent amy harmful effects caused by the radioactive slug in its travel from the reactor to the storage tank. With the disclosed apparatus, all the slugs in the reactor core can be conveyed to the storage tank shortly after shutdown by remotely located operating personnel.

NASA Reactor Facility Hazards Summary. Volume 1

NASA Technical Reports Server (NTRS)

1959-01-01

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

Aerosol reactor production of uniform submicron powders

NASA Technical Reports Server (NTRS)

Flagan, Richard C. (Inventor); Wu, Jin J. (Inventor)

1991-01-01

A method of producing submicron nonagglomerated particles in a single stage reactor includes introducing a reactant or mixture of reactants at one end while varying the temperature along the reactor to initiate reactions at a low rate. As homogeneously small numbers of seed particles generated in the initial section of the reactor progress through the reactor, the reaction is gradually accelerated through programmed increases in temperature along the length of the reactor to promote particle growth by chemical vapor deposition while minimizing agglomerate formation by maintaining a sufficiently low number concentration of particles in the reactor such that coagulation is inhibited within the residence time of particles in the reactor. The maximum temperature and minimum residence time is defined by a combination of temperature and residence time that is necessary to bring the reaction to completion. In one embodiment, electronic grade silane and high purity nitrogen are introduced into the reactor and temperatures of approximately 770.degree. K. to 1550.degree. K. are employed. In another embodiment silane and ammonia are employed at temperatures from 750.degree. K. to 1800.degree. K.

Aerosol reactor production of uniform submicron powders

DOEpatents

Flagan, Richard C.; Wu, Jin J.

1991-02-19

A method of producing submicron nonagglomerated particles in a single stage reactor includes introducing a reactant or mixture of reactants at one end while varying the temperature along the reactor to initiate reactions at a low rate. As homogeneously small numbers of seed particles generated in the initial section of the reactor progress through the reactor, the reaction is gradually accelerated through programmed increases in temperature along the length of the reactor to promote particle growth by chemical vapor deposition while minimizing agglomerate formation by maintaining a sufficiently low number concentration of particles in the reactor such that coagulation is inhibited within the residence time of particles in the reactor. The maximum temperature and minimum residence time is defined by a combination of temperature and residence time that is necessary to bring the reaction to completion. In one embodiment, electronic grade silane and high purity nitrogen are introduced into the reactor and temperatures of approximately 770.degree. K. to 1550.degree. K. are employed. In another embodiment silane and ammonia are employed at temperatures from 750.degree. K. to 1800.degree. K.

Strengthening IAEA Safeguards for Research Reactors

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

Reid, Bruce D.; Anzelon, George A.; Budlong-Sylvester, Kory

During their December 10-11, 2013, workshop in Grenoble France, which focused on the history and future of safeguarding research reactors, the United States, France and the United Kingdom (UK) agreed to conduct a joint study exploring ways to strengthen the IAEA’s safeguards approach for declared research reactors. This decision was prompted by concerns about: 1) historical cases of non-compliance involving misuse (including the use of non-nuclear materials for production of neutron generators for weapons) and diversion that were discovered, in many cases, long after the violations took place and as part of broader pattern of undeclared activities in half amore » dozen countries; 2) the fact that, under the Safeguards Criteria, the IAEA inspects some reactors (e.g., those with power levels under 25 MWt) less than once per year; 3) the long-standing precedent of States using heavy water research reactors (HWRR) to produce plutonium for weapons programs; 4) the use of HEU fuel in some research reactors; and 5) various technical characteristics common to some types of research reactors that could provide an opportunity for potential proliferators to misuse the facility or divert material with low probability of detection by the IAEA. In some research reactors it is difficult to detect diversion or undeclared irradiation. In addition, infrastructure associated with research reactors could pose a safeguards challenge. To strengthen the effectiveness of safeguards at the State level, this paper advocates that the IAEA consider ways to focus additional attention and broaden its safeguards toolbox for research reactors. This increase in focus on the research reactors could begin with the recognition that the research reactor (of any size) could be a common path element on a large number of technically plausible pathways that must be considered when performing acquisition pathway analysis (APA) for developing a State Level Approach (SLA) and Annual Implementation Plan

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

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

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

2016-08-31

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

Approach to developing reliable space reactor power systems

NASA Technical Reports Server (NTRS)

Mondt, Jack F.; Shinbrot, Charles H.

1991-01-01

During Phase II, the Engineering Development Phase, the SP-100 Project has defined and is pursuing a new approach to developing reliable power systems. The approach to developing such a system during the early technology phase is described along with some preliminary examples to help explain the approach. Developing reliable components to meet space reactor power system requirements is based on a top-down systems approach which includes a point design based on a detailed technical specification of a 100-kW power system. The SP-100 system requirements implicitly recognize the challenge of achieving a high system reliability for a ten-year lifetime, while at the same time using technologies that require very significant development efforts. A low-cost method for assessing reliability, based on an understanding of fundamental failure mechanisms and design margins for specific failure mechanisms, is being developed as part of the SP-100 Program.

Cooling system for a nuclear reactor

DOEpatents

Amtmann, Hans H.

1982-01-01

A cooling system for a gas-cooled nuclear reactor is disclosed which includes at least one primary cooling loop adapted to pass coolant gas from the reactor core and an associated steam generator through a duct system having a main circulator therein, and at least one auxiliary cooling loop having communication with the reactor core and adapted to selectively pass coolant gas through an auxiliary heat exchanger and circulator. The main and auxiliary circulators are installed in a common vertical cavity in the reactor vessel, and a common return duct communicates with the reactor core and intersects the common cavity at a junction at which is located a flow diverter valve operative to effect coolant flow through either the primary or auxiliary cooling loops.

