WorldWideScience

Sample records for source reactor facility

  1. Facility for a Low Power Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chalker, R. G.

    1949-09-14

    Preliminary investigation indicates that a reactor facility with ample research provisions for use by University or other interested groups, featuring safety in design, can be economically constructed in the Los Angeles area. The complete installation, including an underground gas-tight reactor building, with associated storage and experiment assembly building, administration offices, two general laboratory buildings, hot latoratory and lodge, can be constructed for approxinately $1,500,000. This does not include the cost of the reactor itself or of its auxiliary equipment,

  2. Neutron dosimetry at the reactor facility VENUS

    Energy Technology Data Exchange (ETDEWEB)

    Deboodt, P.; Vermeersch, F.; Vanhavere, F.; Minsart, G. [SCK-CEN, Belgian Nuclear Research Centre, Mol (Belgium)

    1997-09-01

    The reactor VENUS is a zero-power research reactor mainly devoted to studies on light water fuels. The need for undertaking a neutron spectrometric and dosimetric study became apparent when locally high neutron dose rates were measured. The spectrometric study is based on two approaches. The first is an experimental one in which the neutron spectrum was measured at three positions around the facility. The second is a theoretical one in which a numerical modelling of the neutron transport at the reactor site was performed in order to determine neutron spectra and fluence rates at different positions around the site. The measured and calculated spectra are interpreted in terms of the responses of different individual and environmental dosemeters. These responses are confronted with the in situ measurements. The impact of the ICRP 60 recommendations on the determined dose rates is also studied. (author).

  3. THE WHITE SANDS MISSILE RANGE PULSED REACTOR FACILITY, MAY 1963

    Energy Technology Data Exchange (ETDEWEB)

    Long, Robert L.; Boor, R.A.; Cole, W.M.; Elder, G.E.

    1963-05-15

    A brief statement of the mission of the White Sands Missile Range Nuclear Effects Laboratory is given. The new Nuclear Effects Laboratory Facility is described. This facility consists of two buildings-a laboratory and a reactor building. The White Sands Missile Range bare critical assembly, designated as the MoLLY-G, is described. The MoLLY-G, an unreflected, unmoderated right circular cylinder of uranium-molybdenum alloy designed for pulsed operation, will have a maximum burst capability of approximately 2 x 10/sup 17/ fissions with a burst width of 50 microseconds. The reactor construction and operating procedures are described. As designed, the MoLLY-G will provide an intense source of pulsed neutron and gamma radiation for a great variety of experimental and test arrangements. (auth)

  4. Commercial Light Water Reactor Tritium Extraction Facility

    Energy Technology Data Exchange (ETDEWEB)

    McHood, M D

    2000-10-12

    A geotechnical investigation program has been completed for the Commercial Light Water Reactor - Tritium Extraction Facility (CLWR-TEF) at the Savannah River Site (SRS). The program consisted of reviewing previous geotechnical and geologic data and reports, performing subsurface field exploration, field and laboratory testing, and geologic and engineering analyses. The purpose of this investigation was to characterize the subsurface conditions for the CLWR-TEF in terms of subsurface stratigraphy and engineering properties for design and to perform selected engineering analyses. The objectives of the evaluation were to establish site-specific geologic conditions, obtain representative engineering properties of the subsurface and potential fill materials, evaluate the lateral and vertical extent of any soft zones encountered, and perform engineering analyses for slope stability, bearing capacity and settlement, and liquefaction potential. In addition, provide general recommendations for construction and earthwork.

  5. The new neutron imaging facility at TRIGA reactor in Morocco

    Energy Technology Data Exchange (ETDEWEB)

    Ouardi, A.; Alami, R.; Bensitel, A. [Centre National de l' Energie des Science et des Techniques Nucleaires, PB.1382 R.P 10001 Rabat (Morocco)

    2011-07-01

    A new neutron imaging facility is currently developed around 2 MW TRIGA MARK-II reactor at Maamora Nuclear research centre (CENM). Neutron imaging combined to X-ray or gamma radiography offers the opportunity to extend Non Destructive Testing (NDT) activities DT in Morocco to new fields of applications such as space and aircraft Moroccan industry, mining, wood industry and Archeology. The facility is planed to be completed in the end of 2011. In order to reduce the gamma-ray content in the neutron beam, the reactor tangential channel is selected. For power of 2 MW, the corresponding thermal neutron flux at the inlet of the tangential channel is around 1.10{sup 13}ncm{sup 2}/s. The facility will be based on a conical neutron collimator with a flight tube of 8m and offers three circular diaphragms with diameters of 1cm, 2 cm and 4 cm corresponding to L/D-ratio varying between 200 and 600. The holes will be housed in the primary shutter. These diaphragms' sizes allow to perform neutron radiography with high resolution (L/D = 600) and high speed (L/D= 200). Monte Carlo calculations by a fully 3D numerical code GEANT4 are used to optimize the whole neutron beam line and to reach a shorten distance between the source and detector and reduce as possible the exposure time. (author)

  6. Radiation facilities for fusion-reactor first-wall and blanket structural-materials development

    Energy Technology Data Exchange (ETDEWEB)

    Klueh, R.L.; Bloom, E.E.

    1981-12-01

    Present and future irradiation facilities for the study of fusion reactor irradiation damage are reviewed. Present studies are centered on irradiation in accelerator-based neutron sources, fast- and mixed-spectrum fission reactors, and ion accelerators. The accelerator-based neutron sources are used to demonstrate damage equivalence between high-energy neutrons and fission reactor neutrons. Once equivalence is demonstrated, the large volume of test space available in fission reactors can be used to study displacement damage, and in some instances, the effects of high-helium concentrations and the interaction of displacement damage and helium on properties. Ion bombardment can be used to study the mechanisms of damage evolution and the interaction of displacement damage and helium. These techniques are reviewed, and typical results obtained from such studies are examined. Finally, future techniques and facilities for developing damage levels that more closely approach those expected in an operating fusion reactor are discussed.

  7. Study of fast reactor safety test facilities. Preliminary report

    Energy Technology Data Exchange (ETDEWEB)

    Bell, G.I.; Boudreau, J.E.; McLaughlin, T.; Palmer, R.G.; Starkovich, V.; Stein, W.E.; Stevenson, M.G.; Yarnell, Y.L.

    1975-05-01

    Included are sections dealing with the following topics: (1) perspective and philosophy of fast reactor safety analysis; (2) status of accident analysis and experimental needs; (3) experiment and facility definitions; (4) existing in-pile facilities; (5) new facility options; and (6) data acquisition methods. (DG)

  8. Program for studying fundamental interactions at the PIK reactor facilities

    Science.gov (United States)

    Serebrov, A. P.; Vassiljev, A. V.; Varlamov, V. E.; Geltenbort, P.; Gridnev, K. A.; Dmitriev, S. P.; Dovator, N. A.; Egorov, A. I.; Ezhov, V. F.; Zherebtsov, O. M.; Zinoviev, V. G.; Ivochkin, V. G.; Ivanov, S. N.; Ivanov, S. A.; Kolomensky, E. A.; Konoplev, K. A.; Krasnoschekova, I. A.; Lasakov, M. S.; Lyamkin, V. A.; Martemyanov, V. P.; Murashkin, A. N.; Neustroev, P. V.; Onegin, M. S.; Petelin, A. L.; Pirozhkov, A. N.; Polyushkin, A. O.; Prudnikov, D. V.; Ryabov, V. L.; Samoylov, R. M.; Sbitnev, S. V.; Fomin, A. K.; Fomichev, A. V.; Zimmer, O.; Cherniy, A. V.; Shoka, I. V.

    2016-05-01

    A research program aimed at studying fundamental interactions by means of ultracold and polarized cold neutrons at the GEK-4-4' channel of the PIK reactor is presented. The apparatus to be used includes a source of cold neutrons in the heavy-water reflector of the reactor, a source of ultracold neutrons based on superfluid helium and installed in a cold-neutron beam extracted from the GEK-4 channel, and a number of experimental facilities in neutron beams. An experiment devoted to searches for the neutron electric dipole moment and an experiment aimed at a measurement the neutron lifetime with the aid of a large gravitational trap are planned to be performed in a beam of ultracold neutrons. An experiment devoted to measuring neutron-decay asymmetries with the aid of a superconducting solenoid is planned in a beam of cold polarized neutrons from the GEK-4' channel. The second ultracold-neutron source and an experiment aimed at measuring the neutron lifetime with the aid of a magnetic trap are planned in the neutron-guide system of the GEK-3 channel. In the realms of neutrino physics, an experiment intended for sterile-neutrino searches is designed. The state of affairs around the preparation of the experimental equipment for this program is discussed.

  9. Computational analysis of irradiation facilities at the JSI TRIGA reactor.

    Science.gov (United States)

    Snoj, Luka; Zerovnik, Gašper; Trkov, Andrej

    2012-03-01

    Characterization and optimization of irradiation facilities in a research reactor is important for optimal performance. Nowadays this is commonly done with advanced Monte Carlo neutron transport computer codes such as MCNP. However, the computational model in such calculations should be verified and validated with experiments. In the paper we describe the irradiation facilities at the JSI TRIGA reactor and demonstrate their computational characterization to support experimental campaigns by providing information on the characteristics of the irradiation facilities. Copyright © 2011 Elsevier Ltd. All rights reserved.

  10. High flux diffractometers on reactor neutron sources

    Science.gov (United States)

    Hewat, Alan W.

    2006-11-01

    Continuous neutron sources such as reactors can deliver a very high time-averaged flux to the sample using a relatively wide band of wavelengths, while still retaining good resolution. For example, the D20 diffractometer at ILL Grenoble, the world's highest flux neutron powder machine, can collect complete patterns at 100 ms intervals, and this has been important for the real time study of explosive SHS reactions. New very large 2D detectors, such as those recently installed on D2B and D19 at ILL, are up to an order of magnitude larger than previous designs, and will provide unmatched speed of data collection from very small samples, opening up new scientific perspectives for powder and single crystal diffraction. We will discuss future reactor based diffractometers designed for rapid data collection from small samples in special environments.

  11. Review of Transient Testing of Fast Reactor Fuels in the Transient REActor Test Facility (TREAT)

    Energy Technology Data Exchange (ETDEWEB)

    Jensen, C.; Wachs, D.; Carmack, J.; Woolstenhulme, N.

    2017-01-01

    The restart of the Transient REActor Test (TREAT) facility provides a unique opportunity to engage the fast reactor fuels community to reinitiate in-pile experimental safety studies. Historically, the TREAT facility played a critical role in characterizing the behavior of both metal and oxide fast reactor fuels under off-normal conditions, irradiating hundreds of fuel pins to support fast reactor fuel development programs. The resulting test data has provided validation for a multitude of fuel performance and severe accident analysis computer codes. This paper will provide a review of the historical database of TREAT experiments including experiment design, instrumentation, test objectives, and salient findings. Additionally, the paper will provide an introduction to the current and future experiment plans of the U.S. transient testing program at TREAT.

  12. Preconceptual design of the new production reactor circulator test facility

    Energy Technology Data Exchange (ETDEWEB)

    Thurston, G.

    1990-06-01

    This report presents the results of a study of a new circulator test facility for the New Production Reactor Modular High-Temperature Gas-Cooled Reactor. The report addresses the preconceptual design of a stand-alone test facility with all the required equipment to test the Main Circulator/shutoff valve and Shutdown Cooling Circulator/shutoff valve. Each type of circulator will be tested in its own full flow, full power helium test loop. Testing will cover the entire operating range of each unit. The loop will include a test vessel, in which the circulator/valve will be mounted, and external piping. The external flow piping will include a throttle valve, flowmeter, and heat exchanger. Subsystems will include helium handling, helium purification, and cooling water. A computer-based data acquisition and control system will be provided. The estimated costs for the design and construction of this facility are included. 2 refs., 15 figs.

