Sample records for yonggwang-1 reactor

  1. 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

  2. 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

  3. 10 CFR 1.44 - Office of New Reactors.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 10 Energy 1 2011-01-01 2011-01-01 false Office of New Reactors. 1.44 Section 1.44 Energy NUCLEAR... Office of New Reactors. The Office of New Reactors— (a) Develops, promulgates and implements regulations... safeguarding of nuclear reactor facilities licensed under part 52 of this chapter prior to initial commencement...

  4. 10 CFR 1.44 - Office of New Reactors.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 10 Energy 1 2010-01-01 2010-01-01 false Office of New Reactors. 1.44 Section 1.44 Energy NUCLEAR... Office of New Reactors. The Office of New Reactors— (a) Develops, promulgates and implements regulations... safeguarding of nuclear reactor facilities licensed under part 52 of this chapter prior to initial commencement...

  5. 10 CFR 1.44 - Office of New Reactors.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... 10 Energy 1 2013-01-01 2013-01-01 false Office of New Reactors. 1.44 Section 1.44 Energy NUCLEAR... safeguarding of nuclear reactor facilities licensed under part 52 of this chapter prior to initial commencement... Office of New Reactors. The Office of New Reactors— (a) Develops, promulgates and implements regulations...

  6. 10 CFR 1.44 - Office of New Reactors.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... 10 Energy 1 2014-01-01 2014-01-01 false Office of New Reactors. 1.44 Section 1.44 Energy NUCLEAR... safeguarding of nuclear reactor facilities licensed under part 52 of this chapter prior to initial commencement... Office of New Reactors. The Office of New Reactors— (a) Develops, promulgates and implements regulations...

  7. 10 CFR 1.44 - Office of New Reactors.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... 10 Energy 1 2012-01-01 2012-01-01 false Office of New Reactors. 1.44 Section 1.44 Energy NUCLEAR... safeguarding of nuclear reactor facilities licensed under part 52 of this chapter prior to initial commencement... Office of New Reactors. The Office of New Reactors— (a) Develops, promulgates and implements regulations...

  8. 10 CFR 1.43 - Office of Nuclear Reactor Regulation.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... 10 Energy 1 2014-01-01 2014-01-01 false Office of Nuclear Reactor Regulation. 1.43 Section 1.43 Energy NUCLEAR REGULATORY COMMISSION STATEMENT OF ORGANIZATION AND GENERAL INFORMATION Headquarters Program Offices § 1.43 Office of Nuclear Reactor Regulation. The Office of Nuclear Reactor Regulation— (a...

  9. 10 CFR 1.43 - Office of Nuclear Reactor Regulation.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... 10 Energy 1 2012-01-01 2012-01-01 false Office of Nuclear Reactor Regulation. 1.43 Section 1.43 Energy NUCLEAR REGULATORY COMMISSION STATEMENT OF ORGANIZATION AND GENERAL INFORMATION Headquarters Program Offices § 1.43 Office of Nuclear Reactor Regulation. The Office of Nuclear Reactor Regulation— (a...

  10. 10 CFR 1.43 - Office of Nuclear Reactor Regulation.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... 10 Energy 1 2013-01-01 2013-01-01 false Office of Nuclear Reactor Regulation. 1.43 Section 1.43 Energy NUCLEAR REGULATORY COMMISSION STATEMENT OF ORGANIZATION AND GENERAL INFORMATION Headquarters Program Offices § 1.43 Office of Nuclear Reactor Regulation. The Office of Nuclear Reactor Regulation— (a...

  11. 10 CFR 1.43 - Office of Nuclear Reactor Regulation.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 10 Energy 1 2010-01-01 2010-01-01 false Office of Nuclear Reactor Regulation. 1.43 Section 1.43 Energy NUCLEAR REGULATORY COMMISSION STATEMENT OF ORGANIZATION AND GENERAL INFORMATION Headquarters Program Offices § 1.43 Office of Nuclear Reactor Regulation. The Office of Nuclear Reactor Regulation— (a...

  12. 10 CFR 1.43 - Office of Nuclear Reactor Regulation.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 10 Energy 1 2011-01-01 2011-01-01 false Office of Nuclear Reactor Regulation. 1.43 Section 1.43 Energy NUCLEAR REGULATORY COMMISSION STATEMENT OF ORGANIZATION AND GENERAL INFORMATION Headquarters Program Offices § 1.43 Office of Nuclear Reactor Regulation. The Office of Nuclear Reactor Regulation— (a...

  13. 10 CFR 1.13 - Advisory Committee on Reactor Safeguards.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 10 Energy 1 2011-01-01 2011-01-01 false Advisory Committee on Reactor Safeguards. 1.13 Section 1... Headquarters Panels, Boards, and Committees § 1.13 Advisory Committee on Reactor Safeguards. The Advisory Committee on Reactor Safeguards (ACRS) was established by section 29 of the Atomic Energy Act of 1954, as...

  14. 10 CFR 1.13 - Advisory Committee on Reactor Safeguards.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 10 Energy 1 2010-01-01 2010-01-01 false Advisory Committee on Reactor Safeguards. 1.13 Section 1... Headquarters Panels, Boards, and Committees § 1.13 Advisory Committee on Reactor Safeguards. The Advisory Committee on Reactor Safeguards (ACRS) was established by section 29 of the Atomic Energy Act of 1954, as...

  15. Equipment for neutron measurements at VR-1 Sparrow training reactor.

    PubMed

    Kolros, Antonin; Huml, Ondrej; Kríz, Martin; Kos, Josef

    2010-01-01

    The VR-1 sparrow reactor is an experimental nuclear facility for training, student education and teaching purposes. The sparrow reactor is an educational platform for the basic experiments at the reactor physic and dosimetry. The aim of this article is to describe the new experimental equipment EMK310 features and possibilities for neutron detection by different gas filled detectors at VR-1 reactor. Among the EMK310 equipment typical attributes belong precise set-up, simple control, resistance to electromagnetic interference, high throughput (counting rate), versatility and remote controllability. The methods for non-linearity correction of pulse neutron detection system and reactimeter application are presented. Copyright 2009. Published by Elsevier Ltd.

  16. SL-1 Accident Briefing Report - 1961 Nuclear Reactor Meltdown Educational Documentary

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

    None

    2013-09-25

    U.S. Atomic Energy Commission (Idaho Operations Office) briefing about the SL-1 Nuclear Reactor Meltdown. The SL-1, or Stationary Low-Power Reactor Number One, was a United States Army experimental nuclear power reactor which underwent a steam explosion and meltdown on January 3, 1961, killing its three operators. The direct cause was the improper withdrawal of the central control rod, responsible for absorbing neutrons in the reactor core. The event is the only known fatal reactor accident in the United States. The accident released about 80 curies (3.0 TBq) of Iodine-131, which was not considered significant due to its location in amore » remote desert of Idaho. About 1,100 curies (41 TBq) of fission products were released into the atmosphere. The facility, located at the National Reactor Testing Station approximately 40 miles (64 km) west of Idaho Falls, Idaho, was part of the Army Nuclear Power Program and was known as the Argonne Low Power Reactor (ALPR) during its design and build phase. It was intended to provide electrical power and heat for small, remote military facilities, such as radar sites near the Arctic Circle, and those in the DEW Line. The design power was 3 MW (thermal). Operating power was 200 kW electrical and 400 kW thermal for space heating. In the accident, the core power level reached nearly 20 GW in just four milliseconds, precipitating the reactor accident and steam explosion.« less

  17. SL-1 Accident Briefing Report - 1961 Nuclear Reactor Meltdown Educational Documentary

    ScienceCinema

    None

    2018-01-16

    U.S. Atomic Energy Commission (Idaho Operations Office) briefing about the SL-1 Nuclear Reactor Meltdown. The SL-1, or Stationary Low-Power Reactor Number One, was a United States Army experimental nuclear power reactor which underwent a steam explosion and meltdown on January 3, 1961, killing its three operators. The direct cause was the improper withdrawal of the central control rod, responsible for absorbing neutrons in the reactor core. The event is the only known fatal reactor accident in the United States. The accident released about 80 curies (3.0 TBq) of Iodine-131, which was not considered significant due to its location in a remote desert of Idaho. About 1,100 curies (41 TBq) of fission products were released into the atmosphere. The facility, located at the National Reactor Testing Station approximately 40 miles (64 km) west of Idaho Falls, Idaho, was part of the Army Nuclear Power Program and was known as the Argonne Low Power Reactor (ALPR) during its design and build phase. It was intended to provide electrical power and heat for small, remote military facilities, such as radar sites near the Arctic Circle, and those in the DEW Line. The design power was 3 MW (thermal). Operating power was 200 kW electrical and 400 kW thermal for space heating. In the accident, the core power level reached nearly 20 GW in just four milliseconds, precipitating the reactor accident and steam explosion.

  18. 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.

  19. 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

  20. 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

  1. Determine Operating Reactor to Use for the 2016 PCI Level 1 Milestone

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

    Clarno, Kevin T.

    2016-01-30

    The Consortium for Advanced Simulation of Light Water Reactors (LWRs) (CASL) Level 1 milestone to “Assess the analysis capability for core-wide [pressurized water reactor] PWR Pellet- Clad Interaction (PCI) screening and demonstrate detailed 3-D analysis on selected sub-region” (L1:CASL.P13.03) requires a particular type of nuclear power plant for the assessment. This report documents the operating reactor and cycles chosen for this assessment in completion of the physics integration (PHI) milestone to “Determine Operating Reactor to use for PCI L1 Milestone” (L3:PHI.CMD.P12.02). Watts Bar Unit 1 experienced (at least) one fuel rod failure in each of cycles 6 and 7, andmore » at least one was deemed to be duty related rather than being primarily related to a manufacturing defect or grid effects. This brief report documents that the data required to model cycles 1–12 of Watts Bar Unit 1 using VERA-CS contains sufficient data to model the PHI portion of the PCI challenge problem. A list of additional data needs is also provided that will be important for verification and validation of the BISON results.« less

  2. 75 FR 36126 - Office of New Reactors; Proposed Revision to Standard Review Plan Section 13.6.1, Revision 1 on...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-06-24

    ... NUCLEAR REGULATORY COMMISSION [NRC-2010-0228] Office of New Reactors; Proposed Revision to Standard Review Plan Section 13.6.1, Revision 1 on Physical Security--Combined License and Operating...), Section 13.6.1 on ``Physical Security--Combined License and Operating Reactors,'' (Agencywide Documents...

  3. Neutronic calculation of fast reactors by the EUCLID/V1 integrated code

    NASA Astrophysics Data System (ADS)

    Koltashev, D. A.; Stakhanova, A. A.

    2017-01-01

    This article considers neutronic calculation of a fast-neutron lead-cooled reactor BREST-OD-300 by the EUCLID/V1 integrated code. The main goal of development and application of integrated codes is a nuclear power plant safety justification. EUCLID/V1 is integrated code designed for coupled neutronics, thermomechanical and thermohydraulic fast reactor calculations under normal and abnormal operating conditions. EUCLID/V1 code is being developed in the Nuclear Safety Institute of the Russian Academy of Sciences. The integrated code has a modular structure and consists of three main modules: thermohydraulic module HYDRA-IBRAE/LM/V1, thermomechanical module BERKUT and neutronic module DN3D. In addition, the integrated code includes databases with fuel, coolant and structural materials properties. Neutronic module DN3D provides full-scale simulation of neutronic processes in fast reactors. Heat sources distribution, control rods movement, reactivity level changes and other processes can be simulated. Neutron transport equation in multigroup diffusion approximation is solved. This paper contains some calculations implemented as a part of EUCLID/V1 code validation. A fast-neutron lead-cooled reactor BREST-OD-300 transient simulation (fuel assembly floating, decompression of passive feedback system channel) and cross-validation with MCU-FR code results are presented in this paper. The calculations demonstrate EUCLID/V1 code application for BREST-OD-300 simulating and safety justification.

  4. Study of dietary supplements compositions by neutron activation analysis at the VR-1 training reactor

    NASA Astrophysics Data System (ADS)

    Stefanik, Milan; Rataj, Jan; Huml, Ondrej; Sklenka, Lubomir

    2017-11-01

    The VR-1 training reactor operated by the Czech Technical University in Prague is utilized mainly for education of students and training of various reactor staff; however, R&D is also carried out at the reactor. The experimental instrumentation of the reactor can be used for the irradiation experiments and neutron activation analysis. In this paper, the neutron activation analysis (NAA) is used for a study of dietary supplements containing the zinc (one of the essential trace elements for the human body). This analysis includes the dietary supplement pills of different brands; each brand is represented by several different batches of pills. All pills were irradiated together with the standard activation etalons in the vertical channel of the VR-1 reactor at the nominal power (80 W). Activated samples were investigated by the nuclear gamma-ray spectrometry technique employing the semiconductor HPGe detector. From resulting saturated activities, the amount of mineral element (Zn) in the pills was determined using the comparative NAA method. The results show clearly that the VR-1 training reactor is utilizable for neutron activation analysis experiments.

  5. Future Reactor Neutrino Experiments (RRNOLD)1

    NASA Astrophysics Data System (ADS)

    Jaffe, David E.

    The prospects for future reactor neutrino experiments that would use tens of kilotons of liquid scintillator with a ∼ 50 km baseline are discussed. These experiments are generically dubbed "RRNOLD" for Radical Reactor Neutrino Oscillation Liquid scintillator Detector experiment. Such experiments are designed to resolve the neutrino mass hierarchy and make sub-percent measurements sin2θ12, Δm232 and Δm122 . RRNOLD would also be sensitive to neutrinos from other sources and have notable sensitivity to proton decay.

  6. THE COOLING REQUIREMENTS AND PROCESS SYSTEMS OF THE SOUTH AFRICAN RESEARCH REACTOR, SAFARI 1

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

    Colley, J.R.

    1962-12-01

    The SAFARI 1 research reactor is cooled and moderated by light water. There are three process systems, a primary water system which cools the reactor core and surroundings, a pool water system, and a secondary water system which removes the heat from the primary and pool systems. The cooling requirements for the reactor core and experimental facilities are outlined, and the cooling and purification functions of the three process systems are described. (auth)

  7. 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

  8. Advance High Temperature Inspection Capabilities for Small Modular Reactors: Part 1 - Ultrasonics

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

    Bond, Leonard J.; Bowler, John R.

    The project objective was to investigate the development non-destructive evaluation techniques for advanced small modular reactors (aSMR), where the research sought to provide key enabling inspection technologies needed to support the design and maintenance of reactor component performance. The project tasks for the development of inspection techniques to be applied to small modular reactor are being addressed through two related activities. The first is focused on high temperature ultrasonic transducers development (this report Part 1) and the second is focused on an advanced eddy current inspection capability (Part 2). For both inspection techniques the primary aim is to develop in-servicemore » inspection techniques that can be carried out under standby condition in a fast reactor at a temperature of approximately 250°C in the presence of liquid sodium. The piezoelectric material and the bonding between layers have been recognized as key factors fundamental for development of robust ultrasonic transducers. Dielectric constant characterization of bismuth scantanate-lead titanate ((1-x)BiScO 3-xPbTiO 3) (BS-PT) has shown a high Curie temperature in excess of 450°C , suitable for hot stand-by inspection in liquid metal reactors. High temperature pulse-echo contact measurements have been performed with BS-PT bonded to 12.5 mm thick 1018-low carbon steel plate from 20C up to 260 C. High temperature air-backed immersion transducers have been developed with BS-PT, high temperature epoxy and quarter wavlength nickel plate, needed for wetting ability in liquid sodium. Ultrasonic immersion measurements have been performed in water up to 92C and in silicone oil up to 140C. Physics based models have been validated with room temperature experimental data with benchmark artifical defects.« less

  9. 153. ARAIII Reactor building (ARA608) Foundation plan. Aerojetgeneral 880area/GCRE608S1. Date: ...

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

    153. ARA-III Reactor building (ARA-608) Foundation plan. Aerojet-general 880-area/GCRE-608-S-1. Date: February 1958. Ineel index code no. 063-0608-60-013-102653. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

  10. COMMODITY SCALE SYNTHESIS OF 1-METHYLIMIDAZOLE BASED IONIC LIQUIDS USING A SPINNING TUBE-IN-TUBE REACTOR

    EPA Science Inventory

    The continuous large-scale preparation of several 1-methylimidazole based ionic liquids was carried out using a Spinning Tube-in-Tube (STT) reactor (manufactured by Kreido Laboratories). This reactor, which embodies and facilitates the use of Green Chemistry principles and Proce...

  11. Health Physics State of the RA-1 Reactor, Period Between 1-1-62 and 10-8- 62. Report No. 87; ESTADO RADIOSANIT ARIO DEL REACTOR R.A. 1, PERIODO COMPRENDIDO ENTRE EL 1-1-62 Y EL 10-8-62. Informe No. 87

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

    Gaspar, R.; Moll, O.; Hermelo, C.

    1963-01-01

    The methods used to measure the irradiation levels and the radiation exposure of personnel of the RA-1 reactor are described. The criteria used to evaluate the risks from this exposure are reported. Typical graphs are shown of the radiation levels measured in the control room. (J.S.R.)

  12. Reversal of OFI and CHF in Research Reactors Operating at 1 to 50 Bar. Version 1.0

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

    Kalimullah, M.; Olson, A. P.; Dionne, B.

    2014-02-28

    The conditions at which the critical heat flux (CHF) and the heat flux at the onset of Ledinegg flow instability (OFI) are equal, are determined for a coolant channel with uniform heat flux as a function of five independent parameters: the channel exit pressure (P), heated length (Lh) , heated diameter (Dh), inlet temperature (Tin), and mass flux (G). A diagram is made by plotting the mass flux and heat flux at the OFI-CHF intersection (reversal from CHF > OFI to CHF < OFI as G increases) as a function of P (1 to 50 bar), for 36 combinations ofmore » the remaining three parameters (Lh , Dh , Tin): Lh = 0.28, 0.61, 1.18 m; Dh = 3, 4, 6, 8 mm; Tin = 30, 50, 70 °C. The use of the diagram to scope whether a research reactor is OFI-limited (below the curve) or CHF-limited based on the five parameters of its coolant channel is described. Justification for application of the diagram to research reactors with axially non-uniform heat flux is provided. Due to its limitations (uncertainties not included), the diagram cannot replace the detailed thermal-hydraulic analysis required for a reactor safety analysis. In order to make the OFI-CHF intersection diagram, two world-class CHF prediction methods (the Hall-Mudawar correlation and the extended Groeneveld 2006 table) are compared for 216 combinations of the five independent parameters. The two widely used OFI correlations (the Saha- Zuber and the Whittle-Forgan with η = 32.5) are also compared for the same combinations of the five parameters. The extended Groeneveld table and the Whittle-Forgan OFI correlation are selected for use in making the diagram. Using the above five design parameters, a research reactor can be represented by a point on the reversal diagram, and the diagram can be used to scope, without a thermal-hydraulic calculation, whether the OFI will occur before the CHF, or the CHF will occur before the OFI when the reactor power is increased keeping the five parameters fixed.« less

  13. Superfund record of decision (EPA region 10): Idaho National Engineering Lab, (USDOE) Operable Unit 26 (Stationary Low-Power Reactor-1 and Boiling Water Reactor Experiment-I Burial Grounds), Idaho Falls, ID, December 1, 1995

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

    NONE

    1997-03-01

    This document presents the selected remedial action for the Stationary Low-Power Reactor-1 (SL-1) burial ground, the Boiling Water Reactor Experiment-I (BORAX-I) burial ground, and 10 no action sites in Waste Area Group 5. Actual or threatened releases of hazardous substances from the SL-1 and BORAX-I burial grounds, if not addressed by implementing the response action selected in this Record of Decision, may present a current or potential threat to public health, welfare, or the environment. The 10 no action sites do not present a threat to human health or the environment.

  14. 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

  15. Pm-1 Reactor Core Final Design Report

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

    Bagley, R. O.; Cox, F. H.; Carnasale, A.

    1962-01-01

    The PM-1 water cooled and moderated core contains 741 highly enriched stainless steel cermet tubular fuel elements and 90 lumped B stainless steel burnable poison elements, and it is controlled by 6 Y-shaped europium titanate movable control rods. The core has a lifetime of 1.95 years when operated at its design power level of 9.37 mw of thermal energy. The control of the core is designed so that there is a positive shutdown margin at all times with either one rod stuck completely out or the core or with two rods stuck in the operating condition. The core power ismore » removed by 2125 gpm of pressurized water at an average temperature of 463 deg F and pressure of 1300 psia. In reactors of this type, the core is stable with a negative temperature coefficient of approximately 2.5 x 10/sup -4/ DELTA K/K/ deg F.« less

  16. ARMY GAS-COOLED REACTOR SYSTEMS PROGRAM. Quarterly Progress Report, October 1-December 31, 1963

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

    None

    1964-02-15

    The ML-1 power plant did not operate during the report period; low power reactor physics and shielding experiments were conducted with the ML-1 reactor. Evaluation of moderate corrosion observed on aluminum parts exposed to the ML-1 shield solution indicated no loss of performance capability. Preliminary tests showed that the corrosion probably was caused by heavy metal ions or chlorides in the solution, Massive corrosion observed on the ML-1 fuel element lower spiders was attributed to sub-standard material; failure of some spiders was attributed to a combination of corrosion and sub-standard fabrication. Evaluation indicated that the upper spiders will perform satisfactorilymore » for the design lifetime. Modification, repair, and reassembly of the CSN-1A t-c set was completed. Operation demonstrated bearing stability, but showed that the turbine effective flow area was too large. A bypass flow path in the turbine was being corrected. The TCS-670 t-c set will be stored indefinitely. Since a commercial alternator will be used for the ML-1A, further development of the brushless alternator was postponed indefinitely. Evaluation revealed that the ML-1 improved precooler design was not compatible with ML-1A requirements. Operntion of the IB-17R-2 and -3 test elements in the GETR continued without incident. Preliminary design of the ML-1A power plant was initiated. Design of modifications to the GCRE facility to adapt it to testing the ML-1 reactor skid was initiated. (auth)« less

  17. 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

  18. Lifetime Neutron Fluence Analysis of the Ringhals Unit 1 Boiling Water Reactor

    NASA Astrophysics Data System (ADS)

    Kulesza, Joel A.; Roudén, Jenny; Green, Eva-Lena

    2016-02-01

    This paper describes a neutron fluence assessment considering the entire commercial operating history (35 cycles or ˜ 25 effective full power years) of the Ringhals Unit 1 reactor pressure vessel beltline region. In this assessment, neutron (E >1.0 MeV) fluence and iron atom displacement distributions were calculated on the moderator tank and reactor pressure vessel structures. To validate those calculations, five in-vessel surveillance chain dosimetry sets were evaluated as well as material samples taken from the upper core grid and wide range neutron monitor tubes to act as a form of retrospective dosimetry. During the analysis, it was recognized that delays in characterizing the retrospective dosimetry samples reduced the amount of reactions available to be counted and complicated the material composition determination. However, the comparisons between the surveillance chain dosimetry measurements (M) and calculated (C) results show similar and consistent results with the linear average M/C ratio of 1.13 which is in good agreement with the resultant least squares best estimate (BE)/C ratios of 1.10 for both neutron (E >1.0 MeV) flux and iron atom displacement rate.

  19. Nuclear Engineering Computer Modules: Reactor Dynamics, RD-1 and RD-2.

    ERIC Educational Resources Information Center

    Onega, Ronald J.

    The objective of the Reactor Dynamics Module, RD-1, is to obtain the kinetics equation without feedback and solve the kinetics equations numerically for one to six delayed neutron groups for time varying reactivity insertions. The computer code FUMOKI (Fundamental Mode Kinetics) will calculate the power as a function of time for either uranium or…

  20. 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)

  1. Ethanol Production from Wet-Exploded Wheat Straw Hydrolysate by Thermophilic Anaerobic Bacterium Thermoanaerobacter BG1L1 in a Continuous Immobilized Reactor

    NASA Astrophysics Data System (ADS)

    Georgieva, Tania I.; Mikkelsen, Marie J.; Ahring, Birgitte K.

    Thermophilic ethanol fermentation of wet-exploded wheat straw hydrolysate was investigated in a continuous immobilized reactor system. The experiments were carried out in a lab-scale fluidized bed reactor (FBR) at 70°C. Undetoxified wheat straw hydrolysate was used (3-12% dry matter), corresponding to sugar mixtures of glucose and xylose ranging from 12 to 41 g/1. The organism, thermophilic anaerobic bacterium Thermoanaerobacter BG1L1, exhibited significant resistance to high levels of acetic acid (up to 10 g/1) and other metabolic inhibitors present in the hydrolysate. Although the hydrolysate was not detoxified, ethanol yield in a range of 0.39-0.42 g/g was obtained. Overall, sugar efficiency to ethanol was 68-76%. The reactor was operated continuously for approximately 143 days, and no contamination was seen without the use of any agent for preventing bacterial infections. The tested microorganism has considerable potential to be a novel candidate for lignocellulose bioconversion into ethanol. The work reported here also demonstrates that the use of FBR configuration might be a viable approach for thermophilic anaerobic ethanol fermentation.

  2. 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.

  3. IEA-R1 Nuclear Research Reactor: 58 Years of Operating Experience and Utilization for Research, Teaching and Radioisotopes Production

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

    Cardenas, Jose Patricio Nahuel; Filho, Tufic Madi; Saxena, Rajendra

    IEA-R1 research reactor at the Instituto de Pesquisas Energeticas e Nucleares (Nuclear and Energy Research Institute) IPEN, Sao Paulo, Brazil is the largest power research reactor in Brazil, with a maximum power rating of 5 MWth. It is being used for basic and applied research in the nuclear and neutron related sciences, for the production of radioisotopes for medical and industrial applications, and for providing services of neutron activation analysis, real time neutron radiography, and neutron transmutation doping of silicon. IEA-R1 is a swimming pool reactor, with light water as the coolant and moderator, and graphite and beryllium as reflectors.more » The reactor was commissioned on September 16, 1957 and achieved its first criticality. It is currently operating at 4.5 MWth with a 60-hour cycle per week. In the early sixties, IPEN produced {sup 131}I, {sup 32}P, {sup 198}Au, {sup 24}Na, {sup 35}S, {sup 51}Cr and labeled compounds for medical use. During the past several years, a concerted effort has been made in order to upgrade the reactor power to 5 MWth through refurbishment and modernization programs. One of the reasons for this decision was to produce {sup 99}Mo at IPEN. The reactor cycle will be gradually increased to 120 hours per week continuous operation. It is anticipated that these programs will assure the safe and sustainable operation of the IEA-R1 reactor for several more years, to produce important primary radioisotopes {sup 99}Mo, {sup 125}I, {sup 131}I, {sup 153}Sm and {sup 192}Ir. Currently, all aspects of dealing with fuel element fabrication, fuel transportation, isotope processing, and spent fuel storage are handled by IPEN at the site. The reactor modernization program is slated for completion by 2015. This paper describes 58 years of operating experience and utilization of the IEA-R1 research reactor for research, teaching and radioisotopes production. (authors)« less

  4. 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.

  5. The EUCLID/V1 Integrated Code for Safety Assessment of Liquid Metal Cooled Fast Reactors. Part 1: Basic Models

    NASA Astrophysics Data System (ADS)

    Mosunova, N. A.

    2018-05-01

    The article describes the basic models included in the EUCLID/V1 integrated code intended for safety analysis of liquid metal (sodium, lead, and lead-bismuth) cooled fast reactors using fuel rods with a gas gap and pellet dioxide, mixed oxide or nitride uranium-plutonium fuel under normal operation, under anticipated operational occurrences and accident conditions by carrying out interconnected thermal-hydraulic, neutronics, and thermal-mechanical calculations. Information about the Russian and foreign analogs of the EUCLID/V1 integrated code is given. Modeled objects, equation systems in differential form solved in each module of the EUCLID/V1 integrated code (the thermal-hydraulic, neutronics, fuel rod analysis module, and the burnup and decay heat calculation modules), the main calculated quantities, and also the limitations on application of the code are presented. The article also gives data on the scope of functions performed by the integrated code's thermal-hydraulic module, using which it is possible to describe both one- and twophase processes occurring in the coolant. It is shown that, owing to the availability of the fuel rod analysis module in the integrated code, it becomes possible to estimate the performance of fuel rods in different regimes of the reactor operation. It is also shown that the models implemented in the code for calculating neutron-physical processes make it possible to take into account the neutron field distribution over the fuel assembly cross section as well as other features important for the safety assessment of fast reactors.

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

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

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

    2015-08-20

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

  7. RAZORBACK - A Research Reactor Transient Analysis Code Version 1.0 - Volume 3: Verification and Validation Report.

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

    Talley, Darren G.

    2017-04-01

    This report describes the work and results of the verification and validation (V&V) of the version 1.0 release of the Razorback code. Razorback is a computer code designed to simulate the operation of a research reactor (such as the Annular Core Research Reactor (ACRR)) by a coupled numerical solution of the point reactor kinetics equations, the energy conservation equation for fuel element heat transfer, the equation of motion for fuel element thermal expansion, and the mass, momentum, and energy conservation equations for the water cooling of the fuel elements. This V&V effort was intended to confirm that the code showsmore » good agreement between simulation and actual ACRR operations.« less

  8. 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.

  9. 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.

  10. Fukushima Daiichi Unit 1 Accident Progression Uncertainty Analysis and Implications for Decommissioning of Fukushima Reactors - Volume I.

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

    Gauntt, Randall O.; Mattie, Patrick D.

    Sandia National Laboratories (SNL) has conducted an uncertainty analysis (UA) on the Fukushima Daiichi unit (1F1) accident progression with the MELCOR code. The model used was developed for a previous accident reconstruction investigation jointly sponsored by the US Department of Energy (DOE) and Nuclear Regulatory Commission (NRC). That study focused on reconstructing the accident progressions, as postulated by the limited plant data. This work was focused evaluation of uncertainty in core damage progression behavior and its effect on key figures-of-merit (e.g., hydrogen production, reactor damage state, fraction of intact fuel, vessel lower head failure). The primary intent of this studymore » was to characterize the range of predicted damage states in the 1F1 reactor considering state of knowledge uncertainties associated with MELCOR modeling of core damage progression and to generate information that may be useful in informing the decommissioning activities that will be employed to defuel the damaged reactors at the Fukushima Daiichi Nuclear Power Plant. Additionally, core damage progression variability inherent in MELCOR modeling numerics is investigated.« less

  11. Remediation of 1,2,3-trichlorobenzene contaminated soil using a combined thermal desorption-molten salt oxidation reactor system.

    PubMed

    Li, Jin-hui; Sun, Xiao-fei; Yao, Zhi-tong; Zhao, Xiang-yang

    2014-02-01

    A combined thermal desorption (TD)-molten salt oxidation (MSO) reactor system was applied to remediate the 1,2,3-trichlorobenzene (1,2,3-TCB) contaminated soil. The TD reactor was used to enrich the contaminant from soil, and its dechlorination of the contaminant was achieved in the MSO reactor. The optimum operating conditions of TD, and the effects of MSO reactor temperatures, additive amounts of the TCB on destruction and removal efficiency (DRE) of TCB and chlorine retention efficiency (CRE) were investigated. The reaction mechanism and pathway were proposed as well. The combined system could remediate the contaminated soil at a large scale of concentration from 5 to 25gkg(-1), and the DRE and CRE reached more than 99% and 95%, respectively, at temperatures above 850°C. The reaction emissions included C6H6, CH4, CO and CO2, and chlorinated species were not detected. It was found that a little increase in the temperature can considerably reduce the emission of C6H6, CH4, and CO, while the CO2 level increased. Copyright © 2014. Published by Elsevier Ltd.

  12. Analysis of Accidents at the Pakistan Research Reactor-1 Using Proposed Mixed-Fuel (HEU and LEU) Core

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

    Bokhari, Ishtiaq H.

    2004-12-15

    The Pakistan Research Reactor-1 (PARR-1) was converted from highly enriched uranium (HEU) to low-enriched uranium (LEU) fuel in 1991. The reactor is running successfully, with an upgraded power level of 10 MW. To save money on the purchase of costly fresh LEU fuel elements, the use of less burnt HEU spent fuel elements along with the present LEU fuel elements is being considered. The proposal calls for the HEU fuel elements to be placed near the thermal column to gain the required excess reactivity. In the present study the safety analysis of a proposed mixed-fuel core has been carried outmore » at a calculated steady-state power level of 9.8 MW. Standard computer codes and correlations were employed to compute various parameters. Initiating events in reactivity-induced accidents involve various modes of reactivity insertion, namely, start-up accident, accidental drop of a fuel element on the core, flooding of a beam tube with water, and removal of an in-pile experiment during reactor operation. For each of these transients, time histories of reactor power, energy released, temperature, and reactivity were determined.« less

  13. Introduction to Reactor Statics Modules, RS-1. Nuclear Engineering Computer Modules.

    ERIC Educational Resources Information Center

    Edlund, Milton C.