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.

KINETICS OF TREAT USED AS A TEST REACTOR

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

Dickerman, C.E.; Johnson, R.D.; Gasidlo, J.

1962-05-01

An analysis is presented concerning the reactor kinetics of TREAT used as a pulsed, engineering test reactor for fast reactor fuel element studies. A description of the reactor performance is given for a wide range of conditions associated with its use as a test reactor. Supplemental information on meltdown experimentation is included. (J.R.D.)

Nuclear reactor vessel fuel thermal insulating barrier

DOEpatents

Keegan, C. Patrick; Scobel, James H.; Wright, Richard F.

2013-03-19

The reactor vessel of a nuclear reactor installation which is suspended from the cold leg nozzles in a reactor cavity is provided with a lower thermal insulating barrier spaced from the reactor vessel that has a hemispherical lower section that increases in volume from the center line of the reactor to the outer extent of the diameter of the thermal insulating barrier and smoothly transitions up the side walls of the vessel. The space between the thermal insulating harrier and the reactor vessel forms a chamber which can be flooded with cooling water through passive valving to directly cool the reactor vessel in the event of a severe accident. The passive inlet valve for the cooling water includes a buoyant door that is normally maintained sealed under its own weight and floats open when the cavity is Hooded. Passively opening steam vents are also provided.

Acceptability of reactors in space

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

Buden, D.

1981-04-01

Reactors are the key to our future expansion into space. However, there has been some confusion in the public as to whether they are a safe and acceptable technology for use in space. The answer to these questions is explored. The US position is that when reactors are the preferred technical choice, that they can be used safely. In fact, it dies not appear that reactors add measurably to the risk associated with the Space Transportation System.

REACTOR AND NOVEL METHOD

DOEpatents

Young, G.J.; Ohlinger, L.A.

1958-06-24

A nuclear reactor of the type which uses a liquid fuel and a method of controlling such a reactor are described. The reactor is comprised essentially of a tank for containing the liquid fuel such as a slurry of discrete particles of fissionnble material suspended in a heavy water moderator, and a control means in the form of a disc of neutron absorbirg material disposed below the top surface of the slurry and parallel thereto. The diameter of the disc is slightly smaller than the diameter of the tank and the disc is perforated to permit a flow of the slurry therethrough. The function of the disc is to divide the body of slurry into two separate portions, the lower portion being of a critical size to sustain a nuclear chain reaction and the upper portion between the top surface of the slurry and the top surface of the disc being of a non-critical size. The method of operation is to raise the disc in the reactor until the lower portion of the slurry has reached a critical size when it is desired to initiate the reaction, and to lower the disc in the reactor to reduce the size of the lower active portion the slurry to below criticality when it is desired to stop the reaction.

Entropy Production in Chemical Reactors

NASA Astrophysics Data System (ADS)

Kingston, Diego; Razzitte, Adrián C.

2017-06-01

We have analyzed entropy production in chemically reacting systems and extended previous results to the two limiting cases of ideal reactors, namely continuous stirred tank reactor (CSTR) and plug flow reactor (PFR). We have found upper and lower bounds for the entropy production in isothermal systems and given expressions for non-isothermal operation and analyzed the influence of pressure and temperature in entropy generation minimization in reactors with a fixed volume and production. We also give a graphical picture of entropy production in chemical reactions subject to constant volume, which allows us to easily assess different options. We show that by dividing a reactor into two smaller ones, operating at different temperatures, the entropy production is lowered, going as near as 48 % less in the case of a CSTR and PFR in series, and reaching 58 % with two CSTR. Finally, we study the optimal pressure and temperature for a single isothermal PFR, taking into account the irreversibility introduced by a compressor and a heat exchanger, decreasing the entropy generation by as much as 30 %.

SNAP 10A FS-3 reactor performance

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

Hawley, J.P.; Johnson, R.A.

1966-08-15

SNAP 10FS-3 was the first flight-qualified SNAP reactor system to be operated in a simulated space environment. Prestart-up qualification testing, automatic start-up, endurance period performance, extended operation test and reactor shutdown are described as they affected, or were affected by, overall reactor performance. Performance of the reactor control system and the diagnostic instrumentation is critically evaluted.

Nuclear reactor cavity floor passive heat removal system

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

Edwards, Tyler A.; Neeley, Gary W.; Inman, James B.

A nuclear reactor includes a reactor core disposed in a reactor pressure vessel. A radiological containment contains the nuclear reactor and includes a concrete floor located underneath the nuclear reactor. An ex vessel corium retention system includes flow channels embedded in the concrete floor located underneath the nuclear reactor, an inlet in fluid communication with first ends of the flow channels, and an outlet in fluid communication with second ends of the flow channels. In some embodiments the inlet is in fluid communication with the interior of the radiological containment at a first elevation and the outlet is in fluidmore » communication with the interior of the radiological containment at a second elevation higher than the first elevation. The radiological containment may include a reactor cavity containing a lower portion of the pressure vessel, wherein the concrete floor located underneath the nuclear reactor is the reactor cavity floor.« less

Using thermal balance model to determine optimal reactor volume and insulation material needed in a laboratory-scale composting reactor.