  13. Reed Reactor Facility final report, September 1, 1995--August 31, 1996

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-09-01

    This report covers the period from September 1, 1995 to August 31, 1996. This report is intended to fulfill several purposes including the reporting requirements of the US Nuclear Regulatory Commission, the US Department of Energy, and the Oregon Department of Energy. Highlights of the last year include: student participation in the program is very high; the facility continues its success in obtaining donated equipment from the Portland General Electric, US Department of Energy, and other sources; the facility is developing more paid work; progress is being made in a collaborative project with Pacific Northwest National Laboratory on isotope production for medical purposes. There were over 1,500 individual visits to the Reactor Facility during the year. Most were students in classes at Reed College or area universities, colleges, and high schools. Including tours and research conducted at the facility, the Reed Reactor Facility contributed to the educational programs of six colleges and universities in addition to eighteen pre-college groups. During the year, the reactor was operated almost three hundred separate times. The total energy production was over 23 MW-hours. The reactor staff consists of a Director, an Associated Director, a contract Health Physicist, and approximately twenty Reed College undergraduate students as hourly employees. All radiation exposures to individuals during this year were well below 5% of the federal limits.

  14. Heat barrier for use in a nuclear reactor facility

    Science.gov (United States)

    Keegan, Charles P.

    1988-01-01

    A thermal barrier for use in a nuclear reactor facility is disclosed herein. Generally, the thermal barrier comprises a flexible, heat-resistant web mounted over the annular space between the reactor vessel and the guard vessel in order to prevent convection currents generated in the nitrogen atmosphere in this space from entering the relatively cooler atmosphere of the reactor cavity which surrounds these vessels. Preferably, the flexible web includes a blanket of heat-insulating material formed from fibers of a refractory material, such as alumina and silica, sandwiched between a heat-resistant, metallic cloth made from stainless steel wire. In use, the web is mounted between the upper edges of the guard vessel and the flange of a sealing ring which surrounds the reactor vessel with a sufficient enough slack to avoid being pulled taut as a result of thermal differential expansion between the two vessels. The flexible web replaces the rigid and relatively complicated structures employed in the prior art for insulating the reactor cavity from the convection currents generated between the reactor vessel and the guard vessel.

  15. Safety requirements, facility user needs, and reactor concepts for a new Broad Application Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ryskamp, J.M. [ed.; Liebenthal, J.L.; Denison, A.B.; Fletcher, C.D.

    1992-07-01

    This report describes the EG&G Laboratory Directed Research and Development Program (LDRD) Broad Application Test Reactor (BATR) Project that was conducted in fiscal year 1991. The scope of this project was divided into three phases: a project process definition phase, a requirements development phase, and a preconceptual reactor design and evaluation phase. Multidisciplinary teams of experts conducted each phase. This report presents the need for a new test reactor, the project process definition, a set of current and projected regulatory compliance and safety requirements, a set of facility user needs for a broad range of projected testing missions, and descriptions of reactor concepts capable of meeting these requirements. This information can be applied to strategic planning to provide the Department of Energy with management options.

  16. Safety requirements, facility user needs, and reactor concepts for a new Broad Application Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ryskamp, J.M. (ed.); Liebenthal, J.L.; Denison, A.B.; Fletcher, C.D.

    1992-07-01

    This report describes the EG G Laboratory Directed Research and Development Program (LDRD) Broad Application Test Reactor (BATR) Project that was conducted in fiscal year 1991. The scope of this project was divided into three phases: a project process definition phase, a requirements development phase, and a preconceptual reactor design and evaluation phase. Multidisciplinary teams of experts conducted each phase. This report presents the need for a new test reactor, the project process definition, a set of current and projected regulatory compliance and safety requirements, a set of facility user needs for a broad range of projected testing missions, and descriptions of reactor concepts capable of meeting these requirements. This information can be applied to strategic planning to provide the Department of Energy with management options.

  17. High Flux Isotope Reactor cold neutron source reference design concept

    Energy Technology Data Exchange (ETDEWEB)

    Selby, D.L.; Lucas, A.T.; Hyman, C.R. [and others

    1998-05-01

    In February 1995, Oak Ridge National Laboratory`s (ORNL`s) deputy director formed a group to examine the need for upgrades to the High Flux Isotope Reactor (HFIR) system in light of the cancellation of the Advanced neutron Source Project. One of the major findings of this study was that there was an immediate need for the installation of a cold neutron source facility in the HFIR complex. In May 1995, a team was formed to examine the feasibility of retrofitting a liquid hydrogen (LH{sub 2}) cold source facility into an existing HFIR beam tube. The results of this feasibility study indicated that the most practical location for such a cold source was the HB-4 beam tube. This location provides a potential flux environment higher than the Institut Laue-Langevin (ILL) vertical cold source and maximizes the space available for a future cold neutron guide hall expansion. It was determined that this cold neutron beam would be comparable, in cold neutron brightness, to the best facilities in the world, and a decision was made to complete a preconceptual design study with the intention of proceeding with an activity to install a working LH{sub 2} cold source in the HFIR HB-4 beam tube. During the development of the reference design the liquid hydrogen concept was changed to a supercritical hydrogen system for a number of reasons. This report documents the reference supercritical hydrogen design and its performance. The cold source project has been divided into four phases: (1) preconceptual, (2) conceptual design and testing, (3) detailed design and procurement, and (4) installation and operation. This report marks the conclusion of the conceptual design phase and establishes the baseline reference concept.

  18. 1996 environmental monitoring report for the Naval Reactors Facility

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-12-31

    The results of the radiological and nonradiological environmental monitoring programs for 1996 at the Naval Reactors Facility (NRF) are presented in this report. The NRF is located on the Idaho National Engineering and Environmental Laboratory and contains three naval reactor prototypes and the Expended Core Facility, which examines developmental nuclear fuel material samples, spent naval fuel, and irradiated reactor plant components/materials. The results obtained from the environmental monitoring programs verify that releases to the environment from operations at NRF were in accordance with state and federal regulations. Evaluation of the environmental data confirms that the operation of NRF continues to have no adverse effect on the quality of the environment or the health and safety of the general public. Furthermore, a conservative assessment of radiation exposure to the general public as a result of NRF operations demonstrated that the dose received by any member of the public was well below the most restrictive dose limits prescribed by the Environmental Protection Agency (EPA) and the Department of Energy (DOE).

  19. 77 FR 26321 - Reed College, Reed Research Nuclear Reactor, Renewed Facility Operating License No. R-112

    Science.gov (United States)

    2012-05-03

    ... COMMISSION Reed College, Reed Research Nuclear Reactor, Renewed Facility Operating License No. R-112 AGENCY..., Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Rockville, MD 20852. Telephone..., Research and Test Reactors Licensing Branch, Division of Policy and Rulemaking, Office of Nuclear Reactor...

  20. Backup power sources for DOE facilities

    Energy Technology Data Exchange (ETDEWEB)

    1994-09-01

    This standard establishes fundamental requirements and guidance for backup power sources at DOE facilities. Purpose is to document good engineering practices for installation, testing, and maintenance of these backup power sources, which also covers emergency power sources. Examples are those which supply power to nuclear safety systems, radiation monitors and alarms, fire protection systems, security systems, and emergency lighting.

  1. Facility Configuration Study of the High Temperature Gas-Cooled Reactor Component Test Facility

    Energy Technology Data Exchange (ETDEWEB)

    S. L. Austad; L. E. Guillen; D. S. Ferguson; B. L. Blakely; D. M. Pace; D. Lopez; J. D. Zolynski; B. L. Cowley; V. J. Balls; E.A. Harvego, P.E.; C.W. McKnight, P.E.; R.S. Stewart; B.D. Christensen

    2008-04-01

    A test facility, referred to as the High Temperature Gas-Cooled Reactor Component Test Facility or CTF, will be sited at Idaho National Laboratory for the purposes of supporting development of high temperature gas thermal-hydraulic technologies (helium, helium-Nitrogen, CO2, etc.) as applied in heat transport and heat transfer applications in High Temperature Gas-Cooled Reactors. Such applications include, but are not limited to: primary coolant; secondary coolant; intermediate, secondary, and tertiary heat transfer; and demonstration of processes requiring high temperatures such as hydrogen production. The facility will initially support completion of the Next Generation Nuclear Plant. It will secondarily be open for use by the full range of suppliers, end-users, facilitators, government laboratories, and others in the domestic and international community supporting the development and application of High Temperature Gas-Cooled Reactor technology. This pre-conceptual facility configuration study, which forms the basis for a cost estimate to support CTF scoping and planning, accomplishes the following objectives: • Identifies pre-conceptual design requirements • Develops test loop equipment schematics and layout • Identifies space allocations for each of the facility functions, as required • Develops a pre-conceptual site layout including transportation, parking and support structures, and railway systems • Identifies pre-conceptual utility and support system needs • Establishes pre-conceptual electrical one-line drawings and schedule for development of power needs.

  2. Reed Reactor Facility final report, September 1, 1994--August 31, 1995

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-09-01

    This report covers the period from September 1, 1994 to August 31, 1995. Information contained in this report is intended to fulfill several purposes including the reporting requirements of the US Nuclear Regulatory Commission (USNRC), the US Department of Energy (USDOE), and the Oregon Department of Energy (ODOE). Highlights of the last year include: student participation in the program is very high; the facility has been extraordinarily successful in obtaining donated equipment from Portland General Electric, US Department of Energy, Precision Castparts, Tektronix, and other sources; the facility is developing more paid work. There were 1,115 visits of the Reactor Facility by individuals during the year. Most of these visitors were students in classes at Reed College or area universities, colleges, and high schools. During the year, the reactor was operated 225 separate times on 116 days. The total energy production was 24.6 MW-hours. The reactor staff consists of a Director, an Associate Director, a contract Health Physicist, and approximately fifteen Reed College undergraduate students as hourly employees. All radiation exposures to individuals during this year were well below 1% of the federal limits. There were no releases of liquid radioactive material from the facility and airborne releases (primarily {sup 41}Ar) were well within regulatory limits.

  3. The TRIGA Reactor Facility at the Armed Forces Radiobiology Research Institute: A Simplified Technical Description.

    Science.gov (United States)

    1986-05-01

    AD-AiS68 238 THE TRIGA REACTOR FACILITY AT THE ARMED FORCESI! RADIOBIOLOGY RESEARCH INST..(U) ARMED FORCES RADIOBIOLOGY RESEARCH INST BETHESDA NO...medium-power exposure. The reactor is also used to train military personnel in reactor operations., The AFRRI TRIGA Mark-F reactor facility is within the...AFRRI complex on the grounds ’of the Naval Medical Command National Capital Region, in Bethesda, Maryland.> TRIGA is an acronym for Training, Research

  4. Reed Reactor Facility annual report, September 1, 1994--August 31, 1995

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-12-31

    This report covers the period from September 1, 1994 to August 31, 1995. Information contained in this report is intended to fulfill several purposes including the reporting requirements of the US Nuclear Regulatory Commission (USNRC), the US Department of Energy (USDOE), and the Oregon Department of Energy (ODOE). Highlights of the last year include: (1) The number of new licensed student operators more than replaced the number of graduating seniors. Seven Reed College seniors used the reactor as part of their thesis projects. (2) The facility has been extraordinarily successful in obtaining donated equipment from Portland General Electric, US Department of Energy, Precision Castparts, Tektronix, and other sources. Battelle (Pacific Northwest Laboratory) has been generous in lending valuable equipment to the college. (3) The facility is developing more paid work. Income in the past academic year was much greater than the previous year, and next year should increase by even more. Additionally, the US Department of Energy`s Reactor-Use Sharing grant increased significantly this year. During the year, the reactor was operated 225 separate times on 116 days. The total energy production was 24.6 MW-hours. The reactor staff consists of a Director, an Assistant Director, a contract Health Physicist, and approximately fifteen Reed College undergraduate students as hourly employees. All radiation exposures to individuals during this year were well below one percent of the federal limits. There were no releases of liquid radioactive material from the facility and airborne releases (primarily {sup 41}Ar) were well within regulatory limits. No radioactive waste was shipped from the facility during this period.

  5. Naval Reactors Facility Environmental Monitoring Report, Calendar Year 2003

    Energy Technology Data Exchange (ETDEWEB)

    None

    2003-12-31

    The results of the radiological and nonradiological environmental monitoring programs for 2003 at the Naval Reactors Facility are presented in this report. The results obtained from the environmental monitoring programs verify that releases to the environment from operations at NRF were in accordance with Federal and State regulations. Evaluation of the environmental data confirms that the operation of NRF continues to have no adverse effect on the quality of the environment or the health and safety of the general public. Furthermore, a conservative assessment of radiation exposure to the general public as a result of NRF operations demonstrated that the dose received by any member of the public was well below the most restrictive dose limits prescribed by the U.S. Environmental Protection Agency and the U.S. Department of Energy.