    The nine Reactor Statics Modules are designed to introduce students to the use of numerical methods and digital computers for calculation of neutron flux distributions in space and energy which are needed to calculate criticality, power distribution, and fuel burn-up for both slow neutron and fast neutron fission reactors. The diffusion…

  14. A search for neutrino oscillations using the CHOOZ 1 km baseline reactor neutrino experiment

    NASA Astrophysics Data System (ADS)

    George, Jean

    1999-10-01

    Neutrino oscillation searches are an active field of research due to the implications their discovery may have for the solar neutrino anomaly as well as for the atmospheric neutrino anomaly. Their discovery may also have broad ramifications for the Standard Model of Particle Physics as a whole. Results from an oscillation search using the CHOOZ long baseline reactor neutrino experiment are presented in this thesis. These results are based on the data taken from June 1997 through April 1998 when the two reactors ran at combined thermal power levels ranging from zero power to their full power level of 8.5 GW. Electron flavored antineutrinos emanating from the reactors were detected through the inverse beta decay channel using a liquid scintillating calorimeter located at a distance of approximately 1 km from the reactor sources. The underground experimental site (300 MWE) provided natural shielding from the background of cosmic ray muons-leading to a background rate more than an order of magnitude lower than the full power signal rate. From the agreement between the detected and expected neutrino event rates no evidence for neutrino oscillations was found (at the 90% C.L.) for the oscillation parameter space governed by Δm 2 > 0.8 × 10-3 eV2 for maximal mixing and by sin2 2Θ > 0.18 for large values of Δm2.

  15. 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

  16. 9 CFR 78.31 - Brucellosis reactor swine.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 9 Animals and Animal Products 1 2011-01-01 2011-01-01 false Brucellosis reactor swine. 78.31... Restrictions on Interstate Movement of Swine Because of Brucellosis § 78.31 Brucellosis reactor swine. (a) Destination. Brucellosis reactor swine may be moved interstate only for immediate slaughter as follows: (1...

  17. 9 CFR 78.22 - Brucellosis reactor bison.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 9 Animals and Animal Products 1 2011-01-01 2011-01-01 false Brucellosis reactor bison. 78.22... Restrictions on Interstate Movement of Bison Because of Brucellosis § 78.22 Brucellosis reactor bison. (a) Destination. Brucellosis reactor bison may be moved interstate only for immediate slaughter as follows: (1...

  18. 9 CFR 78.31 - Brucellosis reactor swine.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 9 Animals and Animal Products 1 2010-01-01 2010-01-01 false Brucellosis reactor swine. 78.31... Restrictions on Interstate Movement of Swine Because of Brucellosis § 78.31 Brucellosis reactor swine. (a) Destination. Brucellosis reactor swine may be moved interstate only for immediate slaughter as follows: (1...

  19. 9 CFR 78.7 - Brucellosis reactor cattle.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

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

  20. 9 CFR 78.7 - Brucellosis reactor cattle.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

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

  1. 9 CFR 78.22 - Brucellosis reactor bison.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 9 Animals and Animal Products 1 2010-01-01 2010-01-01 false Brucellosis reactor bison. 78.22... Restrictions on Interstate Movement of Bison Because of Brucellosis § 78.22 Brucellosis reactor bison. (a) Destination. Brucellosis reactor bison may be moved interstate only for immediate slaughter as follows: (1...

  2. 9 CFR 78.31 - Brucellosis reactor swine.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... 9 Animals and Animal Products 1 2012-01-01 2012-01-01 false Brucellosis reactor swine. 78.31... Restrictions on Interstate Movement of Swine Because of Brucellosis § 78.31 Brucellosis reactor swine. (a) Destination. Brucellosis reactor swine may be moved interstate only for immediate slaughter as follows: (1...

  3. 9 CFR 78.31 - Brucellosis reactor swine.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... 9 Animals and Animal Products 1 2013-01-01 2013-01-01 false Brucellosis reactor swine. 78.31... Restrictions on Interstate Movement of Swine Because of Brucellosis § 78.31 Brucellosis reactor swine. (a) Destination. Brucellosis reactor swine may be moved interstate only for immediate slaughter as follows: (1...

  4. 9 CFR 78.31 - Brucellosis reactor swine.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... 9 Animals and Animal Products 1 2014-01-01 2014-01-01 false Brucellosis reactor swine. 78.31... Restrictions on Interstate Movement of Swine Because of Brucellosis § 78.31 Brucellosis reactor swine. (a) Destination. Brucellosis reactor swine may be moved interstate only for immediate slaughter as follows: (1...

  5. Reactor operations informal monthly report, May 1, 1995--May 31, 1995

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

    Hauptman, H.M.; Petro, J.N.; Jacobi, O.

    1995-05-01

    This document is an informal progress report for the operational performance of the Brookhaven Medical Research Reactor, and the Brookhaven High Flux Beam Reactor, for the month of May, 1995. Both machines ran well during this period, with no reportable instrumentation problems, all scheduled maintenance performed, and only one reportable occurance, involving a particle on Vest Button, Personnel Radioactive Contamination.

  6. A Neutronic Program for Critical and Nonequilibrium Study of Mobile Fuel Reactors: The Cinsf1D Code

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

    Lecarpentier, David; Carpentier, Vincent

    2003-01-15

    Molten salt reactors (MSRs) have the distinction of having a liquid fuel that is also the coolant. The transport of delayed-neutron precursors by the fuel modifies the precursors' equation. As a consequence, it is necessary to adapt the methods currently used for solid fuel reactors to achieve critical or kinetics calculations for an MSR. A program is presented for which this adaptation has been carried out within the framework of the two-energy-group diffusion theory with one dimension of space. This program has been called Cinsf1D (Cinetique pour reacteur a sels fondus 1D)

  7. Validation of light water reactor calculation methods and JEF-1-based data libraries by TRX and BAPL critical experiments

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

    Paratte, J.M.; Pelloni, S.; Grimm, P.

    1991-04-01

    This paper analyzes the capability of various code systems and JEF-1-based nuclear data libraries to compute light water reactor lattices by comparing calculations with results from thermal reactor benchmark experiments TRX and BAPL and with previously published values. With the JEF-1 evaluation, eigenvalues are generally well predicted within 8 mk (1 mk = 0.001) or less by all code systems, and all methods give reasonable results for the measured reaction rate ratios within, or not too far from, the experimental uncertainty.

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

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

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

    1962-10-01

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

  9. Potential use of sugar binding proteins in reactors for regeneration of CO2 fixation acceptor D-Ribulose-1,5-bisphosphate

    PubMed Central

    Mahato, Sourav; De, Debojyoti; Dutta, Debajyoti; Kundu, Moloy; Bhattacharya, Sumana; Schiavone, Marc T; Bhattacharya, Sanjoy K

    2004-01-01

    Sugar binding proteins and binders of intermediate sugar metabolites derived from microbes are increasingly being used as reagents in new and expanding areas of biotechnology. The fixation of carbon dioxide at emission source has recently emerged as a technology with potentially significant implications for environmental biotechnology. Carbon dioxide is fixed onto a five carbon sugar D-ribulose-1,5-bisphosphate. We present a review of enzymatic and non-enzymatic binding proteins, for 3-phosphoglycerate (3PGA), 3-phosphoglyceraldehyde (3PGAL), dihydroxyacetone phosphate (DHAP), xylulose-5-phosphate (X5P) and ribulose-1,5-bisphosphate (RuBP) which could be potentially used in reactors regenerating RuBP from 3PGA. A series of reactors combined in a linear fashion has been previously shown to convert 3-PGA, (the product of fixed CO2 on RuBP as starting material) into RuBP (Bhattacharya et al., 2004; Bhattacharya, 2001). This was the basis for designing reactors harboring enzyme complexes/mixtures instead of linear combination of single-enzyme reactors for conversion of 3PGA into RuBP. Specific sugars in such enzyme-complex harboring reactors requires removal at key steps and fed to different reactors necessitating reversible sugar binders. In this review we present an account of existing microbial sugar binding proteins and their potential utility in these operations. PMID:15175111

  10. Developments and Tendencies in Fission Reactor Concepts

    NASA Astrophysics Data System (ADS)

    Adamov, E. O.; Fuji-Ie, Y.

    This chapter describes, in two parts, new-generation nuclear energy systems that are required to be in harmony with nature and to make full use of nuclear resources. The issues of transmutation and containment of radioactive waste will also be addressed. After a short introduction to the first part, Sect. 58.1.2 will detail the requirements these systems must satisfy on the basic premise of peaceful use of nuclear energy. The expected designs themselves are described in Sect. 58.1.3. The subsequent sections discuss various types of advanced reactor systems. Section 58.1.4 deals with the light water reactor (LWR) whose performance is still expected to improve, which would extend its application in the future. The supercritical-water-cooled reactor (SCWR) will also be shortly discussed. Section 58.1.5 is mainly on the high temperature gas-cooled reactor (HTGR), which offers efficient and multipurpose use of nuclear energy. The gas-cooled fast reactor (GFR) is also included. Section 58.1.6 focuses on the sodium-cooled fast reactor (SFR) as a promising concept for advanced nuclear reactors, which may help both to achieve expansion of energy sources and environmental protection thus contributing to the sustainable development of mankind. The molten-salt reactor (MSR) is shortly described in Sect. 58.1.7. The second part of the chapter deals with reactor systems of a new generation, which are now found at the research and development (R&D) stage and in the medium term of 20-30 years can shape up as reliable, economically efficient, and environmentally friendly energy sources. They are viewed as technologies of cardinal importance, capable of resolving the problems of fuel resources, minimizing the quantities of generated radioactive waste and the environmental impacts, and strengthening the regime of nonproliferation of the materials suitable for nuclear weapons production. Particular attention has been given to naturally safe fast reactors with a closed fuel cycle (CFC

  11. Knowledge and abilities catalog for nuclear power plant operators: Boiling water reactors, Revision 1

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

    NONE

    1995-08-01

    The Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Boiling-Water Reactors (BWRs) (NUREG-1123, Revision 1) provides the basis for the development of content-valid licensing examinations for reactor operators (ROs) and senior reactor operators (SROs). The examinations developed using the BWR Catalog along with the Operator Licensing Examiner Standards (NUREG-1021) and the Examiner`s Handbook for Developing Operator Licensing Written Examinations (NUREG/BR-0122), will cover the topics listed under Title 10, Code of Federal Regulations, Part 55 (10 CFR 55). The BWR Catalog contains approximately 7,000 knowledge and ability (K/A) statements for ROs and SROs at BWRs. The catalog is organized intomore » six major sections: Organization of the Catalog, Generic Knowledge and Ability Statements, Plant Systems grouped by Safety Functions, Emergency and Abnormal Plant Evolutions, Components, and Theory. Revision 1 to the BWR Catalog represents a modification in form and content of the original catalog. The K/As were linked to their applicable 10 CFR 55 item numbers. SRO level K/As were identified by 10 CFR 55.43 item numbers. The plant-wide generic and system generic K/As were combined in one section with approximately one hundred new K/As. Component Cooling Water and Instrument Air Systems were added to the Systems Section. Finally, High Containment Hydrogen Concentration and Plant Fire On Site evolutions added to the Emergency and Abnormal Plant Evolutions section.« less

  12. 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.

  13. 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.

  14. International Research Reactor Decommissioning Project

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

    Leopando, Leonardo; Warnecke, Ernst

    2008-01-15

    Many research reactors have been or will be shut down and are candidates for decommissioning. Most of the respective countries neither have a decommissioning policy nor the required expertise and funds to effectively implement a decommissioning project. The IAEA established the Research Reactor Decommissioning Demonstration Project (R{sup 2}D{sup 2}P) to help answer this need. It was agreed to involve the Philippine Research Reactor (PRR-1) as model reactor to demonstrate 'hands-on' experience as it is just starting the decommissioning process. Other facilities may be included in the project as they fit into the scope of R{sup 2}D{sup 2}P and complement tomore » the PRR-1 decommissioning activities. The key outcome of the R{sup 2}D{sup 2}P will be the decommissioning of the PRR-1 reactor. On the way to this final goal the preparation of safety related documents (i.e., decommissioning plan, environmental impact assessment, safety analysis report, health and safety plan, cost estimate, etc.) and the licensing process as well as the actual dismantling activities could provide a model to other countries involved in the project. It is expected that the R{sup 2}D{sup 2}P would initiate activities related to planning and funding of decommissioning activities in the participating countries if that has not yet been done.« less

  15. 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.

  16. Small and medium power reactors 1987

    NASA Astrophysics Data System (ADS)

    1987-12-01

    This TECDOC follows the publication of TECDOC-347: Small and Medium Power Reactors (SMPR) Project Initiation Study, Phase 1, published in 1985 and TECDOC-376: Small and Medium Power Reactors 1985 published in 1986. It is mainly intended for decision makers in Developing Member States interested in embarking on a nuclear power program. It consists of two parts: (1) guidelines for the introduction of small and medium power reactors in developing countries. These Guidelines were established during the Advisory Group Meeting held in Vienna from 11 to 15 May 1987. Their purpose is to review key aspects relating to the introduction of small and medium power reactors in developing countries; (2) up-dated information on SMPR Concepts Contributed by Supplier Industries. According to the recommendations of the Second Technical Committee Meeting on SMPRs held in Vienna in March 1985, this part contains the up-dated information formerly published in Annex 1 of the above mentioned TECDOC-347.

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

    PubMed

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

    2007-10-01

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

  18. 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.

  19. Design of boron carbide-shielded irradiation channel of the outer irradiation channel of the Ghana Research Reactor-1 using MCNP.

    PubMed

    Abrefah, R G; Sogbadji, R B M; Ampomah-Amoako, E; Birikorang, S A; Odoi, H C; Nyarko, B J B

    2011-01-01

    The MCNP model for the Ghana Research Reactor-1 was redesigned to incorporate a boron carbide-shielded irradiation channel in one of the outer irradiation channels. Extensive investigations were made before arriving at the final design of only one boron carbide covered outer irradiation channel; as all the other designs that were considered did not give desirable results of neutronic performance. The concept of redesigning a new MCNP model, which has a boron carbide-shielded channel is to equip the Ghana Research Reactor-1 with the means of performing efficient epithermal neutron activation analysis. After the simulation, a comparison of the results from the original MCNP model for the Ghana Research Reactor-1 and the new redesigned model of the boron carbide shielded channel was made. The final effective criticality of the original MCNP model for the GHARR-1 was recorded as 1.00402 while that of the new boron carbide designed model was recorded as 1.00282. Also, a final prompt neutron lifetime of 1.5245 × 10(-4)s was recorded for the new boron carbide designed model while a value of 1.5571 × 10(-7)s was recorded for the original MCNP design of the GHARR-1. Copyright © 2010 Elsevier Ltd. All rights reserved.

  20. 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.

  1. Detecting Dark Photons with Reactor Neutrino Experiments

    NASA Astrophysics Data System (ADS)

    Park, H. K.

    2017-08-01

    We propose to search for light U (1 ) dark photons, A', produced via kinetically mixing with ordinary photons via the Compton-like process, γ e-→A'e-, in a nuclear reactor and detected by their interactions with the material in the active volumes of reactor neutrino experiments. We derive 95% confidence-level upper limits on ɛ , the A'-γ mixing parameter, ɛ , for dark-photon masses below 1 MeV of ɛ <1.3 ×10-5 and ɛ <2.1 ×10-5, from NEOS and TEXONO experimental data, respectively. This study demonstrates the applicability of nuclear reactors as potential sources of intense fluxes of low-mass dark photons.

  2. 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.

  3. 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…

  4. The 235U Prompt Fission Neutron Spectrum in the BR1 Reactor at SCK•CEN

    NASA Astrophysics Data System (ADS)

    Wagemans, Jan; Malambu, Edouard; Borms, Luc; Fiorito, Luca

    2016-02-01

    The BR1 research reactor at SCK•CEN has a spherical cavity in the graphite above the reactor core. In this cavity an accurately characterised Maxwellian thermal neutron field is present. Different converters can be loaded in the cavity in order to obtain other types of neutron (and gamma) irradiation fields. Inside the so-called MARK III converter a fast 235U(n,f) prompt fission neutron field can be obtained. With the support of MCNP calculations, irradiations in MARK III can be directly related to the pure 235U(n,f) prompt fission neutron spectrum. For this purpose MARK III spectrum averaged cross sections for the most relevant fluence dosimetry reactions have been determined. A calibration factor for absolute measurements has been determined applying activation dosimetry following ISO/IEC 17025 standards.

  5. ``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

  6. 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.

  7. Development of Neutron Imaging System for Neutron Tomography at Thai Research Reactor TRR-1/M1

    NASA Astrophysics Data System (ADS)

    Wonglee, S.; Khaweerat, S.; Channuie, J.; Picha, R.; Liamsuwan, T.; Ratanatongchai, W.

    2017-09-01

    The neutron imaging is a powerful non-destructive technique to investigate the internal structure and provides the information which is different from the conventional X-ray/Gamma radiography. By reconstruction of the obtained 2-dimentional (2D) images from the taken different angle around the specimen, the tomographic image can be obtained and it can provide the information in more detail. The neutron imaging system at Thai Research Reactor TRR-1/M1 of Thailand Institute of Nuclear Technology (Public Organization) has been developed to conduct the neutron tomography since 2014. The primary goal of this work is to serve the investigation of archeological samples, however, this technique can also be applied to various fields, such as investigation of industrial specimen and others. This research paper presents the performance study of a compact neutron camera manufactured by Neutron Optics such as speed and sensitivity. Furthermore, the 3-dimentional (3D) neutron image was successfully reconstructed at the developed neutron imaging system of TRR-1/M1.

  8. Detecting Dark Photons with Reactor Neutrino Experiments.

    PubMed

    Park, H K

    2017-08-25

    We propose to search for light U(1) dark photons, A^{'}, produced via kinetically mixing with ordinary photons via the Compton-like process, γe^{-}→A^{'}e^{-}, in a nuclear reactor and detected by their interactions with the material in the active volumes of reactor neutrino experiments. We derive 95% confidence-level upper limits on ε, the A^{'}-γ mixing parameter, ε, for dark-photon masses below 1 MeV of ε<1.3×10^{-5} and ε<2.1×10^{-5}, from NEOS and TEXONO experimental data, respectively. This study demonstrates the applicability of nuclear reactors as potential sources of intense fluxes of low-mass dark photons.

  9. 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.

  10. Generic Stellarator-like Magnetic Fusion Reactor

    NASA Astrophysics Data System (ADS)

    Sheffield, John; Spong, Donald

    2015-11-01

    The Generic Magnetic Fusion Reactor paper, published in 1985, has been updated, reflecting the improved science and technology base in the magnetic fusion program. Key changes beyond inflation are driven by important benchmark numbers for technologies and costs from ITER construction, and the use of a more conservative neutron wall flux and fluence in modern fusion reactor designs. In this paper the generic approach is applied to a catalyzed D-D stellarator-like reactor. It is shown that an interesting power plant might be possible if the following parameters could be achieved for a reference reactor: R/ < a > ~ 4 , confinement factor, fren = 0.9-1.15, < β > ~ 8 . 0 -11.5 %, Zeff ~ 1.45 plus a relativistic temperature correction, fraction of fast ions lost ~ 0.07, Bm ~ 14-16 T, and R ~ 18-24 m. J. Sheffield was supported under ORNL subcontract 4000088999 with the University of Tennessee.

  11. Reactor Design and Decommissioning - An Overview of International Activities in Post Fukushima Era1 - 12396

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

    Devgun, Jas S.; Laraia, Michele; Pescatore, Claudio

    Accidents at the Fukushima Dai-ichi reactors as a result of the devastating earthquake and tsunami of March 11, 2011 have not only dampened the nuclear renaissance but have also initiated a re-examination of the design and safety features for the existing and planned nuclear reactors. Even though failures of some of the key site features at Fukushima can be attributed to events that in the past would have been considered as beyond the design basis, the industry as well as the regulatory authorities are analyzing what features, especially passive features, should be designed into the new reactor designs to minimizemore » the potential for catastrophic failures. It is also recognized that since the design of the Fukushima BWR reactors which were commissioned in 1971, many advanced safety features are now a part of the newer reactor designs. As the recovery efforts at the Fukushima site are still underway, decisions with respect to the dismantlement and decommissioning of the damaged reactors and structures have not yet been finalized. As it was with Three Mile Island, it could take several decades for dismantlement, decommissioning and clean up, and the project poses especially tough challenges. Near-term assessments have been issued by several organizations, including the IAEA, the USNRC and others. Results of such investigations will lead to additional improvements in system and site design measures including strengthening of the anti-tsunami defenses, more defense-in-depth features in reactor design, and better response planning and preparation involving reactor sites. The question also arises what would the effect be on the decommissioning scene worldwide, and what would the effect be on the new reactors when they are eventually retired and dismantled. This paper provides an overview of the US and international activities related to recovery and decommissioning including the decommissioning features in the reactor design process and examines these from a new

  12. 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.

  13. 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

  14. Analysis of the Gas Core Actinide Transmutation Reactor (GCATR)

    NASA Technical Reports Server (NTRS)

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

    1977-01-01

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

  15. 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)

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

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

    Maloy, Stuart Andrew

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

  17. Nuclear propulsion apparatus with alternate reactor segments

    DOEpatents

    Szekely, Thomas

    1979-04-03

    1. Nuclear propulsion apparatus comprising: A. means for compressing incoming air; B. nuclear fission reactor means for heating said air; C. means for expanding a portion of the heated air to drive said compressing means; D. said nuclear fission reactor means being divided into a plurality of radially extending segments; E. means for directing a portion of the compressed air for heating through alternate segments of said reactor means and another portion of the compressed air for heating through the remaining segments of said reactor means; and F. means for further expanding the heated air from said drive means and the remaining heated air from said reactor means through nozzle means to effect reactive thrust on said apparatus.

  18. 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.

  19. Feasibility study of a magnetic fusion production reactor

    NASA Astrophysics Data System (ADS)

    Moir, R. W.

    1986-12-01

    A magnetic fusion reactor can produce 10.8 kg of tritium at a fusion power of only 400 MW —an order of magnitude lower power than that of a fission production reactor. Alternatively, the same fusion reactor can produce 995 kg of plutonium. Either a tokamak or a tandem mirror production plant can be used for this purpose; the cost is estimated at about 1.4 billion (1982 dollars) in either case. (The direct costs are estimated at 1.1 billion.) The production cost is calculated to be 22,000/g for tritium and 260/g for plutonium of quite high purity (1%240Pu). Because of the lack of demonstrated technology, such a plant could not be constructed today without significant risk. However, good progress is being made in fusion technology and, although success in magnetic fusion science and engineering is hard to predict with assurance, it seems possible that the physics basis and much of the needed technology could be demonstrated in facilities now under construction. Most of the remaining technology could be demonstrated in the early 1990s in a fusion test reactor of a few tens of megawatts. If the Magnetic Fusion Energy Program constructs a fusion test reactor of approximately 400 MW of fusion power as a next step in fusion power development, such a facility could be used later as a production reactor in a spinoff application. A construction decision in the late 1980s could result in an operating production reactor in the late 1990s. A magnetic fusion production reactor (MFPR) has four potential advantages over a fission production reactor: (1) no fissile material input is needed; (2) no fissioning exists in the tritium mode and very low fissioning exists in the plutonium mode thus avoiding the meltdown hazard; (3) the cost will probably be lower because of the smaller thermal power required; (4) and no reprocessing plant is needed in the tritium mode. The MFPR also has two disadvantages: (1) it will be more costly to operate because it consumes rather than sells

  20. 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.

  1. Treatment of fruit-juice industry wastewater in a two-stage anaerobic hybrid (AH) reactor system followed by a sequencing batch reactor (SBR).

    PubMed

    Tawfik, A; El-Kamah, H

    2012-01-01

    This study has been carried out to assess the performance of a combined system consisting of an anaerobic hybrid (AH) reactor followed by a sequencing batch reactor (SBR) for treatment of fruit-juice industry wastewater at a temperature of 26 degrees C. Three experimental runs were conducted in this investigation. In the first experiment, a single-stage AH reactor was operated at a hydraulic retention time (HRT) of 10.2 h and organic loading rate (OLR) of 11.8 kg COD m(-3) d(-1). The reactor achieved a removal efficiency of 42% for chemical oxygen demand (COD), 50.8% for biochemical oxygen demand (BOD5), 50.3% for volatile fatty acids (VFA) and 56.4% for total suspended solids (TSS). In the second experiment, two AH reactors connected in series achieved a higher removal efficiency for COD (67.4%), BOD5 (77%), and TSS (71.5%) at a total HRT of 20 h and an OLR of 5.9 kg COD m(-3) d(-1). For removal of the remaining portions of COD, BOD5 and TSS from the effluent of the two-stage AH system, a sequencing batch reactor (SBR) was investigated as a post-treatment unit. The reactor achieved a substantial reduction in total COD, resulting in an average effluent concentration of 50 mg L(-1) at an HRT of 11 h and OLR of 5.3 kg COD m(-3) d(-1). Almost complete removal of total BOD5 and oil and grease was achieved, i.e. 10 mg L(-1) and 1.2 mg L(-1), respectively, remained in the final effluent of the SBR.

  2. Evaluation of power density on the bioethanol production using mesoscale oscillatory baffled reactor and stirred tank reactor

    NASA Astrophysics Data System (ADS)

    Yussof, H. W.; Bahri, S. S.; Mazlan, N. A.

    2018-03-01

    A recent development in oscillatory baffled reactor technology is down-scaling the reactor, so that it can be used for production of small-scale bioproduct. In the present study, a mesoscale oscillatory baffled reactor (MOBR) with central baffle system was developed. The reactor performance of the MOBR was compared with conventional stirred tank reactor (STR) to evaluate the performance of bioethanol fermentation using Saccharomyces cerevisiae. Evaluation was made at similar power density of 24.21, 57.38, 112.35 and 193.67 Wm-3 by varying frequency (f), amplitude (xo) and agitation speed (rpm). It was found that the MOBR improved the mixing intensity resulted in lower glucose concentration (0.988 gL-1) and higher bioethanol concentration (38.98 gL-1) after 12 hours fermentation at power density of 193.67 Wm-3. Based on the results, the bioethanol yield obtained using MOBR was 39% higher than the maximum achieved in STR. Bioethanol production using MOBR proved to be feasible as it is not only able to compete with conventional STR but also offers advantages of straight-forward scale-up, whereas it is complicated and difficult in STR. Overall, MOBR offers great prospective over the conventional STR.

  3. Inactive enzymatic mutant proteins (phosphoglycerate mutase and enolase) as sugar binders for ribulose-1,5-bisphosphate regeneration reactors

    PubMed Central

    De, Debojyoti; Dutta, Debajyoti; Kundu, Moloy; Mahato, Sourav; Schiavone, Marc T; Chaudhuri, Surabhi; Giri, Ashok; Gupta, Vidya; Bhattacharya, Sanjoy K

    2005-01-01

    Background Carbon dioxide fixation bioprocess in reactors necessitates recycling of D-ribulose1,5-bisphosphate (RuBP) for continuous operation. A radically new close loop of RuBP regenerating reactor design has been proposed that will harbor enzyme-complexes instead of purified enzymes. These reactors will need binders enabling selective capture and release of sugar and intermediate metabolites enabling specific conversions during regeneration. In the current manuscript we describe properties of proteins that will act as potential binders in RuBP regeneration reactors. Results We demonstrate specific binding of 3-phosphoglycerate (3PGA) and 3-phosphoglyceraldehyde (3PGAL) from sugar mixtures by inactive mutant of yeast enzymes phosphoglycerate mutase and enolase. The reversibility in binding with respect to pH and EDTA has also been shown. No chemical conversion of incubated sugars or sugar intermediate metabolites were found by the inactive enzymatic proteins. The dissociation constants for sugar metabolites are in the micromolar range, both proteins showed lower dissociation constant (Kd) for 3-phosphoglycerate (655–796 μM) compared to 3-phosphoglyceraldehyde (822–966 μM) indicating higher affinity for 3PGA. The proteins did not show binding to glucose, sucrose or fructose within the sensitivity limits of detection. Phosphoglycerate mutase showed slightly lower stability on repeated use than enolase mutants. Conclusions The sugar and their intermediate metabolite binders may have a useful role in RuBP regeneration reactors. The reversibility of binding with respect to changes in physicochemical factors and stability when subjected to repeated changes in these conditions are expected to make the mutant proteins candidates for in-situ removal of sugar intermediate metabolites for forward driving of specific reactions in enzyme-complex reactors. PMID:15689239

  4. 78 FR 64028 - Decommissioning of Nuclear Power Reactors

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-10-25

    ... NUCLEAR REGULATORY COMMISSION [NRC-2012-0035] Decommissioning of Nuclear Power Reactors AGENCY... Commission (NRC) is issuing Revision 1 of regulatory guide (RG) 1.184 ``Decommissioning of Nuclear Power... the NRC's regulations relating to the decommissioning process for nuclear power reactors. The revision...

  5. EBT reactor systems analysis and cost code: description and users guide (Version 1)

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

    Santoro, R.T.; Uckan, N.A.; Barnes, J.M.

    1984-06-01

    An ELMO Bumpy Torus (EBT) reactor systems analysis and cost code that incorporates the most recent advances in EBT physics has been written. The code determines a set of reactors that fall within an allowed operating window determined from the coupling of ring and core plasma properties and the self-consistent treatment of the coupled ring-core stability and power balance requirements. The essential elements of the systems analysis and cost code are described, along with the calculational sequences leading to the specification of the reactor options and their associated costs. The input parameters, the constraints imposed upon them, and the operatingmore » range over which the code provides valid results are discussed. A sample problem and the interpretation of the results are also presented.« less

  6. 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

  7. Health physics aspects of advanced reactor licensing reviews

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

    Hinson, C.S.

    1995-03-01

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

  8. 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

  9. 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

  10. 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.

  11. 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

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

    PubMed Central

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

    2014-01-01

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

  13. 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

  14. 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

  15. An investigation of reactivity effect due to inadvertent filling of the irradiation channels with water in NIRR-1 Nigeria Research Reactor-1.

    PubMed

    Iliyasu, U; Ibrahim, Y V; Umar, Sadiq; Agbo, S A; Jibrin, Y

    2017-05-01

    Investigation of reactivity variation due to flooding of the irradiation channels of Nigeria Research Reactor (NIRR-1) a low power miniature neutron source reactor (MNSR) located at the Centre for Energy Research and Training, Ahmadu Bello University, Zaria Nigeria using the MCNP code for High Enrich Uranium (HEU) and Low Enrich Uranium (LEU) core has been simulated in this present study. In this work, the excess reactivity worth of flooding HEU core for 1 inner, 2 inner, 3 inner, 4 inner and all inner are 0.318mk, 0.577mk, 0.318mk, 1.204mk and 1.503mk respectively, and outer irradiation channels are 0.119mk, 0.169mk, 0.348mk, 0.438mk and 0.418mk respectively, the highest excess reactivity result from flooding both inner and outer irradiation channels is 2.04mk (±1.72×10 -7 ), the excess reactivity for LEU core was 0.299mk, 0.568mk, 0.896mk, 1.195mk and 1.524mk in the inner irradiation channels, and the outer irradiation channels are 0.129mk, 0.189mk, 0.219mk, 0.269mk and 0.548mk where the highest excess reactivity was 1.942mk (±1.64×10 -7 ) resulting from flooding inner and outer irradiation channels. The reactivity induced by flooding of the irradiation channels of NIRR-1 with water is within design safety limit enshrined in Safety Analysis Report of NIRR-1. The results also compare well with literature. Copyright © 2017 Elsevier Ltd. All rights reserved.

  16. 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. .

  17. 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.

  18. Pressurized-water reactor internals aging degradation study. Phase 1

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

    Luk, K.H.

    1993-09-01

    This report documents the results of a Phase I study on the effects of aging degradations on pr internals. Primary stressers for internals an generated by the primary coolant flow in the they include unsteady hydrodynamic forces and pump-generated pressure pulsations. Other stressors are applied loads, manufacturing processes, impurities in the coolant and exposures to fast neutron fluxes. A survey of reported aging-related failure information indicates that fatigue, stress corrosion cracking (SCC) and mechanical wear are the three major aging-related degradation mechanisms for PWR internals. Significant reported failures include thermal shield flow-induced vibration problems, SCC in guide tube support pinsmore » and core support structure bolts, fatigue-induced core baffle water-jet impingement problems and excess wear in flux thimbles. Many of the reported problems have been resolved by accepted engineering practices. Uncertainties remain in the assessment of long-term neutron irradiation effects and environmental factors in high-cycle fatigue failures. Reactor internals are examined by visual inspections and the technique is access limited. Improved inspection methods, especially one with an early failure detection capability, can enhance the safety and efficiency of reactor operations.« less

  19. 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.

  20. 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

  1. 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.

  2. 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.