PubMed

Wang, Yongjiang; Pang, Li; Liu, Xinyu; Wang, Yuansheng; Zhou, Kexun; Luo, Fei

2016-04-01

A comprehensive model of thermal balance and degradation kinetics was developed to determine the optimal reactor volume and insulation material. Biological heat production and five channels of heat loss were considered in the thermal balance model for a representative reactor. Degradation kinetics was developed to make the model applicable to different types of substrates. Simulation of the model showed that the internal energy accumulation of compost was the significant heat loss channel, following by heat loss through reactor wall, and latent heat of water evaporation. Lower proportion of heat loss occurred through the reactor wall when the reactor volume was larger. Insulating materials with low densities and low conductive coefficients were more desirable for building small reactor systems. Model developed could be used to determine the optimal reactor volume and insulation material needed before the fabrication of a lab-scale composting system. Copyright © 2016 Elsevier Ltd. All rights reserved.

The Simulator Development for RDE Reactor

NASA Astrophysics Data System (ADS)

Subekti, Muhammad; Bakhri, Syaiful; Sunaryo, Geni Rina

2018-02-01

BATAN is proposing the construction of experimental power reactor (RDE reactor) for increasing the public acceptance on NPP development plan, proofing the safety level of the most advanced reactor by performing safety demonstration on the accidents such as Chernobyl and Fukushima, and owning the generation fourth (G4) reactor technology. For owning the reactor technology, the one of research activities is RDE’s simulator development that employing standard equation. The development utilizes standard point kinetic and thermal equation. The examination of the simulator carried out comparison in which the simulation’s calculation result has good agreement with assumed parameters and ChemCAD calculation results. The transient simulation describes the characteristic of the simulator to respond the variation of power increase of 1.5%/min, 2.5%/min, and 3.5%/min.

NEUTRONIC REACTOR

DOEpatents

Wigner, E.P.; Weinberg, A.W.; Young, G.J.

1958-04-15

A nuclear reactor which uses uranium in the form of elongated tubes as fuel elements and liquid as a coolant is described. Elongated tubular uranium bodies are vertically disposed in an efficient neutron slowing agent, such as graphite, for example, to form a lattice structure which is disposed between upper and lower coolant tanks. Fluid coolant tubes extend through the uranium bodies and communicate with the upper and lower tanks and serve to convey the coolant through the uranium body. The reactor is also provided with means for circulating the cooling fluid through the coolant tanks and coolant tubes, suitable neutron and gnmma ray shields, and control means.

Safety control circuit for a neutronic reactor

DOEpatents

Ellsworth, Howard C.

2004-04-27

A neutronic reactor comprising an active portion containing material fissionable by neutrons of thermal energy, means to control a neutronic chain reaction within the reactor comprising a safety device and a regulating device, a safety device including means defining a vertical channel extending into the reactor from an aperture in the upper surface of the reactor, a rod containing neutron-absorbing materials slidably disposed within the channel, means for maintaining the safety rod in a withdrawn position relative to the active portion of the reactor including means for releasing said rod on actuation thereof, a hopper mounted above the active portion of the reactor having a door disposed at the bottom of the hopper opening into the vertical channel, a plurality of bodies of neutron-absorbing materials disposed within the hopper, and means responsive to the failure of the safety rod on actuation thereof to enter the active portion of the reactor for opening the door in the hopper.

Nuclear reactor fuel containment safety structure

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

Rosewell, M.P.

A nuclear reactor fuel containment safety structure is disclosed and is shown to include an atomic reactor fuel shield with a fuel containment chamber and exhaust passage means, and a deactivating containment base attached beneath the fuel reactor shield and having exhaust passages, manifold, and fluxing and control material and vessels. 1 claim, 8 figures.

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

Application of Reactor Antineutrinos: Neutrinos for Peace

NASA Astrophysics Data System (ADS)

Suekane, F.

2013-02-01

In nuclear reactors, 239Pu are produced along with burn-up of nuclear fuel. 239Pu is subject of safeguard controls since it is an explosive component of nuclear weapon. International Atomic Energy Agency (IAEA) is watching undeclared operation of reactors to prevent illegal production and removal of 239Pu. In operating reactors, a huge numbers of anti electron neutrinos (ν) are produced. Neutrino flux is approximately proportional to the operating power of reactor in short term and long term decrease of the neutrino flux per thermal power is proportional to the amount of 239Pu produced. Thus rector ν's carry direct and real time information useful for the safeguard purposes. Since ν can not be hidden, it could be an ideal medium to monitor the reactor operation. IAEA seeks for novel technologies which enhance their ability and reactor neutrino monitoring is listed as one of such candidates. Currently neutrino physicists are performing R&D of small reactor neutrino detectors to use specifically for the safeguard use in response to the IAEA interest. In this proceedings of the neutrino2012 conference, possibilities of such reactor neutrinos application and current world-wide R&D status are described.

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

Dynamics of heat-pipe reactors

NASA Technical Reports Server (NTRS)

Niederauer, G. F.

1971-01-01

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

NEUTRONIC REACTOR BURIAL ASSEMBLY

DOEpatents

Treshow, M.

1961-05-01

A burial assembly is shown whereby an entire reactor core may be encased with lead shielding, withdrawn from the reactor site and buried. This is made possible by a five-piece interlocking arrangement that may be easily put together by remote control with no aligning of bolt holes or other such close adjustments being necessary.

Reactor core isolation cooling system

DOEpatents

Cooke, F.E.

1992-12-08

A reactor core isolation cooling system includes a reactor pressure vessel containing a reactor core, a drywell vessel, a containment vessel, and an isolation pool containing an isolation condenser. A turbine is operatively joined to the pressure vessel outlet steamline and powers a pump operatively joined to the pressure vessel feedwater line. In operation, steam from the pressure vessel powers the turbine which in turn powers the pump to pump makeup water from a pool to the feedwater line into the pressure vessel for maintaining water level over the reactor core. Steam discharged from the turbine is channeled to the isolation condenser and is condensed therein. The resulting heat is discharged into the isolation pool and vented to the atmosphere outside the containment vessel for removing heat therefrom. 1 figure.