  6. Commercial Light Water Reactor Tritium Extraction Facility Geotechnical Summary Report

    Energy Technology Data Exchange (ETDEWEB)

    Lewis, M R

    2000-01-11

    A geotechnical investigation program has been completed for the Circulating Light Water Reactor - Tritium Extraction Facility (CLWR-TEF) at the Savannah River Site (SRS). The program consisted of reviewing previous geotechnical and geologic data and reports, performing subsurface field exploration, field and laboratory testing and geologic and engineering analyses. The purpose of this investigation was to characterize the subsurface conditions for the CLWR-TEF in terms of subsurface stratigraphy and engineering properties for design and to perform selected engineering analyses. The objectives of the evaluation were to establish site-specific geologic conditions, obtain representative engineering properties of the subsurface and potential fill materials, evaluate the lateral and vertical extent of any soft zones encountered, and perform engineering analyses for slope stability, bearing capacity and settlement, and liquefaction potential. In addition, provide general recommendations for construction and earthwork.

  7. Naval Reactors Facility environmental monitoring report, calendar year 2000

    Energy Technology Data Exchange (ETDEWEB)

    None

    2001-12-01

    The results of the radiological and nonradiological environmental monitoring programs for 2000 at the Naval Reactors Facility (NRF) are presented in this report. The results obtained from the environmental monitoring programs verify that releases to the environment from operations at NRF were in accordance with Federal and State regulations. Evaluation of the environmental data confirms that the operation of NRF continues to have no adverse effect on the quality of the environment or the health and safety of the general public. Furthermore, a conservative assessment of radiation exposure to the general public as a result of NRF operations demonstrated that the dose received by any member of the public was well below the most restrictive dose limits prescribed by the U.S. Environmental Protection Agency (EPA) and the U.S. Department of Energy (DOE).

  8. 1997 environmental monitoring report for the Naval Reactors Facility

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-12-31

    The results of the radiological and nonradiological environmental monitoring programs for 1997 at the Naval Reactors Facility (NRF) are presented in this report. The results obtained from the environmental monitoring programs verify that releases to the environment from operations at NRF were in accordance with state and federal regulations. Evaluation of the environmental data confirms that the operation of NRF continues to have no adverse effect on the quality of the environment or the health and safety of the general public. Furthermore, a conservative assessment of radiation exposure to the general public as a result of NRF operations demonstrated that the dose received by any member of the public was well below the most restrictive dose limits prescribed by the Environmental Protection Agency (EPA) and the Department of Energy (DOE).

  9. Naval Reactors Facility environmental monitoring report, calendar year 2001

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2002-12-31

    The results of the radiological and nonradiological environmental monitoring programs for 2001 at the Naval Reactors Facility are presented in this report. The results obtained from the environmental monitoring programs verify that releases to the environment from operations at NRF were in accordance with Federal and State regulations. Evaluation of the environmental data confirms that the operation of NRF continues to have no adverse effect on the quality of the environment or the health and safety of the general public. Furthermore, a conservative assessment of radiation exposure to the general public as a result of NRF operations demonstrated that the dose received by any member of the public was well below the most restrictive dose limits prescribed by the U. S. Environmental Protection Agency and the U. S. Department of Energy.

  10. Naval Reactors Facility environmental monitoring report, calendar year 1999

    Energy Technology Data Exchange (ETDEWEB)

    None

    2000-12-01

    The results of the radiological and nonradiological environmental monitoring programs for 1999 at the Naval Reactors Facility (NRF) are presented in this report. The results obtained from the environmental monitoring programs verify that releases to the environment from operations at NRF were in accordance with Federal and State regulations. Evaluation of the environmental data confirms that the operation of NRF continues to have no adverse effect on the quality of the environment or the health and safety of the general public. Furthermore, a conservative assessment of radiation exposure to the general public as a result of NRF operations demonstrated that the dose received by any member of the public was well below the most restrictive dose limits prescribed by the U.S. Environmental Protection Agency (EPA) and the U.S. Department of Energy (DOE).

  11. 1993 environmental monitoring report for the naval reactors facility

    Energy Technology Data Exchange (ETDEWEB)

    1994-11-01

    The results of the radiological and nonradiological environmental monitoring programs for 1993 at the Naval Reactors Facility (NRF) are presented in this report. The results obtained from the environmental monitoring programs verify that releases to the environment from operations at NRF were in accordance with state and federal regulations. Evaluation of the environmental data confirms that the operation of NRF continues to have no adverse effect on the quality of the environment or the health and safety of the general public. Furthermore, a conservative assessment of radiation exposure to the general public as a result of NRF operations demonstrated that the dose received by any member of the public was well below the most restrictive dose limits prescribed by the Environmental Protection Agency (EPA) and the Department of Energy (DOE).

  12. Commercial Light Water Reactor -Tritium Extraction Facility Process Waste Assessment (Project S-6091)

    Energy Technology Data Exchange (ETDEWEB)

    Hsu, R.H.; Delley, A.O.; Alexander, G.J.; Clark, E.A.; Holder, J.S.; Lutz, R.N.; Malstrom, R.A.; Nobles, B.R. [Westinghouse Savannah River Co., Aiken, SC (United States); Carson, S.D. [Sandia National Laboratories, New Mexico, NM (United States); Peterson, P.K. [Sandia National Laboratories, New Mexico, NM (United States)

    1997-11-30

    The Savannah River Site (SRS) has been tasked by the Department of Energy (DOE) to design and construct a Tritium Extraction Facility (TEF) to process irradiated tritium producing burnable absorber rods (TPBARs) from a Commercial Light Water Reactor (CLWR). The plan is for the CLWR-TEF to provide tritium to the SRS Replacement Tritium Facility (RTF) in Building 233-H in support of DOE requirements. The CLWR-TEF is being designed to provide 3 kg of new tritium per year, from TPBARS and other sources of tritium (Ref. 1-4).The CLWR TPBAR concept is being developed by Pacific Northwest National Laboratory (PNNL). The TPBAR assemblies will be irradiated in a Commercial Utility light water nuclear reactor and transported to the SRS for tritium extraction and processing at the CLWR-TEF. A Conceptual Design Report for the CLWR-TEF Project was issued in July 1997 (Ref. 4).The scope of this Process Waste Assessment (PWA) will be limited to CLWR-TEF processing of CLWR irradiated TPBARs. Although the CLWR- TEF will also be designed to extract APT tritium-containing materials, they will be excluded at this time to facilitate timely development of this PWA. As with any process, CLWR-TEF waste stream characteristics will depend on process feedstock and contaminant sources. If irradiated APT tritium-containing materials are to be processed in the CLWR-TEF, this PWA should be revised to reflect the introduction of this contaminant source term.

  13. Advanced sodium fast reactor accident source terms :

    Energy Technology Data Exchange (ETDEWEB)

    Powers, Dana Auburn; Clement, Bernard; Denning, Richard; Ohno, Shuji; Zeyen, Roland

    2010-09-01

    An expert opinion elicitation has been used to evaluate phenomena that could affect releases of radionuclides during accidents at sodium-cooled fast reactors. The intent was to identify research needed to develop a mechanistic model of radionuclide release for licensing and risk assessment purposes. Experts from the USA, France, the European Union, and Japan identified phenomena that could affect the release of radionuclides under hypothesized accident conditions. They qualitatively evaluated the importance of these phenomena and the need for additional experimental research. The experts identified seven phenomena that are of high importance and have a high need for additional experimental research: High temperature release of radionuclides from fuel during an energetic event Energetic interactions between molten reactor fuel and sodium coolant and associated transfer of radionuclides from the fuel to the coolant Entrainment of fuel and sodium bond material during the depressurization of a fuel rod with breached cladding Rates of radionuclide leaching from fuel by liquid sodium Surface enrichment of sodium pools by dissolved and suspended radionuclides Thermal decomposition of sodium iodide in the containment atmosphere Reactions of iodine species in the containment to form volatile organic iodides. Other issues of high importance were identified that might merit further research as development of the mechanistic model of radionuclide release progressed.

  14. The negative ion source test facility ELISE

    Energy Technology Data Exchange (ETDEWEB)

    Heinemann, B., E-mail: bernd.heinemann@ipp.mpg.de [Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, Postfach 1533, D-85748 Garching (Germany); Falter, H.-D.; Fantz, U.; Franzen, P.; Froeschle, M.; Kraus, W.; Martens, C.; Nocentini, R.; Riedl, R.; Speth, E.; Staebler, A. [Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, Postfach 1533, D-85748 Garching (Germany)

    2011-10-15

    The ITER neutral beam system is using inductively coupled radio frequency (RF) ion sources, that have demonstrated the required ITER parameters on (small) sources with extraction areas up to 200 cm{sup 2}. As a next step towards the full size ITER source IPP is presently constructing the test facility ELISE ('Extraction from a Large Ion Source Experiment') operating with a 'half-size' source which has approximately the width but only half the height of the ITER source. The modular driver concept is expected to allow a further extrapolation to the full size in one direction to be made. The main aim of this experiment is to demonstrate the production of a large uniform negative ion beam with ITER relevant parameters in stable conditions up to one hour. Plasma operation of the source is foreseen to be performed continuously for 1 h; extraction and acceleration of negative ions up to 60 kV is only possible in pulsed mode (10 s every 180 s) due to limitations of the existing IPP HV system. The design of the source and extraction system implements a high experimental flexibility and a good diagnostic access while still staying as close as possible to the ITER design. The main differences are the source operating in air and the use of a large gate valve between the source and the target chamber. ELISE is expected to start operation at the end of 2011 and is an important step for the development of the ITER NBI system; the experience gained early will support the design as well as the commissioning and operating phases of the PRIMA NBI test facilities and the ITER neutral beam system.

  15. Iranian Light Source Facility, A third generation light source laboratory

    Directory of Open Access Journals (Sweden)

    J Rahighi

    2015-09-01

    Full Text Available The Iranian Light Source Facility (ILSF project is the first large scale accelerator facility which is currently under planning in Iran. On the basis of the present design, circumference of the 3 GeV storage ring is 528 m. Beam current and natural beam emittance are 400 mA and 0.477 nm.rad, respectively. Some prototype accelerator components such as high power solid state radio frequency amplifiers, low level RF system, thermionic RF gun, H-type dipole and quadruple magnets, magnetic measurement laboratory and highly stable magnet power supplies have been constructed at ILSF R&D laboratory

  16. Parametric Thermal Models of the Transient Reactor Test Facility (TREAT)

    Energy Technology Data Exchange (ETDEWEB)

    Bradley K. Heath

    2014-03-01

    This work supports the restart of transient testing in the United States using the Department of Energy’s Transient Reactor Test Facility at the Idaho National Laboratory. It also supports the Global Threat Reduction Initiative by reducing proliferation risk of high enriched uranium fuel. The work involves the creation of a nuclear fuel assembly model using the fuel performance code known as BISON. The model simulates the thermal behavior of a nuclear fuel assembly during steady state and transient operational modes. Additional models of the same geometry but differing material properties are created to perform parametric studies. The results show that fuel and cladding thermal conductivity have the greatest effect on fuel temperature under the steady state operational mode. Fuel density and fuel specific heat have the greatest effect for transient operational model. When considering a new fuel type it is recommended to use materials that decrease the specific heat of the fuel and the thermal conductivity of the fuel’s cladding in order to deal with higher density fuels that accompany the LEU conversion process. Data on the latest operating conditions of TREAT need to be attained in order to validate BISON’s results. BISON’s models for TREAT (material models, boundary convection models) are modest and need additional work to ensure accuracy and confidence in results.

  17. A neutron tomography facility at a low power research reactor

    CERN Document Server

    Körner, S; Von Tobel, P; Rauch, H

    2001-01-01

    Neutron radiography (NR) provides a very efficient tool in the field of non-destructive testing as well as for many applications in fundamental research. A neutron beam penetrating a specimen is attenuated by the sample material and detected by a two-dimensional (2D) imaging device. The image contains information about materials and structure inside the sample because neutrons are attenuated according to the basic law of radiation attenuation. Contrary to X-rays, neutrons can be attenuated by some light materials, as for example, hydrogen and boron, but penetrate many heavy materials. Therefore, NR can yield important information not obtainable by more traditional methods. Nevertheless, there are many aspects of structure, both quantitative and qualitative, that are not accessible from 2D transmission images. Hence, there is an interest in three-dimensional neutron imaging. At the 250 kW TRIGA Mark II reactor of the Atominstitut in Austria a neutron tomography facility has been installed. The neutron flux at ...