  3. The Experimental Breeder Reactor II seismic probabilistic risk assessment

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

    Roglans, J; Hill, D J

    1994-02-01

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

  4. 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

  5. Evaluation of a Method for Remote Detection of Fuel Relocation Outside the Original Core Volumes of Fukushima Reactor Units 1-3

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

    Douglas W. Akers; Edwin A. Harvego

    2012-08-01

    This paper presents the results of a study to evaluate the feasibility of remotely detecting and quantifying fuel relocation from the core to the lower head, and to regions outside the reactor vessel primary containment of the Fukushima 1-3 reactors. The goals of this study were to determine measurement conditions and requirements, and to perform initial radiation transport sensitivity analyses for several potential measurement locations inside the reactor building. The radiation transport sensitivity analyses were performed based on reactor design information for boiling water reactors (BWRs) similar to the Fukushima reactors, ORIGEN2 analyses of 3-cycle BWR fuel inventories, and datamore » on previously molten fuel characteristics from TMI- 2. A 100 kg mass of previously molten fuel material located on the lower head of the reactor vessel was chosen as a fuel interrogation sensitivity target. Two measurement locations were chosen for the transport analyses, one inside the drywell and one outside the concrete biological shield surrounding the drywell. Results of these initial radiation transport analyses indicate that the 100 kg of previously molten fuel material may be detectable at the measurement location inside the drywell, but that it is highly unlikely that any amount of fuel material inside the RPV will be detectable from a location outside the concrete biological shield surrounding the drywell. Three additional fuel relocation scenarios were also analyzed to assess detection sensitivity for varying amount of relocated material in the lower head of the reactor vessel, in the control rods perpendicular to the detector system, and on the lower head of the drywell. Results of these analyses along with an assessment of background radiation effects and a discussion of measurement issues, such as the detector/collimator design, are included in the paper.« less

  6. 10 CFR 140.11 - Amounts of financial protection for certain reactors.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ...,000,000 for each nuclear reactor he is authorized to operate at a thermal power level not exceeding ten kilowatts; (2) In the amount of $1,500,000 for each nuclear reactor he is authorized to operate at... amount of $2,500,000 for each nuclear reactor other than a testing reactor or a reactor licensed under...

  7. 10 CFR 140.11 - Amounts of financial protection for certain reactors.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ...,000,000 for each nuclear reactor he is authorized to operate at a thermal power level not exceeding ten kilowatts; (2) In the amount of $1,500,000 for each nuclear reactor he is authorized to operate at... amount of $2,500,000 for each nuclear reactor other than a testing reactor or a reactor licensed under...

  8. 10 CFR 140.11 - Amounts of financial protection for certain reactors.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ...,000,000 for each nuclear reactor he is authorized to operate at a thermal power level not exceeding ten kilowatts; (2) In the amount of $1,500,000 for each nuclear reactor he is authorized to operate at... amount of $2,500,000 for each nuclear reactor other than a testing reactor or a reactor licensed under...

  9. 76 FR 23630 - Office of New Reactors; Proposed Revision 2 to Standard Review Plan, Section 1.0 on Introduction...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-04-27

    ... Standard Review Plan, Section 1.0 on Introduction and Interfaces AGENCY: Nuclear Regulatory Commission (NRC... Revision 2 to Standard Review Plan (SRP), Section 1.0, ``Introduction and Interfaces'' (Agencywide Documents Access and Management System (ADAMS) Accession No. ML110110573). The Office of New Reactors (NRO...

  10. Reactor for exothermic reactions

    DOEpatents

    Smith, Jr., Lawrence A.; Hearn, Dennis; Jones, Jr., Edward M.

    1993-01-01

    A liquid phase process for oligomerization of C.sub.4 and C.sub.5 isoolefins or the etherification thereof with C.sub.1 to C.sub.6 alcohols wherein the reactants are contacted in a reactor with a fixed bed acid cation exchange resin catalyst at an LHSV of 5 to 20, pressure of 0 to 400 psig and temperature of 120.degree. to 300.degree. F. Wherein the improvement is the operation of the reactor at a pressure to maintain the reaction mixture at its boiling point whereby at least a portion but less than all of the reaction mixture is vaporized. By operating at the boiling point and allowing a portion of the reaction mixture to vaporize, the exothermic heat of reaction is dissipated by the formation of more boil up and the temperature in the reactor is controlled.

  11. Reactor for exothermic reactions

    DOEpatents

    Smith, L.A. Jr.; Hearn, D.; Jones, E.M. Jr.

    1993-03-02

    A liquid phase process is described for oligomerization of C[sub 4] and C[sub 5] isoolefins or the etherification thereof with C[sub 1] to C[sub 6] alcohols wherein the reactants are contacted in a reactor with a fixed bed acid cation exchange resin catalyst at an LHSV of 5 to 20, pressure of 0 to 400 psig and temperature of 120 to 300 F. Wherein the improvement is the operation of the reactor at a pressure to maintain the reaction mixture at its boiling point whereby at least a portion but less than all of the reaction mixture is vaporized. By operating at the boiling point and allowing a portion of the reaction mixture to vaporize, the exothermic heat of reaction is dissipated by the formation of more boil up and the temperature in the reactor is controlled.

  12. Biological hydrogen production by Clostridium acetobutylicum in an unsaturated flow reactor.

    PubMed

    Zhang, Husen; Bruns, Mary Ann; Logan, Bruce E

    2006-02-01

    A mesophilic unsaturated flow (trickle bed) reactor was designed and tested for H2 production via fermentation of glucose. The reactor consisted of a column packed with glass beads and inoculated with a pure culture (Clostridium acetobutylicum ATCC 824). A defined medium containing glucose was fed at a flow rate of 1.6 mL/min (0.096 L/h) into the capped reactor, producing a hydraulic retention time of 2.1 min. Gas-phase H2 concentrations were constant, averaging 74 +/- 3% for all conditions tested. H2 production rates increased from 89 to 220 mL/hL of reactor when influent glucose concentrations were varied from 1.0 to 10.5 g/L. Specific H2 production rate ranged from 680 to 1270 mL/g glucose per liter of reactor (total volume). The H2 yield was 15-27%, based on a theoretical limit by fermentation of 4 moles of H2 from 1 mole of glucose. The major fermentation by-products in the liquid effluent were acetate and butyrate. The reactor rapidly (within 60-72 h) became clogged with biomass, requiring manual cleaning of the system. In order to make long-term operation of the reactor feasible, biofilm accumulation in the reactor will need to be controlled through some process such as backwashing. These tests using an unsaturated flow reactor demonstrate the feasibility of the process to produce high H2 gas concentrations in a trickle-bed type of reactor. A likely application of this reactor technology could be H2 gas recovery from pre-treatment of high carbohydrate-containing wastewaters.

  13. Nitrate removal with lateral flow sulphur autotrophic denitrification reactor.

    PubMed

    Lv, Xiaomei; Shao, Mingfei; Li, Ji; Xie, Chuanbo

    2014-01-01

    An innovative lateral flow sulphur autotrophic denitrification (LFSAD) reactor was developed in this study; the treatment performance was evaluated and compared with traditional sulphur/limestone autotrophic denitrification (SLAD) reactor. Results showed that nitrite accumulation in the LFSAD reactor was less than 1.0 mg/L during the whole operation. Denitrification rate increased with the increased initial alkalinity and was approaching saturation when initial alkalinity exceeded 2.5 times the theoretical value. Higher influent nitrate concentration could facilitate nitrate removal capacity. In addition, denitrification efficiency could be promoted under an appropriate reflux ratio, and the highest nitrate removal percentage was achieved under reflux ratio of 200%, increased by 23.8% than that without reflux. Running resistance was only about 1/9 of that in SLAD reactor with equal amount of nitrate removed, which was the prominent excellence of the new reactor. In short, this study indicated that the developed reactor was feasible for nitrate removal from waters with lower concentrations, including contaminated surface water, groundwater or secondary effluent of municipal wastewater treatment with fairly low running resistance. The innovation in reactor design in this study may bring forth new ideas of reactor development of sulphur autotrophic denitrification for nitrate-contaminated water treatment.

  14. 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.

  15. ETRCF, TRA654, INTERIOR. REACTOR OPERATED IN WATERFILLED TANK. CAMERA LOOKS ...

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

    ETR-CF, TRA-654, INTERIOR. REACTOR OPERATED IN WATER-FILLED TANK. CAMERA LOOKS DOWN FROM ABOVE UPON LATER (NON-NUCLEAR) EXPERIMENTAL GEAR. INL NEGATIVE NO. HD24-1-1. Mike Crane, Photographer, ca. 2003 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

  16. 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.

  17. 9 CFR 78.22 - Brucellosis reactor bison.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... 9 Animals and Animal Products 1 2014-01-01 2014-01-01 false Brucellosis reactor bison. 78.22... AGRICULTURE INTERSTATE TRANSPORTATION OF ANIMALS (INCLUDING POULTRY) AND ANIMAL PRODUCTS BRUCELLOSIS Restrictions on Interstate Movement of Bison Because of Brucellosis § 78.22 Brucellosis reactor bison. (a...

  18. 9 CFR 78.22 - Brucellosis reactor bison.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... 9 Animals and Animal Products 1 2013-01-01 2013-01-01 false Brucellosis reactor bison. 78.22... AGRICULTURE INTERSTATE TRANSPORTATION OF ANIMALS (INCLUDING POULTRY) AND ANIMAL PRODUCTS BRUCELLOSIS Restrictions on Interstate Movement of Bison Because of Brucellosis § 78.22 Brucellosis reactor bison. (a...

  19. 9 CFR 78.7 - Brucellosis reactor cattle.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

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

  20. 9 CFR 78.7 - Brucellosis reactor cattle.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

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

  1. 9 CFR 78.7 - Brucellosis reactor cattle.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

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

  2. 9 CFR 78.22 - Brucellosis reactor bison.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... 9 Animals and Animal Products 1 2012-01-01 2012-01-01 false Brucellosis reactor bison. 78.22... AGRICULTURE INTERSTATE TRANSPORTATION OF ANIMALS (INCLUDING POULTRY) AND ANIMAL PRODUCTS BRUCELLOSIS Restrictions on Interstate Movement of Bison Because of Brucellosis § 78.22 Brucellosis reactor bison. (a...

  3. Target-fueled nuclear reactor for medical isotope production

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

    Coats, Richard L.; Parma, Edward J.

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

  4. Pentachlorophenol (PCP) dechlorination in horizontal-flow anaerobic immobilized biomass (HAIB) reactors.

    PubMed

    Damianovic, M H R Z; Moraes, E M; Zaiat, M; Foresti, E

    2009-10-01

    This study verifies the potential applicability of horizontal-flow anaerobic immobilized biomass (HAIB) reactors to pentachlorophenol (PCP) dechlorination. Two bench-scale HAIB reactors (R1 and R2) were filled with cubic polyurethane foam matrices containing immobilized anaerobic sludge. The reactors were then continuously fed with synthetic wastewater consisting of PCP, glucose, acetic acid, and formic acid as co-substrates for PCP anaerobic degradation. Before being immobilized in polyurethane foam matrices, the biomass was exposed to wastewater containing PCP in reactors fed at a semi-continuous rate of 2.0 microg PCP g(-1) VS. The applied PCP loading rate was increased from 0.05 to 2.59 mg PCP l(-1)day(-1) for R1, and from 0.06 to 4.15 mg PCP l(-1)day(-1) for R2. The organic loading rates (OLR) were 1.1 and 1.7 kg COD m(-3)day(-1) at hydraulic retention times (HRT) of 24h for R1 and 18 h for R2. Under such conditions, chemical oxygen demand (COD) removal efficiencies of up to 98% were achieved in the HAIB reactors. Both reactors exhibited the ability to remove 97% of the loaded PCP. Dichlorophenol (DCP) was the primary chlorophenol detected in the effluent. The adsorption of PCP and metabolites formed during PCP degradation in the packed bed was negligible for PCP removal efficiency.

  5. Nuclear reactors built, being built, or planned, 1991

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

    Simpson, B.

    1992-07-01

    This document contains unclassified information about facilities built, being built, or planned in the United States for domestic use or export as of December 31, 1991. The book is divided into three major sections: Section 1 consists of a reactor locator map and reactor tables; Section 2 includes nuclear reactors that are operating, being built, or planned; and Section 3 includes reactors that have been shut down permanently or dismantled. Sections 2 and 3 contain the following classification of reactors: Civilian, Production, Military, Export, and Critical Assembly. Export reactor refers to a reactor for which the principal nuclear contractor ismore » an American company -- working either independently or in cooperation with a foreign company (Part 4, in each section). Critical assembly refers to an assembly of fuel and assembly of fuel and moderator that requires an external source of neutrons to initiate and maintain fission. A critical assembly is used for experimental measurements (Part 5).« less

  6. SPERT1. Contextual aerial view of SPERTI Reactor Pit Building (PER605) ...

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

    SPERT-1. Contextual aerial view of SPERT-I Reactor Pit Building (PER-605) at top of view, and its accessories: the earth-shielded instrument cell (PER-606) immediately adjacent to it; the Guard House (PER-607) to its right; and the Terminal Building in lower center of view (PER-604). Camera faces west. Road and buried line leaving view at right lead to Control Building (PER-601) out of view. Sagebrush vegetation has been scraped from around buildings. Photographer: R.G. Larsen. Date: June 6, 1955. INEEL negative no. 55-1477. - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

  7. Risk Management for Sodium Fast Reactors.

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

    Denman, Matthew R.; Groth, Katrina; Cardoni, Jeffrey N.

    2015-01-01

    Accident management is an important component to maintaining risk at acceptable levels for all complex systems, such as nuclear power plants. With the introduction of self - correcting, or inherently safe, reactor designs the focus has shifted from management by operators to allowing the syste m's design to manage the accident. While inherently and passively safe designs are laudable, extreme boundary conditions can interfere with the design attributes which facilitate inherent safety , thus resulting in unanticipated and undesirable end states. This report examines an inherently safe and small sodium fast reactor experiencing a beyond design basis seismic event withmore » the intend of exploring two issues : (1) can human intervention either improve or worsen the potential end states and (2) can a Bayes ian Network be constructed to infer the state of the reactor to inform (1). ACKNOWLEDGEMENTS The author s would like to acknowledge the U.S. Department of E nergy's Office of Nuclear Energy for funding this research through Work Package SR - 14SN100303 under the Advanced Reactor Concepts program. The authors also acknowledge the PRA teams at A rgonne N ational L aborator y , O ak R idge N ational L aborator y , and I daho N ational L aborator y for their continue d contributions to the advanced reactor PRA mission area.« less

  8. 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.

  9. 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

  10. In-reactor performance of LWR-type tritium target rods

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

    Lanning, D.D.; Paxton, M.M.; Crumbaugh, L.

    Pacific Northwest Laboratory has conducted several 1-yr irradiation tests of light water reactor-type tritium target rods. These tests have been sponsored by the U.S. Department of Energy's Office of New Production Reactors. The first test, designated water capsule-1 (WC-1), was conducted in the Advanced Test Reactor (ATR) at the Idaho National Engineering Laboratory from November 1989 to December 1990. The test vehicle contained a single 4-ft target rod within a pressurized water capsule. The capsule maintained the rod at pressurized water reactor (PWR)-type water temperature and pressure conditions. Posttest nondestructive examinations of the WC-1 rod involved visual examinations, dimensional checks,more » gamma scanning, and neutron radiography. The results indicate that the rod maintained the integrity of its pressure seal and was otherwise unaltered both mechanically and dimensionally by its irradiation and posttest handling.« less

  11. 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.

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

    PubMed

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

    2016-05-01

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

  13. 10 CFR 50.72 - Immediate notification requirements for operating nuclear power reactors.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... power reactors. 50.72 Section 50.72 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF... notification requirements for operating nuclear power reactors. (a) General requirements. 1 (1) Each nuclear... requirements for immediate notification of the NRC by licensed operating nuclear power reactors are contained...

  14. 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)

  15. 10 CFR 140.11 - Amounts of financial protection for certain reactors.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 false Amounts of financial protection for certain reactors. 140... reactors. (a) Each licensee is required to have and maintain financial protection: (1) In the amount of $1,000,000 for each nuclear reactor he is authorized to operate at a thermal power level not exceeding...

  16. 10 CFR 140.11 - Amounts of financial protection for certain reactors.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 10 Energy 2 2011-01-01 2011-01-01 false Amounts of financial protection for certain reactors. 140... reactors. (a) Each licensee is required to have and maintain financial protection: (1) In the amount of $1,000,000 for each nuclear reactor he is authorized to operate at a thermal power level not exceeding...

  17. 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 %).

  18. Characterization of cartridge filters from the IEA-R1 Nuclear Reactor

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

    NONE

    The management of radioactive waste ensures safety to human health and the environment nowadays and for the future, without overwhelming the upcoming generations. The primary characterization of radioactive waste is one of the main steps in the management of radioactive waste. This step permits to choose the best treatment for the radioactive waste before forwarding it to its final disposal. The aim of the present work is the primary characterization of cartridge filters from the IEA-R1 nuclear reactor utilizing gamma-ray spectrometry, and the method of Monte Carlo for calibration. The IEA-R1 is located in the Nuclear and Energy Research Institutemore » (IPEN - CNEN) in the city of Sao Paulo, Brazil. Cartridge filters are used for purification of the cooling water that is pumped through the core of the pool type nuclear research reactors. Once worn out, these filters are replaced and then become radioactive waste. Determination of the radioactive inventory is of paramount importance in the management of such radioactive waste, and one of the main methods for doing so is the gamma-ray spectrometry, which can identify and quantify high energy photon emitters. The technique chosen for the characterization of radioactive waste in the present work is the gamma-ray spectrometry with High purity Germanium (HPGe) detectors. From the energy identified in the experimental spectrum, three radioisotopes were identified in the cartridge filter: {sup 108m}Ag, {sup 110m}Ag, {sup 60}Co. For the estimated activity of the filter, the calibration in efficiency was made utilizing the MCNP4C code of the Monte Carlo method. Such method was chosen because there is no standard source available in the same geometry of the cartridge filter, therefore a simulation had to be developed in order to reach a calibration equation, necessary to estimate the activity of the radioactive waste. The results presented an activity value in the order of MBq for all radioisotopes. (authors)« less

  19. 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.

  20. Membrane-aerated biofilm reactor for the removal of 1,2-dichloroethane by Pseudomonas sp. strain DCA1.

    PubMed

    Hage, J C; Van Houten, R T; Tramper, J; Hartmans, S

    2004-06-01

    A membrane-aerated biofilm reactor (MBR) with a biofilm of Pseudomonas sp. strain DCA1 was studied for the removal of 1,2-dichloroethane (DCA) from water. A hydrophobic membrane was used to create a barrier between the liquid and the gas phase. Inoculation of the MBR with cells of strain DCA1 grown in a continuous culture resulted in the formation of a stable and active DCA-degrading biofilm on the membrane. The maximum removal rate of the MBR was reached at a DCA concentration of approximately 80 micro M. Simulation of the DCA fluxes into the biofilm showed that the MBR performance at lower concentrations was limited by the DCA diffusion rate rather than by kinetic constraints of strain DCA1. Aerobic biodegradation of DCA present in anoxic water could be achieved by supplying oxygen solely from the gas phase to the biofilm grown on the liquid side of the membrane. As a result, direct aeration of the water, which leads to undesired coagulation of iron oxides, could be avoided.

  1. 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.

  2. Extending the maximum operation time of the MNSR reactor.

    PubMed

    Dawahra, S; Khattab, K; Saba, G

    2016-09-01

    An effective modification to extend the maximum operation time of the Miniature Neutron Source Reactor (MNSR) to enhance the utilization of the reactor has been tested using the MCNP4C code. This modification consisted of inserting manually in each of the reactor inner irradiation tube a chain of three polyethylene-connected containers filled of water. The total height of the chain was 11.5cm. The replacement of the actual cadmium absorber with B(10) absorber was needed as well. The rest of the core structure materials and dimensions remained unchanged. A 3-D neutronic model with the new modifications was developed to compare the neutronic parameters of the old and modified cores. The results of the old and modified core excess reactivities (ρex) were: 3.954, 6.241 mk respectively. The maximum reactor operation times were: 428, 1025min and the safety reactivity factors were: 1.654 and 1.595 respectively. Therefore, a 139% increase in the maximum reactor operation time was noticed for the modified core. This increase enhanced the utilization of the MNSR reactor to conduct a long time irradiation of the unknown samples using the NAA technique and increase the amount of radioisotope production in the reactor. Copyright © 2016 Elsevier Ltd. All rights reserved.

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

    USGS Publications Warehouse

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

    2006-01-01

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

  4. Future Scenarios for Fission Based Reactors

    NASA Astrophysics Data System (ADS)

    David, S.

    2005-04-01

    The coming century will see the exhaustion of standard fossil fuels, coal, gas and oil, which today represent 75% of the world energy production. Moreover, their use will have caused large-scale emission of greenhouse gases (GEG), and induced global climate change. This problem is exacerbated by a growing world energy demand. In this context, nuclear power is the only GEG-free energy source available today capable of responding significantly to this demand. Some scenarios consider a nuclear energy production of around 5 Gtoe in 2050, wich would represent a 20% share of the world energy supply. Present reactors generate energy from the fission of U-235 and require around 200 tons of natural Uranium to produce 1GWe.y of energy, equivalent to the fission of one ton of fissile material. In a scenario of a significant increase in nuclear energy generation, these standard reactors will consume the whole of the world's estimated Uranium reserves in a few decades. However, natural Uranium or Thorium ore, wich are not themselves fissile, can produce a fissile material after a neutron capture ( 239Pu and 233U respectively). In a breeder reactor, the mass of fissile material remains constant, and the fertile ore is the only material to be consumed. In this case, only 1 ton of natural ore is needed to produce 1GWe.y. Thus, the breeding concept allows optimal use of fertile ore and development of sustainable nuclear energy production for several thousand years into the future. Different sustainable nuclear reactor concepts are studied in the international forum "generation IV". Different types of coolant (Na, Pb and He) are studied for fast breeder reactors based on the Uranium cycle. The thermal Thorium cycle requires the use of a liquid fuel, which can be reprocessed online in order to extract the neutron poisons. This paper presents these different sustainable reactors, based on the Uranium or Thorium fuel cycles and will compare the different options in term of fissile

  5. 76 FR 39922 - Office of New Reactors; Proposed Revision 4 to Standard Review Plan Section 8.1 on Electric Power...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-07-07

    ... the proposed Revision 4 to Standard Review Plan Section 8.1 on ``Electric Power-- Introduction.'' The... NUCLEAR REGULATORY COMMISSION [NRC-2011-0119] Office of New Reactors; Proposed Revision 4 to Standard Review Plan Section 8.1 on Electric Power--Introduction, Correction AGENCY: Nuclear Regulatory...

  6. Reactor Monitoring with Antineutrinos - A Progress Report

    NASA Astrophysics Data System (ADS)

    Bernstein, Adam

    2012-08-01

    The Reactor Safeguards regime is the name given to a set of protocols and technologies used to monitor the consumption and production of fissile materials in nuclear reactors. The Safeguards regime is administered by the International Atomic Energy Agency (IAEA), and is an essential component of the global Treaty on Nuclear Nonproliferation, recently renewed by its 189 remaining signators. (The 190th, North Korea, withdrew from the Treaty in 2003). Beginning in Russia in the 1980s, a number of researchers worldwide have experimentally demonstrated the potential of cubic meter scale antineutrino detectors for non-intrusive real-time monitoring of fissile inventories and power output of reactors. The detectors built so far have operated tens of meters from a reactor core, outside of the containment dome, largely unattended and with remote data acquisition for an entire 1.5 year reactor cycle, and have achieved levels of sensitivity to fissile content of potential interest for the IAEA safeguards regime. In this article, I will describe the unique advantages of antineutrino detectors for cooperative monitoring, consider the prospects and benefits of increasing the range of detectability for small reactors, and provide a partial survey of ongoing global research aimed at improving near-field and far field monitoring and discovery of nuclear reactors.

  7. PARTIAL ECONOMIC STUDY OF STEAM COOLED HEAVY WATER MODERATED REACTORS

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

    None

    1960-04-01

    Steam-cooled reactors are compared with CAHDU for costs of Calandria tubes, pressure tubes. heavy water moderator, heavy water reflector, fuel supply, heat exchanger, and turbine generator. A direct-cycle lightsteam-cooled heavy- water-moderated pressure-tube reactor formed the basic reactor design for the study. Two methods of steam circulation through the reactor were examined. In both cases the steam was generated outside the reactor and superheated in the reactor core. One method consisted of a series of reactor and steam generator passes. The second method consisted of the Loeffler cycle and its modifications. The fuel was assumed to be natural cylindrical UO/sub 2/more » pellets sheathed in a hypothetical material with the nuclear properties of Zircaloy, but able to function at temperatures to 900 deg F. For the conditions assumed, the longer the rod, the higher the outlet temperature and therefore the higher the efficiency. The turbine cycle efficiency was calculated on the assumption that suitable steam generators are available. As the neutron losses to the pressure tubes were significant, an economic analysis of insulated pressure tubes is included. A description of the physics program for steam-cooled reactors is included. Results indicated that power from the steam-cooled reactor would cost 1.4 mills/ kwh compared with 1.25 mills/kwh for CANDU. (M.C.G.)« less

  8. 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.

  9. 10 CFR 50.72 - Immediate notification requirements for operating nuclear power reactors.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... power reactors. 50.72 Section 50.72 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF... notification requirements for operating nuclear power reactors. (a) General requirements. 1 (1) Each nuclear power reactor licensee licensed under §§ 50.21(b) or 50.22 holding an operating license under this part...

  10. 10 CFR 50.72 - Immediate notification requirements for operating nuclear power reactors.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... power reactors. 50.72 Section 50.72 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF... notification requirements for operating nuclear power reactors. (a) General requirements. 1 (1) Each nuclear power reactor licensee licensed under §§ 50.21(b) or 50.22 holding an operating license under this part...

  11. 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.

  12. 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.

  13. 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.

  14. 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

  15. 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)

  16. Determination of neutron spectra within the energy of 1 keV to 1 MeV by means of reactor dosimetry

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

    Sergeyeva, Victoria; Destouches, Christophe; Lyoussi, Abdallah

    2015-07-01

    The standard procedure for neutron reactor dosimetry is based on neutron irradiation of a target and its post-irradiation analysis by Gamma and/or X-ray spectrometry. Nowadays, the neutron spectra can be easily characterized for thermal and fast energies (respectively 0.025 eV and >1 MeV). In this work we propose a new target and an innovating post-irradiation technique of analysis in order to detect the neutron spectra within the energy of 1 keV to 1 MeV. This article will present the calculations performed for the selection of a suitable nuclear reaction and isotope, the results predicted by simulations, the irradiation campaign thatmore » is proposed and the post-irradiation technique of analysis. (authors)« less

  17. Nuclear reactors built, being built, or planned 1993

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

    Not Available

    1993-08-01

    Nuclear Reactors Built, Being Built, or Planned contains unclassified information about facilities built, being built, or planned in the United States for domestic use or export as of December 31, 1993. The Office of Scientific and Technical Information, US Department of Energy, gathers this information annually from Washington headquarters and field offices of DOE; from the US Nuclear Regulatory Commission (NRC); from the US reactor manufacturers who are the principal nuclear embassies; and from foreign governmental nuclear departments. The book consists of three divisions, as follows: (1) a commercial reactor locator map and tables of the characteristic and statistical datamore » that follow; a table of abbreviations; (2) tables of data for reactors operating, being built, or planned; and (3) tables of data for reactors that have been shut down permanently or dismantled. The reactors are subdivided into the following parts: civilian, production, military, export, and critical assembly.« less

  18. Boiling-Water Reactor internals aging degradation study. Phase 1

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

    Luk, K.H.

    1993-09-01

    This report documents the results of an aging assessment study for boiling water reactor (BWR) internals. Major stressors for BWR internals are related to unsteady hydrodynamic forces generated by the primary coolant flow in the reactor vessel. Welding and cold-working, dissolved oxygen and impurities in the coolant, applied loads and exposures to fast neutron fluxes are other important stressors. Based on results of a component failure information survey, stress corrosion cracking (SCC) and fatigue are identified as the two major aging-related degradation mechanisms for BWR internals. Significant reported failures include SCC in jet-pump holddown beams, in-core neutron flux monitor drymore » tubes and core spray spargers. Fatigue failures were detected in feedwater spargers. The implementation of a plant Hydrogen Water Chemistry (HWC) program is considered as a promising method for controlling SCC problems in BWR. More operating data are needed to evaluate its effectiveness for internal components. Long-term fast neutron irradiation effects and high-cycle fatigue in a corrosive environment are uncertainty factors in the aging assessment process. BWR internals are examined by visual inspections and the method is access limited. The presence of a large water gap and an absence of ex-core neutron flux monitors may handicap the use of advanced inspection methods, such as neutron noise vibration measurements, for BWR.« less

  19. NETL - Chemical Looping Reactor

    ScienceCinema

    None

    2018-02-14

    NETL's Chemical Looping Reactor unit is a high-temperature integrated CLC process with extensive instrumentation to improve computational simulations. A non-reacting test unit is also used to study solids flow at ambient temperature. The CLR unit circulates approximately 1,000 pounds per hour at temperatures around 1,800 degrees Fahrenheit.

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

    NASA Astrophysics Data System (ADS)

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

    2014-11-01

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

  1. NUCLEAR REACTOR COOLANT

    DOEpatents

    Colichman, E.L.

    1959-10-20

    The formation of new reactor coolants which suppress polymerization resulting from pyrolitic and radiation decomposition is described. The coolants consist of polyphenyls and condensed ring compounds having from two to about four carbon rings and from 0.1 to about 10% of an alkall metal dispersed in the hydrocarbon.

  2. NUCLEAR REACTOR COOLANT

    DOEpatents

    Colichman, E.L.

    1959-10-20

    The formation of new reactor coolants which suppress polymerization resulting from pyrolytic and radiation decomposition is described. The coolants consist of polyphenyls and condensed ring compounds having from two to about four carbon rings and from 0.1 to about 5% of beryllium or magnesium dispersed in the hydrocarbon.

  3. Laboratory instrumentation modernization at the WPI Nuclear Reactor Facility

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

    Not Available

    1995-01-01

    With partial funding from the Department of Energy (DOE) University Reactor Instrumentation Program several laboratory instruments utilized by students and researchers at the WPI Nuclear Reactor Facility have been upgraded or replaced. 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 undergraduate 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 low power output of the reactor and an ergonomicmore » facility design make it an ideal tool for undergraduate nuclear engineering education and other training. The reactor, its control system, and the associate laboratory equipment are all located in the same room. Over the years, several important milestones have taken place at the WPI reactor. In 1969, the reactor power level was upgraded from 1 kW to 10 kW. The reactor`s Nuclear Regulatory Commission operating license was renewed for 20 years in 1983. In 1988, under DOE Grant No. DE-FG07-86ER75271, the reactor was converted to low-enriched uranium fuel. In 1992, again with partial funding from DOE (Grant No. DE-FG02-90ER12982), the original control console was replaced.« less

  4. 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

  5. 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.

  6. 10 CFR 50.46a - Acceptance criteria for reactor coolant system venting systems.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 10 Energy 1 2010-01-01 2010-01-01 false Acceptance criteria for reactor coolant system venting... criteria for reactor coolant system venting systems. Each nuclear power reactor must be provided with high point vents for the reactor coolant system, for the reactor vessel head, and for other systems required...

  7. 10 CFR 50.46a - Acceptance criteria for reactor coolant system venting systems.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 10 Energy 1 2011-01-01 2011-01-01 false Acceptance criteria for reactor coolant system venting... criteria for reactor coolant system venting systems. Each nuclear power reactor must be provided with high point vents for the reactor coolant system, for the reactor vessel head, and for other systems required...

  8. 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.

  9. Two conceptual designs of helical fusion reactor FFHR-d1A based on ITER technologies and challenging ideas

    NASA Astrophysics Data System (ADS)

    Sagara, A.; Miyazawa, J.; Tamura, H.; Tanaka, T.; Goto, T.; Yanagi, N.; Sakamoto, R.; Masuzaki, S.; Ohtani, H.; The FFHR Design Group

    2017-08-01

    The Fusion Engineering Research Project (FERP) at the National Institute for Fusion Science (NIFS) is conducting conceptual design activities for the LHD-type helical fusion reactor FFHR-d1A. This paper newly defines two design options, ‘basic’ and ‘challenging.’ Conservative technologies, including those that will be demonstrated in ITER, are chosen in the basic option in which two helical coils are made of continuously wound cable-in-conduit superconductors of Nb3Sn strands, the divertor is composed of water-cooled tungsten monoblocks, and the blanket is composed of water-cooled ceramic breeders. In contrast, new ideas that would possibly be beneficial for making the reactor design more attractive are boldly included in the challenging option in which the helical coils are wound by connecting high-temperature REBCO superconductors using mechanical joints, the divertor is composed of a shower of molten tin jets, and the blanket is composed of molten salt FLiNaBe including Ti powers to increase hydrogen solubility. The main targets of the challenging option are early construction and easy maintenance of a large and three-dimensionally complicated helical structure, high thermal efficiency, and, in particular, realistic feasibility of the helical reactor.