Reactor core isolation cooling system

DOEpatents

Cooke, Franklin E.

1992-01-01

A reactor core isolation cooling system includes a reactor pressure vessel containing a reactor core, a drywell vessel, a containment vessel, and an isolation pool containing an isolation condenser. A turbine is operatively joined to the pressure vessel outlet steamline and powers a pump operatively joined to the pressure vessel feedwater line. In operation, steam from the pressure vessel powers the turbine which in turn powers the pump to pump makeup water from a pool to the feedwater line into the pressure vessel for maintaining water level over the reactor core. Steam discharged from the turbine is channeled to the isolation condenser and is condensed therein. The resulting heat is discharged into the isolation pool and vented to the atmosphere outside the containment vessel for removing heat therefrom.

Nuclear Reactors and Technology

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

Cason, D.L.; Hicks, S.C.

1992-01-01

This publication Nuclear Reactors and Technology (NRT) announces on a monthly basis the current worldwide information available from the open literature on nuclear reactors and technology, including all aspects of power reactors, components and accessories, fuel elements, control systems, and materials. This publication contains the abstracts of DOE reports, journal articles, conference papers, patents, theses, and monographs added to the Energy Science and Technology Database during the past month. Also included are US information obtained through acquisition programs or interagency agreements and international information obtained through the International Energy Agency`s Energy Technology Data Exchange or government-to-government agreements. The digests inmore » NRT and other citations to information on nuclear reactors back to 1948 are available for online searching and retrieval on the Energy Science and Technology Database and Nuclear Science Abstracts (NSA) database. Current information, added daily to the Energy Science and Technology Database, is available to DOE and its contractors through the DOE Integrated Technical Information System. Customized profiles can be developed to provide current information to meet each user`s needs.« less

Magnetic Flux Compression Reactor Concepts for Spacecraft Propulsion and Power (MSFC Center Director's Discretionary Fund; Project No. 99-24). Part 1

NASA Technical Reports Server (NTRS)

Litchford, R. J.; Robertson, G. A.; Hawk, C. W.; Turner, M. W.; Koelfgen, S.; Litchford, Ron J. (Technical Monitor)

2001-01-01

This technical publication (TP) examines performance and design issues associated with magnetic flux compression reactor concepts for nuclear/chemical pulse propulsion and power. Assuming that low-yield microfusion detonations or chemical detonations using high-energy density matter can eventually be realized in practice, various magnetic flux compression concepts are conceivable. In particular, reactors in which a magnetic field would be compressed between an expanding detonation-driven plasma cloud and a stationary structure formed from a high-temperature superconductor are envisioned. Primary interest is accomplishing two important functions: (1) Collimation and reflection of a hot diamagnetic plasma for direct thrust production, and (2) electric power generation for fusion standoff drivers and/or dense plasma formation. In this TP, performance potential is examined, major technical uncertainties related to this concept accessed, and a simple performance model for a radial-mode reactor developed. Flux trapping effectiveness is analyzed using a skin layer methodology, which accounts for magnetic diffusion losses into the plasma armature and the stationary stator. The results of laboratory-scale experiments on magnetic diffusion in bulk-processed type II superconductors are also presented.

``Sleeping reactor`` irradiations: Shutdown reactor determination of short-lived activation products

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

Jerde, E.A.; Glasgow, D.C.

1998-09-01

At the High-Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory, the principal irradiation system has a thermal neutron flux ({phi}) of {approximately} 4 {times} 10{sup 14} n/cm{sup 2} {center_dot} s, permitting the detection of elements via irradiation of 60 s or less. Irradiations of 6 or 7 s are acceptable for detection of elements with half-lives of as little as 30 min. However, important elements such as Al, Mg, Ti, and V have half-lives of only a few minutes. At HFIR, these can be determined with irradiation times of {approximately} 6 s, but the requirement of immediate countingmore » leads to increased exposure to the high activity produced by irradiation in the high flux. In addition, pneumatic system timing uncertainties (about {+-} 0.5 s) make irradiations of < 6 s less reliable. Therefore, the determination of these ultra-short-lived species in mixed matrices has not generally been made at HFIR. The authors have found that very short lived activation products can be produced easily during the period after reactor shutdown (SCRAM), but prior to the removal of spent fuel elements. During this 24- to 36-h period (dubbed the ``sleeping reactor``), neutrons are produced in the beryllium reflector by the reaction {sup 9}Be({gamma},n){sup 8}Be, the gamma rays principally originating in the spent fuel. Upon reactor SCRAM, the flux drops to {approximately} 1 {times} 10{sup 10} n/cm{sup 2} {center_dot} s within 1 h. By the time the fuel elements are removed, the flux has dropped to {approximately} 6 {times} 10{sup 8}. Such fluxes are ideal for the determination of short-lived elements such as Al, Ti, Mg, and V. An important feature of the sleeping reactor is a flux that is not constant.« less

DENSITY CONTROL IN A REACTOR

DOEpatents

Marshall, J. Jr.

1961-10-24

A reactor is described in which natural-uranium bodies are located in parallel channels which extend through the graphite mass in a regular lattice. The graphite mass has additional channels that are out of the lattice and contain no uranium. These additional channels decrease in number per unit volume of graphite from the center of the reactor to the exterior and have the effect of reducing the density of the graphite more at the center than at the exterior, thereby spreading neutron activity throughout the reactor. (AEC)

High-rate treatment of molasses wastewater by combination of an acidification reactor and a USSB reactor.