  18. Experiment Needs and Facilities Study Appendix A Transient Reactor Test Facility (TREAT) Upgrade

    Energy Technology Data Exchange (ETDEWEB)

    None

    1976-09-01

    The TREAT Upgrade effort is designed to provide significant new capabilities to satisfy experiment requirements associated with key LMFBR Safety Issues. The upgrade consists of reactor-core modifications to supply the physics performance needed for the new experiments, an Advanced TREAT loop with size and thermal-hydraulics capabilities needed for the experiments, associated interface equipment for loop operations and handling, and facility modifications necessary to accommodate operations with the Loop. The costs and schedules of the tasks to be accomplished under the TREAT Upgrade project are summarized. Cost, including contingency, is about 10 million dollars (1976 dollars). A schedule for execution of 36 months has been established to provide the new capabilities in order to provide timely support of the LMFBR national effort. A key requirement for the facility modifications is that the reactor availability will not be interrupted for more than 12 weeks during the upgrade. The Advanced TREAT loop is the prototype for the STF small-bundle package loop. Modified TREAT fuel elements contain segments of graphite-matrix fuel with graded uranium loadings similar to those of STF. In addition, the TREAT upgrade provides for use of STF-like stainless steel-UO{sub 2} TREAT fuel for tests of fully enriched fuel bundles. This report will introduce the Upgrade study by presenting a brief description of the scope, performance capability, safety considerations, cost schedule, and development requirements. This work is followed by a "Design Description". Because greatly upgraded loop performance is central to the upgrade, a description is given of Advanced TREAT loop requirements prior to description of the loop concept. Performance requirements of the upgraded reactor system are given. An extensive discussion of the reactor physics calculations performed for the Upgrade concept study is provided. Adequate physics performance is essential for performance of experiments with

  19. Operating manual for the High Flux Isotope Reactor. Volume I. Description of the facility

    Energy Technology Data Exchange (ETDEWEB)

    1982-09-01

    This volume contains a comprehensive description of the High Flux Isotope Reactor Facility. Its primary purpose is to supplement the detailed operating procedures, providing the reactor operators with background information on the various HFIR systems. The detailed operating procdures are presented in another report.

  20. Proceedings of the 1984 DOE nuclear reactor and facility safety conference. Volume II

    Energy Technology Data Exchange (ETDEWEB)

    1984-01-01

    This report is a collection of papers on reactor safety. The report takes the form of proceedings from the 1984 DOE Nuclear Reactor and Facility Safety Conference, Volume II of two. These proceedings cover Safety, Accidents, Training, Task/Job Analysis, Robotics and the Engineering Aspects of Man/Safety interfaces.

  1. Operation of N Reactor and Fuels Fabrication Facilities, Hanford Reservation, Richland, Benton County, Washington: Environmental assessment

    Energy Technology Data Exchange (ETDEWEB)

    1980-08-01

    Environmental data, calculations and analyses show no significant adverse radiological or nonradiological impacts from current or projected future operations resulting from N Reactor, Fuels Fabrication and Spent Fuel Storage Facilities. Nonoccupational radiation exposures resulting from 1978 N Reactor operations are summarized and compared to allowable exposure limits.

  2. Sandia Pulsed Reactor Facility (SPRF) calculator-assisted pulse analysis and display system

    Energy Technology Data Exchange (ETDEWEB)

    Estes, B.F.; Berry, D.T.

    1980-02-01

    Two solid-metal fast burst type reactors (SPR II and SPR III) are operated at the Sandia Pulsed Reactor Facility. Since startup of the reactors, oscilloscope traces have been used to record (by camera) the pulse (power) shape while log N systems have measured initial reactor period. Virtually no other pulse information is available. A decision was made to build a system that could collect the basic input data available from the reactor - fission chambers, photodiodes, and thermocouples - condition the signals and output the various parameters such as power, energy, temperature, period and lifetime on hard copy that would provide a record for operations personnel as well as the experimenter. Because the reactors operate in short time frames - pulse operation - it is convenient to utilize the classical Nordheim-Fuchs approximation of the diffusion equation to describe reactor behavior. This report describes the work performed to date in developing the calculator system and analytical models for computing the desired parameters.

  3. 10 CFR 50.83 - Release of part of a power reactor facility or site for unrestricted use.

    Science.gov (United States)

    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...

  4. A Design of Alarm System in a Research Reactor Facility

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jaekwan; Jang, Gwisook; Seo, Sangmun; Suh, Yongsuk [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-05-15

    The digital alarm system has become an indispensable design to process a large amount of alarms of power plants. Korean research reactor operated for decades maintains a hybrid alarm system with both an analog annunciator and a digital alarm display. In this design, several alarms are indicated on an analog panel and digital display, respectively, and it requires more attention and effort of the operators. As proven in power plants, a centralized alarm system design is necessary for a new research reactor. However, the number of alarms and operators in a research reactor is significantly lesser than power plants. Thus, simplification should be considered as an important factor for the operation efficiency. This paper introduces a simplified alarm system. As advances in information technology, fully digitalized alarm systems have been applied to power plants. In a new research reactor, it will be more useful than an analog or hybrid configuration installed in research reactors decades ago. However, the simplification feature should be considered as an important factor because the number of alarms and number of operators in a research reactor is significantly lesser than in power plants.

  5. Design, construction, and demonstration of a neutron beamline and a neutron imaging facility at a Mark-I TRIGA reactor

    Science.gov (United States)

    Craft, Aaron E.

    The fleet of research and training reactors is aging, and no new research reactors are planned in the United States. Thus, there is a need to expand the capabilities of existing reactors to meet users' needs. While many research reactors have beam port facilities, the original design of the United States Geological Survey TRIGA Reactor (GSTR) did not include beam ports. The MInes NEutron Radiography (MINER) facility developed by this thesis and installed at the GSTR provides new capabilities for both researchers and students at the Colorado School of Mines. The facility consists of a number of components, including a neutron beamline and beamstop, an optical table, an experimental enclosure and associated interlocks, a computer control system, a multi-channel plate imaging detector, and the associated electronics. The neutron beam source location, determined through Monte Carlo modeling, provides the best mixture of high neutron flux, high thermal neutron content, and low gamma radiation content. A Monte Carlo n-Particle (MCNP) model of the neutron beam provides researchers with a tool for designing experiments before placing objects in the neutron beam. Experimental multi-foil activation results, compared to calculated multi-foil activation results, verify the model. The MCNP model predicts a neutron beamline flux of 2.2*106 +/- 6.4*105 n/cm2-s based on a source particle rate determined from the foil activation experiments when the reactor is operating at a power of 950 kWt with the beam shutter fully open. The average cadmium ratio of the beamline is 7.4, and the L/D of the neutron beam is approximately 200+/-10. Radiographs of a sensitivity indicator taken using both the digital detector and the transfer foil method provide one demonstration of the radiographic capabilities of the new facility. Calibration fuel pins manufactured using copper and stainless steel surrogate fuel pellets provide additional specimens for demonstration of the new facility and offer a

  6. Beamlines for Iranian Light Source Facility

    Directory of Open Access Journals (Sweden)

    A Gholampour Azhir

    2015-09-01

    Full Text Available This paper describes day-one beamlines of the Iranian Light Source Facility and design concept of powder diffraction and spectromicroscopy beamlines as the most priorities of each synchrotron that cover the research requirements in the fields of physics, chemistry, nano-science, etc. For powder diffraction beamline energy range is 6-30 keV, resolution: 10-4, flux: 1012(ph/s/0.1%B.W. and spot size at sample is 0.1×0.1-1×10 mm2. For spectromicroscopy beamline energy range is 90-2500 eV, flux: 3×1015(ph/s/0.1%B.W.@96eV, resolving power of 1820 at 1000 eV and spot size at sample is 4×2-27×74 µm2

  7. Advanced Test Reactor National Scientific User Facility 2010 Annual Report

    Energy Technology Data Exchange (ETDEWEB)

    Mary Catherine Thelen; Todd R. Allen

    2011-05-01

    This is the 2010 ATR National Scientific User Facility Annual Report. This report provides an overview of the program for 2010, along with individual project reports from each of the university principal investigators. The report also describes the capabilities offered to university researchers here at INL and at the ATR NSUF partner facilities.

  8. Fabrication development for the Advanced Neutron Source Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Pace, B.W. [Babcock and Wilcox, Lynchburg, VA (United States); Copeland, G.L. [Oak Ridge National Lab., TN (United States)

    1995-08-01

    This report presents the fuel fabrication development for the Advanced Neutron Source (ANS) reactor. The fuel element is similar to that successfully fabricated and used in the High Flux Isotope Reactor (HFIR) for many years, but there are two significant differences that require some development. The fuel compound is U{sub 3}Si{sub 2} rather than U{sub 3}O{sub 8}, and the fuel is graded in the axial as well as the radial direction. Both of these changes can be accomplished with a straightforward extension of the HFIR technology. The ANS also requires some improvements in inspection technology and somewhat more stringent acceptance criteria. Early indications were that the fuel fabrication and inspection technology would produce a reactor core meeting the requirements of the ANS for the low volume fraction loadings needed for the highly enriched uranium design (up to 1.7 Mg U/m{sup 3}). Near the end of the development work, higher volume fractions were fabricated that would be required for a lower- enrichment uranium core. Again, results look encouraging for loadings up to {approx}3.5 Mg U/m{sup 3}; however, much less evaluation was done for the higher loadings.

  9. Upgrading of neutron radiography/tomography facility at research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Abd El Bar, Waleed; Mongy, Tarek [Atomic Energy Authority, Cairo (Egypt). ETRR-2; Kardjilov, Nikolay [Helmholtz Zentrum Berlin (HZB) for Materials and Energy, Berlin (Germany)

    2014-03-15

    A state-of-the-art neutron tomography imaging system was set up at the neutron radiography beam tube at the Egypt Second Research Reactor (ETRR-2) and was successfully commissioned in 2013. This study presents a set of tomographic experiments that demonstrate a high quality tomographic image formation. A computer technique for data processing and 3D image reconstruction was used to see inside a copy module of an ancient clay article provided by the International Atomic Energy Agency (IAEA). The technique was also able to uncover tomographic imaging details of a mummified fish and provided a high resolution tomographic image of a defective fire valve. (orig.)

  10. Optimally moderated nuclear fission reactor and fuel source therefor

    Science.gov (United States)

    Ougouag, Abderrafi M [Idaho Falls, ID; Terry, William K [Shelley, ID; Gougar, Hans D [Idaho Falls, ID

    2008-07-22

    An improved nuclear fission reactor of the continuous fueling type involves determining an asymptotic equilibrium state for the nuclear fission reactor and providing the reactor with a moderator-to-fuel ratio that is optimally moderated for the asymptotic equilibrium state of the nuclear fission reactor; the fuel-to-moderator ratio allowing the nuclear fission reactor to be substantially continuously operated in an optimally moderated state.

  11. TREAT (Transient Reactor Test Facility) reactor control rod scram system simulations and testing

    Energy Technology Data Exchange (ETDEWEB)

    Solbrig, C.W.; Stevens, W.W.

    1990-01-01

    Air cylinders moving heavy components (100 to 300 lbs) at high speeds (above 300 in/sec) present a formidable end-cushion-shock problem. With no speed control, the moving components can reach over 600 in/sec if the air cylinder has a 5 ft stroke. This paper presents an overview of a successful upgrade modification to an existing reactor control rod drive design using a computer model to simulate the modified system performance for system design analysis. This design uses a high speed air cylinder to rapidly insert control rods (278 lb moved 5 ft in less than 300 msec) to scram an air-cooled test reactor. Included is information about the computer models developed to simulate high-speed air cylinder operation and a unique new speed control and end cushion design. A patent application is pending with the US Patent Trade Mark Office for this system (DOE case number S-68,622). The evolution of the design, from computer simulations thru operational testing in a test stand (simulating in-reactor operating conditions) to installation and use in the reactor, is also described. 6 figs.

  12. Collaborative Physical and Biological Dosimetry Studies for Neutron Capture Therapy at the RA-1 Research Reactor Facility

    Energy Technology Data Exchange (ETDEWEB)

    Nigg, D.W.; Schwint, A.E.; Hartwell, J.K.; Heber, E.M.; Trivillin, V.; Castillo, J.; Wentzeis, L.; Sloan, P.; Wemple, C.A.