  10. Nuclear reactors built, being built, or planned: 1995

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

    NONE

    1996-08-01

    This report contains unclassified information about facilities built, being built, or planned in the US for domestic use or export as of December 31, 1995. The Office of Scientific and Technical Information, US Department of Energy, gathers this information annually from Washington headquarters and field offices of DOE; from the US Nuclear Regulatory Commission (NRC); from the US reactor manufacturers who are the principal nuclear contractors for foreign reactor locations; from US and foreign embassies; and from foreign governmental nuclear departments. The book consists of three divisions, as follows: (1) a commercial reactor locator map and tables of the characteristicmore » and statistical data that follow; a table of abbreviations; (2) tables of data for reactors operating, being built, or planned; and (3) tables of data for reactors that have been shut down permanently or dismantled. The reactors are subdivided into the following parts: Civilian, Production, Military, Export, and Critical Assembly. Export reactor refers to a reactor for which the principal nuclear contractor is a US company--working either independently or in cooperation with a foreign company (Part 4). Critical assembly refers to an assembly of fuel and moderator that requires an external source of neutrons to initiate and maintain fission. A critical assembly is used for experimental measurements (Part 5).« less

  11. 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."

  12. 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

  13. Micro-reactors for characterization of nanostructure-based sensors.

    PubMed

    Savu, R; Silveira, J V; Flacker, A; Vaz, A R; Joanni, E; Pinto, A C; Gobbi, A L; Santos, T E A; Rotondaro, A L P; Moshkalev, S A

    2012-05-01

    Fabrication and testing of micro-reactors for the characterization of nanosensors is presented in this work. The reactors have a small volume (100 μl) and are equipped with gas input/output channels. They were machined from a single piece of kovar in order to avoid leaks in the system due to additional welding. The contact pins were electrically insulated from the body of the reactor using a borosilicate sealing glass and the reactor was hermetically sealed using a lid and an elastomeric o-ring. One of the advantages of the reactor lies in its simple assembly and ease of use with any vacuum/gas system, allowing the connection of more than one device. Moreover, the lid can be modified in order to fit a window for in situ optical characterization. In order to prove its versatility, carbon nanotube-based sensors were tested using this micro-reactor. The devices were fabricated by depositing carbon nanotubes over 1 μm thick gold electrodes patterned onto Si/SiO(2) substrates. The sensors were tested using oxygen and nitrogen atmospheres, in the pressure range between 10(-5) and 10(-1) mbar. The small chamber volume allowed the measurement of fast sensor characteristic times, with the sensors showing good sensitivity towards gas and pressure as well as high reproducibility.

  14. Micro-reactors for characterization of nanostructure-based sensors

    NASA Astrophysics Data System (ADS)

    Savu, R.; Silveira, J. V.; Flacker, A.; Vaz, A. R.; Joanni, E.; Pinto, A. C.; Gobbi, A. L.; Santos, T. E. A.; Rotondaro, A. L. P.; Moshkalev, S. A.

    2012-05-01

    Fabrication and testing of micro-reactors for the characterization of nanosensors is presented in this work. The reactors have a small volume (100 μl) and are equipped with gas input/output channels. They were machined from a single piece of kovar in order to avoid leaks in the system due to additional welding. The contact pins were electrically insulated from the body of the reactor using a borosilicate sealing glass and the reactor was hermetically sealed using a lid and an elastomeric o-ring. One of the advantages of the reactor lies in its simple assembly and ease of use with any vacuum/gas system, allowing the connection of more than one device. Moreover, the lid can be modified in order to fit a window for in situ optical characterization. In order to prove its versatility, carbon nanotube-based sensors were tested using this micro-reactor. The devices were fabricated by depositing carbon nanotubes over 1 μm thick gold electrodes patterned onto Si/SiO2 substrates. The sensors were tested using oxygen and nitrogen atmospheres, in the pressure range between 10-5 and 10-1 mbar. The small chamber volume allowed the measurement of fast sensor characteristic times, with the sensors showing good sensitivity towards gas and pressure as well as high reproducibility.

  15. 40 CFR 63.1407 - Non-reactor batch process vent provisions.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 40 Protection of Environment 11 2010-07-01 2010-07-01 true Non-reactor batch process vent... § 63.1407 Non-reactor batch process vent provisions. (a) Emission standards. (1) Owners or operators of non-reactor batch process vents located at new or existing affected sources with 0.25 tons per year (0...

  16. 40 CFR 63.1407 - Non-reactor batch process vent provisions.

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ... 40 Protection of Environment 11 2011-07-01 2011-07-01 false Non-reactor batch process vent... § 63.1407 Non-reactor batch process vent provisions. (a) Emission standards. (1) Owners or operators of non-reactor batch process vents located at new or existing affected sources with 0.25 tons per year (0...

  17. Results on the neutron energy distribution measurements at the RECH-1 Chilean nuclear reactor

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

    Aguilera, P., E-mail: paguilera87@gmail.com; Romero-Barrientos, J.; Universidad de Chile, Dpto. de Física, Facultad de Ciencias, Las Palmeras 3425, Nuñoa, Santiago

    2016-07-07

    Neutron activations experiments has been perform at the RECH-1 Chilean Nuclear Reactor to measure its neutron flux energy distribution. Samples of pure elements was activated to obtain the saturation activities for each reaction. Using - ray spectroscopy we identify and measure the activity of the reaction product nuclei, obtaining the saturation activities of 20 reactions. GEANT4 and MCNP was used to compute the self shielding factor to correct the cross section for each element. With the Expectation-Maximization algorithm (EM) we were able to unfold the neutron flux energy distribution at dry tube position, near the RECH-1 core. In this work,more » we present the unfolding results using the EM algorithm.« less

  18. Convective cooling in a pool-type research reactor

    NASA Astrophysics Data System (ADS)

    Sipaun, Susan; Usman, Shoaib

    2016-01-01

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

  19. 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

  20. Performance of intermittent aeration reactor on NH4-N removal from groundwater resources.

    PubMed

    Khanitchaidecha, W; Nakamura, T; Sumino, T; Kazama, F

    2010-01-01

    To study the effect of intermittent aeration period on ammonium-nitrogen (NH4-N) removal from groundwater resources, synthetic groundwater was prepared and three reactors were operated under different conditions--"reactor A" under continuous aeration, "reactor B" under 6 h intermittent aeration, and "reactor C" under 2 h intermittent aeration. To facilitate denitrification simultaneously with nitrification, "acetate" was added as an external carbon source with step-wise increase from 0.5 to 1.5 C/N ratio, where C stands for total carbon content in the system, and N for NH4-N concentration in the synthetic groundwater. Results show that complete NH4-N removal was obtained in "reactor B" and "reactor C" at 1.3 and 1.5 C/N ratio respectively; and partial NH4-N removal in "reactor A". These results suggest that intermittent aeration at longer interval could enhance the reactor performance on NH4-N removal in terms of efficiency and low external carbon requirement. Because of consumption of internal carbon by the process, less amount of external carbon is required. Further increase in carbon in a form of acetate (1.5 to 2.5 C/N ratios) increases removal rate (represented by reaction rate coefficient (k) of kinetic equation) as well as occurrence of free cells. It suggests that the operating condition at reactor B with 1.3 C/N ratio is more appropriate for long-term operation at a pilot-scale.

  1. Measurement and Analysis of Structural Integrity of Reactor Core Support Structure in Pressurized Water Reactor (PWR) Plant

    NASA Astrophysics Data System (ADS)

    Ansari, Saleem A.; Haroon, Muhammad; Rashid, Atif; Kazmi, Zafar

    2017-02-01

    Extensive calculation and measurements of flow-induced vibrations (FIV) of reactor internals were made in a PWR plant to assess the structural integrity of reactor core support structure against coolant flow. The work was done to meet the requirements of the Fukushima Response Action Plan (FRAP) for enhancement of reactor safety, and the regulatory guide RG-1.20. For the core surveillance measurements the Reactor Internals Vibration Monitoring System (IVMS) has been developed based on detailed neutron noise analysis of the flux signals from the four ex-core neutron detectors. The natural frequencies, displacement and mode shapes of the reactor core barrel (CB) motion were determined with the help of IVMS. The random pressure fluctuations in reactor coolant flow due to turbulence force have been identified as the predominant cause of beam-mode deflection of CB. The dynamic FIV calculations were also made to supplement the core surveillance measurements. The calculational package employed the computational fluid dynamics, mode shape analysis, calculation of power spectral densities of flow & pressure fields and the structural response to random flow excitation forces. The dynamic loads and stiffness of the Hold-Down Spring that keeps the core structure in position against upward coolant thrust were also determined by noise measurements. Also, the boron concentration in primary coolant at any time of the core cycle has been determined with the IVMS.

  2. 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)

  3. 77 FR 64563 - Advisory Committee on Reactor Safeguards; Notice of Meeting

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-10-22

    ... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards; Notice of Meeting In... Advisory Committee on Reactor Safeguards (ACRS) will hold a meeting on November 1-3, 2012, 11545 Rockville...-Term Core Cooling Approach for the Advanced Boiling Water Reactor (ABWR) Design for South Texas Project...

  4. 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

  5. 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.

  6. 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.

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

    NASA Astrophysics Data System (ADS)

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

    2017-03-01

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

  8. Neutronic reactor construction

    DOEpatents

    Huston, Norman E.

    1976-07-06

    1. A neutronic reactor comprising a moderator including horizontal layers formed of horizontal rows of graphite blocks, alternate layers of blocks having the rows extending in one direction, the remaining alternate layers having the rows extending transversely to the said one direction, alternate rows of blocks in one set of alternate layers having longitudinal ducts, the moderator further including slotted graphite tubes positioned in the ducts, the reactor further comprising an aluminum coolant tube positioned within the slotted tube in spaced relation thereto, bodies of thermal-neutron-fissionable material, and jackets enclosing the bodies and being formed of a corrosion-resistant material having a low neutron-capture cross section, the bodies and jackets being positioned within the coolant tube so that the jackets are spaced from the coolant tube.

  9. 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.

  10. 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.

  11. Neutron Fluence And DPA Rate Analysis In Pebble-Bed HTR Reactor Vessel Using MCNP

    NASA Astrophysics Data System (ADS)

    Hamzah, Amir; Suwoto; Rohanda, Anis; Adrial, Hery; Bakhri, Syaiful; Sunaryo, Geni Rina

    2018-02-01

    In the Pebble-bed HTR reactor, the distance between the core and the reactor vessel is very close and the media inside are carbon and He gas. Neutron moderation capability of graphite material is theoretically lower than that of water-moderated reactors. Thus, it is estimated much more the fast neutrons will reach the reactor vessel. The fast neutron collisions with the atoms in the reactor vessel will result in radiation damage and could be reducing the vessel life. The purpose of this study was to obtain the magnitude of neutron fluence in the Pebble-bed HTR reactor vessel. Neutron fluence calculations in the pebble-bed HTR reactor vessel were performed using the MCNP computer program. By determining the tally position, it can be calculated flux, spectrum and neutron fluence in the position of Pebble-bed HTR reactor vessel. The calculations results of total neutron flux and fast neutron flux in the reactor vessel of 1.82x108 n/cm2/s and 1.79x108 n/cm2/s respectively. The fast neutron fluence in the reactor vessel is 3.4x1017 n/cm2 for 60 years reactor operation. Radiation damage in stainless steel material caused by high-energy neutrons (> 1.0 MeV) will occur when it has reached the neutron flux level of 1.0x1024 n/cm2. The neutron fluence results show that there is no radiation damage in the Pebble-bed HTR reactor vessel, so it is predicted that it will be safe to operate at least for 60 years.

  12. 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.

  13. Research and proposal on selective catalytic reduction reactor optimization for industrial boiler.

    PubMed

    Yang, Yiming; Li, Jian; He, Hong

    2017-08-24

    The advanced computational fluid dynamics (CFD) software STAR-CCM+ was used to simulate a denitrification (De-NOx) project for a boiler in this paper, and the simulation result was verified based on a physical model. Two selective catalytic reduction (SCR) reactors were developed: reactor 1 was optimized and reactor 2 was developed based on reactor 1. Various indicators, including gas flow field, ammonia concentration distribution, temperature distribution, gas incident angle, and system pressure drop were analyzed. The analysis indicated that reactor 2 was of outstanding performance and could simplify developing greatly. Ammonia injection grid (AIG), the core component of the reactor, was studied; three AIGs were developed and their performances were compared and analyzed. The result indicated that AIG 3 was of the best performance. The technical indicators were proposed for SCR reactor based on the study. Flow filed distribution, gas incident angle, and temperature distribution are subjected to SCR reactor shape to a great extent, and reactor 2 proposed in this paper was of outstanding performance; ammonia concentration distribution is subjected to ammonia injection grid (AIG) shape, and AIG 3 could meet the technical indicator of ammonia concentration without mounting ammonia mixer. The developments above on the reactor and the AIG are both of great application value and social efficiency.

  14. 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.

  15. Noncatalytic hydrogenation of decene-1 with hydrogen accumulated in a hybrid carbon nanostructure in nanosized membrane reactors

    NASA Astrophysics Data System (ADS)

    Soldatov, A. P.

    2014-08-01

    Studies on the creation of nanosized membrane reactors (NMRs) of a new generation with accumulated hydrogen and a regulated volume of reaction zone were continued at the next stage. Hydrogenation was performed in the pores of ceramic membranes with hydrogen preliminarily adsorbed in mono- and multilayered orientated carbon nanotubes with graphene walls (OCNTGs)—a new hybrid carbon nanostructure formed on the inner pore surface. Quantitative determination of hydrogen adsorption in OCNTGs was performed using TRUMEM ultrafiltration membranes with D av = 50 and 90 nm and showed that hydrogen adsorption was up to ˜1.5% of the mass of OCNTG. The instrumentation and procedure for noncatalytic hydrogenation of decene-1 at 250-350°C using hydrogen accumulated and stored in OCNTG were developed. The conversion of decene-1 into decane was ˜0.2-1.8% at hydrogenation temperatures of 250 and 350°C, respectively. The rate constants and activation energy of hydrogenation were determined. The latter was found to be 94.5 kJ/mol, which is much smaller than the values typical for noncatalytic hydrogenations and very close to the values characteristic for catalytic reactions. The quantitative distribution of the reacting compounds in each pore regarded as a nanosized membrane reactor was determined. The activity of hydrogen adsorbed in a 2D carbon nanostructure was evaluated. Possible mechanisms of noncatalytic hydrogenation were discussed.

  16. Control console replacement at the WPI Reactor. [Final report

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

    Not Available

    1992-12-31

    With partial funding from the Department of Energy (DOE) University Reactor Instrumentation Upgrade Program (DOE Grant No. DE-FG02-90ER12982), the original control console at the Worcester Polytechnic Institute (WPI) Reactor has been replaced with a modern system. The new console maintains the original design bases and functionality while utilizing current technology. An advanced remote monitoring system has been added to augment the educational capabilities of the reactor. Designed and built by General Electric in 1959, the open pool nuclear training reactor at WPI was one of the first such facilities in the nation located on a university campus. Devoted to undergraduatemore » use, the reactor and its related facilities have been since used to train two generations of nuclear engineers and scientists for the nuclear industry. The reactor power level was upgraded from 1 to 10 kill in 1969, and its operating license was renewed for 20 years in 1983. In 1988, the reactor was converted to low enriched uranium. The low power output of the reactor and ergonomic facility design make it an ideal tool for undergraduate nuclear engineering education and other training.« less

  17. 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

  18. Nuclear reactors built, being built, or planned 1996

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

    NONE

    1997-08-01

    This publication contains unclassified information about facilities, built, being built, or planned in the United States for domestic use or export as of December 31, 1996. The Office of Scientific and Technical Information, U.S. Department of Energy, gathers this information annually from Washington headquarters, and field offices of DOE; from the U.S. Nuclear Regulatory Commission (NRC); from the U. S. reactor manufacturers who are the principal nuclear contractors for foreign reactor locations; from U.S. and foreign embassies; and from foreign governmental nuclear departments. The book consists of three divisions, as follows: (1) a commercial reactor locator map and tables ofmore » the characteristic and statistical data that follow; a table of abbreviations; (2) tables of data for reactors operating, being built, or planned; and (3) tables of data for reactors that have been shut down permanently or dismantled.« less

  19. 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.

  20. 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.

  1. Catalyzed D-D stellarator reactor

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

    Sheffield, John; Spong, Donald A.

    The advantages of using the catalyzed deuterium-deuterium (D-D) approach for a fusion reactor—lower and less energetic neutron flux and no need for a tritium breeding blanket—have been evaluated in previous papers, giving examples of both tokamak and stellarator reactors. This paper presents an update for the stellarator example, taking account of more recent empirical transport scaling results and design studies of lower-aspect-ratio stellarators. We use a modified version of the Generic Magnetic Fusion Reactor model to cost a stellarator-type reactor. Recently, this model has been updated to reflect the improved science and technology base and costs in the magnetic fusionmore » program. Furthermore, it is shown that an interesting catalyzed D-D, stellarator power plant might be possible if the following parameters could be achieved: R/ ≈ 4, required improvement factor to ISS04 scaling, F R = 0.9 to 1.15, ≈ 8.0% to 11.5%, Z eff ≈ 1.45 plus a relativistic temperature correction, fraction of fast ions lost ≈ 0.07, B m ≈ 14 to 16 T, and R ≈ 18 to 24 m.« less

  2. Catalyzed D-D stellarator reactor

    DOE PAGES

    Sheffield, John; Spong, Donald A.

    2016-05-12

    The advantages of using the catalyzed deuterium-deuterium (D-D) approach for a fusion reactor—lower and less energetic neutron flux and no need for a tritium breeding blanket—have been evaluated in previous papers, giving examples of both tokamak and stellarator reactors. This paper presents an update for the stellarator example, taking account of more recent empirical transport scaling results and design studies of lower-aspect-ratio stellarators. We use a modified version of the Generic Magnetic Fusion Reactor model to cost a stellarator-type reactor. Recently, this model has been updated to reflect the improved science and technology base and costs in the magnetic fusionmore » program. Furthermore, it is shown that an interesting catalyzed D-D, stellarator power plant might be possible if the following parameters could be achieved: R/ ≈ 4, required improvement factor to ISS04 scaling, F R = 0.9 to 1.15, ≈ 8.0% to 11.5%, Z eff ≈ 1.45 plus a relativistic temperature correction, fraction of fast ions lost ≈ 0.07, B m ≈ 14 to 16 T, and R ≈ 18 to 24 m.« less

  3. 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

  4. 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.

  5. NEUTRONIC REACTOR CONTROL ROD DRIVE APPARATUS

    DOEpatents

    Oakes, L.C.; Walker, C.S.

    1959-12-15

    ABS>A suspension mechanism between a vertically movable nuclear reactor control rod and a rod extension, which also provides information for the operator or an automatic control signal, is described. A spring connects the rod extension to a drive shift. The extension of the spring indicates whether (1) the rod is at rest on the reactor, (2) the rod and extension are suspended, or (3) the extension alone is suspended, the spring controlling a 3-position electrical switch.

  6. Silicon production in a fluidized bed reactor

    NASA Technical Reports Server (NTRS)

    Rohatgi, N. K.

    1986-01-01

    Part of the development effort of the JPL in-house technology involved in the Flat-Plate Solar Array (FSA) Project was the investigation of a low-cost process to produce semiconductor-grade silicon for terrestrial photovoltaic cell applications. The process selected was based on pyrolysis of silane in a fluidized-bed reactor (FBR). Following initial investigations involving 1- and 2-in. diameter reactors, a 6-in. diameter, engineering-scale FBR was constructed to establish reactor performance, mechanism of silicon deposition, product morphology, and product purity. The overall mass balance for all experiments indicates that more than 90% of the total silicon fed into the reactor is deposited on silicon seed particles and the remaining 10% becomes elutriated fines. Silicon production rates were demonstrated of 1.5 kg/h at 30% silane concentration and 3.5 kg/h at 80% silane concentration. The mechanism of silicon deposition is described by a six-path process: heterogeneous deposition, homogeneous decomposition, coalescence, coagulation, scavenging, and heterogeneous growth on fines. The bulk of the growth silicon layer appears to be made up of small diameter particles. This product morphology lends support to the concept of the scavenging of homogeneously nucleated silicon.

  7. 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

  8. DR Reactor VSR channel damage

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

    Kempf, F.J.; Rawlins, J.K.

    1961-10-30

    On July 11, 1961 the Ball 3X System at DR Reactor was inadventently tripped. All vertical safety rods dropped and all channels were filled with balls. This report has the twofold purpose of documenting borescope observations of ten vertical rod channels at DR Reactor and recording the estimated extent of graphite damage resulting from the above incident. Channel damage data are presented on appended drawings. With suitable notations, the tracings of these drawings may be revised to reflect any future graphite damage. All vertical rod channels at DR Reactor were visually examined with a closed circuit television system during ballmore » removal efforts. Typical photographs of trapped balls and ledges, as viewed on the television monitor, are shown. Photographs of typical graphite damage, obtained through the borescope are also included in this report. 3 refs., 8 figs., 1 tab.« less

  9. 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.

  10. 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.

  11. 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.

  12. 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.

  13. 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.

  14. Convective cooling in a pool-type research reactor

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

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

    2016-01-22

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

  15. Neutronic reactor thermal shield

    DOEpatents

    Wende, Charles W. J.

    1976-06-15

    1. The method of operating a water-cooled neutronic reactor having a graphite moderator which comprises flowing a gaseous mixture of carbon dioxide and helium, in which the helium comprises 40-60 volume percent of the mixture, in contact with the graphite moderator.

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

    NASA Astrophysics Data System (ADS)

    Syarip; Po, L. C. C.

    2018-05-01

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

  17. Fischer-Tropsch Slurry Reactor modeling

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

    Soong, Y.; Gamwo, I.K.; Harke, F.W.

    1995-12-31

    This paper reports experimental and theoretical results on hydrodynamic studies. The experiments were conducted in a hot-pressurized Slurry-Bubble Column Reactor (SBCR). It includes experimental results of Drakeol-10 oil/nitrogen/glass beads hydrodynamic study and the development of an ultrasonic technique for measuring solids concentration. A model to describe the flow behavior in reactors was developed. The hydrodynamic properties in a 10.16 cm diameter bubble column with a perforated-plate gas distributor were studied at pressures ranging from 0.1 to 1.36 MPa, and at temperatures from 20 to 200{degrees}C, using a dual hot-wire probe with nitrogen, glass beads, and Drakeol-10 oil as the gas,more » solid, and liquid phase, respectively. It was found that the addition of 20 oil wt% glass beads in the system has a slight effect on the average gas holdup and bubble size. A well-posed three-dimensional model for bed dynamics was developed from an ill-posed model. The new model has computed solid holdup distributions consistent with experimental observations with no artificial {open_quotes}fountain{close_quotes} as predicted by the earlier model. The model can be applied to a variety of multiphase flows of practical interest. An ultrasonic technique is being developed to measure solids concentration in a three-phase slurry reactor. Preliminary measurements have been made on slurries consisting of molten paraffin wax, glass beads, and nitrogen bubbles at 180 {degrees}C and 0.1 MPa. The data show that both the sound speed and attenuation are well-defined functions of both the solid and gas concentrations in the slurries. The results suggest possibilities to directly measure solids concentration during the operation of an autoclave reactor containing molten wax.« less

  18. Enhanced biodegradation of hexachlorocyclohexane in upflow anaerobic sludge blanket reactor using methanol as an electron donor.

    PubMed

    Bhatt, Praveena; Kumar, M Suresh; Mudliar, Sandeep; Chakrabarti, Tapan

    2008-05-01

    Anaerobic dechlorination of technical grade hexachlorocyclohexane (THCH) was studied in a continuous upflow anaerobic sludge blanket (UASB) reactor with methanol as a supplementary substrate and electron donor. A reactor without methanol served as the experimental control. The inlet feed concentration of THCH in both the experimental and the control UASB reactor was 100 mg l(-1). After 60 days of continuous operation, the removal of THCH was >99% in the methanol-supplemented reactor as compared to 20-35% in the control reactor. THCH was completely dechlorinated in the methanol fed reactor at 48 h HRT after 2 months of continuous operation. This period was also accompanied by increase in biomass in the reactor, which was not observed in the experimental control. Batch studies using other supplementary substrates as well as electron donors namely acetate, butyrate, formate and ethanol showed lower % dechlorination (<85%) and dechlorination rates (<3 mg g(-1)d(-1)) as compared to methanol (98%, 5 mg g(-1)d(-1)). The optimum concentration of methanol required, for stable dechlorination of THCH (100 mg l(-1)) in the UASB reactor, was found to be 500 mg l(-1). Results indicate that addition of methanol as electron donor enhances dechlorination of THCH at high inlet concentration, and is also required for stable UASB reactor performance.

  19. 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.

  20. VERA Core Simulator methodology for pressurized water reactor cycle depletion

    DOE PAGES

    Kochunas, Brendan; Collins, Benjamin; Stimpson, Shane; ...

    2017-01-12

    This paper describes the methodology developed and implemented in the Virtual Environment for Reactor Applications Core Simulator (VERA-CS) to perform high-fidelity, pressurized water reactor (PWR), multicycle, core physics calculations. Depletion of the core with pin-resolved power and nuclide detail is a significant advance in the state of the art for reactor analysis, providing the level of detail necessary to address the problems of the U.S. Department of Energy Nuclear Reactor Simulation Hub, the Consortium for Advanced Simulation of Light Water Reactors (CASL). VERA-CS has three main components: the neutronics solver MPACT, the thermal-hydraulic (T-H) solver COBRA-TF (CTF), and the nuclidemore » transmutation solver ORIGEN. This paper focuses on MPACT and provides an overview of the resonance self-shielding methods, macroscopic-cross-section calculation, two-dimensional/one-dimensional (2-D/1-D) transport, nuclide depletion, T-H feedback, and other supporting methods representing a minimal set of the capabilities needed to simulate high-fidelity models of a commercial nuclear reactor. Results are presented from the simulation of a model of the first cycle of Watts Bar Unit 1. The simulation is within 16 parts per million boron (ppmB) reactivity for all state points compared to cycle measurements, with an average reactivity bias of <5 ppmB for the entire cycle. Comparisons to cycle 1 flux map data are also provided, and the average 2-D root-mean-square (rms) error during cycle 1 is 1.07%. To demonstrate the multicycle capability, a state point at beginning of cycle (BOC) 2 was also simulated and compared to plant data. The comparison of the cycle 2 BOC state has a reactivity difference of +3 ppmB from measurement, and the 2-D rms of the comparison in the flux maps is 1.77%. Lastly, these results provide confidence in VERA-CS’s capability to perform high-fidelity calculations for practical PWR reactor problems.« less

  1. 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.

  2. 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.

  3. 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.

  4. 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

  5. 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.

  6. Nuclear Engineering Computer Modules, Thermal-Hydraulics, TH-1: Pressurized Water Reactors.

    ERIC Educational Resources Information Center

    Reihman, Thomas C.

    This learning module is concerned with the temperature field, the heat transfer rates, and the coolant pressure drop in typical pressurized water reactor (PWR) fuel assemblies. As in all of the modules of this series, emphasis is placed on developing the theory and demonstrating its use with a simplified model. The heart of the module is the PWR…

  7. 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

  8. REUSABLE ADSORBENTS FOR DILUTE SOLUTIONS SEPARATION. 6. BATCH AND CONTINUOUS REACTORS FOR ADSORPTION AND DEGRADATION OF 1,2-DICHLOROBENZENE FROM DILUTE WASTEWATER STREAMS USING TITANIA AS A PHOTOCATALYST. (R828598C753)

    EPA Science Inventory

    Two types of external lamp reactors were investigated for the titania catalyzed photodegradation of 1,2-dichlorobenzene (DCB) from a dilute water stream. The first one was a batch mixed slurry reactor and the second one was a semi-batch reactor with continuous feed recycle wit...

  9. Controlled nitric oxide production via O(1D) + N2O reactions for use in oxidation flow reactor studies

    NASA Astrophysics Data System (ADS)

    Lambe, Andrew; Massoli, Paola; Zhang, Xuan; Canagaratna, Manjula; Nowak, John; Daube, Conner; Yan, Chao; Nie, Wei; Onasch, Timothy; Jayne, John; Kolb, Charles; Davidovits, Paul; Worsnop, Douglas; Brune, William

    2017-06-01

    Oxidation flow reactors that use low-pressure mercury lamps to produce hydroxyl (OH) radicals are an emerging technique for studying the oxidative aging of organic aerosols. Here, ozone (O3) is photolyzed at 254 nm to produce O(1D) radicals, which react with water vapor to produce OH. However, the need to use parts-per-million levels of O3 hinders the ability of oxidation flow reactors to simulate NOx-dependent secondary organic aerosol (SOA) formation pathways. Simple addition of nitric oxide (NO) results in fast conversion of NOx (NO + NO2) to nitric acid (HNO3), making it impossible to sustain NOx at levels that are sufficient to compete with hydroperoxy (HO2) radicals as a sink for organic peroxy (RO2) radicals. We developed a new method that is well suited to the characterization of NOx-dependent SOA formation pathways in oxidation flow reactors. NO and NO2 are produced via the reaction O(1D) + N2O → 2NO, followed by the reaction NO + O3 → NO2 + O2. Laboratory measurements coupled with photochemical model simulations suggest that O(1D) + N2O reactions can be used to systematically vary the relative branching ratio of RO2 + NO reactions relative to RO2 + HO2 and/or RO2 + RO2 reactions over a range of conditions relevant to atmospheric SOA formation. We demonstrate proof of concept using high-resolution time-of-flight chemical ionization mass spectrometer (HR-ToF-CIMS) measurements with nitrate (NO3-) reagent ion to detect gas-phase oxidation products of isoprene and α-pinene previously observed in NOx-influenced environments and in laboratory chamber experiments.

  10. Controlled Nitric Oxide Production via O(1D) + N2O Reactions for Use in Oxidation Flow Reactor Studies

    NASA Technical Reports Server (NTRS)

    Lambe, Andrew; Massoli, Paola; Zhang, Xuan; Canagaratna, Manjula; Nowak, John; Daube, Conner; Yan, Chao; Nie, Wei; Onasch, Timothy; Jayne, John; hide

    2017-01-01

    Oxidation flow reactors that use low-pressure mercury lamps to produce hydroxyl (OH) radicals are an emerging technique for studying the oxidative aging of organic aerosols. Here, ozone (O3) is photolyzed at 254 nm to produce O(1D) radicals, which react with water vapor to produce OH. However, the need to use parts-per-million levels of O3 hinders the ability of oxidation flow reactors to simulate NOx-dependent secondary organic aerosol (SOA) formation pathways. Simple addition of nitric oxide (NO) results in fast conversion of NOx (NO+NO2) to nitric acid (HNO3), making it impossible to sustain NOx at levels that are sufficient to compete with hydroperoxy (HO2) radicals as a sink for organic peroxy (RO2) radicals. We developed a new method that is well suited to the characterization of NOx-dependent SOA formation pathways in oxidation flow reactors. NO and NO2 are produced via the reaction O(1D)+N2O->2NO, followed by the reaction NO+O3->NO2+O2. Laboratory measurements coupled with photochemical model simulations suggest that O(1D)+N2O reactions can be used to systematically vary the relative branching ratio of RO2 +NO reactions relative to RO2 +HO2 and/or RO2+RO2 reactions over a range of conditions relevant to atmospheric SOA formation. We demonstrate proof of concept using high-resolution time-of-flight chemical ionization mass spectrometer (HR-ToF-CIMS) measurements with nitrate (NO-3 ) reagent ion to detect gas-phase oxidation products of isoprene and -pinene previously observed in NOx-influenced environments and in laboratory chamber experiments.

  11. Carbofuran removal in continuous-photocatalytic reactor: Reactor optimization, rate-constant determination and carbofuran degradation pathway analysis.

    PubMed

    Vishnuganth, M A; Remya, Neelancherry; Kumar, Mathava; Selvaraju, N

    2017-05-04

    Carbofuran (CBF) removal in a continuous-flow photocatalytic reactor with granular activated carbon supported titanium dioxide (GAC-TiO 2 ) catalyst was investigated. The effects of feed flow rate, TiO 2 concentration and addition of supplementary oxidants on CBF removal were investigated. The central composite design (CCD) was used to design the experiments and to estimate the effects of feed flow rate and TiO 2 concentration on CBF removal. The outcome of CCD experiments demonstrated that reactor performance was influenced mainly by feed flow rate compared to TiO 2 concentration. A second-order polynomial model developed based on CCD experiments fitted the experimental data with good correlation (R 2 ∼ 0.964). The addition of 1 mL min -1 hydrogen peroxide has shown complete CBF degradation and 76% chemical oxygen demand removal under the following operating conditions of CBF ∼50 mg L -1 , TiO 2 ∼5 mg L -1 and feed flow rate ∼82.5 mL min -1 . Rate constant of the photodegradation process was also calculated by applying the kinetic data in pseudo-first-order kinetics. Four major degradation intermediates of CBF were identified using GC-MS analysis. As a whole, the reactor system and GAC-TiO 2 catalyst used could be constructive in cost-effective CBF removal with no impact to receiving environment through getaway of photocatalyst.

  12. 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.