PubMed

Onodera, Takashi; Sase, Shinya; Choeisai, Pairaya; Yoochatchaval, Wilasinee; Sumino, Haruhiko; Yamaguchi, Takashi; Ebie, Yoshitaka; Xu, Kaiqin; Tomioka, Noriko; Syutsubo, Kazuaki

2011-01-01

A combination of an acidification reactor and an up-flow staged sludge bed (USSB) reactor was applied for treatment of molasses wastewater containing a large amount of organic compounds and sulfate. The USSB reactor had three gas-solid separators (GSS) along the height of the reactor. The combined system was continuously operated at mesophilic temperature over 400 days. In the acidification reactor, acid formation and sulfate reduction were effectively carried out. The sugars contained in the influent wastewater were mostly acidified into acetate, propionate, and n-butyrate. In addition, 10-30% of influent sulfur was removed from the acidification reactor by means of sulfate reduction followed by stripping of hydrogen sulfide. The USSB achieved a high organic loading rate (OLR) of 30 kgCOD m(-3) day(-1) with 82% COD removal. Vigorous biogas production was observed at a rate of 15 Nm(3) biogas m(-3) reactor day(-1). The produced biogas, including hydrogen sulfide, was removed from the wastewater mostly via the GSS. The GSS provided a moderate superficial biogas flux and low sulfide concentration in the sludge bed, resulting in the prevention of sludge washout and sulfide inhibition of methanogens. By advantages of this feature, the USSB may have been responsible for achieving sufficient retention (approximately 60 gVSS L(-1)) of the granular sludge with high methanogenic activity (0.88 gCOD gVSS(-1) day(-1) for acetate and as high as 2.6 gCOD gVSS(-1) day(-1) for H(2)/CO(2)). Analysis of the microbial community revealed that sugar-degrading acid-forming bacteria proliferated in the sludge of the USSB as well as the acidification reactor at high OLR conditions.

Review of Transient Testing of Fast Reactor Fuels in the Transient REActor Test Facility (TREAT)

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

Jensen, C.; Wachs, D.; Carmack, J.

The restart of the Transient REActor Test (TREAT) facility provides a unique opportunity to engage the fast reactor fuels community to reinitiate in-pile experimental safety studies. Historically, the TREAT facility played a critical role in characterizing the behavior of both metal and oxide fast reactor fuels under off-normal conditions, irradiating hundreds of fuel pins to support fast reactor fuel development programs. The resulting test data has provided validation for a multitude of fuel performance and severe accident analysis computer codes. This paper will provide a review of the historical database of TREAT experiments including experiment design, instrumentation, test objectives, andmore » salient findings. Additionally, the paper will provide an introduction to the current and future experiment plans of the U.S. transient testing program at TREAT.« less

Thermal-hydraulic interfacing code modules for CANDU reactors

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

Liu, W.S.; Gold, M.; Sills, H.

1997-07-01

The approach for CANDU reactor safety analysis in Ontario Hydro Nuclear (OHN) and Atomic Energy of Canada Limited (AECL) is presented. Reflecting the unique characteristics of CANDU reactors, the procedure of coupling the thermal-hydraulics, reactor physics and fuel channel/element codes in the safety analysis is described. The experience generated in the Canadian nuclear industry may be useful to other types of reactors in the areas of reactor safety analysis.

Characteristics and Dose Levels for Spent Reactor Fuels

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

Coates, Cameron W

2007-01-01

Current guidance considers highly radioactive special nuclear materials to be those materials that, unshielded, emit a radiation dose [rate] measured at 1 m which exceeds 100 rem/h. Smaller, less massive fuel assemblies from research reactors can present a challenge from the point of view of self protection because of their size (lower dose, easier to handle) and the desirability of higher enrichments; however, a follow-on study to cross-compare dose trends of research reactors and power reactors was deemed useful to confirm/verify these trends. This paper summarizes the characteristics and dose levels of spent reactor fuels for both research reactors andmore » power reactors and extends previous studies aimed at quantifying expected dose rates from research reactor fuels worldwide.« less

Dynamic bed reactor

DOEpatents

Stormo, Keith E.

1996-07-02

A dynamic bed reactor is disclosed in which a compressible open cell foam matrix is periodically compressed and expanded to move a liquid or fluid through the matrix. In preferred embodiments, the matrix contains an active material such as an enzyme, biological cell, chelating agent, oligonucleotide, adsorbent or other material that acts upon the liquid or fluid passing through the matrix. The active material may be physically immobilized in the matrix, or attached by covalent or ionic bonds. Microbeads, substantially all of which have diameters less than 50 microns, can be used to immobilize the active material in the matrix and further improve reactor efficiency. A particularly preferred matrix is made of open cell polyurethane foam, which adsorbs pollutants such as polychlorophenol or o-nitrophenol. The reactors of the present invention allow unidirectional non-laminar flow through the matrix, and promote intimate exposure of liquid reactants to active agents such as microorganisms immobilized in the matrix.

BOILER-SUPERHEATED REACTOR

DOEpatents

Heckman, T.P.

1961-05-01

A nuclear power reactor of the type in which a liquid moderator-coolant is transformed by nuclear heating into a vapor that may be used to drive a turbo- generator is described. The core of this reactor comprises a plurality of freely suspended tubular fuel elements, called fuel element trains, within which nonboiling pressurized liquid moderator-coolant is preheated and sprayed through orifices in the walls of the trains against the outer walls thereof to be converted into vapor. Passage of the vapor ovcr other unwetted portions of the outside of the fuel elements causes the steam to be superheated. The moderatorcoolant within the fuel elements remains in the liqUid state, and that between the fuel elements remains substantiaily in the vapor state. A unique liquid neutron-absorber control system is used. Advantages expected from the reactor design include reduced fuel element failure, increased stability of operation, direct response to power demand, and circulation of a minimum amount of liquid moderatorcoolant. (A.G.W.)