    2004-10-04

    Initial physical dosimetry measurements have been completed using activation spectrometry and thermoluminiscent dosimeters to characterize the BNCT irradiation facility developed at the RA-1 research reactor operated by the Argentine National Atomic Energy Commission in Buenos Aires. Some biological scoping irradiations have also been completed using a small-animal (hamster) oral mucosa tumor model. Results indicate that the RA-1 neutron source produces useful dose rates but that some improvements in the initial configuration will be needed to optimize the spectrum for thermal-neutron BNCT research applications.

  13. Collaborative Physical and Biological Dosimetry Studies for Neutron Capture Therapy at the RA-1 Research Reactor Facility

    Energy Technology Data Exchange (ETDEWEB)

    David W. Nigg; Amanda E. Schwint; John K. Hartwell; Elisa M. Heber; Veronica Trivillin; Jorge Castillo; Luis Wentzeis; Patrick Sloan; Charles A. Wemple

    2004-10-01

    Initial physical dosimetry measurements have been completed using activation spectrometry and thermoluminiscent dosimeters to characterize the BNCT irradiation facility developed at the RA-1 research reactor operated by the Argentine National Atomic Energy Commission in Buenos Aires. Some biological scoping irradiations have also been completed using a small-animal (hamster) oral mucosa tumor model. Results indicate that the RA-1 neutron source produces useful dose rates but that some improvements in the initial configuration will be needed to optimize the spectrum for thermal-neutron BNCT research applications.

  14. A comparison of different neutron spectroscopy systems at the reactor facility VENUS

    CERN Document Server

    Vanhavere, F; Chartier, J L; Itie, C; Rosenstock, W; Koeble, T; D'Errico, F

    2002-01-01

    The VENUS facility is a zero-power research reactor mainly devoted to studies on LWR fuels. Localised high-neutron rates were found around the reactor, with a neutron/gamma dose equivalent rate ratio as high as three. Therefore, a study of the neutron dosimetry around the reactor was started some years ago. During this study, several methods of neutron spectroscopy were employed and a study of individual and ambient dosemeters was performed. A first spectrometric measurement was done with the IPSN multisphere spectrometer in three positions around the reactor. Secondly, the ROSPEC spectrometer from the Fraunhofer Institut was used. The spectra were also measured with the bubble interactive neutron spectrometer. These measurements were compared with a numerical simulation of the neutron field made with the code TRIPOLI-3. Dosimetric measurements were made with three types of personal neutron dosemeters: an albedo type, a track etch detector and a bubble detector.

  15. Advanced neutron source reactor probabilistic flow blockage assessment

    Energy Technology Data Exchange (ETDEWEB)

    Ramsey, C.T.

    1995-08-01

    The Phase I Level I Probabilistic Risk Assessment (PRA) of the conceptual design of the Advanced Neutron Source (ANS) Reactor identified core flow blockage as the most likely internal event leading to fuel damage. The flow blockage event frequency used in the original ANS PRA was based primarily on the flow blockage work done for the High Flux Isotope Reactor (HFIR) PRA. This report examines potential flow blockage scenarios and calculates an estimate of the likelihood of debris-induced fuel damage. The bulk of the report is based specifically on the conceptual design of ANS with a 93%-enriched, two-element core; insights to the impact of the proposed three-element core are examined in Sect. 5. In addition to providing a probability (uncertainty) distribution for the likelihood of core flow blockage, this ongoing effort will serve to indicate potential areas of concern to be focused on in the preliminary design for elimination or mitigation. It will also serve as a loose-parts management tool.

  16. Performance of the solid deuterium ultra-cold neutron source at the pulsed reactor TRIGA Mainz

    Science.gov (United States)

    Karch, J.; Sobolev, Yu.; Beck, M.; Eberhardt, K.; Hampel, G.; Heil, W.; Kieser, R.; Reich, T.; Trautmann, N.; Ziegner, M.

    2014-04-01

    The performance of the solid deuterium ultra-cold neutron (UCN) source at the pulsed reactor TRIGA Mainz with a maximum peak energy of 10MJ is described. The solid deuterium converter with a volume of cm3 (8mol), which is exposed to a thermal neutron fluence of n/cm2, delivers up to 240000 UCN ( m/s) per pulse outside the biological shield at the experimental area. UCN densities of 10 cm3 are obtained in stainless-steel bottles of 10 L. The measured UCN yields compare well with the predictions from a Monte Carlo simulation developed to model the source and to optimize its performance for the upcoming upgrade of the TRIGA Mainz into a user facility for UCN physics.

  17. Final report of the HFIR (High Flux Isotope Reactor) irradiation facilities improvement project

    Energy Technology Data Exchange (ETDEWEB)

    Montgomery, B.H.; Thoms, K.R.; West, C.D.

    1987-09-01

    The High-Flux Isotope Reactor (HFIR) has outstanding neutronics characteristics for materials irradiation, but some relatively minor aspects of its mechanical design severely limited its usefulness for that purpose. In particular, though the flux trap region in the center of the annular fuel elements has a very high neutron flux, it had no provision for instrumentation access to irradiation capsules. The irradiation positions in the beryllium reflector outside the fuel elements also have a high flux; however, although instrumented, they were too small and too few to replace the facilities of a materials testing reactor. To address these drawbacks, the HFIR Irradiation Facilities Improvement Project consisted of modifications to the reactor vessel cover, internal structures, and reflector. Two instrumented facilities were provided in the flux trap region, and the number of materials irradiation positions in the removable beryllium (RB) was increased from four to eight, each with almost twice the available experimental space of the previous ones. The instrumented target facilities were completed in August 1986, and the RB facilities were completed in June 1987.

  18. Operational Philosophy for the Advanced Test Reactor National Scientific User Facility

    Energy Technology Data Exchange (ETDEWEB)

    J. Benson; J. Cole; J. Jackson; F. Marshall; D. Ogden; J. Rempe; M. C. Thelen

    2013-02-01

    In 2007, the Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF). At its core, the ATR NSUF Program combines access to a portion of the available ATR radiation capability, the associated required examination and analysis facilities at the Idaho National Laboratory (INL), and INL staff expertise with novel ideas provided by external contributors (universities, laboratories, and industry). These collaborations define the cutting edge of nuclear technology research in high-temperature and radiation environments, contribute to improved industry performance of current and future light-water reactors (LWRs), and stimulate cooperative research between user groups conducting basic and applied research. To make possible the broadest access to key national capability, the ATR NSUF formed a partnership program that also makes available access to critical facilities outside of the INL. Finally, the ATR NSUF has established a sample library that allows access to pre-irradiated samples as needed by national research teams.

  19. Experimental facility for development of high-temperature reactor technology: instrumentation needs and challenges

    Directory of Open Access Journals (Sweden)

    Sabharwall Piyush

    2015-01-01

    Full Text Available A high-temperature, multi-fluid, multi-loop test facility is under development at the Idaho National Laboratory for support of thermal hydraulic materials, and system integration research for high-temperature reactors. The experimental facility includes a high-temperature helium loop, a liquid salt loop, and a hot water/steam loop. The three loops will be thermally coupled through an intermediate heat exchanger (IHX and a secondary heat exchanger (SHX. Research topics to be addressed include the characterization and performance evaluation of candidate compact heat exchangers such as printed circuit heat exchangers (PCHEs at prototypical operating conditions. Each loop will also include an interchangeable high-temperature test section that can be customized to address specific research issues associated with each working fluid. This paper also discusses needs and challenges associated with advanced instrumentation for the multi-loop facility, which could be further applied to advanced high-temperature reactors. Based on its relevance to advanced reactor systems, the new facility has been named the Advanced Reactor Technology Integral System Test (ARTIST facility. A preliminary design configuration of the ARTIST facility will be presented with the required design and operating characteristics of the various components. The initial configuration will include a high-temperature (750 °C, high-pressure (7 MPa helium loop thermally integrated with a molten fluoride salt (KF-ZrF4 flow loop operating at low pressure (0.2 MPa, at a temperature of ∼450 °C. The salt loop will be thermally integrated with the steam/water loop operating at PWR conditions. Experiment design challenges include identifying suitable materials and components that will withstand the required loop operating conditions. The instrumentation needs to be highly accurate (negligible drift in measuring operational data for extended periods of times, as data collected will be

  20. Proposed design for the PGAA facility at the TRIGA IPR-R1 research reactor.

    Science.gov (United States)

    Guerra, Bruno T; Jacimovic, Radojko; Menezes, Maria Angela Bc; Leal, Alexandre S

    2013-01-01

    This work presents an initial proposed design of a Prompt Gamma Activation Analysis (PGAA) facility to be installed at the TRIGA IPR-R1, a 60 years old research reactor of the Centre of Development of Nuclear Technology (CDTN) in Brazil. The basic characteristics of the facility and the results of the neutron flux are presented and discussed. The proposed design is based on a quasi vertical tube as a neutron guide from the reactor core, inside the reactor pool, 6 m below the room's level where shall be located the rack containing the set sample/detector/shielding. The evaluation of the thermal and epithermal neutron flux in the sample position was done considering the experimental data obtained from a vertical neutron guide, already existent in the reactor, and the simulated model for the facility. The experimental determination of the neutron flux was obtained through the standard procedure of using Au monitors in different positions of the vertical tube. In order to validate both, this experiment and calculations of the simulated model, the flux was also determined in different positions in the core used for sample irradiation. The model of the system was developed using the Monte Carlo code MCNP5. The preliminary results suggest the possibility of obtaining a beam with minimum thermal flux of magnitude 10(6) cm(-2) s(-1), which confirm the technical feasibility of the installation of PGAA at the TRIGA IPR-R1 reactor. This beam would open new possibilities for enhancing the applications using the reactor.

  1. Testing and Research Capabilities at the Sandia Fast Pulsed Reactor Facility

    Science.gov (United States)

    Berry, Donald T.

    1994-07-01

    A wide variety of space-based system components have been qualified for use through neutron irradiation testing performed at the Sandia Pulsed Reactor (SPR) Facility. The SPR Facility is the operating location for two fast burst reactors, SPR II and SPR III, which have been used to induce neutron and gamma damage in electronic components and other materials for customers in the Department of Energy, Department of Defense, NASA, and the private sector. In addition to the pulse mode of operation, during which peak fluxes of up to 1023 n/m2-s are achieved, the steady state mode allows for the long term irradiation of components and systems in a fast neutron environment at a flux of up to 5×1015 n/m2-s. The SPR reactors are operated in a 9.2 meter diameter exposure cell, or Kiva, suitable for the irradiation of large test articles external to the reactors. Currently, a new upgraded version of SPR III (SPR HIM) is in fabrication; a unique feature of SPR HIM is its 190 mm (usable diameter) central irradiation cavity, the largest of any U.S. fast burst reactor. An improved cooling system permits continuous operation at power levels in excess of 20 kWt. The SPR Facility is also the operating site for a critical assembly which was used to characterize prototypic fuels in arrays appropriate for the Space Nuclear Thermal Propulsion Program. Work continues on use of the facility to design, build, and operate critical assemblies for a diverse customer base.

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

    Energy Technology Data Exchange (ETDEWEB)

    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 the 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-04-17

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

  4. Strategic sourcing and procurement of facilities management services

    DEFF Research Database (Denmark)

    Jensen, Per Anker

    2017-01-01

    Purpose: The purpose it to provide insights into strategic sourcing concerning Facilities Management (FM) and how it can contribute to a sourcing decision that combines the benefits of internal and external provision with consideration of business risk and cost. Design/methodology/approach: The p......Purpose: The purpose it to provide insights into strategic sourcing concerning Facilities Management (FM) and how it can contribute to a sourcing decision that combines the benefits of internal and external provision with consideration of business risk and cost. Design...

  5. Guidelines for preparing criticality safety evaluations at Department of Energy non-reactor nuclear facilities

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-09-01

    This Department of Energy (DOE) is approved for use by all components of DOE. It contains guidelines that should be followed when preparing Criticality Safety Evaluations that will be used to demonstrate the safety of operations performed at DOE Non-Reactor Nuclear Facilities. Adherence with these guidelines will provide consistency and uniformity in Criticality Safety Evaluations (CSEs) across the complex and will document compliance with DOE Order 5480.24 requirements as they pertain to CSEs.