  13. Biofilm reactors for industrial bioconversion processes: employing potential of enhanced reaction rates

    PubMed Central

    Qureshi, Nasib; Annous, Bassam A; Ezeji, Thaddeus C; Karcher, Patrick; Maddox, Ian S

    2005-01-01

    This article describes the use of biofilm reactors for the production of various chemicals by fermentation and wastewater treatment. Biofilm formation is a natural process where microbial cells attach to the support (adsorbent) or form flocs/aggregates (also called granules) without use of chemicals and form thick layers of cells known as "biofilms." As a result of biofilm formation, cell densities in the reactor increase and cell concentrations as high as 74 gL-1 can be achieved. The reactor configurations can be as simple as a batch reactor, continuous stirred tank reactor (CSTR), packed bed reactor (PBR), fluidized bed reactor (FBR), airlift reactor (ALR), upflow anaerobic sludge blanket (UASB) reactor, or any other suitable configuration. In UASB granular biofilm particles are used. This article demonstrates that reactor productivities in these reactors have been superior to any other reactor types. This article describes production of ethanol, butanol, lactic acid, acetic acid/vinegar, succinic acid, and fumaric acid in addition to wastewater treatment in the biofilm reactors. As the title suggests, biofilm reactors have high potential to be employed in biotechnology/bioconversion industry for viable economic reasons. In this article, various reactor types have been compared for the above bioconversion processes. PMID:16122390

  14. 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.

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

    NASA Astrophysics Data System (ADS)

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

    2018-01-01

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

  16. 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

  17. Pebble Bed Reactors Design Optimization Methods and their Application to the Pebble Bed Fluoride Salt Cooled High Temperature Reactor (PB-FHR)

    NASA Astrophysics Data System (ADS)

    Cisneros, Anselmo Tomas, Jr.

    The Fluoride salt cooled High temperature Reactor (FHR) is a class of advanced nuclear reactors that combine the robust coated particle fuel form from high temperature gas cooled reactors, direct reactor auxillary cooling system (DRACS) passive decay removal of liquid metal fast reactors, and the transparent, high volumetric heat capacitance liquid fluoride salt working fluids---flibe (33%7Li2F-67%BeF)---from molten salt reactors. This combination of fuel and coolant enables FHRs to operate in a high-temperature low-pressure design space that has beneficial safety and economic implications. In 2012, UC Berkeley was charged with developing a pre-conceptual design of a commercial prototype FHR---the Pebble Bed- Fluoride Salt Cooled High Temperature Reactor (PB-FHR)---as part of the Nuclear Energy University Programs' (NEUP) integrated research project. The Mark 1 design of the PB-FHR (Mk1 PB-FHR) is 236 MWt flibe cooled pebble bed nuclear heat source that drives an open-air Brayton combine-cycle power conversion system. The PB-FHR's pebble bed consists of a 19.8% enriched uranium fuel core surrounded by an inert graphite pebble reflector that shields the outer solid graphite reflector, core barrel and reactor vessel. The fuel reaches an average burnup of 178000 MWt-d/MT. The Mk1 PB-FHR exhibits strong negative temperature reactivity feedback from the fuel, graphite moderator and the flibe coolant but a small positive temperature reactivity feedback of the inner reflector and from the outer graphite pebble reflector. A novel neutronics and depletion methodology---the multiple burnup state methodology was developed for an accurate and efficient search for the equilibrium composition of an arbitrary continuously refueled pebble bed reactor core. The Burnup Equilibrium Analysis Utility (BEAU) computer program was developed to implement this methodology. BEAU was successfully benchmarked against published results generated with existing equilibrium depletion codes VSOP

  18. 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.

  19. SM-1 REACTOR VESSEL COVER AND FLANGE STRESS ANALYSIS

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

    Sayre, M.F.

    1962-02-19

    The maximum stress calculated for the SMl-1 reactor vessel closure studs occurs during operation at full power. This value is 27,180 psi of which 19,800 psi is tension and 7380 psi bending. This stress does not include a stress concentration factor for effect of threads. It was eonservatively assumed the studs were initially tightened to a code allowable stress of 20,000 psi as specified in the ASME Code rather than the lesser stress obtained by the normal operating procedure. The maximum calculated stress occurs at the outside surface of the cover where the stress ranges from 318 psi in tensionmore » to 90,660 psi in compression. The alternating stress is 50,000 psi. According to the Navy Code for a stress range of 50,000 psi, the eover material ean safely undergo a maximum of 1600 cycles. It was estimated that the SM-1 will go through approximately 000 startup and shutdown cycles during a 20-yr life period, so the calculated stress is regarded as safe. For a transient eondition of 30 deg F/hr during heat-up, approximate temperature differences between the inside and outside surfaces of the cover were obtained. Temperature differentials between the inside and outside surfaces of the cover are increased by roughly 10%; above the steady state condition. More exact calculations of the transient stresses did not appear necessary siuce they would be not more than 10% greater than the steady state thermal stress. (auth)« less

  20. A dual purpose packed-bed reactor for biogas scrubbing and methane-dependent water quality improvement applying to a wastewater treatment system consisting of UASB reactor and trickling filter.

    PubMed

    Tanaka, Yasuo

    2002-08-01

    A wastewater treatment system employing a UASB reactor in temperate regions requires biogas as a heat source for the UASB reactor during low temperature seasons. In this case, removal of H2S in the biogas by means of a scrubber before burning is necessary in order to prevent the boilers from corroding. Heating of the UASB reactor is, however, unnecessary in a warm season, and the scrubber and biogas become useless. Methane-dependent water quality improvement using the scrubber and biogas would be one way to use them efficiently during the warm season. The possible dual-purpose use of a packed-bed reactor was examined, with one of its uses being the scrubbing of biogas during the cold season and the other being the methane-dependent improvement of effluent water quality during the warm season. A bench scale packed-bed filled with plastic latticed-ring media was installed in a livestock wastewater treatment plant consisting of a UASB reactor and a trickling filter for post-treatment. The packed-bed was operated with biogas flowing at a superficial velocity of 0.14-0.39 m h(-1) and the hydraulic loading of trickling filter effluent sprayed onto the media 9.4-26.1 m3 m2 day(-1). H2S in the biogas from the UASB reactor was reduced from 1,200-2,500 ppm to less than 2 ppm by the reactor. Methane-dependent water quality improvement was examined using a laboratory scale reactor to which methane and/or air was supplied from the bottom, while plant effluent was spread from the top of the reactor. When the mixture gas of methane and air (volume ratio 1:3) was added to the reactor, biofilm grew on the surface of the media. Accompanying this growth, ammonium and phosphate in the spread water decreased, probably due to assimilation by the methane-oxidizing bacteria. Though assimilation activity dropped after the accumulation of biomass, it could be reactivated by washing out the excess biomass. Periodical backwash at a rate of more than once a week seemed to efficiently maintain

  1. 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.

  2. 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.

  3. 10 CFR 50.72 - Immediate notification requirements for operating nuclear power reactors.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... 10 Energy 1 2012-01-01 2012-01-01 false Immediate notification requirements for operating nuclear power reactors. 50.72 Section 50.72 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF... notification requirements for operating nuclear power reactors. (a) General requirements. 1 (1) Each nuclear...

  4. 10 CFR 50.72 - Immediate notification requirements for operating nuclear power reactors.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... 10 Energy 1 2013-01-01 2013-01-01 false Immediate notification requirements for operating nuclear power reactors. 50.72 Section 50.72 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF... notification requirements for operating nuclear power reactors. (a) General requirements. 1 (1) Each nuclear...

  5. Design and analysis of a nuclear reactor core for innovative small light water reactors

    NASA Astrophysics Data System (ADS)

    Soldatov, Alexey I.

    In order to address the energy needs of developing countries and remote communities, Oregon State University has proposed the Multi-Application Small Light Water Reactor (MASLWR) design. In order to achieve five years of operation without refueling, use of 8% enriched fuel is necessary. This dissertation is focused on core design issues related with increased fuel enrichment (8.0%) and specific MASLWR operational conditions (such as lower operational pressure and temperature, and increased leakage due to small core). Neutron physics calculations are performed with the commercial nuclear industry tools CASMO-4 and SIMULATE-3, developed by Studsvik Scandpower Inc. The first set of results are generated from infinite lattice level calculations with CASMO-4, and focus on evaluation of the principal differences between standard PWR fuel and MASLWR fuel. Chapter 4-1 covers aspects of fuel isotopic composition changes with burnup, evaluation of kinetic parameters and reactivity coefficients. Chapter 4-2 discusses gadolinium self-shielding and shadowing effects, and subsequent impacts on power generation peaking and Reactor Control System shadowing. The second aspect of the research is dedicated to core design issues, such as reflector design (chapter 4-3), burnable absorber distribution and programmed fuel burnup and fuel use strategy (chapter 4-4). This section also includes discussion of the parameters important for safety and evaluation of Reactor Control System options for the proposed core design. An evaluation of the sensitivity of the proposed design to uncertainty in calculated parameters is presented in chapter 4-5. The results presented in this dissertation cover a new area of reactor design and operational parameters, and may be applicable to other small and large pressurized water reactor designs.

  6. 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.

  7. 148. ARAIII Reactor building (ARA608) Floor plan. Shows location of ...

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

    148. ARA-III Reactor building (ARA-608) Floor plan. Shows location of reactor, heater, and mechanical loop pits; mechanical and electrical equipment rooms; and other work areas. Aerojet-general 880-area/GCRE-608-A-1. Date: February 1958. Ineel index code no. 063-0608-00-013-102612. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

  8. REACTOR SERVICES BUILDING, TRA635, INTERIOR. ALSO KNOWN AS MATERIAL RECEIVING ...

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

    REACTOR SERVICES BUILDING, TRA-635, INTERIOR. ALSO KNOWN AS MATERIAL RECEIVING AREA AND LABORATORY. CAMERA ON FIRST FLOOR FACING NORTH TOWARD MTR BUILDING. MOCK-UP AREA WAS TO THE RIGHT OF VIEW. INL NEGATIVE NO. HD46-10-1. Mike Crane, Photographer, 2/2005 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

  9. 10 CFR 50.44 - Combustible gas control for nuclear power reactors.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... 10 Energy 1 2013-01-01 2013-01-01 false Combustible gas control for nuclear power reactors. 50.44 Section 50.44 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION... for nuclear power reactors. (a) Definitions—(1) Inerted atmosphere means a containment atmosphere with...

  10. 10 CFR 50.44 - Combustible gas control for nuclear power reactors.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 10 Energy 1 2011-01-01 2011-01-01 false Combustible gas control for nuclear power reactors. 50.44 Section 50.44 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION... for nuclear power reactors. (a) Definitions—(1) Inerted atmosphere means a containment atmosphere with...

  11. Spherical torus fusion reactor

    DOEpatents

    Peng, Yueng-Kay M.

    1989-04-04

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

  12. Spherical torus fusion reactor

    DOEpatents

    Peng, Yueng-Kay M.

    1989-01-01

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

  13. Fluorescence-based monitoring of tracer and substrate distribution in an UASB reactor.

    PubMed

    Lou, S J; Tartakovsky, B; Zeng, Y; Wu, P; Guiot, S R

    2006-11-01

    In this work, rhodamine-related fluorescence was measured on-line at four reactor heights in order to study hydrodynamics within an upflow anaerobic sludge bed reactor. A linear dependence of the dispersion coefficient (D) on the upflow velocity was observed, while the influence of the organic loading rate (OLR) was insignificant. Furthermore, the Bodenstein number of the reactor loaded with granulated sludge was found to be position-dependent with the largest values measured at the bottom of the sludge bed. This trend was not observed in the reactor without sludge. Chemical oxygen demand (COD) and volatile fatty acid (VFA) concentrations were measured at the same reactor heights as in rhodamine tests using conventional off-line analytical methods and on-line multiwavelength fluorometry. Significant spatial COD and VFA gradients were observed at organic loading rates above 6g COD l(R)(-1)d(-1) and linear upflow velocities below 0.8m h(-1).

  14. Multicapillary Flow Reactor: Synthesis of 1,2,5-Thiadiazepane 1,1-Dioxide Library Utilizing One-Pot Elimination and Inter-/Intramolecular Double aza-Michael Addition Via Microwave-Assisted, Continuous-Flow Organic Synthesis (MACOS)

    PubMed Central

    Ullah, Farman; Zang, Qin; Javed, Salim; Zhou, Aihua; Knudtson, Christopher A.; Bi, Danse; Hanson, Paul R.; Organ, Michael G.

    2013-01-01

    A microwave-assisted, continuous-flow organic synthesis (MACOS) protocol for the synthesis of functionalized 1,2,5-thiadiazepane 1,1-dioxide library, utilizing a one-pot elimination and inter-/intramolecular double aza-Michael addition strategy is reported. The optimized protocol in MACOS was utilized for scale-out and further extended for library production using a multicapillary flow reactor. A 50-member library of 1,2,5-thiadiazepane 1,1-dioxides was prepared on a 100- to 300-mg scale with overall yields between 50 and 80% and over 90 % purity determined by proton nuclear magnetic resonance (1H-NMR) spectroscopy. PMID:24244871

  15. The effect of cover use on plastic pyrolysis reactor heating process

    NASA Astrophysics Data System (ADS)

    Armadi, Benny H.; Rangkuti, Chalilullah; Fauzi, M. D.; Permatasari, R.

    2017-03-01

    Plastic pyrolysis process to produce liquid fuel is an endothermic process that uses heat from the combustion of fuel as heat source. The reactor used is usually a vertical cylindrical in shape, with LPG fuel combustion under the flat bottom of the reactor, and the combustion gases is dispersed into the surrounding environment, so that heat transferred to the plastic inside the reactor is not effective, causing high LPG consumption. In this study, the reactor is made of stainless steel plate, with a vertical cylindrical shape, with a basic cylindrical conical truncated by a pit pass hot flue gas in the middle that serves to deliver flue gas into the chimney. The contact area between the hot combusted LPG gases to the processed plastic inside the reactor becomes bigger and gets better heat transfer, and required less LPG consumption. For more effective heat transfer process, an outer cover of this reactor was made and the relatively hot combustion gases are used to heat the outside of the reactor by directing the flow of the flue gas from the chimney down along the outer wall of the reactor and out the bottom lid. This construction makes the heating process to be faster and the LPG fuel is used more efficiently. From the measurements, it was found to raise 1°C of temperature inside the covered reactor, the LPG consumed is 0.59 gram, and if the reactor cover is removed, the gas demand will rise nearly threefold to 1.43 grams. With this method, in addition to reducing the rate of heat loss will also help reduce LPG consumption significantly.

  16. FOOD IRRADIATION REACTOR

    DOEpatents

    Leyse, C.F.; Putnam, G.E.

    1961-05-01

    An irradiation apparatus is described. It comprises a pressure vessel, a neutronic reactor active portion having a substantially greater height than diameter in the pressure vessel, an annular tank surrounding and spaced from the pressure vessel containing an aqueous indium/sup 1//sup 1//sup 5/ sulfate solution of approximately 600 grams per liter concentration, means for circulating separate coolants through the active portion and the space between the annular tank and the pressure vessel, radiator means adapted to receive the materials to be irradiated, and means for flowing the indium/sup 1//sup 1//sup 5/ sulfate solution through the radiator means.

  17. Thermal-Hydraulic Design of a Fluoride High-Temperature Demonstration Reactor

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

    Carbajo, Juan J; Qualls, A L

    2016-01-01

    INTRODUCTION The Fluoride High-Temperature Reactor (FHR) named the Demonstration Reactor (DR) is a novel reactor concept using molten salt coolant and TRIstructural ISOtropic (TRISO) fuel that is being developed at Oak Ridge National Laboratory (ORNL). The objective of the FHR DR is to advance the technology readiness level of FHRs. The FHR DR will demonstrate technologies needed to close remaining gaps to commercial viability. The FHR DR has a thermal power of 100 MWt, very similar to the SmAHTR, another FHR ORNL concept (Refs. 1 and 2) with a power of 125 MWt. The FHR DR is also a smallmore » version of the Advanced High Temperature Reactor (AHTR), with a power of 3400 MWt, cooled by a molten salt and also being developed at ORNL (Ref. 3). The FHR DR combines three existing technologies: (1) high-temperature, low-pressure molten salt coolant, (2) high-temperature coated-particle TRISO fuel, (3) and passive decay heat cooling systems by using Direct Reactor Auxiliary Cooling Systems (DRACS). This paper presents FHR DR thermal-hydraulic design calculations.« less

  18. Korean standard nuclear plant ex-vessel neutron dosimetry program Ulchin 4

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

    Duo, J.I.; Chen, J.; Kulesza, J.A.

    2011-07-01

    A comprehensive ex-vessel neutron dosimetry (EVND) surveillance program has been deployed in 16 pressurized water reactors (PWR) in South Korea and EVND dosimetry sets have already been installed and analyzed in Westinghouse reactor designs. In this paper, the unique features of the design, training, and installation in the Korean standard nuclear plant (KSNP) Ulchin Unit 4 are presented. Ulchin Unit 4 Cycle 9 represents the first dosimetry analyzed from the EVND design deployed in KSNP plants: Yonggwang Units 3 through 6 and Ulchin Units 3 through 6. KSNP's cavity configuration precludes a conventional installation from the cavity floor. The solution,more » requiring the installation crew to access the cavity at an elevation of the active core, places a premium on rapid installation due to high area dose rates. Numerous geometrical features warranted the use of a detailed design in true 3D mechanical design software to control interferences. A full-size training mockup maximized the crew ability to correctly install the instrument in minimum time. The analysis of the first dosimetry set shows good agreements between measurement and calculation within the associated uncertainties. A complete EVND system has been successfully designed, installed, and analyzed for a KNSP plant. Current and future EVND analyses will continue supporting the successful operation of PWR units in South Korea. (authors)« less

  19. 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)

  20. 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

  1. Preliminary validation of computational model for neutron flux prediction of Thai Research Reactor (TRR-1/M1)

    NASA Astrophysics Data System (ADS)

    Sabaibang, S.; Lekchaum, S.; Tipayakul, C.

    2015-05-01

    This study is a part of an on-going work to develop a computational model of Thai Research Reactor (TRR-1/M1) which is capable of accurately predicting the neutron flux level and spectrum. The computational model was created by MCNPX program and the CT (Central Thimble) in-core irradiation facility was selected as the location for validation. The comparison was performed with the typical flux measurement method routinely practiced at TRR-1/M1, that is, the foil activation technique. In this technique, gold foil is irradiated for a certain period of time and the activity of the irradiated target is measured to derive the thermal neutron flux. Additionally, the flux measurement with SPND (self-powered neutron detector) was also performed for comparison. The thermal neutron flux from the MCNPX simulation was found to be 1.79×1013 neutron/cm2s while that from the foil activation measurement was 4.68×1013 neutron/cm2s. On the other hand, the thermal neutron flux from the measurement using SPND was 2.47×1013 neutron/cm2s. An assessment of the differences among the three methods was done. The difference of the MCNPX with the foil activation technique was found to be 67.8% and the difference of the MCNPX with the SPND was found to be 27.8%.

  2. Fiber-Optical Sensors: Basics and Applications in Multiphase Reactors

    PubMed Central

    Li, Xiangyang; Yang, Chao; Yang, Shifang; Li, Guozheng

    2012-01-01

    This work presents a brief introduction on the basics of fiber-optical sensors and an overview focused on the applications to measurements in multiphase reactors. The most commonly principle utilized is laser back scattering, which is also the foundation for almost all current probes used in multiphase reactors. The fiber-optical probe techniques in two-phase reactors are more developed than those in three-phase reactors. There are many studies on the measurement of gas holdup using fiber-optical probes in three-phase fluidized beds, but negative interference of particles on probe function was less studied. The interactions between solids and probe tips were less studied because glass beads etc. were always used as the solid phase. The vision probes may be the most promising for simultaneous measurements of gas dispersion and solids suspension in three-phase reactors. Thus, the following techniques of the fiber-optical probes in multiphase reactors should be developed further: (1) online measuring techniques under nearly industrial operating conditions; (2) corresponding signal data processing techniques; (3) joint application with other measuring techniques.

  3. 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.

  4. 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)

  5. 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.

  6. 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.

  7. 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

  8. 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.

  9. Non-equilibrium plasma reactors for organic solvent destruction

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

    Yang, C.L.; Beltran, M.R.; Kravets, Z.

    1997-12-31

    Two non-equilibrium plasma reactors were evaluated for their ability to destroy three widely used organic solvents, i.e., 2-butanone, toluene and ethyl acetate. The catalyzed plasma reactor (CPR) with 6 mm glass beads destroys 98% of 50 ppm toluene in air at 24 kV/cm and space velocities of 1,400 v/v/hr. Eighty-five percent of ethyl acetate and 2-butanone are destroyed under the same conditions. The tubular plasma reactor (TPR) has an efficiency of 10% to 20% lower than that of a CPR under the same conditions. The 1,400 v/v/hr in a CPR is equal to a residence time of 2.6 seconds inmore » a TPR. The operating temperatures, corona characteristics, as well as the kinetics of VOC destruction in both TPR and CPR were studied.« less

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

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

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

    1970-01-01

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

  11. 77 FR 8902 - Draft Regulatory Guide: Issuance, Availability Decommissioning of Nuclear Power Reactors

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-02-15

    ... Decommissioning of Nuclear Power Reactors AGENCY: Nuclear Regulatory Commission. ACTION: Draft regulatory guide... draft regulatory guide (DG) DG-1271 ``Decommissioning of Nuclear Power Reactors.'' This guide describes... Regulatory Guide 1.184, ``Decommissioning of Nuclear Power Reactors,'' dated July 2000. This proposed...

  12. Growth mechanism of GaAs1-xSbx ternary alloy thin film on MOCVD reactor using TMGa, TDMAAs and TDMASb

    NASA Astrophysics Data System (ADS)

    Suhandi, A.; Tayubi, Y. R.; Arifin, P.

    2016-04-01

    Metal Organic Chemical Vapor Deposition (MOCVD) is a method for growing a solid material (in the form of thin films, especially for semiconductor materials) using vapor phase metal organic sources. Studies on the growth mechanism of GaAs1-xSbx ternary alloy thin solid film in the range of miscibility-gap using metal organic sources trimethylgallium (TMGa), trisdimethylaminoarsenic (TDMAAs), and trisdimethylaminoantimony (TDMASb) on MOCVD reactor has been done to understand the physical and chemical processes involved. Knowledge of the processes that occur during alloy formation is very important to determine the couple of growth condition and growth parameters are appropriate for yield high quality GaAs1-xSbx alloy. The mechanism has been studied include decomposition of metal organic sources and chemical reactions that may occur, the incorporation of the alloy elements forming and the contaminants element that are formed in the gown thin film. In this paper presented the results of experimental data on the growth of GaAs1-xSbx alloy using Vertical-MOCVD reactor to demonstrate its potential in growing GaAs1-xSbx alloy in the range of its miscibility gap.

  13. 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.

  14. 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.

  15. 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.

  16. Design of megawatt power level heat pipe reactors

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

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

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

  17. Enrichment of acetogenic bacteria in high rate anaerobic reactors under mesophilic and thermophilic conditions.

    PubMed

    Ryan, P; Forbes, C; McHugh, S; O'Reilly, C; Fleming, G T A; Colleran, E

    2010-07-01

    The objective of the current study was to expand the knowledge of the role of acetogenic Bacteria in high rate anaerobic digesters. To this end, acetogens were enriched by supplying a variety of acetogenic growth supportive substrates to two laboratory scale high rate upflow anaerobic sludge bed (UASB) reactors operated at 37 degrees C (R1) and 55 degrees C (R2). The reactors were initially fed a glucose/acetate influent. Having achieved high operational performance and granular sludge development and activity, both reactors were changed to homoacetogenic bacterial substrates on day 373 of the trial. The reactors were initially fed with sodium vanillate as a sole substrate. Although % COD removal indicated that the 55 degrees C reactor out performed the 37 degrees C reactor, effluent acetate levels from R2 were generally higher than from R1, reaching values as high as 5023 mg l(-1). Homoacetogenic activity in both reactors was confirmed on day 419 by specific acetogenic activity (SAA) measurement, with higher values obtained for R2 than R1. Sodium formate was introduced as sole substrate to both reactors on day 464. It was found that formate supported acetogenic activity at both temperatures. By the end of the trial, no specific methanogenic activity (SMA) was observed against acetate and propionate indicating that the methane produced was solely by hydrogenotrophic Archaea. Higher SMA and SAA values against H(2)/CO(2) suggested development of a formate utilising acetogenic population growing in syntrophy with hydrogenotrophic methanogens. Throughout the formate trial, the mesophilic reactor performed better overall than the thermophilic reactor. Copyright 2010 Elsevier Ltd. All rights reserved.

  18. Controlled nitric oxide production via O( 1D)+N 2O reactions for use in oxidation flow reactor studies

    DOE PAGES

    Lambe, Andrew; Massoli, Paola; Zhang, Xuan; ...

    2017-06-22

    Oxidation flow reactors that use low-pressure mercury lamps to produce hydroxyl (OH) radicals are an emerging technique for studying the oxidative aging of organic aerosols. Here, ozone (O 3) is photolyzed at 254 nm to produce O( 1D) radicals, which react with water vapor to produce OH. However, the need to use parts-per-million levels of O 3 hinders the ability of oxidation flow reactors to simulate NO x-dependent secondary organic aerosol (SOA) formation pathways. Simple addition of nitric oxide (NO) results in fast conversion of NO x (NO+NO 2) to nitric acid (HNO 3), making it impossible to sustain NOmore » x at levels that are sufficient to compete with hydroperoxy (HO 2) radicals as a sink for organic peroxy (RO 2) radicals. We developed a new method that is well suited to the characterization of NO x-dependent SOA formation pathways in oxidation flow reactors. NO and NO 2 are produced via the reaction O( 1D) + N 2O → 2NO, followed by the reaction NO + O 3 → NO 2+O 2. Laboratory measurements coupled with photochemical model simulations suggest that O( 1D) + N 2O reactions can be used to systematically vary the relative branching ratio of RO 2 + NO reactions relative to RO 2 + HO 2 and/or RO 2 + RO 2 reactions over a range of conditions relevant to atmospheric SOA formation. We demonstrate proof of concept using high-resolution time-of-flight chemical ionization mass spectrometer (HR-ToF-CIMS) measurements with nitrate (NO 3 -) reagent ion to detect gas-phase oxidation products of isoprene and α-pinene previously observed in NO x-influenced environments and in laboratory chamber experiments.« less

  19. Controlled nitric oxide production via O( 1D)+N 2O reactions for use in oxidation flow reactor studies

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

    Lambe, Andrew; Massoli, Paola; Zhang, Xuan

    Oxidation flow reactors that use low-pressure mercury lamps to produce hydroxyl (OH) radicals are an emerging technique for studying the oxidative aging of organic aerosols. Here, ozone (O 3) is photolyzed at 254 nm to produce O( 1D) radicals, which react with water vapor to produce OH. However, the need to use parts-per-million levels of O 3 hinders the ability of oxidation flow reactors to simulate NO x-dependent secondary organic aerosol (SOA) formation pathways. Simple addition of nitric oxide (NO) results in fast conversion of NO x (NO+NO 2) to nitric acid (HNO 3), making it impossible to sustain NOmore » x at levels that are sufficient to compete with hydroperoxy (HO 2) radicals as a sink for organic peroxy (RO 2) radicals. We developed a new method that is well suited to the characterization of NO x-dependent SOA formation pathways in oxidation flow reactors. NO and NO 2 are produced via the reaction O( 1D) + N 2O → 2NO, followed by the reaction NO + O 3 → NO 2+O 2. Laboratory measurements coupled with photochemical model simulations suggest that O( 1D) + N 2O reactions can be used to systematically vary the relative branching ratio of RO 2 + NO reactions relative to RO 2 + HO 2 and/or RO 2 + RO 2 reactions over a range of conditions relevant to atmospheric SOA formation. We demonstrate proof of concept using high-resolution time-of-flight chemical ionization mass spectrometer (HR-ToF-CIMS) measurements with nitrate (NO 3 -) reagent ion to detect gas-phase oxidation products of isoprene and α-pinene previously observed in NO x-influenced environments and in laboratory chamber experiments.« less

  20. 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

  1. 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.

  2. 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.

  3. 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.

  4. 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.)

  5. MATERIALS TESTING REACTOR (MTR) BUILDING, TRA603. CONTEXTUAL VIEW OF MTR ...

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

    MATERIALS TESTING REACTOR (MTR) BUILDING, TRA-603. CONTEXTUAL VIEW OF MTR BUILDING SHOWING NORTH SIDES OF THE HIGH-BAY REACTOR BUILDING, ITS SECOND/THIRD FLOOR BALCONY LEVEL, AND THE ATTACHED ONE-STORY OFFICE/LABORATORY BUILDING, TRA-604. CAMERA FACING SOUTHEAST. VERTICAL CONCRETE-SHROUDED BEAMS SUPPORT PRECAST CONCRETE PANELS. CONCRETE PROJECTION FORMED AS A BUNKER AT LEFT OF VIEW IS TRA-657, PLUG STORAGE BUILDING. INL NEGATIVE NO. HD46-42-1. Mike Crane, Photographer, 4/2005 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

  6. Reactor Dosimetry State of the Art 2008

    NASA Astrophysics Data System (ADS)

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

    2009-08-01

    Oral session 1: Retrospective dosimetry. Retrospective dosimetry of VVER 440 reactor pressure vessel at the 3rd unit of Dukovany NPP / M. Marek ... [et al.]. Retrospective dosimetry study at the RPV of NPP Greifswald unit 1 / J. Konheiser ... [et al.]. Test of prototype detector for retrospective neutron dosimetry of reactor internals and vessel / K. Hayashi ... [et al.]. Neutron doses to the concrete vessel and tendons of a magnox reactor using retrospective dosimetry / D. A. Allen ... [et al.]. A retrospective dosimetry feasibility study for Atucha I / J. Wagemans ... [et al.]. Retrospective reactor dosimetry with zirconium alloy samples in a PWR / L. R. Greenwood and J. P. Foster -- Oral session 2: Experimental techniques. Characterizing the Time-dependent components of reactor n/y environments / P. J. Griffin, S. M. Luker and A. J. Suo-Anttila. Measurements of the recoil-ion response of silicon carbide detectors to fast neutrons / F. H. Ruddy, J. G. Seidel and F. Franceschini. Measurement of the neutron spectrum of the HB-4 cold source at the high flux isotope reactor at Oak Ridge National Laboratory / J. L. Robertson and E. B. Iverson. Feasibility of cavity ring-down laser spectroscopy for dose rate monitoring on nuclear reactor / H. Tomita ... [et al.]. Measuring transistor damage factors in a non-stable defect environment / D. B. King ... [et al.]. Neutron-detection based monitoring of void effects in boiling water reactors / J. Loberg ... [et al.] -- Poster session 1: Power reactor surveillance, retrospective dosimetry, benchmarks and inter-comparisons, adjustment methods, experimental techniques, transport calculations. Improved diagnostics for analysis of a reactor pulse radiation environment / S. M. Luker ... [et al.]. Simulation of the response of silicon carbide fast neutron detectors / F. Franceschini, F. H. Ruddy and B. Petrović. NSV A-3: a computer code for least-squares adjustment of neutron spectra and measured dosimeter responses / J. G

  7. Pressurized fluidized bed reactor and a method of operating the same

    DOEpatents

    Isaksson, J.

    1996-02-20

    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.

  8. 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.

  9. 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.

  10. A brief history of design studies on innovative nuclear reactors

    NASA Astrophysics Data System (ADS)

    Sekimoto, Hiroshi

    2014-09-01

    In a short period after the success of CP1, many types of nuclear reactors were proposed and investigated. However, soon only a small number of reactors were selected for practical use. Around 1970, only LWRs with small number of CANDUs were operated in the western world, and FBRs were under development. It was about the time when Apollo moon landing was accomplished. However, at the same time, the future of human being was widely considered pessimistic and Limits to Growth was published. In the end of 1970's the TMI accident occurred and many nuclear reactor contracts were cancelled in USA and any more contracts had not been concluded until recent years. From the reflection of this accident, many Inherent Safe Reactors (ISRs) were proposed, though none of them were constructed. A common idea of ISRs is smallness of their size. Tokyo Institute of Technology (TokyoTech) held a symposium on small reactors, SR/TIT, in 1991, where many types of small ISRs were presented. Recently small reactors attract interest again. The most ideas employed in these reactors were the same discussed in SR/TIT. In 1980's the radioactive wastes from fuel cycle became a severe problem around the world. In TokyoTech, this issue was discussed mainly from the viewpoint of nuclear transmutations. The neutron economy became inevitable for these innovative nuclear reactors especially small long-life reactors and transmutation reactors.

  11. A brief history of design studies on innovative nuclear reactors

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

    Sekimoto, Hiroshi, E-mail: hsekimot@gmail.com

    2014-09-30

    In a short period after the success of CP1, many types of nuclear reactors were proposed and investigated. However, soon only a small number of reactors were selected for practical use. Around 1970, only LWRs with small number of CANDUs were operated in the western world, and FBRs were under development. It was about the time when Apollo moon landing was accomplished. However, at the same time, the future of human being was widely considered pessimistic and Limits to Growth was published. In the end of 1970’s the TMI accident occurred and many nuclear reactor contracts were cancelled in USAmore » and any more contracts had not been concluded until recent years. From the reflection of this accident, many Inherent Safe Reactors (ISRs) were proposed, though none of them were constructed. A common idea of ISRs is smallness of their size. Tokyo Institute of Technology (TokyoTech) held a symposium on small reactors, SR/TIT, in 1991, where many types of small ISRs were presented. Recently small reactors attract interest again. The most ideas employed in these reactors were the same discussed in SR/TIT. In 1980’s the radioactive wastes from fuel cycle became a severe problem around the world. In TokyoTech, this issue was discussed mainly from the viewpoint of nuclear transmutations. The neutron economy became inevitable for these innovative nuclear reactors especially small long-life reactors and transmutation reactors.« less

  12. Assessment of nuclear reactor concepts for low power space applications

    NASA Technical Reports Server (NTRS)

    Klein, Andrew C.; Gedeon, Stephen R.; Morey, Dennis C.