Partitionable-space enhanced coagulation (PEC) reactor and its working mechanism: a new prospective chemical technology for phosphorus pollution control.

PubMed

Zhang, Meng; Zheng, Ping; Abbas, Ghulam; Chen, Xiaoguang

2014-02-01

Phosphorus pollution control and phosphorus recycling, simultaneously, are focus of attention in the wastewater treatment. In this work, a novel reactor named partitionable-space enhanced coagulation (PEC) was invented for phosphorus control. The working performance and process mechanism of PEC reactor were investigated. The results showed that the PEC technology was highly efficient and cost-effective. The volumetric removal rate (VRR) reached up to 2.86 ± 0.04 kg P/(m(3) d) with a phosphorus removal rate of over 97%. The precipitant consumption was reduced to 2.60-2.76 kg Fe(II)/kg P with low operational cost of $ 0.632-0.673/kg P. The peak phosphorus content in precipitate was up to 30.44% by P2O5, which reveal the benefit of the recycling phosphorus resource. The excellent performance of PEC technology was mainly attributed to the partitionable-space and 'flocculation filter'. The partition limited the trans-regional back-mixing of reagents along the reactor, which promoted the precipitation reaction. The 'flocculation filter' retained the microflocs, enhancing the flocculation process. Copyright © 2013 Elsevier Ltd. All rights reserved.

Determining Pu-239 content by resonance transmission analysis using a filtered reactor beam.

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

Klann, R. T.

A novel technique has been developed at Argonne National Laboratory to determine the {sup 239}Pu content in EBR-II blanket elements using resonance transmission analysis (RTA) with a filtered reactor beam. The technique uses cadmium and gadolinium filters along with a {sup 239}Pu fission chamber to isolate the 0.3 eV resonance in {sup 239}Pu. In the energy range from 0.1 to 0.5 eV, the total microscopic cross-section of {sup 239}Pu is significantly larger than the cross-sections of {sup 238}U and {sup 235}U. This large difference in cross-section allows small amounts of {sup 239}Pu to be detected in uranium samples. Tests usingmore » a direct beam from a 250 kW TRIGA reactor have been performed with stacks of depleted uranium and {sup 239}Pu foils. Preliminary measurement results are in good agreement with the predicted results up to about two weight percent of {sup 239}Pu in the sample. In addition, measured {sup 239}Pu masses were in agreement with actual sample masses with uncertainties less than 3.8 percent.« less

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

Rapid starting methanol reactor system

DOEpatents

Chludzinski, Paul J.; Dantowitz, Philip; McElroy, James F.

1984-01-01

The invention relates to a methanol-to-hydrogen cracking reactor for use with a fuel cell vehicular power plant. The system is particularly designed for rapid start-up of the catalytic methanol cracking reactor after an extended shut-down period, i.e., after the vehicular fuel cell power plant has been inoperative overnight. Rapid system start-up is accomplished by a combination of direct and indirect heating of the cracking catalyst. Initially, liquid methanol is burned with a stoichiometric or slightly lean air mixture in the combustion chamber of the reactor assembly. The hot combustion gas travels down a flue gas chamber in heat exchange relationship with the catalytic cracking chamber transferring heat across the catalyst chamber wall to heat the catalyst indirectly. The combustion gas is then diverted back through the catalyst bed to heat the catalyst pellets directly. When the cracking reactor temperature reaches operating temperature, methanol combustion is stopped and a hot gas valve is switched to route the flue gas overboard, with methanol being fed directly to the catalytic cracking reactor. Thereafter, the burner operates on excess hydrogen from the fuel cells.

Dechlorination of trichloroethylene formed from 1,1,2,2-tetrachloroethane by dehydrochlorination in Portland cement slurry including Fe(II).

PubMed

Jung, Bahngmi; Batchelor, Bill

2008-03-01

Transformation of 1,1,2,2-tetrachloroethane (1,1,2,2-TeCA) by Fe(II) in 10% cement slurries was characterized using a batch reactor system. 1,1,2,2-TeCA was completely converted to trichloroethylene (TCE) within 1h in all experiments, even in controls with cement that did not include Fe(II). Therefore, complete degradation of 1,1,2,2-TeCA depends on the behavior of TCE. The half-life of TCE was observed to be 15d when concentrations of Fe(II) and 1,1,2,2-TeCA were 98mM and 0.245mM, respectively. The kinetics of TCE removal was observed to be dependent on Fe(II) dose, pH and initial substrate concentration. Pseudo-first-order rate constants linearly increased with Fe(II) dose up to 198mM when initial target concentration was 0.245mM. Pseudo-first-order kinetics generally described the degradation reactions of TCE at a specific initial concentration, but a modified Langmuir-Hinshelwood model was necessary to describe the degradation kinetics of TCE over a wide range of initial concentrations. A surface reaction of TCE on active solids, which were formed from Fe(II) and products of cement hydration appears to control observed TCE degradation kinetics.

Combustion synthesis continuous flow reactor

DOEpatents

Maupin, G.D.; Chick, L.A.; Kurosky, R.P.