  6. The SPES3 Experimental Facility Design for the IRIS Reactor Simulation

    Directory of Open Access Journals (Sweden)

    Mario Carelli

    2009-01-01

    Full Text Available IRIS is an advanced integral pressurized water reactor, developed by an international consortium led by Westinghouse. The licensing process requires the execution of integral and separate effect tests on a properly scaled reactor simulator for reactor concept, safety system verification, and code assessment. Within the framework of an Italian R&D program on Nuclear Fission, managed by ENEA and supported by the Ministry of Economic Development, the SPES3 facility is under design and will be built and operated at SIET laboratories. SPES3 simulates the primary, secondary, and containment systems of IRIS with 1 : 100 volume scale, full elevation, and prototypical thermal-hydraulic conditions. The simulation of the facility with the RELAP5 code and the execution of the tests will provide a reliable tool for data extrapolation and safety analyses of the final IRIS design. This paper summarises the main design steps of the SPES3 integral test facility, underlying choices and phases that lead to the final design.

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

    Science.gov (United States)

    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 %).

  8. Advanced Test Reactor National Scientific User Facility (ATR NSUF) Monthly Report November 2014

    Energy Technology Data Exchange (ETDEWEB)

    Soelberg, Renae [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-11-01

    Advanced Test Reactor National Scientific User Facility (ATR NSUF) Monthly Report November 2014 Highlights Rory Kennedy and Sarah Robertson attended the American Nuclear Society Winter Meeting and Nuclear Technology Expo in Anaheim, California, Nov. 10-13. ATR NSUF exhibited at the technology expo where hundreds of meeting participants had an opportunity to learn more about ATR NSUF. Dr. Kennedy briefed the Nuclear Engineering Department Heads Organization (NEDHO) on the workings of the ATR NSUF. • Rory Kennedy, James Cole and Dan Ogden participated in a reactor instrumentation discussion with Jean-Francois Villard and Christopher Destouches of CEA and several members of the INL staff. • ATR NSUF received approval from the NE-20 office to start planning the annual Users Meeting. The meeting will be held at INL, June 22-25. • Mike Worley, director of the Office of Innovative Nuclear Research (NE-42), visited INL Nov. 4-5. Milestones Completed • Recommendations for the Summer Rapid Turnaround Experiment awards were submitted to DOE-HQ Nov. 12 (Level 2 milestone due Nov. 30). Major Accomplishments/Activities • The University of California, Santa Barbara 2 experiment was unloaded from the GE-2000 at HFEF. The experiment specimen packs will be removed and shipped to ORNL for PIE. • The Terrani experiment, one of three FY 2014 new awards, was completed utilizing the Advanced Photon Source MRCAT beamline. The experiment investigated the chemical state of Ag and Pd in SiC shell of irradiated TRISO particles via X-ray Absorption Fine Structure (XAFS) spectroscopy. Upcoming Meetings/Events • The ATR NSUF program review meeting will be held Dec. 9-10 at L’Enfant Plaza. In addition to NSUF staff and users, NE-4, NE-5 and NE-7 representatives will attend the meeting. Awarded Research Projects Boise State University Rapid Turnaround Experiments (14-485 and 14-486) Nanoindentation and TEM work on the T91, HT9, HCM12A and 9Cr ODS specimens has been completed at

  9. Calculation and comparisons with measurement of fast neutron fluxes in the material testing facilities of the NRU research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Leung, T.C. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2012-07-01

    The NRU reactor at Chalk River provides three irradiation facilities to study the effects of fast neutrons (E> 1 MeV) on reactor materials for assessing material damage and deformation. The facilities comprise two types of fast neutron rods (Mark 4 and Mark 7), and a Material Test Bundle (MTB) irradiated in a loop site. This paper describes the neutronic simulation of these testing facilities using the WIMS-AECL and TRIAD codes, and comparisons with the fast neutron flux measurements using iron-wire activation techniques. It also provides comparisons of flux levels, neutron spectra, and size limitations of the experimental cavities between these test facilities. (author)

  10. Advanced Test Reactor National Scientific User Facility: Addressing advanced nuclear materials research

    Energy Technology Data Exchange (ETDEWEB)

    John Jackson; Todd Allen; Frances Marshall; Jim Cole

    2013-03-01

    The Advanced Test Reactor National Scientific User Facility (ATR NSUF), based at the Idaho National Laboratory in the United States, is supporting Department of Energy and industry research efforts to ensure the properties of materials in light water reactors are well understood. The ATR NSUF is providing this support through three main efforts: establishing unique infrastructure necessary to conduct research on highly radioactive materials, conducting research in conjunction with industry partners on life extension relevant topics, and providing training courses to encourage more U.S. researchers to understand and address LWR materials issues. In 2010 and 2011, several advanced instruments with capability focused on resolving nuclear material performance issues through analysis on the micro (10-6 m) to atomic (10-10 m) scales were installed primarily at the Center for Advanced Energy Studies (CAES) in Idaho Falls, Idaho. These instruments included a local electrode atom probe (LEAP), a field-emission gun scanning transmission electron microscope (FEG-STEM), a focused ion beam (FIB) system, a Raman spectrometer, and an nanoindentor/atomic force microscope. Ongoing capability enhancements intended to support industry efforts include completion of two shielded, irradiation assisted stress corrosion cracking (IASCC) test loops, the first of which will come online in early calendar year 2013, a pressurized and controlled chemistry water loop for the ATR center flux trap, and a dedicated facility intended to house post irradiation examination equipment. In addition to capability enhancements at the main site in Idaho, the ATR NSUF also welcomed two new partner facilities in 2011 and two new partner facilities in 2012; the Oak Ridge National Laboratory, High Flux Isotope Reactor (HFIR) and associated hot cells and the University California Berkeley capabilities in irradiated materials analysis were added in 2011. In 2012, Purdue University’s Interaction of Materials

  11. Post 9-11 Security Issues for Non-Power Reactor Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Zaffuts, P. J.

    2003-02-25

    This paper addresses the legal and practical issues arising out of the design and implementation of a security-enhancement program for non power reactor nuclear facilities. The security enhancements discussed are derived from the commercial nuclear power industry's approach to security. The nuclear power industry's long and successful experience with protecting highly sensitive assets provides a wealth of information and lessons that should be examined by other industries contemplating security improvements, including, but not limited to facilities using or disposing of nuclear materials. This paper describes the nuclear industry's approach to security, the advantages and disadvantages of its constituent elements, and the legal issues that facilities will need to address when adopting some or all of these elements in the absence of statutory or regulatory requirements to do so.

  12. Effect of Sulfur Source on the Performance and Metal Retention of Methanol-Fed UASB Reactors

    NARCIS (Netherlands)

    Zandvoort, M.H.; Hullebusch, van E.D.; Gieteling, J.; Lettinga, G.; Lens, P.N.L.

    2005-01-01

    The effect of a sulfur source on the performance and metal retention of methanol-fed upflow anaerobic sludge bed (UASB) reactors was investigated. For this purpose, two UASB reactors were operated with cobalt preloaded granular sludge (1 mM CoCl2; 30° C; 24 h) at an organic loading rate (OLR) of 5 g

  13. Neutronic design study of accelerator driven system (ADS) for Jordan subcritical reactor as a neutron source for nuclear research.

    Science.gov (United States)

    Xoubi, Ned

    2018-01-01

    In this paper, a preliminary neutronic design study of an accelerator driven subcritical system for Jordan Subcritical Assembly (JSA) is presented. The conceptual design of coupling the JSA core with proton accelerator and spallation target is investigated, and its feasibility as a neutron source for nuclear research, and possibly for target irradiation and isotope production evaluated. 3D MCNPX model of the JSA reactor, the accelerator beam, and the Pb target was developed, based on actual reactor parameters. MCNPX calculations were carried out to estimate the absolute radial and axial neutron flux in the reactor, and to calculate the multiplication factor K eff and heat generated in the reactor. Numerical results showed an enormous increase in the neutron flux, by seven orders of magnitude, compared to the current JSA core design using Pu-Be source. In this research the results obtained are discussed and compared with those of the JSA, and do confirm the feasibility of utilizing the JSA as a viable nuclear research facility with adequate neutron flux. Copyright © 2017 Elsevier Ltd. All rights reserved.

  14. Innovation Sources and Role of ICT in Facilities Services

    DEFF Research Database (Denmark)

    Scupola, Ada; Holzweber, Markus; Tuunainen, VK.

    2010-01-01

    In this paper, we investigate innovation sources in facilities services and the role that ICT has in supporting such innovation processes. Based on literature review, we propose a conceptual framework, which is then used to analyze empirical data. The empirical data was collected through seven semi...... structured interviews with main actors in the facilities management market. Our main finding is that facilities services innovation is mainly driven by management. However, employees also contribute to innovation. Customers have also a role in co-creation and as customer pull, even though these roles seem...

  15. Safety evaluation report related to the renewal of the facility license for the research reactor at the Dow Chemical Company

    Energy Technology Data Exchange (ETDEWEB)

    1989-04-01

    This safety evaluation report for the application filed by the Dow Chemical Company for renewal of facility Operating License R-108 to continue to operate its research reactor at an increased operating power level has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is located on the grounds of the Michigan Division of the Dow Chemical Company in Midland, Michigan. The staff concludes that the Dow Chemical Company can continue to operate its reactor without endangering the health and safety of the public.

  16. Gas cooled fast breeder reactor design for a circulator test facility (modified HTGR circulator test facility)

    Energy Technology Data Exchange (ETDEWEB)

    1979-10-01

    A GCFR helium circulator test facility sized for full design conditions is proposed for meeting the above requirements. The circulator will be mounted in a large vessel containing high pressure helium which will permit testing at the same power, speed, pressure, temperature and flow conditions intended in the demonstration plant. The electric drive motor for the circulator will obtain its power from an electric supply and distribution system in which electric power will be taken from a local utility. The conceptual design decribed in this report is the result of close interaction between the General Atomic Company (GA), designer of the GCFR, and The Ralph M. Parson Company, architect/engineer for the test facility. A realistic estimate of total project cost is presented, together with a schedule for design, procurement, construction, and inspection.

  17. Summary engineering description of underwater fuel storage facility for foreign research reactor spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Dahlke, H.J.; Johnson, D.A.; Rawlins, J.K.; Searle, D.K.; Wachs, G.W.

    1994-10-01

    This document is a summary description for an Underwater Fuel Storage Facility (UFSF) for foreign research reactor (FRR) spent nuclear fuel (SNF). A FRR SNF environmental Impact Statement (EIS) is being prepared and will include both wet and dry storage facilities as storage alternatives. For the UFSF presented in this document, a specific site is not chosen. This facility can be sited at any one of the five locations under consideration in the EIS. These locations are the Idaho National Engineering Laboratory, Savannah River Site, Hanford, Oak Ridge National Laboratory, and Nevada Test Site. Generic facility environmental impacts and emissions are provided in this report. A baseline fuel element is defined in Section 2.2, and the results of a fission product analysis are presented. Requirements for a storage facility have been researched and are summarized in Section 3. Section 4 describes three facility options: (1) the Centralized-UFSF, which would store the entire fuel element quantity in a single facility at a single location, (2) the Regionalized Large-UFSF, which would store 75% of the fuel element quantity in some region of the country, and (3) the Regionalized Small-UFSF, which would store 25% of the fuel element quantity, with the possibility of a number of these facilities in various regions throughout the country. The operational philosophy is presented in Section 5, and Section 6 contains a description of the equipment. Section 7 defines the utilities required for the facility. Cost estimates are discussed in Section 8, and detailed cost estimates are included. Impacts to worker safety, public safety, and the environment are discussed in Section 9. Accidental releases are presented in Section 10. Standard Environmental Impact Forms are included in Section 11.