    1988-01-01

    The results of a preliminary small reactor concepts feasibility and safety evaluation designed to provide a first order validation of the nuclear feasibility and safety of six small reactor concepts are given. These small reactor concepts have potential space applications for missions in the 1 to 20 kWe power output range. It was concluded that low power concepts are available from the U.S. nuclear industry that have the potential for meeting both the operational and launch safety space mission requirements. However, each design has its uncertainties, and further work is required. The reactor concepts must be mated to a power conversion technology that can offer safe and reliable operation.

  13. 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.

  14. Computer study of emergency shutdowns of a 60-kilowatt reactor Brayton space power system

    NASA Technical Reports Server (NTRS)

    Tew, R. C.; Jefferies, K. S.

    1974-01-01

    A digital computer study of emergency shutdowns of a 60-kWe reactor Brayton power system was conducted. Malfunctions considered were (1) loss of reactor coolant flow, (2) loss of Brayton system gas flow, (3)turbine overspeed, and (4) a reactivity insertion error. Loss of reactor coolant flow was the most serious malfunction for the reactor. Methods for moderating the reactor transients due to this malfunction are considered.

  15. Deployment history and design considerations for space reactor power systems

    NASA Astrophysics Data System (ADS)

    El-Genk, Mohamed S.

    2009-05-01

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

  16. Neutron transmutation doping of silicon in the SAFARI-1 research reactor

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

    Louw, P.A.; Robertson, D.G.; Strydom, W.J.

    1994-12-31

    The SAFARI-1 research reactor has operated with an exemplary safety record since commissioning in 1965. As part of a commercialisation effort a silicon irradiation facility (SILIRAD) has been installed in the poolside region of SAFARI-1 for Neutron Transmutation Doping (NTD) of silicon. Commissioning of the facility took place in the last quarter of 1992 with a series of trial irradiations which were performed in close collaboration with Wacker Chemitronic of Germany. A methodology for the determination of irradiation times necessary to achieve the target resistivities was verified on the basis of the results from the trial irradiations. All production activitiesmore » are controlled by quality assurance procedures. To date some hundred and twelve silicon ingots (103 mm diameter) have been successfully irradiated on a commercial contract basis. The observed axial and radial variations in the resistivity profile of the ingots are very small compared to the profiles associated with conventionally doped silicon and small tolerances on target resistivities are attained. In this paper an overview of the design and characterisation of SILIRAD is given and the methods applied that ensure a quality product are described. Results obtained from trial and production irradiations are presented and the envisaged future modifications to SILIRAD discussed.« less

  17. 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.

  18. 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.

  19. 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.

  20. Recycled and virgin plastic carriers in hybrid reactors for wastewater treatment.

    PubMed

    Paul, Etienne; Wolff, Delmira Beatriz; Ochoa, Juan Carlos; da Costa, Rejane Helena Ribeiro

    2007-07-01

    The reduction of organic and nitrogen pollution of wastewater was investigated in two hybrid reactors and compared with the reduction obtained by using a conventional activated sludge reactor (ASR) run as a control. Both HR-1 and HR-2 were activated sludge systems where a low-density carrier, P1 (polyethylene) for HR-1 and P2 (recycled plastics) for HR-2, was added. Firstly, the three reactors were operated at 10 days Suspended Solid Retention Time (SRT(SS)), leading to a complete nitrification. Secondly, the SRT(SS) for each reactor was lowered to 3 days. Nitrification was lost for the ASR but remained complete for HR's. Respirometric techniques were used to measure fixed or suspended biomass activities for heterotrophic and autotrophic biomass. More than 90% of the autotrophic activity was found on the supports whatever the SRT(SS) used. The results may underline the role of the carrier geometry or surface characteristics on the autotrophic/heterotrophic microorganism distribution.

  1. D-He-3 spherical torus fusion reactor system study

    NASA Astrophysics Data System (ADS)

    Macon, William A., Jr.

    1992-04-01

    This system study extrapolates present physics knowledge and technology to predict the anticipated characteristics of D-He3 spherical torus fusion reactors and their sensitivity to uncertainties in important parameters. Reference cases for steady-state 1000 MWe reactors operating in H-mode in both the 1st stability regime and the 2nd stability regime were developed and assessed quantitatively. These devices would a very small aspect ratio (A=1,2), a major radius of about 2.0 m, an on-axis magnetic field less than 2 T, a large plasma current (80-120 MA) dominated by the bootstrap effect, and high plasma beta (greater than O.6). The estimated cost of electricity is in the range of 60-90 mills/kW-hr, assuming the use of a direct energy conversion system. The inherent safety and environmental advantages of D-He3 fusion indicate that this reactor concept could be competitive with advanced fission breeder reactors and large-scale solar electric plants by the end of the 21st century if research and development can produce the anticipated physics and technology advances.

  2. 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.

  3. 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)

  4. Reactor

    DOEpatents

    Evans, Robert M.

    1976-10-05

    1. A neutronic reactor having a moderator, coolant tubes traversing the moderator from an inlet end to an outlet end, bodies of material fissionable by neutrons of thermal energy disposed within the coolant tubes, and means for circulating water through said coolant tubes characterized by the improved construction wherein the coolant tubes are constructed of aluminum having an outer diameter of 1.729 inches and a wall thickness of 0.059 inch, and the means for circulating a liquid coolant through the tubes includes a source of water at a pressure of approximately 350 pounds per square inch connected to the inlet end of the tubes, and said construction including a pressure reducing orifice disposed at the inlet ends of the tubes reducing the pressure of the water by approximately 150 pounds per square inch.

  5. 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.

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

    PubMed

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

    2010-11-01

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

  7. 10 CFR 71.97 - Advance notification of shipment of irradiated reactor fuel and nuclear waste.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... notification of shipment of irradiated reactor fuel and nuclear waste. (a)(1) As specified in paragraphs (b... shipment of irradiated reactor fuel or nuclear waste must contain the following information: (1) The name... nuclear waste shipment; (2) A description of the irradiated reactor fuel or nuclear waste contained in the...

  8. 10 CFR 71.97 - Advance notification of shipment of irradiated reactor fuel and nuclear waste.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... notification of shipment of irradiated reactor fuel and nuclear waste. (a)(1) As specified in paragraphs (b... shipment of irradiated reactor fuel or nuclear waste must contain the following information: (1) The name... nuclear waste shipment; (2) A description of the irradiated reactor fuel or nuclear waste contained in the...

  9. FMDP reactor alternative summary report. Volume 1 - existing LWR alternative

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

    Greene, S.R.; Bevard, B.B.

    1996-10-07

    Significant quantities of weapons-usable fissile materials [primarily plutonium and highly enriched uranium (HEU)] are becoming surplus to national defense needs in both the United States and Russia. These stocks of fissile materials pose significant dangers to national and international security. The dangers exist not only in the potential proliferation of nuclear weapons but also in the potential for environmental, safety, and health (ES&H) consequences if surplus fissile materials are not properly managed. This document summarizes the results of analysis concerned with existing light water reactor plutonium disposition alternatives.

  10. The role of nuclear reactors in space exploration and development

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

    Lipinski, R.J.

    2000-07-01

    The United States has launched more than 20 radioisotopic thermoelectric generators (RTGs) into space over the past 30 yr but has launched only one nuclear reactor, and that was in 1965. Russia has launched more than 30 reactors. The RTGs use the heat of alpha decay of {sup 238}Pu for power and typically generate <1 kW of electricity. Apollo, Pioneer, Voyager, Viking, Galileo, Ulysses, and Cassini all used RTGs. Space reactors use the fission energy of {sup 235}U; typical designs are for 100 to 1000 kW of electricity. The only US space reactor launch (SNAP-10A) was a demonstration mission. Onemore » reason for the lack of space reactor use by the United States was the lack of space missions that required high power. But, another was the assumed negative publicity that would accompany a reactor launch. The net result is that all space reactor programs after 1970 were terminated before an operating space reactor could be developed, and they are now many years from recovering the ability to build them. Two major near-term needs for space reactors are the human exploration of Mars and advanced missions to and beyond the orbit of Jupiter. To help obtain public acceptance of space reactors, one must correct some of the misconceptions concerning space reactors and convey the following facts to the public and to decision makers: Space reactors are 1000 times smaller in power and size than a commercial power reactor. A space reactor at launch is only as radioactive as a pile of dirt 60 m (200 ft) across. A space reactor contains no plutonium at launch. It does not become significantly radioactive until it is turned on, and it will be engineered so that no launch accident can turn it on, even if that means fueling it after launch. The reactor will not be turned on until it is in a high stable orbit or even on an earth-escape trajectory for some missions. The benefits of space reactors are that they give humanity a stairway to the planets and perhaps the stars. They open a

  11. Monte-Carlo Simulations of the Nuclear Energy Deposition Inside the CARMEN-1P Differential Calorimeter Irradiated into OSIRIS Reactor

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

    Amharrak, H.; Reynard-Carette, C.; Carette, M.

    The nuclear heating measurements in Material Testing Reactors (MTRs) are crucial for the study of nuclear materials and fuels under irradiation. The reference measurements of this nuclear heating are especially performed by a differential calorimeter including a graphite sample material. These measurements are then used for other experimental conditions in order to predict the nuclear heating and thermal conditions induced in the irradiation devices. Nuclear heating is a great deal of interest at the moment as the measurement of such heating is an important issue for MTRs reactors. This need is especially generated by the new Jules Horowitz Reactor (JHR),more » under construction at CEA/Cadarache 'French Alternative Energies and Atomic Energy Commission'. This new reactor, that will be operational in late 2019, is a new facility for the nuclear research on materials and fuels. Indeed the expected nuclear heating rate is about 20 W/g for nominal capacity of 100 MW. The present Monte Carlo calculation works belong to the IN-CORE (Instrumentation for Nuclear radiation and Calorimetry On line in Reactor): a joint research program between the CEA and Aix- Marseille University in 2009. One scientific aim of this program is to design and develop a multi-sensors device, called CARMEN, dedicated to the measurements of main physical parameters simultaneously encountered inside JHR's experimental channels (core and reflector) such as neutron fluxes, photon fluxes, temperature, and nuclear heating. A first prototype was already developed. This prototype includes two mock-ups dedicated respectively to neutronic measurements (CARMEN-1N) and to photonic measurements (CARMEN-1P) with in particular a specific differential calorimeter. Two irradiation campaigns were performed successfully in the periphery of OSIRIS reactor (a MTR located at Saclay, France) in 2012 for nuclear heating levels up to 2 W/g. First Monte Carlo calculations reduced to the graphite sample of the

  12. 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.

  13. Implications of Fast Reactor Transuranic Conversion Ratio

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

    Steven J. Piet; Edward A. Hoffman; Samuel E. Bays

    2010-11-01

    Theoretically, the transuranic conversion ratio (CR), i.e. the transuranic production divided by transuranic destruction, in a fast reactor can range from near zero to about 1.9, which is the average neutron yield from Pu239 minus 1. In practice, the possible range will be somewhat less. We have studied the implications of transuranic conversion ratio of 0.0 to 1.7 using the fresh and discharge fuel compositions calculated elsewhere. The corresponding fissile breeding ratio ranges from 0.2 to 1.6. The cases below CR=1 (“burners”) do not have blankets; the cases above CR=1 (“breeders”) have breeding blankets. The burnup was allowed to floatmore » while holding the maximum fluence to the cladding constant. We graph the fuel burnup and composition change. As a function of transuranic conversion ratio, we calculate and graph the heat, gamma, and neutron emission of fresh fuel; whether the material is “attractive” for direct weapon use using published criteria; the uranium utilization and rate of consumption of natural uranium; and the long-term radiotoxicity after fuel discharge. For context, other cases and analyses are included, primarily once-through light water reactor (LWR) uranium oxide fuel at 51 MWth-day/kg-iHM burnup (UOX-51). For CR<1, the heat, gamma, and neutron emission increase as material is recycled. The uranium utilization is at or below 1%, just as it is in thermal reactors as both types of reactors require continuing fissile support. For CR>1, heat, gamma, and neutron emission decrease with recycling. The uranium utilization exceeds 1%, especially as all the transuranic elements are recycled. exceeds 1%, especially as all the transuranic elements are recycled. At the system equilibrium, heat and gamma vary by somewhat over an order of magnitude as a function of CR. Isotopes that dominate heat and gamma emission are scattered throughout the actinide chain, so the modest impact of CR is unsurprising. Neutron emitters are preferentially

  14. ENGINEERING TEST REACTOR (ETR) BUILDING, TRA642. CONTEXTUAL VIEW, CAMERA FACING ...

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

    ENGINEERING TEST REACTOR (ETR) BUILDING, TRA-642. CONTEXTUAL VIEW, CAMERA FACING EAST. VERTICAL METAL SIDING. ROOF IS SLIGHTLY ELEVATED AT CENTER LINE FOR DRAINAGE. WEST SIDE OF ETR COMPRESSOR BUILDING, TRA-643, PROJECTS TOWARD LEFT AT FAR END OF ETR BUILDING. INL NEGATIVE NO. HD46-37-1. Mike Crane, Photographer, 4/2005 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

  15. 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.

  16. Microprocessor tester for the treat upgrade reactor trip system

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

    Lenkszus, F.R.; Bucher, R.G.

    1984-01-01

    The upgrading of the Transient Reactor Test (TREAT) Facility at ANL-Idaho has been designed to provide additional experimental capabilities for the study of core disruptive accident (CDA) phenomena. In addition, a programmable Automated Reactor Control System (ARCS) will permit high-power transients up to 11,000 MW having a controlled reactor period of from 15 to 0.1 sec. These modifications to the core neutronics will improve simulation of LMFBR accident conditions. Finally, a sophisticated, multiply-redundant safety system, the Reactor Trip System (RTS), will provide safe operation for both steady state and transient production operating modes. To insure that this complex safety systemmore » is functioning properly, a Dedicated Microprocessor Tester (DMT) has been implemented to perform a thorough checkout of the RTS prior to all TREAT operations.« less

  17. Titer-plate formatted continuous flow thermal reactors: Design and performance of a nanoliter reactor

    PubMed Central

    Chen, Pin-Chuan; Park, Daniel S.; You, Byoung-Hee; Kim, Namwon; Park, Taehyun; Soper, Steven A.; Nikitopoulos, Dimitris E.; Murphy, Michael C.

    2010-01-01

    Arrays of continuous flow thermal reactors were designed, configured, and fabricated in a 96-device (12 × 8) titer-plate format with overall dimensions of 120 mm × 96 mm, with each reactor confined to a 8 mm × 8 mm footprint. To demonstrate the potential, individual 20-cycle (740 nL) and 25-cycle (990 nL) reactors were used to perform the continuous flow polymerase chain reaction (CFPCR) for amplification of DNA fragments of different lengths. Since thermal isolation of the required temperature zones was essential for optimal biochemical reactions, three finite element models, executed with ANSYS (v. 11.0, Canonsburg, PA), were used to characterize the thermal performance and guide system design: (1) a single device to determine the dimensions of the thermal management structures; (2) a single CFPCR device within an 8 mm × 8 mm area to evaluate the integrity of the thermostatic zones; and (3) a single, straight microchannel representing a single loop of the spiral CFPCR device, accounting for all of the heat transfer modes, to determine whether the PCR cocktail was exposed to the proper temperature cycling. In prior work on larger footprint devices, simple grooves between temperature zones provided sufficient thermal resistance between zones. For the small footprint reactor array, 0.4 mm wide and 1.2 mm high fins were necessary within the groove to cool the PCR cocktail efficiently, with a temperature gradient of 15.8°C/mm, as it flowed from the denaturation zone to the renaturation zone. With temperature tolerance bands of ±2°C defined about the nominal temperatures, more than 72.5% of the microchannel length was located within the desired temperature bands. The residence time of the PCR cocktail in each temperature zone decreased and the transition times between zones increased at higher PCR cocktail flow velocities, leading to less time for the amplification reactions. Experiments demonstrated the performance of the CFPCR devices as a function of flow

  18. Station Blackout Analysis of HTGR-Type Experimental Power Reactor

    NASA Astrophysics Data System (ADS)

    Syarip; Zuhdi, Aliq; Falah, Sabilul

    2018-01-01

    The National Nuclear Energy Agency of Indonesia has decided to build an experimental power reactor of high-temperature gas-cooled reactor (HTGR) type located at Puspiptek Complex. The purpose of this project is to demonstrate a small modular nuclear power plant that can be operated safely. One of the reactor safety characteristics is the reliability of the reactor to the station blackout (SBO) event. The event was observed due to relatively high disturbance frequency of electricity network in Indonesia. The PCTRAN-HTR functional simulator code was used to observe fuel and coolant temperature, and coolant pressure during the SBO event. The reactor simulated at 10 MW for 7200 s then the SBO occurred for 1-3 minutes. The analysis result shows that the reactor power decreases automatically as the temperature increase during SBO accident without operator’s active action. The fuel temperature increased by 36.57 °C every minute during SBO and the power decreased by 0.069 MW every °C fuel temperature rise at the condition of anticipated transient without reactor scram. Whilst, the maximum coolant (helium) temperature and pressure are 1004 °C and 9.2 MPa respectively. The maximum fuel temperature is 1282 °C, this value still far below the fuel temperature limiting condition i.e. 1600 °C, its mean that the HTGR has a very good inherent safety system.

  19. Microchannel Reactor System for Catalytic Hydrogenation

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

    Adeniyi Lawal; Woo Lee; Ron Besser

    2010-12-22

    We successfully demonstrated a novel process intensification concept enabled by the development of microchannel reactors, for energy efficient catalytic hydrogenation reactions at moderate temperature, and pressure, and low solvent levels. We designed, fabricated, evaluated, and optimized a laboratory-scale microchannel reactor system for hydrogenation of onitroanisole and a proprietary BMS molecule. In the second phase of the program, as a prelude to full-scale commercialization, we designed and developed a fully-automated skid-mounted multichannel microreactor pilot plant system for multiphase reactions. The system is capable of processing 1 – 10 kg/h of liquid substrate, and an industrially relevant immiscible liquid-liquid was successfully demonstratedmore » on the system. Our microreactor-based pilot plant is one-of-akind. We anticipate that this process intensification concept, if successfully demonstrated, will provide a paradigm-changing basis for replacing existing energy inefficient, cost ineffective, environmentally detrimental slurry semi-batch reactor-based manufacturing practiced in the pharmaceutical and fine chemicals industries.« less

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

    DTIC Science & Technology

    2009-06-01

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

  1. 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)

  2. Characteristics of potential repository wastes: Volume 4, Appendix 4A, Nuclear reactors at educational institutions of the United States; Appendix 4B, Data sheets for nuclear reactors at educational institutions; Appendix 4C, Supplemental data for Fort St. Vrain spent fuel; Appendix 4D, Supplemental data for Peach Bottom 1 spent fuel; Appendix 4E, Supplemental data for Fast Flux Test Facility

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

    Not Available

    1992-07-01

    Volume 4 contains the following appendices: nuclear reactors at educational institutions in the United States; data sheets for nuclear reactors at educational institutions in the United States(operational reactors and shut-down reactors); supplemental data for Fort St. Vrain spent fuel; supplemental data for Peach Bottom 1 spent fuel; and supplemental data for Fast Flux Test Facility.

  3. 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.

  4. 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.

  5. 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.

  6. 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.

  7. Fuel supply of nuclear power industry with the introduction of fast reactors

    NASA Astrophysics Data System (ADS)

    Muraviev, E. V.

    2014-12-01

    The results of studies conducted for the validation of the updated development strategy for nuclear power industry in Russia in the 21st century are presented. Scenarios with different options for the reprocessing of spent fuel of thermal reactors and large-scale growth of nuclear power industry based on fast reactors of inherent safety with a breeding ratio of ˜1 in a closed nuclear fuel cycle are considered. The possibility of enhanced fuel breeding in fast reactors is also taken into account in the analysis. The potential to establish a large-scale nuclear power industry that covers 100% of the increase in electric power requirements in Russia is demonstrated. This power industry may be built by the end of the century through the introduction of fast reactors (replacing thermal ones) with a gross uranium consumption of up to ˜1 million t and the termination of uranium mining even if the reprocessing of spent fuel of thermal reactors is stopped or suffers a long-term delay.

  8. Method of operating a neutronic reactor

    DOEpatents

    Woods, Wallace K.

    1976-05-25

    1. A method of operating a reactor having an active portion of a given length between a charging end and a discharging end, a first end region of the reactor extending from the charging end for one-quarter to one-third of said given length, a second end region extending from the discharging end for one-quarter to one-third of said given length, and a middle region extending between said end regions, said method comprising the steps of inserting end to end in the active region through the charging end a first group of bodies filling the middle region and a second group of bodies filling the first end region, irradiating the first and second groups of bodies while in the middle and first end regions, removing the first group from the reactor through the second end region, shifting the second group through the middle region to the second end region, inserting new first and second groups of bodies through the charging face into the middle and first end regions of the reactor, respectively, and irradiating the original second group and the new first and second groups while in the second end, middle, and first end regions, respectively, removing the original second group and the new first group from the reactor through the second end region, shifting the new second group through the middle region to the second end region, and irradiating the new second group again, whereby the first groups of bodies are irradiated only once and the second groups are irradiated twice.

  9. 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.

  10. Liquid fuel molten salt reactors for thorium utilization

    DOE PAGES

    Gehin, Jess C.; Powers, Jeffrey J.

    2016-04-08

    Molten salt reactors (MSRs) represent a class of reactors that use liquid salt, usually fluoride- or chloride-based, as either a coolant with a solid fuel (such as fluoride salt-cooled high temperature reactors) or as a combined coolant and fuel with fuel dissolved in a carrier salt. For liquid-fuelled MSRs, the salt can be processed online or in a batch mode to allow for removal of fission products as well as introduction of fissile fuel and fertile materials during reactor operation. The MSR is most commonly associated with the 233U/thorium fuel cycle, as the nuclear properties of 233U combined with themore » online removal of parasitic absorbers allow for the ability to design a thermal-spectrum breeder reactor; however, MSR concepts have been developed using all neutron energy spectra (thermal, intermediate, fast, and mixed-spectrum zoned concepts) and with a variety of fuels including uranium, thorium, plutonium, and minor actinides. Early MSR work was supported by a significant research and development (R&D) program that resulted in two experimental systems operating at ORNL in the 1960s, the Aircraft Reactor Experiment and the Molten Salt Reactor Experiment. Subsequent design studies in the 1970s focusing on thermal-spectrum thorium-fueled systems established reference concepts for two major design variants: (1) a molten salt breeder reactor (MSBR), with multiple configurations that could breed additional fissile material or maintain self-sustaining operation; and (2) a denatured molten salt reactor (DMSR) with enhanced proliferation-resistance. T MSRs has been selected as one of six most promising Generation IV systems and development activities have been seen in fast-spectrum MSRs, waste-burning MSRs, MSRs fueled with low-enriched uranium (LEU), as well as more traditional thorium fuel cycle-based MSRs. This study provides an historical background of MSR R&D efforts, surveys and summarizes many of the recent development, and provides analysis comparing

  11. 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.

  12. 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.

  13. Cyclic crack growth behavior of reactor pressure vessel steels in light water reactor environments

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

    Van Der Sluys, W.A.; Emanuelson, R.H.

    1986-01-01

    During normal operation light water reactor (LWR) pressure vessels are subjected to a variety of transients resulting in time varying stresses. Consequently, fatigue and environmentally assisted fatigue are growth mechanisms relevant to flaws in these pressure vessels. In order to provide a better understanding of the resistance of nuclear pressure vessel steels to flaw growth process, a series of fracture mechanics experiments were conducted to generate data on the rate of cyclic crack growth in SA508-2 and SA533b-1 steels in simulated 550/sup 0/F boiling water reactor (BWR) and 550/sup 0/F pressurized water reactor (PWR) environments. Areas investigated over the coursemore » of the test program included the effects of loading frequency and r ratio (Kmin-Kmax) on crack growth rate as a function of the stress intensity factor (deltaK) range. In addition, the effect of sulfur content of the test material on the cyclic crack growth rate was studied. Cyclic crack growth rates were found to be controlled by deltaK, R ratio, and loading frequency. The sulfur impurity content of the reactor pressure vessel steels studied had a significant effect on the cyclic crack growth rates. The higher growth rates were always associated with materials of higher sulfur content. For a given level of sulfur, growth rates were in a 550/sup 0/F simulated BWR environment than in a 550/sup 0/F simulated PWR environment. In both environments cyclic crack growth rates were a strong function of the loading frequency.« less

  14. Kr-85m activity as burnup measurement indicator in a pebble bed reactor based on ORIGEN2.1 Computer Simulation

    NASA Astrophysics Data System (ADS)

    Husnayani, I.; Udiyani, P. M.; Bakhri, S.; Sunaryo, G. R.

    2018-02-01

    Pebble Bed Reactor (PBR) is a high temperature gas-cooled reactor which employs graphite as a moderator and helium as a coolant. In a multi-pass PBR, burnup of the fuel pebble must be measured in each cycle by online measurement in order to determine whether the fuel pebble should be reloaded into the core for another cycle or moved out of the core into spent fuel storage. One of the well-known methods for measuring burnup is based on the activity of radionuclide decay inside the fuel pebble. In this work, the activity and gamma emission of Kr-85m were studied in order to investigate the feasibility of Kr-85m as burnup measurement indicator in a PBR. The activity and gamma emission of Kr-85 were estimated using ORIGEN2.1 computer code. The parameters of HTR-10 were taken as a case study in performing ORIGEN2.1 simulation. The results show that the activity revolution of Kr-85m has a good relationship with the burnup of the pebble fuel in each cycle. The Kr-85m activity reduction in each burnup step,in the range of 12% to 4%, is considered sufficient to show the burnup level in each cycle. The gamma emission of Kr-85m is also sufficiently high which is in the order of 1010 photon/second. From these results, it can be concluded that Kr-85m is suitable to be used as burnup measurement indicator in a pebble bed reactor.

  15. 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.

  16. The slightly-enriched spectral shift control reactor

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

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

    1991-11-01

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

  17. Isotopic composition and neutronics of the Okelobondo natural reactor

    NASA Astrophysics Data System (ADS)

    Palenik, Christopher Samuel

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

  18. 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.

  19. 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.

  20. 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.

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

    NASA Astrophysics Data System (ADS)

    Chongkum, Somporn

    2002-01-01

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

  2. Electrons to Reactors Multiscale Modeling: Catalytic CO Oxidation over RuO 2

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

    Sutton, Jonathan E.; Lorenzi, Juan M.; Krogel, Jaron T.

    First-principles kinetic Monte Carlo (1p-kMC) simulations for CO oxidation on two RuO 2 facets, RuO 2(110) and RuO 2(111), were coupled to the computational fluid dynamics (CFD) simulations package MFIX, and reactor-scale simulations were then performed. 1p-kMC coupled with CFD has recently been shown as a feasible method for translating molecular scale mechanistic knowledge to the reactor scale, enabling comparisons to in situ and online experimental measurements. Only a few studies with such coupling have been published. This work incorporates multiple catalytic surface facets into the scale-coupled simulation, and three possibilities were investigated: the two possibilities of each facet individuallymore » being the dominant phase in the reactor, and also the possibility that both facets were present on the catalyst particles in the ratio predicted by an ab initio thermodynamics-based Wulff construction. When lateral interactions between adsorbates were included in the 1p-kMC simulations, the two surfaces, RuO 2(110) and RuO 2(111), were found to be of similar order-of-magnitude in activity for the pressure range of 1 × 10 –4 bar to 1 bar, with the RuO 2(110) surface-termination showing more simulated activity than the RuO 2(111) surface-termination. Coupling between the 1p-kMC and CFD was achieved with a lookup table generated by the error-based modified Shepard interpolation scheme. Isothermal reactor scale simulations were performed and compared to two separate experimental studies, conducted with reactant partial pressures of ≤0.1 bar. Simulations without an isothermality restriction were also conducted and showed that the simulated temperature gradient across the catalytic reactor bed is <0.5 K, which validated the use of the isothermality restriction for investigating the reactor-scale phenomenological temperature dependences. The approach with the Wulff construction based reactor simulations reproduced a trend similar to one experimental data set

  3. Electrons to Reactors Multiscale Modeling: Catalytic CO Oxidation over RuO 2

    DOE PAGES

    Sutton, Jonathan E.; Lorenzi, Juan M.; Krogel, Jaron T.; ...

    2018-04-20

    First-principles kinetic Monte Carlo (1p-kMC) simulations for CO oxidation on two RuO 2 facets, RuO 2(110) and RuO 2(111), were coupled to the computational fluid dynamics (CFD) simulations package MFIX, and reactor-scale simulations were then performed. 1p-kMC coupled with CFD has recently been shown as a feasible method for translating molecular scale mechanistic knowledge to the reactor scale, enabling comparisons to in situ and online experimental measurements. Only a few studies with such coupling have been published. This work incorporates multiple catalytic surface facets into the scale-coupled simulation, and three possibilities were investigated: the two possibilities of each facet individuallymore » being the dominant phase in the reactor, and also the possibility that both facets were present on the catalyst particles in the ratio predicted by an ab initio thermodynamics-based Wulff construction. When lateral interactions between adsorbates were included in the 1p-kMC simulations, the two surfaces, RuO 2(110) and RuO 2(111), were found to be of similar order-of-magnitude in activity for the pressure range of 1 × 10 –4 bar to 1 bar, with the RuO 2(110) surface-termination showing more simulated activity than the RuO 2(111) surface-termination. Coupling between the 1p-kMC and CFD was achieved with a lookup table generated by the error-based modified Shepard interpolation scheme. Isothermal reactor scale simulations were performed and compared to two separate experimental studies, conducted with reactant partial pressures of ≤0.1 bar. Simulations without an isothermality restriction were also conducted and showed that the simulated temperature gradient across the catalytic reactor bed is <0.5 K, which validated the use of the isothermality restriction for investigating the reactor-scale phenomenological temperature dependences. The approach with the Wulff construction based reactor simulations reproduced a trend similar to one experimental data set

  4. 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…

  5. Modeling residence-time distribution in horizontal screw hydrolysis reactors

    DOE PAGES

    Sievers, David A.; Stickel, Jonathan J.

    2017-10-12

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

  6. Modeling residence-time distribution in horizontal screw hydrolysis reactors

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

    Sievers, David A.; Stickel, Jonathan J.

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

  7. 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

  8. 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.

  9. 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)

  10. Pilot plant operation of a nonadiabatic methanation reactor. [15 refs. ; Raney nickel catalyst

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

    Schehl, R.R.; Pennline, H.W.; Strakey, J.P.

    The design and operation of a pilot plant scale hybrid methanation reactor is discussed. The hybrid methanator, utilizing a finned, Raney nickel coated insert, consolidates features of the tube-wall and hot-gas-recycle methanation reactors. Data are presented from four tests lasting from 3/sup 1///sub 2/ weeks to three months. Topics discussed include conversion, product yields, catalyst properties, and reactor temperature profiles. A one-dimensional mathematical model capable of explaining reactor performance trends is employed.

  11. Design of a laboratory scale fluidized bed reactor

    NASA Astrophysics Data System (ADS)

    Wikström, E.; Andersson, P.; Marklund, S.