1998-01-06

The present invention is a reactor for combustion synthesis of inorganic powders. The reactor includes a reaction vessel having a length and a first end and a second end. The reaction vessel further has a solution inlet and a carrier gas inlet. The reactor further has a heater for heating both the solution and the carrier gas. In a preferred embodiment, the reaction vessel is heated and the solution is in contact with the heated reaction vessel. It is further preferred that the reaction vessel be cylindrical and that the carrier gas is introduced tangentially into the reaction vessel so that the solution flows helically along the interior wall of the reaction vessel. As the solution evaporates and combustion produces inorganic material powder, the carrier gas entrains the powder and carries it out of the reactor. 10 figs.

Combustion synthesis continuous flow reactor

DOEpatents

Maupin, Gary D.; Chick, Lawrence A.; Kurosky, Randal P.

1998-01-01

The present invention is a reactor for combustion synthesis of inorganic powders. The reactor includes a reaction vessel having a length and a first end and a second end. The reaction vessel further has a solution inlet and a carrier gas inlet. The reactor further has a heater for heating both the solution and the carrier gas. In a preferred embodiment, the reaction vessel is heated and the solution is in contact with the heated reaction vessel. It is further preferred that the reaction vessel be cylindrical and that the carrier gas is introduced tangentially into the reaction vessel so that the solution flows helically along the interior wall of the reaction vessel. As the solution evaporates and combustion produces inorganic material powder, the carrier gas entrains the powder and carries it out of the reactor.

Cyber security evaluation of II&C technologies

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

Thomas, Ken

The Light Water Reactor Sustainability (LWRS) Program is a research and development program sponsored by the Department of Energy, which is conducted in close collaboration with industry to provide the technical foundations for licensing and managing the long-term, safe and economical operation of current nuclear power plants The LWRS Program serves to help the US nuclear industry adopt new technologies and engineering solutions that facilitate the continued safe operation of the plants and extension of the current operating licenses. Within the LWRS Program, the Advanced Instrumentation, Information, and Control (II&C) Systems Technologies Pathway conducts targeted research and development (R&D) tomore » address aging and reliability concerns with the legacy instrumentation and control and related information systems of the U.S. operating light water reactor (LWR) fleet. The II&C Pathway is conducted by Idaho National Laboratory (INL). Cyber security is a common concern among nuclear utilities and other nuclear industry stakeholders regarding the digital technologies that are being developed under this program. This concern extends to the point of calling into question whether these types of technologies could ever be deployed in nuclear plants given the possibility that the information in them can be compromised and the technologies themselves can potentially be exploited to serve as attack vectors for adversaries. To this end, a cyber security evaluation has been conducted of these technologies to determine whether they constitute a threat beyond what the nuclear plants already manage within their regulatory-required cyber security programs. Specifically, the evaluation is based on NEI 08-09, which is the industry’s template for cyber security programs and evaluations, accepted by the Nuclear Regulatory Commission (NRC) as responsive to the requirements of the nuclear power plant cyber security regulation found in 10 CFR 73.54. The evaluation was conducted

NEUTRONIC REACTOR

DOEpatents

Wade, E.J.

1958-09-16

This patent relates to a reflector means for a neutronic reactor. A reflector comprised of a plurality of vertically movable beryllium control members is provided surrounding the sides of the reactor core. An absorber of fast neutrons comprised of natural uramum surrounds the reflector. An absorber of slow neutrons surrounds the absorber of fast neutrons and is formed of a plurality of beryllium blocks having natural uranium members distributcd therethrough. in addition, a movable body is positioned directly below the core and is comprised of a beryllium reflector and an absorbing member attached to the botiom thereof, the absorbing member containing a substance selected from the goup consisting of natural urantum and Th/sup 232/.

Fuel Fabrication and Nuclear Reactors

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

Karpius, Peter Joseph

2017-02-02

The uranium from the enrichment plant is still in the form of UF 6. UF 6 is not suitable for use in a reactor due to its highly corrosive chemistry as well as its phase diagram. UF 6 is converted into UO 2 fuel pellets, which are in turn placed in fuel rods and assemblies. Reactor designs are variable in moderators, coolants, fuel, performance etc.The dream of energy ‘too-cheap to meter’ is no more, and now the nuclear power industry is pushing ahead with advanced reactor designs.

A magnetostructural study of linear NiII MnIII NiII, NiII CrIII NiII and triangular Ni(II)3 species containing (pyridine-2-aldoximato)nickel(II) unit as a building block.

PubMed

Weyhermüller, Thomas; Wagner, Rita; Khanra, Sumit; Chaudhuri, Phalguni

2005-08-07

Three trinuclear complexes, NiII MnIII NiII, NiII CrIII NiII and Ni(II)3 based on (pyridine-2-aldoximato)nickel(II) units are described. Two of them, and , contain metal-centers in linear arrangement, as is revealed by X-ray diffraction. Complex is a homonuclear complex in which the three nickel(II) centers are disposed in a triangular fashion. The compounds were characterized by various physical methods including cyclic voltammetric and variable-temperature (2-290 K) susceptibility measurements. Complexes and display antiferromagnetic exchange coupling of the neighbouring metal centers, while weak ferromagnetic spin exchange between the adjacent Ni II and Cr III ions in is observed. The experimental magnetic data were simulated by using appropriate models.

TREAT Reactor Control and Protection System

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

Lipinski, W.C.; Brookshier, W.K.; Burrows, D.R.

1985-01-01

The main control algorithm of the Transient Reactor Test Facility (TREAT) Automatic Reactor Control System (ARCS) resides in Read Only Memory (ROM) and only experiment specific parameters are input via keyboard entry. Prior to executing an experiment, the software and hardware of the control computer is tested by a closed loop real-time simulation. Two computers with parallel processing are used for the reactor simulation and another computer is used for simulation of the control rod system. A monitor computer, used as a redundant diverse reactor protection channel, uses more conservative setpoints and reduces challenges to the Reactor Trip System (RTS).more » The RTS consists of triplicated hardwired channels with one out of three logic. The RTS is automatically tested by a digital Dedicated Microprocessor Tester (DMT) prior to the execution of an experiment. 6 refs., 5 figs., 1 tab.« less

REFLECTOR FOR NEUTRONIC REACTORS

DOEpatents

Fraas, A.P.