  18. Plant model of KIPT neutron source facility simulator

    Energy Technology Data Exchange (ETDEWEB)

    Cao, Yan [Argonne National Lab. (ANL), Argonne, IL (United States); Wei, Thomas Y. [Argonne National Lab. (ANL), Argonne, IL (United States); Grelle, Austin L. [Argonne National Lab. (ANL), Argonne, IL (United States); Gohar, Yousry [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-02-01

    Argonne National Laboratory (ANL) of the United States and Kharkov Institute of Physics and Technology (KIPT) of Ukraine are collaborating on constructing a neutron source facility at KIPT, Kharkov, Ukraine. The facility has 100-kW electron beam driving a subcritical assembly (SCA). The electron beam interacts with a natural uranium target or a tungsten target to generate neutrons, and deposits its power in the target zone. The total fission power generated in SCA is about 300 kW. Two primary cooling loops are designed to remove 100-kW and 300-kW from the target zone and the SCA, respectively. A secondary cooling system is coupled with the primary cooling system to dispose of the generated heat outside the facility buildings to the atmosphere. In addition, the electron accelerator has a low efficiency for generating the electron beam, which uses another secondary cooling loop to remove the generated heat from the accelerator primary cooling loop. One of the main functions the KIPT neutron source facility is to train young nuclear specialists; therefore, ANL has developed the KIPT Neutron Source Facility Simulator for this function. In this simulator, a Plant Control System and a Plant Protection System were developed to perform proper control and to provide automatic protection against unsafe and improper operation of the facility during the steady-state and the transient states using a facility plant model. This report focuses on describing the physics of the plant model and provides several test cases to demonstrate its capabilities. The plant facility model uses the PYTHON script language. It is consistent with the computer language of the plant control system. It is easy to integrate with the simulator without an additional interface, and it is able to simulate the transients of the cooling systems with system control variables changing on real-time.

  19. 77 FR 68155 - The Armed Forces Radiobiology Research Institute TRIGA Reactor: Facility Operating License No. R-84

    Science.gov (United States)

    2012-11-15

    ... COMMISSION The Armed Forces Radiobiology Research Institute TRIGA Reactor: Facility Operating License No. R... Operating License No. R-84 (Application), which currently authorizes the Armed Forces Radiobiology Research... the renewal of Facility Operating License No. R-84, which currently authorizes the licensee to operate...

  20. 77 FR 7613 - Dow Chemical Company; Dow Chemical TRIGA Research Reactor; Facility Operating License No. R-108

    Science.gov (United States)

    2012-02-13

    ... COMMISSION Dow Chemical Company; Dow Chemical TRIGA Research Reactor; Facility Operating License No. R-108... renewal of Facility Operating License No. R-108 (``Application''), which currently authorizes the Dow... Operating License No. R-108 for the DTRR. The application contains SUNSI. Based on its initial review of the...

  1. Awareness, Preference, Utilization, and Messaging Research for the Spallation Neutron Source and High Flux Isotope Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bryant, Rebecca [Bryant Research, LLC; Kszos, Lynn A [ORNL

    2011-03-01

    Oak Ridge National Laboratory (ORNL) offers the scientific community unique access to two types of world-class neutron sources at a single site - the Spallation Neutron Source (SNS) and the High Flux Isotope Reactor (HFIR). The 85-MW HFIR provides one of the highest steady-state neutron fluxes of any research reactor in the world, and the SNS is one of the world's most intense pulsed neutron beams. Management of these two resources is the responsibility of the Neutron Sciences Directorate (NScD). NScD commissioned this survey research to develop baseline information regarding awareness of and perceptions about neutron science. Specific areas of investigative interest include the following: (1) awareness levels among those in the scientific community about the two neutron sources that ORNL offers; (2) the level of understanding members of various scientific communities have regarding benefits that neutron scattering techniques offer; and (3) any perceptions that negatively impact utilization of the facilities. NScD leadership identified users of two light sources in North America - the Advanced Photon Source (APS) at Argonne National Laboratory and the National Synchrotron Light Source (NSLS) at Brookhaven National Laboratory - as key publics. Given the type of research in which these scientists engage, they would quite likely benefit from including the neutron techniques available at SNS and HFIR among their scientific investigation tools. The objective of the survey of users of APS, NSLS, SNS, and HFIR was to explore awareness of and perceptions regarding SNS and HFIR among those in selected scientific communities. Perceptions of SNS and FHIR will provide a foundation for strategic communication plan development and for developing key educational messages. The survey was conducted in two phases. The first phase included qualitative methods of (1) key stakeholder meetings; (2) online interviews with user administrators of APS and NSLS; and (3) one

  2. Decommissioning an Active Historical Reactor Facility at the Savannah River Site - 13453

    Energy Technology Data Exchange (ETDEWEB)

    Bergren, Christopher L.; Long, J. Tony; Blankenship, John K. [Savannah River Nuclear Solutions, LLC, Bldg. 730-4B, Aiken, SC 29808 (United States); Adams, Karen M. [United States Department of Energy, Bldg. 730-B, Aiken, SC 29808 (United States)

    2013-07-01

    The Savannah River Site (SRS) is an 802 square-kilometer United States Department of Energy (US DOE) nuclear facility located along the Savannah River near Aiken, South Carolina, where Management and Operations are performed by Savannah River Nuclear Solutions (SRNS). In 2004, DOE recognized SRS as structure within the Cold War Historic District of national, state and local significance composed of the first generation of facilities constructed and operated from 1950 through 1989 to produce plutonium and tritium for our nation's defense. DOE agreed to manage the SRS 105-C Reactor Facility as a potentially historic property due to its significance in supporting the U.S. Cold War Mission and for potential for future interpretation. This reactor has five primary areas within it, including a Disassembly Basin (DB) that received irradiated materials from the reactor, cooled them and prepared the components for loading and transport to a Separation Canyon for processing. The 6,317 square meter area was divided into numerous work/storage areas. The walls between the individual basin compartments have narrow vertical openings called 'slots' that permit the transfer of material from one section to another. Data indicated there was over 830 curies of radioactivity associated with the basin sediments and approximately 9.1 M liters of contaminated water, not including a large quantity of activated reactor equipment, scrap metal, and debris on the basin floor. The need for an action was identified in 2010 to reduce risks to personnel in the facility and to eliminate the possible release of contaminants into the environment. The release of DB water could potentially migrate to the aquifer and contaminate groundwater. DOE, its regulators [U. S. Environmental Protection Agency (USEPA)-Region 4 and the South Carolina Department of Health and Environmental Control (SCDHEC)] and the SC Historical Preservation Office (SHPO) agreed/concurred to perform a non

  3. Passive and Active Radiation Measurements Capability at the INL Zero Power Physics Reactor (ZPPR) Facility

    Energy Technology Data Exchange (ETDEWEB)

    Robert Neibert; John Zabriskie; Collin Knight; James L. Jones

    2010-12-01

    The Zero Power Physics Reactor (ZPPR) facility is a Department of Energy facility located in the Idaho National Laboratory’s (INL) Materials and Fuels Complex. It contains various nuclear and non-nuclear materials that are available to support many radiation measurement assessments. User-selected, single material, nuclear and non-nuclear materials can be readily utilized with ZPPR clamshell containers with almost no criticality concerns. If custom, multi-material configurations are desired, the ZPPR clamshell or an approved aluminum Inspection Object (IO) Box container may be utilized, yet each specific material configuration will require a criticality assessment. As an example of the specialized material configurations possible, the National Nuclear Security Agency’s Office of Nuclear Verification (NNSA/NA 243) has sponsored the assembly of six material configurations. These are shown in the Appendixes and have been designated for semi-permanent storage that can be available to support various radiation measurement applications.

  4. Medical Isotope Production Analyses In KIPT Neutron Source Facility

    Energy Technology Data Exchange (ETDEWEB)

    Talamo, Alberto [Argonne National Lab. (ANL), Argonne, IL (United States); Gohar, Yousry [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-01-01

    Medical isotope production analyses in Kharkov Institute of Physics and Technology (KIPT) neutron source facility were performed to include the details of the irradiation cassette and the self-shielding effect. An updated detailed model of the facility was used for the analyses. The facility consists of an accelerator-driven system (ADS), which has a subcritical assembly using low-enriched uranium fuel elements with a beryllium-graphite reflector. The beryllium assemblies of the reflector have the same outer geometry as the fuel elements, which permits loading the subcritical assembly with different number of fuel elements without impacting the reflector performance. The subcritical assembly is driven by an external neutron source generated from the interaction of 100-kW electron beam with a tungsten target. The facility construction was completed at the end of 2015, and it is planned to start the operation during the year of 2016. It is the first ADS in the world, which has a coolant system for removing the generated fission power. Argonne National Laboratory has developed the design concept and performed extensive design analyses for the facility including its utilization for the production of different radioactive medical isotopes. 99Mo is the parent isotope of 99mTc, which is the most commonly used medical radioactive isotope. Detailed analyses were performed to define the optimal sample irradiation location and the generated activity, for several radioactive medical isotopes, as a function of the irradiation time.

  5. AREAL test facility for advanced accelerator and radiation source concepts

    Science.gov (United States)

    Tsakanov, V. M.; Amatuni, G. A.; Amirkhanyan, Z. G.; Aslyan, L. V.; Avagyan, V. Sh.; Danielyan, V. A.; Davtyan, H. D.; Dekhtiarov, V. S.; Gevorgyan, K. L.; Ghazaryan, N. G.; Grigoryan, B. A.; Grigoryan, A. H.; Hakobyan, L. S.; Haroutiunian, S. G.; Ivanyan, M. I.; Khachatryan, V. G.; Laziev, E. M.; Manukyan, P. S.; Margaryan, I. N.; Markosyan, T. M.; Martirosyan, N. V.; Mehrabyan, Sh. A.; Mkrtchyan, T. H.; Muradyan, L. Kh.; Nikogosyan, G. H.; Petrosyan, V. H.; Sahakyan, V. V.; Sargsyan, A. A.; Simonyan, A. S.; Toneyan, H. A.; Tsakanian, A. V.; Vardanyan, T. L.; Vardanyan, A. S.; Yeremyan, A. S.; Zakaryan, S. V.; Zanyan, G. S.

    2016-09-01

    Advanced Research Electron Accelerator Laboratory (AREAL) is a 50 MeV electron linear accelerator project with a laser driven RF gun being constructed at the CANDLE Synchrotron Research Institute. In addition to applications in life and materials sciences, the project aims as a test facility for advanced accelerator and radiation source concepts. In this paper, the AREAL RF photoinjector performance, the facility design considerations and its highlights in the fields of free electron laser, the study of new high frequency accelerating structures, the beam microbunching and wakefield acceleration concepts are presented.

  6. Facilities for small-molecule crystallography at synchrotron sources.

    Science.gov (United States)

    Barnett, Sarah A; Nowell, Harriott; Warren, Mark R; Wilcox, Andrian; Allan, David R

    2016-01-01

    Although macromolecular crystallography is a widely supported technique at synchrotron radiation facilities throughout the world, there are, in comparison, only very few beamlines dedicated to small-molecule crystallography. This limited provision is despite the increasing demand for beamtime from the chemical crystallography community and the ever greater overlap between systems that can be classed as either small macromolecules or large small molecules. In this article, a very brief overview of beamlines that support small-molecule single-crystal diffraction techniques will be given along with a more detailed description of beamline I19, a dedicated facility for small-molecule crystallography at Diamond Light Source.

  7. The DRAGON aerosol research facility to study aerosol behaviour for reactor safety applications

    Energy Technology Data Exchange (ETDEWEB)

    Suckow, Detlef; Guentay, Salih [Paul Scherrer Institut, Villigen (Switzerland)

    2008-07-01

    During a severe accident in a nuclear power plant fission products are expected to be released in form of aerosol particles and droplets. To study the behaviour of safety relevant reactor components under aerosol loads and prototypical severe accident conditions the multi-purpose aerosol generation facility DRAGON is used since 1994 for several projects. DRAGON can generate aerosol particles by the evaporation-condensation technique using a plasma torch system, fluidized bed and atomization of particles suspended in a liquid. Soluble, hygroscopic aerosol (i.e. CsOH) and insoluble aerosol particles (i.e. SnO{sub 2}, TiO{sub 2}) or mixtures of them can be used. DRAGON uses state-of-the-art thermal-hydraulic, data acquisition and aerosol measurement techniques and is mainly composed of a mixing chamber, the plasma torch system, a steam generator, nitrogen gas and compressed air delivery systems, several aerosol delivery piping, gas heaters and several auxiliary systems to provide vacuum, coolant and off-gas treatment. The facility can be operated at system pressure of 5 bars, temperatures of 300 deg. C, flow rates of non-condensable gas of 900 kg/h and steam of 270 kg/h, respectively. A test section under investigation is attached to DRAGON. The paper summarizes and demonstrates with the help of two project examples the capabilities of DRAGON for reactor safety studies. (authors)

  8. Summary of estimated doses and risks resulting from routine radionuclide releases from fast breeder reactor fuel cycle facilities

    Energy Technology Data Exchange (ETDEWEB)

    Miller, C.W.; Meyer, H.R.