    1998-04-01

    The aim of this project was to construct a laboratory scale fluidized bed reactor that simulates the behavior of full scale municipal solid waste combustors. The design of this reactor is thoroughly described. The size of the laboratory scale fluidized bed reactor is 5 kW, which corresponds to a fuel-feeding rate of approximately 1 kg/h. The reactor system consists of four parts: a bed section, a freeboard section, a convector (postcombustion zone), and an air pollution control (APC) device system. The inside diameter of the reactor is 100 mm at the bed section and it widens to 200 mm in diameter in the freeboard section; the total height of the reactor is 1760 mm. The convector part consists of five identical sections; each section is 2700 mm long and has an inside diameter of 44.3 mm. The reactor is flexible regarding the placement and number of sampling ports. At the beginning of the first convector unit and at the end of each unit there are sampling ports for organic micropollutants (OMP). This makes it possible to study the composition of the flue gases at various residence times. Sampling ports for inorganic compounds and particulate matter are also placed in the convector section. All operating parameters, reactor temperatures, concentrations of CO, CO2, O2, SO2, NO, and NO2 are continuously measured and stored at selected intervals for further evaluation. These unique features enable full control over the fuel feed, air flows, and air distribution as well as over the temperature profile. Elaborate details are provided regarding the configuration of the fuel-feeding systems, the fluidized bed, the convector section, and the APC device. This laboratory reactor enables detailed studies of the formation mechanisms of OMP, such as polychlorinated dibenzo-p-dioxins (PCDDs), polychlorinated dibenzofurans (PCDFs), poly-chlorinated biphenyls (PCBs), and polychlorinated benzenes (PCBzs). With this system formation mechanisms of OMP occurring in both the combustion

  12. Void effect analysis of Pb-208 of fast reactors with modified CANDLE burn-up scheme

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

    Widiawati, Nina, E-mail: nina-widiawati28@yahoo.com; Su’ud, Zaki, E-mail: szaki@fi.itb.ac.id

    Void effect analysis of Pb-208 as coolant of fast reactors with modified candle burn-up scheme has been conducted. Lead cooled fast reactor (LFR) is one of the fourth-generation reactor designs. The reactor is designed with a thermal power output of 500 MWt. Modified CANDLE burn-up scheme allows the reactor to have long life operation by supplying only natural uranium as fuel cycle input. This scheme introducing discrete region, the fuel is initially put in region 1, after one cycle of 10 years of burn up it is shifted to region 2 and region 1 is filled by fresh natural uraniummore » fuel. The reactor is designed for 100 years with 10 regions arranged axially. The results of neutronic calculation showed that the void coefficients ranged from −0.6695443 % at BOC to −0.5273626 % at EOC for 500 MWt reactor. The void coefficients of Pb-208 more negative than Pb-nat. The results showed that the reactors with Pb-208 coolant have better level of safety than Pb-nat.« less

  13. 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.

  14. 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.

  15. 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.

  16. 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.

  17. 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

  18. Summary of space nuclear reactor power systems, 1983 - 1992

    NASA Astrophysics Data System (ADS)

    Buden, D.

    1993-08-01

    This report summarizes major developments in the last ten years which have greatly expanded the space nuclear reactor power systems technology base. In the SP-100 program, after a competition between liquid-metal, gas-cooled, thermionic, and heat pipe reactors integrated with various combinations of thermoelectric thermionic, Brayton, Rankine, and Stirling energy conversion systems, three concepts were selected for further evaluation. In 1985, the high-temperature (1,350 K), lithium-cooled reactor with thermoelectric conversion was selected for full scale development. Since then, significant progress has been achieved including the demonstration of a 7-y-life uranium nitride fuel pin. Progress on the lithium-cooled reactor with thermoelectrics has progressed from a concept, through a generic flight system design, to the design, development, and testing of specific components. Meanwhile, the USSR in 1987-88 orbited a new generation of nuclear power systems beyond the, thermoelectric plants on the RORSAT satellites. The US has continued to advance its own thermionic fuel element development, concentrating on a multicell fuel element configuration. Experimental work has demonstrated a single cell operating time of about 1 1/2-y. Technology advances have also been made in the Stirling engine; an advanced engine that operates at 1,050 K is ready for testing. Additional concepts have been studied and experiments have been performed on a variety of systems to meet changing needs; such as powers of tens-to-hundreds of megawatts and highly survivable systems of tens-of-kilowatts power.

  19. Treatment of oilfield wastewater in moving bed biofilm reactors using a novel suspended ceramic biocarrier.

    PubMed

    Dong, Zhiyong; Lu, Mang; Huang, Wenhui; Xu, Xiaochun

    2011-11-30

    In this study, a novel suspended ceramic carrier was prepared, which has high strength, optimum density (close to water), and high porosity. Two different carriers, unmodified and sepiolite-modified suspended ceramic carriers were used to feed two moving bed biofilm reactors (MBBRs) with a filling fraction of 50% to treat oilfield produced water. The hydraulic retention time (HRT) was varied from 36 to 10h. The results, during a monitoring period of 190 days, showed that removal efficiency of chemical oxygen demand was the highest in reactor 3 filled with the sepiolite-modified carriers, followed by reactor 2 filled with the unmodified carriers, with the lowest in reactor 1 (activated sludge reactor), at an HRT of 10h. Similar trends were found in the removal efficiencies of ammonia nitrogen and polycyclic aromatic hydrocarbons. Reactor 3 was more shock resistant than reactors 2 and 1. The results indicate that the suspended ceramic carrier is an excellent MBBR carrier. Copyright © 2011 Elsevier B.V. All rights reserved.

  20. 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.

  1. 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.

  2. 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.

  3. 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.

  4. 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.

  5. Utilization of thorium and U-ZrH1.6 fuels in various heterogeneous cores for TRIGA PUSPATI Reactor (RTP)

    NASA Astrophysics Data System (ADS)

    Damahuri, Abdul Hannan Bin; Mohamed, Hassan; Aziz Mohamed, Abdul; Idris, Faridah

    2018-01-01

    The use of thorium as nuclear fuel has been an appealing prospect for many years and will be great significance to nuclear power generation. There is an increasing need for more research on thorium as Malaysian government is currently active in the national Thorium Flagship Project, which was launched in 2014. The thorium project, which is still in phase 1, focuses on the research and development of the thorium extraction from mineral processing ore. Thus, the aim of the study is to investigate other alternative TRIGA PUSPATI Reactor (RTP) core designs that can fully utilize thorium. Currently, the RTP reactor has an average neutron flux of 2.797 x 1012 cm-2/s-1 and an effective multiplication factor, k eff, of 1.001. The RTP core has a circular array core configuration with six circular rings. Each ring consists of 6, 12, 18, 24, 30 or 36 U-ZrH1.6 fuel rods. There are three main type of uranium weight, namely 8.5, 12 and 20 wt.%. For this research, uranium zirconium hydride (U-ZrH1.6) fuel rods in the RTP core were replaced by thorium (ThO2) fuel rods. Seven core configurations with different thorium fuel rods placements were modelled in a 2D structure and simulated using Monte Carlo n-particle (MCNPX) code. Results show that the highest initial criticality obtained is around 1.35101. Additionally there is a significant discrepancy between results from previous study and the work because of the large estimated leakage probability of approximately 21.7% and 2D model simplification.

  6. Accident analysis of heavy water cooled thorium breeder reactor

    NASA Astrophysics Data System (ADS)

    Yulianti, Yanti; Su'ud, Zaki; Takaki, Naoyuki

    2015-04-01

    Thorium has lately attracted considerable attention because it is accumulating as a by-product of large scale rare earth mining. The objective of research is to analyze transient behavior of a heavy water cooled thorium breeder that is designed by Tokai University and Tokyo Institute of Technology. That is oxide fueled, PWR type reactor with heavy water as primary coolant. An example of the optimized core has relatively small moderator to fuel volume ratio (MFR) of 0.6 and the characteristics of the core are burn-up of 67 GWd/t, breeding ratio of 1.08, burn-up reactivity loss during cycles of < 0.2% dk/k, and negative coolant reactivity coefficient. One of the nuclear reactor accidents types examined here is Unprotected Transient over Power (UTOP) due to withdrawing of the control rod that result in the positive reactivity insertion so that the reactor power will increase rapidly. Another accident type is Unprotected Loss of Flow (ULOF) that caused by failure of coolant pumps. To analyze the reactor accidents, neutron distribution calculation in the nuclear reactor is the most important factor. The best expression for the neutron distribution is the Boltzmann transport equation. However, solving this equation is very difficult so that the space-time diffusion equation is commonly used. Usually, space-time diffusion equation is solved by employing a point kinetics approach. However, this approach is less accurate for a spatially heterogeneous nuclear reactor and the nuclear reactor with quite large reactivity input. Direct method is therefore used to solve space-time diffusion equation which consider spatial factor in detail during nuclear reactor accident simulation. Set of equations that obtained from full implicit finite-difference method is solved by using iterative methods. The indication of UTOP accident is decreasing macroscopic absorption cross-section that results large external reactivity, and ULOF accident is indicated by decreasing coolant flow. The

  7. Gravity Scaling of a Power Reactor Water Shield

    NASA Technical Reports Server (NTRS)

    Reid, Robert S.; Pearson, J. Boise

    2007-01-01

    A similarity analysis on a water-based reactor shield examined the effect of gravity on free convection between a reactor shield inner and outer vessel boundaries. Two approaches established similarity between operation on the Earth and the Moon: 1) direct scaling of Rayleigh number equating gravity-surface heat flux products, 2) temperature difference between the wall and thermal boundary layer held constant. Nusselt number for natural convection (laminar and turbulent) is assumed of form Nu = CRa(sup n).

  8. NEUTRONIC REACTOR CORE

    DOEpatents

    Thomson, W.B.; Corbin, A. Jr.

    1961-07-18

    An improved core for a gas-cooled power reactor which admits gas coolant at high temperatures while affording strong integral supporting structure and efficient moderation of neutrons is described. The multiplicities of fuel elements constituting the critical amassment of fissionable material are supported and confined by a matrix of metallic structure which is interspersed therebetween. Thermal insulation is interposed between substantially all of the metallic matrix and the fuel elements; the insulation then defines the principal conduit system for conducting the coolant gas in heat-transfer relationship with the fuel elements. The metallic matrix itseif comprises a system of ducts through which an externally-cooled hydrogeneous liquid, such as water, is circulated to serve as the principal neutron moderant for the core and conjointly as the principal coolant for the insulated metallic structure. In this way, use of substantially neutron transparent metals, such as aluminum, becomes possible for the supporting structure, despite the high temperatures of the proximate gas. The Aircraft Nuclear Propulsion program's "R-1" reactor design is a preferred embodiment.

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

    NASA Astrophysics Data System (ADS)

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

    2018-01-01

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

  10. Transformation products of clindamycin in moving bed biofilm reactor (MBBR).

    PubMed

    Ooi, Gordon T H; Escola Casas, Monica; Andersen, Henrik R; Bester, Kai

    2017-04-15

    Clindamycin is widely prescribed for its ability to treat a number of common bacterial infections. Thus, clindamycin enters wastewater via human excretion or disposal of unused medication and widespread detection of pharmaceuticals in rivers proves the insufficiency of conventional wastewater treatment plants in removing clindamycin. Recently, it has been discovered that attached biofilm reactors, e.g., moving bed biofilm reactors (MBBRs) obtain a higher removal of pharmaceuticals than conventional sludge wastewater treatment plants. Therefore, this study investigated the capability of MBBRs applied in the effluent of conventional wastewater treatment plants to remove clindamycin. First, a batch experiment was executed with a high initial concentration of clindamycin to identify the transformation products. It was shown that clindamycin can be removed from wastewater by MBBR and the treatment process converts clindamycin into the, possibly persistent, products clindamycin sulfoxide and N-desmethyl clindamycin as well as 3 other mono-oxygenated products. Subsequently, the removal kinetics of clindamycin and the formation of the two identified products were investigated in batch experiments using MBBR carriers from polishing and nitrifying reactors. Additionally, the presence of these two metabolites in biofilm-free wastewater effluent was studied. The nitrifying biofilm reactor had a higher biological activity with k-value of 0.1813 h -1 than the reactor with polishing biofilm (k = 0.0161 h -1 ) which again has a much higher biological activity for removal of clindamycin than of the suspended bacteria (biofilm-free control). Clindamycin sulfoxide was the main transformation product which was found in concentrations exceeding 10% of the initial clindamycin concentration after 1 day of MBBR treatment. Thus, MBBRs should not necessarily be considered as reactors mineralizing clindamycin as they perform transformation reactions at least to some extent. Copyright

  11. 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.

  12. 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.

  13. Mesophilic and thermophilic biotreatment of BTEX-polluted air in reactors.

    PubMed

    Mohammad, Balsam T; Veiga, María C; Kennes, Christian

    2007-08-15

    This study compares the removal of a mixture of benzene, toluene, ethylbenzene, and all three xylene isomers (BTEX) in mesophilic and thermophilic (50 degrees C) bioreactors. In the mesophilic reactor fungi became dominant after long-term operation, while bacteria dominated in the thermophilic unit. Microbial acclimation was achieved by exposing the biofilters to initial BTEX loads of 2-15 g m(-3) h(-1), at an empty bed residence time of 96 s. After adaptation, the elimination capacities ranged from 3 to 188 g m(-3) h(-1), depending on the inlet load, for the mesophilic biofilter with removal efficiencies reaching 96%. On the other hand, in the thermophilic reactor the average removal efficiency was 83% with a maximum elimination capacity of 218 g m(-3) h(-1). There was a clear positive relationship between temperature gradients as well as CO(2) production and elimination capacities across the biofilters. The gas phase was sampled at different depths along the reactors observing that the percentage pollutant removal in each section was strongly dependant on the load applied. The fate of individual alkylbenzene compounds was checked, showing the unusually high biodegradation rate of benzene at high loads under thermophilic conditions (100%) compared to its very low removal in the mesophilic reactor at such load (<10%). Such difference was less pronounced for the other pollutants. After 210 days of operation, the dry biomass content for the mesophilic and thermophilic reactors were 0.300 and 0.114 g g(-1) (support), respectively, reaching higher removals under thermophilic conditions with a lower biomass accumulation, that is, lower pressure drop. (c) 2007 Wiley Periodicals, Inc.

  14. Building on knowledge base of sodium cooled fast spectrum reactors to develop materials technology for fusion reactors

    NASA Astrophysics Data System (ADS)

    Raj, Baldev; Rao, K. Bhanu Sankara

    2009-04-01

    The alloys 316L(N) and Mod. 9Cr-1Mo steel are the major structural materials for fabrication of structural components in sodium cooled fast reactors (SFRs). Various factors influencing the mechanical behaviour of these alloys and different modes of deformation and failure in SFR systems, their analysis and the simulated tests performed on components for assessment of structural integrity and the applicability of RCC-MR code for the design and validation of components are highlighted. The procedures followed for optimal design of die and punch for the near net shape forming of petals of main vessel of 500 MWe prototype fast breeder reactor (PFBR); the safe temperature and strain rate domains established using dynamic materials model for forming of 316L(N) and 9Cr-1Mo steels components by various industrial processes are illustrated. Weldability problems associated with 316L(N) and Mo. 9Cr-1Mo are briefly discussed. The utilization of artificial neural network models for prediction of creep rupture life and delta-ferrite in austenitic stainless steel welds is described. The usage of non-destructive examination techniques in characterization of deformation, fracture and various microstructural features in SFR materials is briefly discussed. Most of the experience gained on SFR systems could be utilized in developing science and technology for fusion reactors. Summary of the current status of knowledge on various aspects of fission and fusion systems with emphasis on cross fertilization of research is presented.

  15. Anaerobic degradation of coconut husk leachate using UASB-reactor.

    PubMed

    Neena, C; Ambily, P S; Jisha, M S

    2007-07-01

    Reffing of coconut husk, the majorprocess in quality coir fibre extraction, causes serious pollution with brackish water lagoons of Kerala. An attempt is made to treat the coconut husk leachate by using a laboratory scale UASB-reactor The experiment was conducted with loading of leachate from 1 kg of fresh coconut husk. The anaerobic treatment was done continuously The parameters like VFA, pH, COD and polyphenols were analysed regularly during the evaluation of the reactor performance. The polyphenol, VFA and COD were diminished gradually with time. The pH of the reactor during the study was found to be in the range of 6-8. The biogas production was increased with loading and about 82% of the total COD/kg husk could be converted to biogas. The maximum polyphenol loading in the reactor was reached to about 298.51 mg/l of husk.

  16. 10 CFR 73.37 - Requirements for physical protection of irradiated reactor fuel in transit.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 10 Energy 2 2011-01-01 2011-01-01 false Requirements for physical protection of irradiated reactor... Requirements for physical protection of irradiated reactor fuel in transit. (a) Performance objectives. (1... of irradiated reactor fuel in excess of 100 grams in net weight of irradiated fuel, exclusive of...

  17. 10 CFR 73.37 - Requirements for physical protection of irradiated reactor fuel in transit.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 false Requirements for physical protection of irradiated reactor... Requirements for physical protection of irradiated reactor fuel in transit. (a) Performance objectives. (1... of irradiated reactor fuel in excess of 100 grams in net weight of irradiated fuel, exclusive of...

  18. 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

  19. 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.

  20. 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.

  1. 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.

  2. 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.

  3. Reactor-released radionuclides in Susquehanna River sediments

    USGS Publications Warehouse

    Olsen, C.R.; Larsen, I.L.; Cutshall, N.H.; Donoghue, J.F.; Bricker, O.P.; Simpson, H.J.

    1981-01-01

    Three Mile Island (TMI) and Peach Bottom (PB) reactors have introduced 137Cs, 134Cs, 60Co, 58Co and several other anthropogenic radionuclides into the lower Susquehanna River. Here we present the release history for these nuclides (Table 1) and radionuclide concentration data (Table 2) for sediment samples collected in the river and upper portions of the Chesapeake Bay (Fig. 1) within a few months after the 28 March 1979 loss-of-coolant-water problem at TMI. Although we found no evidence for nuclides characteristic of a ruptured fuel element, we did find nuclides characteristic of routine operations. Despite the TMI incident, more than 95% of the total 134Cs input to the Susquehanna has been a result of controlled low-level releases from the PB site. 134Cs activity released into the river is effectively trapped by sediments with the major zones of reactor-nuclide accumulation behind Conowingo Dam and in the upper portions of Chesapeake Bay. The reported distributions document the fate of reactor-released radionuclides and their extent of environmental contamination in the Susquehanna-Upper Chesapeake Bay System. ?? 1981 Nature Publishing Group.

  4. 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.

  5. High-performance recombinant protein production with Escherichia coli in continuously operated cascades of stirred-tank reactors.

    PubMed

    Schmideder, Andreas; Weuster-Botz, Dirk

    2017-07-01

    The microbial expression of intracellular, recombinant proteins in continuous bioprocesses suffers from low product concentrations. Hence, a process for the intracellular production of photoactivatable mCherry with Escherichia coli in a continuously operated cascade of two stirred-tank reactors was established to separate biomass formation (first reactor) and protein expression (second reactor) spatially. Cascades of miniaturized stirred-tank reactors were implemented, which enable the 24-fold parallel characterization of cascade processes and the direct scale-up of results to the liter scale. With PAmCherry concentrations of 1.15 g L -1 cascades of stirred-tank reactors improved the process performance significantly compared to production processes in chemostats. In addition, an optimized fed-batch process was outperformed regarding space-time yield (149 mg L -1  h -1 ). This study implicates continuous cascade processes to be a promising alternative to fed-batch processes for microbial protein production and demonstrates that miniaturized stirred-tank reactors can reduce the timeline and costs for cascade process characterization.

  6. 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

  7. 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.

  8. 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)

  9. Internally Heated Screw Pyrolysis Reactor (IHSPR) heat transfer performance study

    NASA Astrophysics Data System (ADS)

    Teo, S. H.; Gan, H. L.; Alias, A.; Gan, L. M.

    2018-04-01

    1.5 billion end-of-life tyres (ELT) were discarded globally each year and pyrolysis is considered the best solution to convert the ELT into valuable high energy-density products. Among all pyrolysis technologies, screw reactor is favourable. However, conventional screw reactor risks plugging issue due to its lacklustre heat transfer performance. An internally heated screw pyrolysis reactor (IHSPR) was developed by local renewable energy industry, which serves as the research subject for heat transfer performance study of this particular paper. Zero-load heating test (ZLHT) was first carried out to obtain the operational parameters of the reactor, followed by the one dimensional steady-state heat transfer analysis carried out using SolidWorks Flow Simulation 2016. Experiments with feed rate manipulations and pyrolysis products analyses were conducted last to conclude the study.

  10. 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)

  11. 10 CFR 50.58 - Hearings and report of the Advisory Committee on Reactor Safeguards.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 10 Energy 1 2011-01-01 2011-01-01 false Hearings and report of the Advisory Committee on Reactor... Hearings and report of the Advisory Committee on Reactor Safeguards. (a) Each application for a....22, or for a testing facility, shall be referred to the Advisory Committee on Reactor Safeguards for...

  12. 10 CFR 50.58 - Hearings and report of the Advisory Committee on Reactor Safeguards.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 10 Energy 1 2010-01-01 2010-01-01 false Hearings and report of the Advisory Committee on Reactor... Hearings and report of the Advisory Committee on Reactor Safeguards. (a) Each application for a....22, or for a testing facility, shall be referred to the Advisory Committee on Reactor Safeguards for...

  13. 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

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

    PubMed

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

    2014-01-01

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

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

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

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

    2016-11-01

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

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

    PubMed

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

    2017-09-16

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

  17. LOFT. Reactor arrives at containment building (TAN650), now being pushed ...

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

    LOFT. Reactor arrives at containment building (TAN-650), now being pushed by locomotive. Camera facing northerly. Note "Hello Dolly" and "PWR MTA No. 1" (pressurized water reactor mobile test assembly) signs. Date: 1973. INEEL negative no. 73-3710 - Idaho National Engineering Laboratory, Test Area North, Scoville, Butte County, ID

  18. 10 CFR 50.44 - Combustible gas control for nuclear power reactors.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... 10 Energy 1 2014-01-01 2014-01-01 false Combustible gas control for nuclear power reactors. 50.44... FACILITIES Standards for Licenses, Certifications, and Regulatory Approvals § 50.44 Combustible gas control... capability for ensuring a mixed atmosphere. (2) Combustible gas control. (i) All boiling water reactors with...

  19. 10 CFR 50.44 - Combustible gas control for nuclear power reactors.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 10 Energy 1 2010-01-01 2010-01-01 false Combustible gas control for nuclear power reactors. 50.44... FACILITIES Standards for Licenses, Certifications, and Regulatory Approvals § 50.44 Combustible gas control... capability for ensuring a mixed atmosphere. (2) Combustible gas control. (i) All boiling water reactors with...

  20. 10 CFR 50.44 - Combustible gas control for nuclear power reactors.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... 10 Energy 1 2012-01-01 2012-01-01 false Combustible gas control for nuclear power reactors. 50.44... FACILITIES Standards for Licenses, Certifications, and Regulatory Approvals § 50.44 Combustible gas control... capability for ensuring a mixed atmosphere. (2) Combustible gas control. (i) All boiling water reactors with...

  1. 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.

  2. From biofilm ecology to reactors: a focused review.

    PubMed

    Boltz, Joshua P; Smets, Barth F; Rittmann, Bruce E; van Loosdrecht, Mark C M; Morgenroth, Eberhard; Daigger, Glen T

    2017-04-01

    Biofilms are complex biostructures that appear on all surfaces that are regularly in contact with water. They are structurally complex, dynamic systems with attributes of primordial multicellular organisms and multifaceted ecosystems. The presence of biofilms may have a negative impact on the performance of various systems, but they can also be used beneficially for the treatment of water (defined herein as potable water, municipal and industrial wastewater, fresh/brackish/salt water bodies, groundwater) as well as in water stream-based biological resource recovery systems. This review addresses the following three topics: (1) biofilm ecology, (2) biofilm reactor technology and design, and (3) biofilm modeling. In so doing, it addresses the processes occurring in the biofilm, and how these affect and are affected by the broader biofilm system. The symphonic application of a suite of biological methods has led to significant advances in the understanding of biofilm ecology. New metabolic pathways, such as anaerobic ammonium oxidation (anammox) or complete ammonium oxidation (comammox) were first observed in biofilm reactors. The functions, properties, and constituents of the biofilm extracellular polymeric substance matrix are somewhat known, but their exact composition and role in the microbial conversion kinetics and biochemical transformations are still to be resolved. Biofilm grown microorganisms may contribute to increased metabolism of micro-pollutants. Several types of biofilm reactors have been used for water treatment, with current focus on moving bed biofilm reactors, integrated fixed-film activated sludge, membrane-supported biofilm reactors, and granular sludge processes. The control and/or beneficial use of biofilms in membrane processes is advancing. Biofilm models have become essential tools for fundamental biofilm research and biofilm reactor engineering and design. At the same time, the divergence between biofilm modeling and biofilm reactor

  3. Cultivation of aerobic granules in a novel configuration of sequencing batch airlift reactor.

    PubMed

    Rezaei, Laya Siroos; Ayati, Bita; Ganjidoust, Hossein

    2012-01-01

    Aerobic granules can be formed in sequencing batch airlift reactors (SBAR) and sequencing batch reactors (SBR). Comparing these two systems, the SBAR has excellent mixing condition, but due to a high height-to-diameter ratio (H/D), there is no performance capability at full scale at the present time. This research examined a novel configuration of SBAR at laboratory scale (with a box structure) for industrial wastewater treatment. To evaluate chemical oxygen demand (COD) removal efficiency and granule formation of the novel reactor (R1), in comparison a conventional SBAR (R2) was operated under similar conditions during the experimental period. R1 and R2 with working volumes of 3.6 L and 4.5 L, respectively, were used to cultivate aerobic granules. Both reactors were operated for 4 h per cycle. Experiments were done at different organic loading rates (OLRs) ranging from 0.6-4.5 kg COD/m3.d for R1 and from 0.72-5.4 kg COD/m3.d for R2. After 150 days of operation, large-sized black filamentous granules with diameters of 0.5-2 mm and 2-11 mm were formed in R1 and R2, respectively. In the second part of the experiment, the efficiency of removal of a toxic substance by aerobic granules was investigated using aniline as a carbon source with a concentration in the range 1.2-6.6 kg COD/m3.d and 1.44-7.92 kg COD/m3.d in R1 and R2, respectively. It was found that COD removal efficiency of the novel airlift reactor was over 97% and 94.5% using glucose and aniline as carbon sources, respectively. Sludge volume index (SVI) was also decreased to 30 mL/g by granulation in the novel airlift reactor.

  4. 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.

  5. 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.

  6. 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)

  7. 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.

  8. ETR, TRA642. ON GROUND FLOOR. THE 60TON ETR REACTOR VESSEL ...

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

    ETR, TRA-642. ON GROUND FLOOR. THE 60-TON ETR REACTOR VESSEL IS DROPPED INTO PLACE OVER PIT. KAISER USED TWO MULTI-BLOCK DRUM PULLEYS WITH A COMBINED CAPACITY OF 100 TONS AND A 100-TON DRUM HOIST. THE VESSEL WAS 35 1/2 FEET LONG. INL NEGATIVE NO. 56-4149. R.G. Larsen, Photographer, 12/18/1956 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

  9. 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.

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

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

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

    1993-07-01

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

  11. Summary of space nuclear reactor power systems, 1983--1992

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

    Buden, D.

    1993-08-11

    This report summarizes major developments in the last ten years which have greatly expanded the space nuclear reactor power systems technology base. In the SP-100 program, after a competition between liquid-metal, gas-cooled, thermionic, and heat pipe reactors integrated with various combinations of thermoelectric thermionic, Brayton, Rankine, and Stirling energy conversion systems, three concepts:were selected for further evaluation. In 1985, the high-temperature (1,350 K), lithium-cooled reactor with thermoelectric conversion was selected for full scale development. Since then, significant progress has been achieved including the demonstration of a 7-y-life uranium nitride fuel pin. Progress on the lithium-cooled reactor with thermoelectrics has progressedmore » from a concept, through a generic flight system design, to the design, development, and testing of specific components. Meanwhile, the USSR in 1987--88 orbited a new generation of nuclear power systems beyond the, thermoelectric plants on the RORSAT satellites. The US has continued to advance its own thermionic fuel element development, concentrating on a multicell fuel element configuration. Experimental work has demonstrated a single cell operating time of about 1 1/2-y. Technology advances have also been made in the Stirling engine; an advanced engine that operates at 1,050 K is ready for testing. Additional concepts have been studied and experiments have been performed on a variety of systems to meet changing needs; such as powers of tens-to-hundreds of megawatts and highly survivable systems of tens-of-kilowatts power.« less

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

    PubMed

    Bian, Wenjuan; Ying, Xiangli; Shi, Junwen

    2009-03-15

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

  13. Psychosocial effects of restarting a TMI reactor

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

    Not Available

    1990-01-01

    ORNL is studying human responses to hazardous environmental phenomena. This study attempts to understand the human behavior associated with the restart of TMI-1 Reactor after a nuclear event occurred at TMI-2.

  14. 10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... light-water nuclear power reactors. 50.46 Section 50.46 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC... reactors. (a)(1)(i) Each boiling or pressurized light-water nuclear power reactor fueled with uranium oxide... behavior of the reactor system during a loss-of-coolant accident. Comparisons to applicable experimental...

  15. FAST NEUTRONIC REACTOR

    DOEpatents

    Snell, A.H.

    1957-12-01

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

  16. 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.

  17. 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

  18. Tritium Mitigation/Control for Advanced Reactor System

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

    Sun, Xiaodong; Christensen, Richard; Saving, John P.

    A tritium removal facility, which is similar to the design used for tritium recovery in fusion reactors, is proposed in this study for fluoride-salt-cooled high-temperature reactors (FHRs) to result in a two-loop FHR design with the elimination of an intermediate loop. Using this approach, an economic benefit can potentially be obtained by removing the intermediate loop, while the safety concern of tritium release can be mitigated. In addition, an intermediate heat exchanger (IHX) that can yield a similar tritium permeation rate to the production rate of 1.9 Ci/day in a 1,000 MWe PWR needs to be designed to prevent themore » residual tritium that is not captured in the tritium removal system from escaping into the power cycle and ultimately the environment. The main focus of this study is to aid the mitigation of tritium permeation issue from the FHR primary side to significantly reduce the concentration of tritium in the secondary side and the process heat application side (if applicable). The goal of the research is to propose a baseline FHR system without the intermediate loop. The specific objectives to accomplish the goals are: To estimate tritium permeation behavior in FHRs; To design a tritium removal system for FHRs; To meet the same tritium permeation level in FHRs as the tritium production rate of 1.9 Ci/day in 1,000 MWe PWRs; To demonstrate economic benefits of the proposed FHR system via comparing with the three-loop FHR system. The objectives were accomplished by designing tritium removal facilities, developing a tritium analysis code, and conducting an economic analysis. In the fusion reactor community, tritium extraction has been widely investigated and researched. Borrowing the experiences from the fusion reactor community, a tritium control and mitigation system was proposed. Based on mass transport theories, a tritium analysis code was developed, and the tritium behaviors were analyzed using the developed code. Tritium removal facilities were

  19. Supplemental Reactor Physics Calculations and Analysis of ELF Mk 1A Fuel

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

    Pope, Michael A.

    2014-10-01

    These calculations supplement previous the reactor physics work evaluating the Enhanced Low Enriched Uranium (LEU) Fuel (ELF) Mk 1A element. This includes various additional comparisons between the current Highly Enriched Uranium (HEU) and LEU along with further characterization of the performance of the ELF fuel. The excess reactivity to be held down at BOC for ELF Mk 1A fuel is estimated to be approximately $2.75 greater than with HEU for a typical cycle. This is a combined effect of the absence of burnable poison in the ELF fuel and the reduced neck shim worth in LEU fuel compared to HEU.more » Burnable poison rods were conceptualized for use in the small B positions containing Gd2O3 absorber. These were shown to provide $2.37 of negative reactivity at BOC and to burn out in less than half of a cycle. The worth of OSCCs is approximately the same between HEU and ELF Mk 1A (LEU) fuels in the representative loading evaluated. This was evaluated by rotating all banks simultaneously. The safety rod worth is relatively unchanged between HEU and ELF Mk 1A (LEU) fuels in the representative loading evaluated. However, this should be reevaluated with different loadings. Neutron flux, both total and fast (>1 MeV), is either the same or reduced upon changing from HEU to ELF Mk 1A (LEU) fuels in the representative loading evaluated. This is consistent with the well-established trend of lower neutron fluxes for a given power in LEU than HEU.The IPT loop void reactivity is approximately the same or less positive with ELF Mk 1A (LEU) fuel than HEU in the representative loading evaluated.« less

  20. 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

  1. 10 CFR 50.83 - Release of part of a power reactor facility or site for unrestricted use.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 10 Energy 1 2011-01-01 2011-01-01 false Release of part of a power reactor facility or site for... of a power reactor facility or site for unrestricted use. (a) Prior written NRC approval is required... release. Nuclear power reactor licensees seeking NRC approval shall— (1) Evaluate the effect of releasing...

  2. 10 CFR 50.83 - Release of part of a power reactor facility or site for unrestricted use.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 10 Energy 1 2010-01-01 2010-01-01 false Release of part of a power reactor facility or site for... of a power reactor facility or site for unrestricted use. (a) Prior written NRC approval is required... release. Nuclear power reactor licensees seeking NRC approval shall— (1) Evaluate the effect of releasing...

  3. 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.

  4. 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.

  5. Study of Pu consumption in advanced light water reactors: Evaluation of GE advanced boiling water reactor plants - compilation of Phase 1B task reports

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

    NONE

    1993-09-15

    This report contains an extensive evaluation of GE advanced boiling water reactor plants prepared for United State Department of Energy. The general areas covered in this report are: core and system performance; fuel cycle; infrastructure and deployment; and safety and environmental approval.