1963-08-01

A reflector for nuclear reactors that comprises an assembly of closely packed graphite rods disposed with their major axes substantially perpendicular to the interface between the reactor core and the reflector is described. Each graphite rod is round in transverse cross section at (at least) its interface end and is provided, at that end, with a coaxial, inwardly tapering hole. (AEC)

Average [O II] nebular emission associated with Mg II absorbers: dependence on Fe II absorption

NASA Astrophysics Data System (ADS)

Joshi, Ravi; Srianand, Raghunathan; Petitjean, Patrick; Noterdaeme, Pasquier

2018-05-01

We investigate the effect of Fe II equivalent width (W2600) and fibre size on the average luminosity of [O II] λλ3727, 3729 nebular emission associated with Mg II absorbers (at 0.55 ≤ z ≤ 1.3) in the composite spectra of quasars obtained with 3 and 2 arcsec fibres in the Sloan Digital Sky Survey. We confirm the presence of strong correlations between [O II] luminosity (L_{[O II]}) and equivalent width (W2796) and redshift of Mg II absorbers. However, we show L_{[O II]} and average luminosity surface density suffer from fibre size effects. More importantly, for a given fibre size, the average L_{[O II]} strongly depends on the equivalent width of Fe II absorption lines and found to be higher for Mg II absorbers with R ≡W2600/W2796 ≥ 0.5. In fact, we show the observed strong correlations of L_{[O II]} with W2796 and z of Mg II absorbers are mainly driven by such systems. Direct [O II] detections also confirm the link between L_{[O II]} and R. Therefore, one has to pay attention to the fibre losses and dependence of redshift evolution of Mg II absorbers on W2600 before using them as a luminosity unbiased probe of global star formation rate density. We show that the [O II] nebular emission detected in the stacked spectrum is not dominated by few direct detections (i.e. detections ≥3σ significant level). On an average, the systems with R ≥ 0.5 and W2796 ≥ 2 Å are more reddened, showing colour excess E(B - V) ˜ 0.02, with respect to the systems with R < 0.5 and most likely trace the high H I column density systems.

MERCHANT MARINE SHIP REACTOR

DOEpatents

Mumm, J.F.; North, D.C. Jr.; Rock, H.R.; Geston, D.K.

1961-05-01

A nuclear reactor is described for use in a merchant marine ship. The reactor is of pressurized light water cooled and moderated design in which three passes of the water through the core in successive regions of low, intermediate, and high heat generation and downflow in a fuel region are made. The foregoing design makes a compact reactor construction with extended core life. The core has an egg-crate lattice containing the fuel elements confined between a lower flow baffle and upper grid plate, with the latter serving also as part of a turn- around manifold from which the entire coolant is distributed into the outer fuel elements for the second pass through the core. The inner fuel elements are cooled in the third pass.

Merchant Marine Ship Reactor

DOEpatents

Sankovich, M. F.; Mumm, J. F.; North, Jr, D. C.; Rock, H. R.; Gestson, D. K.

1961-05-01

A nuclear reactor for use in a merchant marine ship is described. The reactor is of pressurized, light water cooled and moderated design in which three passes of the water through the core in successive regions of low, intermediate, and high heat generation and downflow in a fuel region are made. The design makes a compact reactor construction with extended core life. The core has an egg-crate lattice containing the fuel elements that are confined between a lower flow baffle and upper grid plate, with the latter serving also as part of a turn- around manifold from which the entire coolant is distributed into the outer fuel elements for the second pass through the core. The inner fuel elements are cooled in the third pass. (AEC)

Heat dissipating nuclear reactor

DOEpatents

Hunsbedt, A.; Lazarus, J.D.

1985-11-21

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

Heat dissipating nuclear reactor

DOEpatents

Hunsbedt, Anstein; Lazarus, Jonathan D.

1987-01-01

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

Thermochemical reactor systems and methods

DOEpatents

Lipinski, Wojciech; Davidson, Jane Holloway; Chase, Thomas Richard

2016-11-29

Thermochemical reactor systems that may be used to produce a fuel, and methods of using the thermochemical reactor systems, utilizing a reactive cylindrical element, an optional energy transfer cylindrical element, an inlet gas management system, and an outlet gas management system.

Reactor vibration reduction based on giant magnetostrictive materials

NASA Astrophysics Data System (ADS)

Rongge, Yan; Weiying, Liu; Yuechao, Wu; Menghua, Duan; Xiaohong, Zhang; Lihua, Zhu; Ling, Weng; Ying, Sun

2017-05-01

The vibration of reactors not only produces noise pollution, but also affects the safe operation of reactors. Giant magnetostrictive materials can generate huge expansion and shrinkage deformation in a magnetic field. With the principle of mutual offset between the giant magnetostrictive force produced by the giant magnetostrictive material and the original vibration force of the reactor, the vibration of the reactor can be reduced. In this paper, magnetization and magnetostriction characteristics in silicon steel and the giant magnetostrictive material are measured, respectively. According to the presented magneto-mechanical coupling model including the electromagnetic force and the magnetostrictive force, reactor vibration is calculated. By comparing the vibration of the reactor with different inserted materials in the air gaps between the reactor cores, the vibration reduction effectiveness of the giant magnetostrictive material is validated.

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