  6. SCW Pressure-Channel Nuclear Reactor Some Design Features

    NASA Astrophysics Data System (ADS)

    Pioro, Igor L.; Khan, Mosin; Hopps, Victory; Jacobs, Chris; Patkunam, Ruban; Gopaul, Sandeep; Bakan, Kurtulus

    Concepts of nuclear reactors cooled with water at supercritical pressures were studied as early as the 1950s and 1960s in the USA and Russia. After a 30-year break, the idea of developing nuclear reactors cooled with SuperCritical Water (SCW) became attractive again as the ultimate development path for water cooling. The main objectives of using SCW in nuclear reactors are: 1) to increase the thermal efficiency of modern Nuclear Power Plants (NPPs) from 30-35% to about 45-48%, and 2) to decrease capital and operational costs and hence decrease electrical energy costs (˜1000 US/kW or even less). SCW NPPs will have much higher operating parameters compared to modern NPPs (pressure about 25 MPa and outlet temperature up to 625°C), and a simplified flow circuit, in which steam generators, steam dryers, steam separators, etc., can be eliminated. Also, higher SCW temperatures allow direct thermo-chemical production of hydrogen at low cost, due to increased reaction rates. Pressure-tube or pressure-channel SCW nuclear reactor concepts are being developed in Canada and Russia for some time. Some design features of the Canadian concept related to fuel channels are discussed in this paper. The main conclusion is that the development of SCW pressure-tube nuclear reactors is feasible and significant benefits can be expected over other thermal-energy systems.

  7. Reactor on-off antineutrino measurement with KamLAND

    NASA Astrophysics Data System (ADS)

    Gando, A.; Gando, Y.; Hanakago, H.; Ikeda, H.; Inoue, K.; Ishidoshiro, K.; Ishikawa, H.; Koga, M.; Matsuda, R.; Matsuda, S.; Mitsui, T.; Motoki, D.; Nakamura, K.; Obata, A.; Oki, A.; Oki, Y.; Otani, M.; Shimizu, I.; Shirai, J.; Suzuki, A.; Takemoto, Y.; Tamae, K.; Ueshima, K.; Watanabe, H.; Xu, B. D.; Yamada, S.; Yamauchi, Y.; Yoshida, H.; Kozlov, A.; Yoshida, S.; Piepke, A.; Banks, T. I.; Fujikawa, B. K.; Han, K.; O'Donnell, T.; Berger, B. E.; Learned, J. G.; Matsuno, S.; Sakai, M.; Efremenko, Y.; Karwowski, H. J.; Markoff, D. M.; Tornow, W.; Detwiler, J. A.; Enomoto, S.; Decowski, M. P.

    2013-08-01

    The recent long-term shutdown of Japanese nuclear reactors has resulted in a significantly reduced reactor ν¯e flux at KamLAND. This running condition provides a unique opportunity to confirm and constrain backgrounds for the reactor ν¯e oscillation analysis. The data set also has improved sensitivity for other ν¯e signals, in particular ν¯e’s produced in β-decays from U238 and Th232 within the Earth’s interior, whose energy spectrum overlaps with that of reactor ν¯e’s. Including constraints on θ13 from accelerator and short-baseline reactor neutrino experiments, a combined three-flavor analysis of solar and KamLAND data gives fit values for the oscillation parameters of tan⁡2θ12=0.436-0.025+0.029, Δm212=7.53-0.18+0.18×10-5eV2, and sin⁡2θ13=0.023-0.002+0.002. Assuming a chondritic Th/U mass ratio, we obtain 116-27+28 ν¯e events from U238 and Th232, corresponding to a geo ν¯e flux of 3.4-0.8+0.8×106cm-2s-1 at the KamLAND location. We evaluate various bulk silicate Earth composition models using the observed geo ν¯e rate.

  8. Anaerobic biodegradation of aircraft deicing fluid in UASB reactors.

    PubMed

    Tham, P T Pham thi; Kennedy, K J Kevin J

    2004-05-01

    A central composite design was employed to methodically investigate anaerobic treatment of aircraft deicing fluid (ADF) in bench-scale Upflow Anaerobic Sludge Blanket (UASB) reactors. A total of 23 runs at 17 different operating conditions (0.8% 1.6% ADF (6000-12,000mg/L COD), 12-56h HRT, and 18-36gVSS/L) were conducted in continuous mode. The development of four empirical models describing process responses (i.e. COD removal efficiency, biomass-specific acetoclastic activity, methane production rate, and methane production potential) as functions of ADF concentration, hydraulic retention time, and biomass concentration is presented. Model verification indicated that predicted responses (COD removal efficiencies, biomass-specific acetoclastic activity, and methane production rates and potential) were in good agreement with experimental results. Biomass-specific acetoclastic activity was improved two-fold from 0.23gCOD/gVSS/d for inoculum to a maximum of 0.55gCOD/gVSS/d during ADF treatment in UASB reactors. For the design window, COD removal efficiencies were higher than 90%. The predicted methane production potentials were close to theoretical values, and methane production rates increased as the organic loading rate is increased. ADF toxicity effects were evident for 1.6% ADF at medium organic loadings (SOLR above 0.5gCOD/gVSS/d). In contrast, good reactor stability and excellent COD removal efficiencies were achieved at 1.2% ADF for reactor loadings approaching that of highly loaded systems (0.73gCOD/gVSS/d).

  9. 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.

  10. 10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... light-water nuclear power reactors. 50.46 Section 50.46 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC... reactors. (a)(1)(i) Each boiling or pressurized light-water nuclear power reactor fueled with uranium oxide... evaluation model. This section does not apply to a nuclear power reactor facility for which the...

  11. 10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... light-water nuclear power reactors. 50.46 Section 50.46 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC... reactors. (a)(1)(i) Each boiling or pressurized light-water nuclear power reactor fueled with uranium oxide... evaluation model. This section does not apply to a nuclear power reactor facility for which the...

  12. 10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... light-water nuclear power reactors. 50.46 Section 50.46 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC... reactors. (a)(1)(i) Each boiling or pressurized light-water nuclear power reactor fueled with uranium oxide... evaluation model. This section does not apply to a nuclear power reactor facility for which the...

  13. 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

  14. 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.

  15. Accident analysis of heavy water cooled thorium breeder reactor

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

    Yulianti, Yanti; Su’ud, Zaki; Takaki, Naoyuki

    2015-04-16

    Thorium has lately attracted considerable attention because it is accumulating as a by-product of large scale rare earth mining. The objective of research is to analyze transient behavior of a heavy water cooled thorium breeder that is designed by Tokai University and Tokyo Institute of Technology. That is oxide fueled, PWR type reactor with heavy water as primary coolant. An example of the optimized core has relatively small moderator to fuel volume ratio (MFR) of 0.6 and the characteristics of the core are burn-up of 67 GWd/t, breeding ratio of 1.08, burn-up reactivity loss during cycles of < 0.2% dk/k,more » and negative coolant reactivity coefficient. One of the nuclear reactor accidents types examined here is Unprotected Transient over Power (UTOP) due to withdrawing of the control rod that result in the positive reactivity insertion so that the reactor power will increase rapidly. Another accident type is Unprotected Loss of Flow (ULOF) that caused by failure of coolant pumps. To analyze the reactor accidents, neutron distribution calculation in the nuclear reactor is the most important factor. The best expression for the neutron distribution is the Boltzmann transport equation. However, solving this equation is very difficult so that the space-time diffusion equation is commonly used. Usually, space-time diffusion equation is solved by employing a point kinetics approach. However, this approach is less accurate for a spatially heterogeneous nuclear reactor and the nuclear reactor with quite large reactivity input. Direct method is therefore used to solve space-time diffusion equation which consider spatial factor in detail during nuclear reactor accident simulation. Set of equations that obtained from full implicit finite-difference method is solved by using iterative methods. The indication of UTOP accident is decreasing macroscopic absorption cross-section that results large external reactivity, and ULOF accident is indicated by decreasing coolant flow

  16. Characterization of a continuous agitated cell reactor for oxygen dependent biocatalysis.

    PubMed

    Toftgaard Pedersen, Asbjørn; de Carvalho, Teresa Melo; Sutherland, Euan; Rehn, Gustav; Ashe, Robert; Woodley, John M

    2017-06-01

    Biocatalytic oxidation reactions employing molecular oxygen as the electron acceptor are difficult to conduct in a continuous flow reactor because of the requirement for high oxygen transfer rates. In this paper, the oxidation of glucose to glucono-1,5-lactone by glucose oxidase was used as a model reaction to study a novel continuous agitated cell reactor (ACR). The ACR consists of ten cells interconnected by small channels. An agitator is placed in each cell, which mixes the content of the cell when the reactor body is shaken by lateral movement. Based on tracer experiments, a hydrodynamic model for the ACR was developed. The model consisted of ten tanks-in-series with back-mixing occurring within and between each cell. The back-mixing was a necessary addition to the model in order to explain the observed phenomenon that the ACR behaved as two continuous stirred tank reactors (CSTRs) at low flow rates, while it at high flow rates behaved as the expected ten CSTRs in series. The performance of the ACR was evaluated by comparing the steady state conversion at varying residence times with the conversion observed in a stirred batch reactor of comparable size. It was found that the ACR could more than double the overall reaction rate, which was solely due to an increased oxygen transfer rate in the ACR caused by the intense mixing as a result of the spring agitators. The volumetric oxygen transfer coefficient, k L a, was estimated to be 344 h -1 in the 100 mL ACR, opposed to only 104 h -1 in a batch reactor of comparable working volume. Interestingly, the large deviation from plug flow behavior seen in the tracer experiments was found to have little influence on the conversion in the ACR, since both a plug flow reactor (PFR) model and the backflow cell model described the data sufficiently well. Biotechnol. Bioeng. 2017;114: 1222-1230. © 2017 Wiley Periodicals, Inc. © 2017 Wiley Periodicals, Inc.

  17. Westinghouse Small Modular Reactor nuclear steam supply system design

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

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

    2012-07-01

    generator, and eight reactor coolant pumps (RCP). The containment vessel is 27.1 m (89 ft) long and 9.8 m (32 ft) in diameter, and is designed to withstand pressures up to 1.7 MPa (250 psi). It is completely submerged in a pool of water serving as a heat sink and radiation shield. Housed within the containment are four combined core makeup tanks (CMT)/passive residual heat removal (PRHR) heat exchangers, two in-containment pools (ICP), two ICP tanks and four valves which function as the automatic depressurization system (ADS). The PRHR heat exchangers are thermally connected to two different ultimate heat sink (UHS) tanks which provide transient cooling capabilities. (authors)« less

  18. 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.

  19. First wall for polarized fusion reactors

    DOEpatents

    Greenside, Henry S.; Budny, Robert V.; Post, Jr., Douglass E.

    1988-01-01

    Depolarization mechanisms arising from the recycling of the polarized fuel at the limiter and the first-wall of a fusion reactor are greater than those mechanisms in the plasma. Rapid depolarization of the plasma is prevented by providing a first-wall or first-wall coating formed of a low-Z, non-metallic material having a depolarization rate greater than 1 sec.sup.-1.

  20. PBF (PER620) interior. Detail view across top of reactor tank. ...

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

    PBF (PER-620) interior. Detail view across top of reactor tank. Camera facing northeast. Ait tubing is cleanup equipment. Note projections from reactor structure above water level in tank. Date: May 2004. INEEL negative no. HD-41-5-1 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

  1. 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.

  2. 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.

  3. 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.

  4. Methane and hydrogen sulfide emissions in UASB reactors treating domestic wastewater.

    PubMed

    Souza, C L; Chernicharo, C A L; Melo, G C B

    2012-01-01

    The release of CH(4) and H(2)S in UASB reactors was evaluated with the aim to quantify the emissions from the liquid surfaces (three-phase separator and settler compartment) and also from the reactor's discharge hydraulic structures. The studies were carried out in two pilot- (360 L) and one demo-scale (14 m(3)) UASB reactors treating domestic wastewater. As expected, the release rates were much higher across the gas/liquid interfaces of the three-phase separators (5.4-9.7 kg CH(4) m(-2) d(-1) and 23.0-35.8 g S m(-2) d(-1)) as compared with the quiescent settler surfaces (11.0-17.8 g CH(4) m(-2) d(-1) and 0.21 to 0.37 g S m(-2) d(-1)). The decrease of dissolved methane and dissolved hydrogen sulfide was very large in the discharging hydraulic structures very close to the reactor (>60 and >80%, respectively), largely due to the loss to the atmosphere, indicating that the concentration of these compounds will probably fall to values close to zero in the near downstream structures. The emission factors due to the release of dissolved methane in the discharge structure amounted to around 0.040 g CH(4) g COD(infl)(-1) and 0.060 g CH(4) g COD(rem)(-1), representing around 60% of the methane collected in the three-phase separator.

  5. Comparison of reactivity in a flow reactor and a single cylinder engine

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

    Natelson, Robert H.; Johnson, Rodney O.; Kurman, Matthew S.

    2010-10-15

    The relative reactivity of 2:1:1 and 1:1:1 mixtures of n-decane:n-butylcyclohexane:n-butylbenzene and an average sample of JP-8 were evaluated in a single cylinder engine and compared to results obtained in a pressurized flow reactor. At compression ratios of 14:1, 15:1, and 16:1, inlet temperature of 500 K, inlet pressure of 0.1 MPa, equivalence ratio of 0.23, and engine speed of 800 RPM, the autoignition delay times were, from shortest to longest, the 2:1:1, followed by the 1:1:1, and then the JP-8. This order corresponded with recent results in a pressurized flow reactor, where the preignition oxidation chemistry was monitored at temperaturesmore » of 600-800 K, 0.8 MPa pressure, and an equivalence ratio of 0.30, and where the preignition reactivity from highest to lowest was the 2:1:1, followed by the 1:1:1, and the JP-8. This shows that the relative reactivity at low temperatures in the flow reactor tracks the autoignition tendencies in the engine for these particular fuels. (author) the computed experimental error. (author)« less

  6. 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.

  7. A comparison of mass transfer coefficients between trickle-bed, hollow fiber membrane and stirred tank reactors.

    PubMed

    Orgill, James J; Atiyeh, Hasan K; Devarapalli, Mamatha; Phillips, John R; Lewis, Randy S; Huhnke, Raymond L

    2013-04-01

    Trickle-bed reactor (TBR), hollow fiber membrane reactor (HFR) and stirred tank reactor (STR) can be used in fermentation of sparingly soluble gasses such as CO and H2 to produce biofuels and bio-based chemicals. Gas fermenting reactors must provide high mass transfer capabilities that match the kinetic requirements of the microorganisms used. The present study compared the volumetric mass transfer coefficient (K(tot)A/V(L)) of three reactor types; the TBR with 3 mm and 6 mm beads, five different modules of HFRs, and the STR. The analysis was performed using O2 as the gaseous mass transfer agent. The non-porous polydimethylsiloxane (PDMS) HFR provided the highest K(tot)A/V(L) (1062 h(-1)), followed by the TBR with 6mm beads (421 h(-1)), and then the STR (114 h(-1)). The mass transfer characteristics in each reactor were affected by agitation speed, and gas and liquid flow rates. Furthermore, issues regarding the comparison of mass transfer coefficients are discussed. Copyright © 2013 Elsevier Ltd. All rights reserved.

  8. Preliminary risks associated with postulated tritium release from production reactor operation

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

    O'Kula, K.R.; Horton, W.H.

    1988-01-01

    The Probabilistic Risk Assessment (PRA) of Savannah River Plant (SRP) reactor operation is assessing the off-site risk due to tritium releases during postulated full or partial loss of heavy water moderator accidents. Other sources of tritium in the reactor are less likely to contribute to off-site risk in non-fuel melting accident scenarios. Preliminary determination of the frequency of average partial moderator loss (including incidents with leaks as small as .5 kg) yields an estimate of /approximately/1 per reactor year. The full moderator loss frequency is conservatively chosen as 5 /times/ 10/sup /minus/3/ per reactor year. Conditional consequences, determined with amore » version of the MACCS code modified to handle tritium, are found to be insignificant. The 95th percentile individual cancer risk is 4 /times/ 10/sup /minus/8/ per reactor year within 16 km of the release point. The full moderator loss accident contributes about 75% of the evaluated risks. 13 refs., 4 figs., 5 tabs.« less

  9. Minimizing or eliminating refueling of nuclear reactor

    DOEpatents

    Doncals, Richard A.; Paik, Nam-Chin; Andre, Sandra V.; Porter, Charles A.; Rathbun, Roy W.; Schwallie, Ambrose L.; Petras, Diane S.

    1989-01-01

    Demand for refueling of a liquid metal fast nuclear reactor having a life of 30 years is eliminated or reduced to intervals of at least 10 years by operating the reactor at a low linear-power density, typically 2.5 kw/ft of fuel rod, rather than 7.5 or 15 kw/ft, which is the prior art practice. So that power of the same magnitude as for prior art reactors is produced, the volume of the core is increased. In addition, the height of the core and it diameter are dimensioned so that the ratio of the height to the diameter approximates 1 to the extent practicable considering the requirement of control and that the pressure drop in the coolant shall not be excessive. The surface area of a cylinder of given volume is a minimum if the ratio of the height to the diameter is 1. By minimizing the surface area, the leakage of neutrons is reduced. By reducing the linear-power density, increasing core volume, reducing fissile enrichment and optimizing core geometry, internal-core breeding of fissionable fuel is substantially enhanced. As a result, core operational life, limited by control worth requirements and fuel burnup capability, is extended up to 30 years of continuous power operation.

  10. 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.

  11. 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.

  12. 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

  13. Kinetic Parameter Measurements in the MINERVE Reactor

    NASA Astrophysics Data System (ADS)

    Perret, Grégory; Geslot, Benoit; Gruel, Adrien; Blaise, Patrick; Di-Salvo, Jacques; De Izarra, Grégoire; Jammes, Christian; Hursin, Mathieu; Pautz, Andréas

    2017-01-01

    In the framework of an international collaboration, teams of the PSI and CEA research institutes measure the critical decay constant (α0 = β/A), delayed neutron fraction (β) and generation time (A) of the Minerve reactor using the Feynman-α, Power Spectral Density and Rossi-α neutron noise measurement techniques. These measurements contribute to the experimental database of kinetic parameters used to improve nuclear data files and validate modern methods in Monte Carlo codes. Minerve is a zero-power pool reactor composed of a central experimental test lattice surrounded by a large aluminum buffer and four high-enriched driver regions. Measurements are performed in three slightly subcritical configurations (-2 cents to -30 cents) using two high-efficiency 235U fission chambers in the driver regions. Measurement of α0 and β obtained by the two institutes and with the different techniques are consistent for the configurations envisaged. Slight increases of the β values are observed with the subcriticality level. Best estimate values are obtained with the Cross-Power Spectral Density technique at -2 cents, and are worth: β = 716.9±9.0 pcm, α0 = 79.0±0.6 s-1 and A = 90.7±1.4 μs. The kinetic parameters are predicted with MCNP5-v1.6 and TRIPOLI4.9 and the JEFF-3.1/3.1.1 and ENDF/B-VII.1 nuclear data libraries. The predictions for β and α0 overestimate the experimental results by 3-5% and 10-12%, respectively; that for A underestimate the experimental result by 6-7%. The discrepancies are suspected to come from the driven system nature of Minerve and the location of the detectors in the driver regions, which prevent accounting for the full reactor.

  14. 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.

  15. 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.

  16. An atmospheric pressure flow reactor: Gas phase kinetics and mechanism in tropospheric conditions without wall effects

    NASA Technical Reports Server (NTRS)

    Koontz, Steven L.; Davis, Dennis D.; Hansen, Merrill

    1988-01-01

    A new type of gas phase flow reactor, designed to permit the study of gas phase reactions near 1 atm of pressure, is described. A general solution to the flow/diffusion/reaction equations describing reactor performance under pseudo-first-order kinetic conditions is presented along with a discussion of critical reactor parameters and reactor limitations. The results of numerical simulations of the reactions of ozone with monomethylhydrazine and hydrazine are discussed, and performance data from a prototype flow reactor are presented.

  17. The Ongoing Impact of the U.S. Fast Reactor Integral Experiments Program

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

    John D. Bess; Michael A. Pope; Harold F. McFarlane

    2012-11-01

    The creation of a large database of integral fast reactor physics experiments advanced nuclear science and technology in ways that were unachievable by less capital intensive and operationally challenging approaches. They enabled the compilation of integral physics benchmark data, validated (or not) analytical methods, and provided assurance of future rector designs The integral experiments performed at Argonne National Laboratory (ANL) represent decades of research performed to support fast reactor design and our understanding of neutronics behavior and reactor physics measurements. Experiments began in 1955 with the Zero Power Reactor No. 3 (ZPR-3) and terminated with the Zero Power Physics Reactormore » (ZPPR, originally the Zero Power Plutonium Reactor) in 1990 at the former ANL-West site in Idaho, which is now part of the Idaho National Laboratory (INL). Two additional critical assemblies, ZPR-6 and ZPR-9, operated at the ANL-East site in Illinois. A total of 128 fast reactor assemblies were constructed with these facilities [1]. The infrastructure and measurement capabilities are too expensive to be replicated in the modern era, making the integral database invaluable as the world pushes ahead with development of liquid metal cooled reactors.« less

  18. 76 FR 74630 - Making Changes to Emergency Plans for Nuclear Power Reactors

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-12-01

    ... NUCLEAR REGULATORY COMMISSION 10 CFR Parts 50 and 52 RIN 3150-AI10 [NRC-2008-0122] Making Changes to Emergency Plans for Nuclear Power Reactors AGENCY: Nuclear Regulatory Commission. ACTION... guide (RG) 1.219, ``Guidance on Making Changes to Emergency Plans for Nuclear Power Reactors.'' This...

  19. 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.)

  20. 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.

  1. 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.

  2. 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.

  3. Core Physics and Kinetics Calculations for the Fissioning Plasma Core Reactor

    NASA Technical Reports Server (NTRS)

    Butler, C.; Albright, D.

    2007-01-01

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

  4. Metallurgical failure analysis of MH-1A reactor core hold-down bolts. Final report

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

    Hawthorne, J.R.; Watson, H.E.

    1976-11-01

    The Naval Research Laboratory has performed a failure analysis on two MH-1A reactor core hold-down bolts that broke in service. Adherence to fabrication specifications, post-service properties and possible causes of bolt failure were investigated. The bolt material was verified as 17-4PH precipitation hardening stainless steel. Measured bolt dimensions also were in accordance with fabrication drawing specifications. Bolt failure occurred in the region of a locking pin hole which reduced the bolt net section by 47 percent. The failure analysis indicates that the probable cause of failure was net section overloading resulting from a lateral bending force on the bolt. Themore » analysis indicates that net section overloading could also have resulted from combined tensile stresses (bolt preloading plus differential thermal expansion). Recommendations are made for improved bolting.« less

  5. 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 %.

  6. Optical transmittance investigation of 1-keV ion-irradiated sapphire crystals as potential VUV to NIR window materials of fusion reactors

    NASA Astrophysics Data System (ADS)

    Iwano, Keisuke; Yamanoi, Kohei; Iwasa, Yuki; Mori, Kazuyuki; Minami, Yuki; Arita, Ren; Yamanaka, Takuma; Fukuda, Kazuhito; Empizo, Melvin John F.; Takano, Keisuke; Shimizu, Toshihiko; Nakajima, Makoto; Yoshimura, Masashi; Sarukura, Nobuhiko; Norimatsu, Takayoshi; Hangyo, Masanori; Azechi, Hiroshi; Singidas, Bess G.; Sarmago, Roland V.; Oya, Makoto; Ueda, Yoshio

    2016-10-01

    We investigate the optical transmittances of ion-irradiated sapphire crystals as potential vacuum ultraviolet (VUV) to near-infrared (NIR) window materials of fusion reactors. Under potential conditions in fusion reactors, sapphire crystals are irradiated with hydrogen (H), deuterium (D), and helium (He) ions with 1-keV energy and ˜ 1020-m-2 s-1 flux. Ion irradiation decreases the transmittances from 140 to 260 nm but hardly affects the transmittances from 300 to 1500 nm. H-ion and D-ion irradiation causes optical absorptions near 210 and 260 nm associated with an F-center and an F+-center, respectively. These F-type centers are classified as Schottky defects that can be removed through annealing above 1000 K. In contrast, He-ion irradiation does not cause optical absorptions above 200 nm because He-ions cannot be incorporated in the crystal lattice due to the large ionic radius of He-ions. Moreover, the significant decrease in transmittance of the ion-irradiated sapphire crystals from 140 to 180 nm is related to the light scattering on the crystal surface. Similar to diamond polishing, ion irradiation modifies the crystal surface thereby affecting the optical properties especially at shorter wavelengths. Although the transmittances in the VUV wavelengths decrease after ion irradiation, the transmittances can be improved through annealing above 1000 K. With an optical transmittance in the VUV region that can recover through simple annealing and with a high transparency from the ultraviolet (UV) to the NIR region, sapphire crystals can therefore be used as good optical windows inside modern fusion power reactors in terms of light particle loadings of hydrogen isotopes and helium.

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

    NASA Technical Reports Server (NTRS)

    Mayo, W.

    1973-01-01

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

  8. Investigation of Anaerobic Fluidized Bed Reactor/ Aerobic Moving Bed Bio Reactor (AFBR/MMBR) System for Treatment of Currant Wastewater

    PubMed Central

    JAFARI, Jalil; MESDAGHINIA, Alireza; NABIZADEH, Ramin; FARROKHI, Mehrdad; MAHVI, Amir Hossein

    2013-01-01

    Background: Anaerobic treatment methods are more suitable for the treatment of concentrated wastewater streams, offer lower operating costs, the production of usable biogas product. The aim of this study was to investigate the performance of an Anaerobic Fluidized Bed Reactor (AFBR)-Aerobic Moving Bed Bio Reactor (MBBR) in series arrangement to treat Currant wastewater. Methods: The bed materials of AFBR were cylindrical particles made of PVC with a diameter of 2–2.3 mm, particle density of 1250 kg/m3. The volume of all bed materials was 1.7 liter which expanded to 2.46 liters in fluidized situation. In MBBR, support media was composed of 1.5 liters Bee-Cell 2000 having porosity of 87% and specific surface area of 650m2/m3. Results: When system operated at 35 ºC, chemical oxygen demand (COD) removal efficiencies were achieved to 98% and 81.6% for organic loading rates (OLR) of 9.4 and 24.2 g COD/l.d, and hydraulic retention times (HRT) of 48 and 18 h, in average COD concentration feeding of 18.4 g/l, respectively. Conclusion: The contribution of AFBR in total COD removal efficiency at an organic loading rate (OLR) of 9.4 g COD/l.d was 95%, and gradually decreased to 76.5% in OLR of 24.2 g COD/l.d. Also with increasing in organic loading rate the contribution of aerobic reactor in removing COD gradually decreased. In this system, the anaerobic reactor played the most important role in the removal of COD, and the aerobic MBBR was actually needed to polish the anaerobic treated wastewater. PMID:26056640

  9. 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.

  10. 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

  11. 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.

  12. Direct nn-Scattering Measurement With the Pulsed Reactor YAGUAR.

    PubMed

    Mitchell, G E; Furman, W I; Lychagin, E V; Muzichka, A Yu; Nekhaev, G V; Strelkov, A V; Sharapov, E I; Shvetsov, V N; Chernuhin, Yu I; Levakov, B G; Litvin, V I; Lyzhin, A E; Magda, E P; Crawford, B E; Stephenson, S L; Howell, C R; Tornow, W

    2005-01-01

    Although crucial for resolving the issue of charge symmetry in the nuclear force, direct measurement of nn-scattering by colliding free neutrons has never been performed. At present the Russian pulsed reactor YAGUAR is the best neutron source for performing such a measurement. It has a through channel where the neutron moderator is installed. The neutrons are counted by a neutron detector located 12 m from the reactor. In preliminary experiments an instantaneous value of 1.1 × 10(18)/cm(2)s was obtained for the thermal neutron flux density. The experiment will be performed by the DIANNA Collaboration as International Science & Technology Center (ISTC) project No. 2286.

  13. Direct nn-Scattering Measurement With the Pulsed Reactor YAGUAR

    PubMed Central

    Mitchell, G. E.; Furman, W. I.; Lychagin, E. V.; Muzichka, A. Yu.; Nekhaev, G. V.; Strelkov, A. V.; Sharapov, E. I.; Shvetsov, V. N.; Chernuhin, Yu. I.; Levakov, B. G.; Litvin, V. I.; Lyzhin, A. E.; Magda, E. P.; Crawford, B. E.; Stephenson, S. L.; Howell, C. R.; Tornow, W

    2005-01-01

    Although crucial for resolving the issue of charge symmetry in the nuclear force, direct measurement of nn-scattering by colliding free neutrons has never been performed. At present the Russian pulsed reactor YAGUAR is the best neutron source for performing such a measurement. It has a through channel where the neutron moderator is installed. The neutrons are counted by a neutron detector located 12 m from the reactor. In preliminary experiments an instantaneous value of 1.1 × 1018/cm2s was obtained for the thermal neutron flux density. The experiment will be performed by the DIANNA Collaboration as International Science & Technology Center (ISTC) project No. 2286. PMID:27308126

  14. Performances and microbial features of an aerobic packed-bed biofilm reactor developed to post-treat an olive mill effluent from an anaerobic GAC reactor

    PubMed Central

    Bertin, Lorenzo; Colao, Maria Chiara; Ruzzi, Maurizio; Marchetti, Leonardo; Fava, Fabio

    2006-01-01

    Background Olive mill wastewater (OMW) is the aqueous effluent of olive oil producing processes. Given its high COD and content of phenols, it has to be decontaminated before being discharged. Anaerobic digestion is one of the most promising treatment process for such an effluent, as it combines high decontamination efficiency with methane production. The large scale anaerobic digestion of OMWs is normally conducted in dispersed-growth reactors, where however are generally achieved unsatisfactory COD removal and methane production yields. The possibility of intensifying the performance of the process using a packed bed biofilm reactor, as anaerobic treatment alternative, was demonstrated. Even in this case, however, a post-treatment step is required to further reduce the COD. In this work, a biological post-treatment, consisting of an aerobic biological "Manville" silica bead-packed bed aerobic reactor, was developed, tested for its ability to complete COD removal from the anaerobic digestion effluents, and characterized biologically through molecular tools. Results The aerobic post-treatment was assessed through a 2 month-continuous feeding with the digested effluent at 50.42 and 2.04 gl-1day-1 of COD and phenol loading rates, respectively. It was found to be a stable process, able to remove 24 and 39% of such organic loads, respectively, and to account for 1/4 of the overall decontamination efficiency displayed by the anaerobic-aerobic integrated system when fed with an amended OMW at 31.74 and 1.70 gl-1day-1 of COD and phenol loading rates, respectively. Analysis of 16S rRNA gene sequences of biomass samples from the aerobic reactor biofilm revealed that it was colonized by Rhodobacterales, Bacteroidales, Pseudomonadales, Enterobacteriales, Rhodocyclales and genera incertae sedis TM7. Some taxons occurring in the influent were not detected in the biofilm, whereas others, such as Paracoccus, Pseudomonas, Acinetobacter and Enterobacter, enriched significantly in

  15. 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.

  16. 10 CFR 2.809 - Participation by the Advisory Committee on Reactor Safeguards.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 10 Energy 1 2011-01-01 2011-01-01 false Participation by the Advisory Committee on Reactor Safeguards. 2.809 Section 2.809 Energy NUCLEAR REGULATORY COMMISSION RULES OF PRACTICE FOR DOMESTIC LICENSING PROCEEDINGS AND ISSUANCE OF ORDERS Rulemaking § 2.809 Participation by the Advisory Committee on Reactor...

  17. 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.

  18. Application of 1D and 2D MFR reactor technology for the isolation of insecticidal and anti-microbial properties from pyrolysis bio-oils.

    PubMed

    Hossain, Mohammad M; Scott, Ian M; Berruti, Franco; Briens, Cedric

    2016-12-01

    Valuable chemicals can be separated from agricultural residues by chemical or thermochemical processes. The application of pyrolysis has already been demonstrated as an efficient means to produce a liquid with a high concentration of desired product. The objective of this study was to apply an insect and microorganism bioassay-guided approach to separate and isolate pesticidal compounds from bio-oil produced through biomass pyrolysis. Tobacco leaf (Nicotianata bacum), tomato plant (Solanum lycopersicum), and spent coffee (Coffea arabica) grounds were pyrolyzed at 10°C/min from ambient to 565°C using the mechanically fluidized reactor (MFR). With one-dimensional (1D) MFR pyrolysis, the composition of the product vapors varied as the reactor temperature was raised allowing for the selection of the temperature range that corresponds to vapors with a high concentration of pesticidal properties. Further product separation was performed in a fractional condensation train, or 2D MFR pyrolysis, thus allowing for the separation of vapor components according to their condensation temperature. The 300-400°C tobacco and tomato bio-oil cuts from the 1D MFR showed the highest insecticidal and anti-microbial activity compared to the other bio-oil cuts. The 300-350 and 350-400°C bio-oil cuts produced by 2D MFR had the highest insecticidal activity when the bio-oil was collected from the 210°C condenser. The tobacco and tomato bio-oil had similar insecticidal activity (LC 50 of 2.1 and 2.2 mg/mL) when the bio-oil was collected in the 210°C condenser from the 300-350°C reactor temperature gases. The 2D MFR does concentrate the pesticidal products compared to the 1D MFR and thus can reduce the need for further separation steps such as solvent extraction.

  19. 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.

  20. 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.