    1985-01-01

    A project is underway at Oak Ridge National Laboratory to assess the human health and environment effects associated with operation of Liquid Metal Fast Breeder Reactor fuel cycle. In this first phase of the work, emphasis was focused on routine radionuclide releases from reactor and reprocessing facilities. For this study, sites for fifty 1-GW(e) capacity reactors and three reprocessing plants were selected to develop scenarios representative of US power requirements. For both the reactor and reprocessing facility siting schemes selected, relatively small impacts were calculated for locality-specific populations residing within 100 km. Also, the results of these analyses are being used in the identification of research priorities. 13 refs., 2 figs., 3 tabs.

  9. Regulatory Technology Development Plan Sodium Fast Reactor. Mechanistic Source Term Development

    Energy Technology Data Exchange (ETDEWEB)

    Grabaskas, David S. [Argonne National Lab. (ANL), Argonne, IL (United States); Brunett, Acacia Joann [Argonne National Lab. (ANL), Argonne, IL (United States); Bucknor, Matthew D. [Argonne National Lab. (ANL), Argonne, IL (United States); Sienicki, James J. [Argonne National Lab. (ANL), Argonne, IL (United States); Sofu, Tanju [Argonne National Lab. (ANL), Argonne, IL (United States)

    2015-02-28

    Construction and operation of a nuclear power installation in the U.S. requires licensing by the U.S. Nuclear Regulatory Commission (NRC). A vital part of this licensing process and integrated safety assessment entails the analysis of a source term (or source terms) that represents the release of radionuclides during normal operation and accident sequences. Historically, nuclear plant source term analyses have utilized deterministic, bounding assessments of the radionuclides released to the environment. Significant advancements in technical capabilities and the knowledge state have enabled the development of more realistic analyses such that a mechanistic source term (MST) assessment is now expected to be a requirement of advanced reactor licensing. This report focuses on the state of development of an MST for a sodium fast reactor (SFR), with the intent of aiding in the process of MST definition by qualitatively identifying and characterizing the major sources and transport processes of radionuclides. Due to common design characteristics among current U.S. SFR vendor designs, a metal-fuel, pool-type SFR has been selected as the reference design for this work, with all phenomenological discussions geared toward this specific reactor configuration. This works also aims to identify the key gaps and uncertainties in the current knowledge state that must be addressed for SFR MST development. It is anticipated that this knowledge state assessment can enable the coordination of technology and analysis tool development discussions such that any knowledge gaps may be addressed. Sources of radionuclides considered in this report include releases originating both in-vessel and ex-vessel, including in-core fuel, primary sodium and cover gas cleanup systems, and spent fuel movement and handling. Transport phenomena affecting various release groups are identified and qualitatively discussed, including fuel pin and primary coolant retention, and behavior in the cover gas and

  10. Upgrade of the BATMAN test facility for H- source development

    Science.gov (United States)

    Heinemann, B.; Fröschle, M.; Falter, H.-D.; Fantz, U.; Franzen, P.; Kraus, W.; Nocentini, R.; Riedl, R.; Ruf, B.

    2015-04-01

    The development of a radio frequency (RF) driven source for negative hydrogen ions for the neutral beam heating devices of fusion experiments has been successfully carried out at IPP since 1996 on the test facility BATMAN. The required ITER parameters have been achieved with the prototype source consisting of a cylindrical driver on the back side of a racetrack like expansion chamber. The extraction system, called "Large Area Grid" (LAG) was derived from a positive ion accelerator from ASDEX Upgrade (AUG) using its aperture size (ø 8 mm) and pattern but replacing the first two electrodes and masking down the extraction area to 70 cm2. BATMAN is a well diagnosed and highly flexible test facility which will be kept operational in parallel to the half size ITER source test facility ELISE for further developments to improve the RF efficiency and the beam properties. It is therefore planned to upgrade BATMAN with a new ITER-like grid system (ILG) representing almost one ITER beamlet group, namely 5 × 14 apertures (ø 14 mm). Additionally to the standard three grid extraction system a repeller electrode upstream of the grounded grid can optionally be installed which is positively charged against it by 2 kV. This is designated to affect the onset of the space charge compensation downstream of the grounded grid and to reduce the backstreaming of positive ions from the drift space backwards into the ion source. For magnetic filter field studies a plasma grid current up to 3 kA will be available as well as permanent magnets embedded into a diagnostic flange or in an external magnet frame. Furthermore different source vessels and source configurations are under discussion for BATMAN, e.g. using the AUG type racetrack RF source as driver instead of the circular one or modifying the expansion chamber for a more flexible position of the external magnet frame.

  11. Modernization of the facilities of the TRIGA Mark III reactor of ININ; Modernizacion de las instalaciones del reactor TRIGA Mark III del ININ

    Energy Technology Data Exchange (ETDEWEB)

    Mendez T, D.; Flores C, J., E-mail: dario.mendez@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2016-09-15

    The TRIGA Mark III reactor of the Instituto Nacional de Investigaciones Nucleares (ININ) has been in operation since 1968 under strict maintenance and component replacement programs, which has allowed its safe operation during this time. Under this scheme, the reactor was operating under suitable conditions, taking into account the different requests for operation that were received for the samples irradiation for the radioisotopes production such as the Sm-153, personnel training, basic research, archaeology and environmental studies and nuclear chemistry of the elements. However, a modernization program of its components and laboratories was required, in order to improve safety in the operation of the same and to increase its use in the analysis of samples by neutron activation and in the training of personnel. This program known as Modernization Program of the Reactor Facilities, was proposed alongside the project to replace high-enrichment fuels with low-enrichment fuels at the end of 2011 and early 2012. The central aspects of this program are described in this work, grouped into generic topics that include instrumentation and control, the radiological monitoring system of the area, the cooling system, the ventilation system, the neutron activation analysis laboratory, the manufacture of graphite elements, inspection submersible system of the pool, temporary storage system for irradiated fuels, traveling crane, Reactor support laboratories and technical meetings, courses and seminars for reactor personnel and associated groups. It also describes some of the most relevant components required for each system and the progress that is made in each one of them. As a fundamental result of the implementation of this Modernization Program of the Reactor Facilities, there has been a substantial improvement in the performance of the systems and components of its facilities, in the reliability of its operation and in the safety of the same. (Author)

  12. Extraction of pure thermal neutron beam for the proposed PGNAA facility at the TRIGA research reactor of AERE, Savar, Bangladesh

    Science.gov (United States)

    Alam, Sabina; Zaman, M. A.; Islam, S. M. A.; Ahsan, M. H.

    1993-10-01

    A study on collimators and filters for the design of a spectrometer for prompt gamma neutron activation analysis (PGNAA) at one of the radial beamports of the TRIGA Mark II reactor at AERE, Savar has been carried out. On the basis of this study a collimator and a filter have been designed for the proposed PGNAA facility. Calculations have been done for measuring neutron flux at various positions of the core of the reactor using the computer code TRIGAP. Gamma dose in the core of the reactor has also been measured experimentally using TLD technique in the present work.

  13. STEADY STATE MODELING OF THE MINIMUM CRITICAL CORE OF THE TRANSIENT REACTOR TEST FACILITY

    Energy Technology Data Exchange (ETDEWEB)

    Anthony L. Alberti; Todd S. Palmer; Javier Ortensi; Mark D. DeHart

    2016-05-01

    With the advent of next generation reactor systems and new fuel designs, the U.S. Department of Energy (DOE) has identified the need for the resumption of transient testing of nuclear fuels. The DOE has decided that the Transient Reactor Test Facility (TREAT) at Idaho National Laboratory (INL) is best suited for future testing. TREAT is a thermal neutron spectrum, air-cooled, nuclear test facility that is designed to test nuclear fuels in transient scenarios. These specific scenarios range from simple temperature transients to full fuel melt accidents. DOE has expressed a desire to develop a simulation capability that will accurately model the experiments before they are irradiated at the facility. It is the aim for this capability to have an emphasis on effective and safe operation while minimizing experimental time and cost. The multi physics platform MOOSE has been selected as the framework for this project. The goals for this work are to identify the fundamental neutronics properties of TREAT and to develop an accurate steady state model for future multiphysics transient simulations. In order to minimize computational cost, the effect of spatial homogenization and angular discretization are investigated. It was found that significant anisotropy is present in TREAT assemblies and to capture this effect, explicit modeling of cooling channels and inter-element gaps is necessary. For this modeling scheme, single element calculations at 293 K gave power distributions with a root mean square difference of 0.076% from those of reference SERPENT calculations. The minimum critical core configuration with identical gap and channel treatment at 293 K resulted in a root mean square, total core, radial power distribution 2.423% different than those of reference SERPENT solutions.

  14. Passive Safety Features Evaluation of KIPT Neutron Source Facility

    Energy Technology Data Exchange (ETDEWEB)

    Zhong, Zhaopeng [Argonne National Lab. (ANL), Argonne, IL (United States); Gohar, Yousry [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-06-01

    Argonne National Laboratory (ANL) of the United States and Kharkov Institute of Physics and Technology (KIPT) of Ukraine have cooperated on the development, design, and construction of a neutron source facility. The facility was constructed at Kharkov, Ukraine and its commissioning process is underway. It will be used to conduct basic and applied nuclear research, produce medical isotopes, and train young nuclear specialists. The facility has an electron accelerator-driven subcritical assembly. The electron beam power is 100 kW using 100 MeV electrons. Tungsten or natural uranium is the target material for generating neutrons driving the subcritical assembly. The subcritical assembly is composed of WWR-M2 - Russian fuel assemblies with U-235 enrichment of 19.7 wt%, surrounded by beryllium reflector assembles and graphite blocks. The subcritical assembly is seated in a water tank, which is a part of the primary cooling loop. During normal operation, the water coolant operates at room temperature and the total facility power is ~300 KW. The passive safety features of the facility are discussed in in this study. Monte Carlo computer code MCNPX was utilized in the analyses with ENDF/B-VII.0 nuclear data libraries. Negative reactivity temperature feedback was consistently observed, which is important for the facility safety performance. Due to the design of WWR-M2 fuel assemblies, slight water temperature increase and the corresponding water density decrease produce large reactivity drop, which offset the reactivity gain by mistakenly loading an additional fuel assembly. The increase of fuel temperature also causes sufficiently large reactivity decrease. This enhances the facility safety performance because fuel temperature increase provides prompt negative reactivity feedback. The reactivity variation due to an empty fuel position filled by water during the fuel loading process is examined. Also, the loading mistakes of removing beryllium reflector assemblies and

  15. Offline Ion Source Developments for the BECOLA Facility at NSCL

    Science.gov (United States)

    Klose, Andrew; Mantica, Paul; Minamisono, Kei

    2013-04-01

    Singly-charged beams of the stable isotope(s) of K, Ca, Mn, and Fe have been produced using either a commercial plasma ion source or a home-built electron ionization source for the BEam COoler and LAser spectroscopy (BECOLA) facility at the National Superconducting Cyclotron Laboratory (NSCL) at Michigan State University. For each element, collinear laser spectroscopy was performed to confirm the presence of the respective element. Production of stable ion beams for a given element is necessary to obtain reference hyperfine spectra of species with known electromagnetic moments before such nuclear properties can be deduced from the laser hyperfine-structure measurements of rare isotopes that are planned at NSCL. The results from the commissioning tests of the plasma and electron ionization sources will be presented, and development of a new Penning Ion Gauge (PIG) ion source will be discussed.

  16. Lewis Research Center's coal-fired, pressurized, fluidized-bed reactor test facility

    Science.gov (United States)

    Kobak, J. A.; Rollbuhler, R. J.

    1981-01-01

    A 200-kilowatt-thermal, pressurized, fluidized-bed (PFB) reactor, research test facility was designed, constructed, and operated as part of a NASA-funded project to assess and evaluate the effect of PFB hot-gas effluent on aircraft turbine engine materials that might have applications in stationary-power-plant turbogenerators. Some of the techniques and components developed for this PFB system are described. One of the more important items was the development of a two-in-one, gas-solids separator that removed 95+ percent of the solids in 1600 F to 1900 F gases. Another was a coal and sorbent feed and mixing system for injecting the fuel into the pressurized combustor. Also important were the controls and data-acquisition systems that enabled one person to operate the entire facility. The solid, liquid, and gas sub-systems all had problems that were solved over the 2-year operating time of the facility, which culminated in a 400-hour, hot-gas, turbine test.

  17. Toward a Mechanistic Source Term in Advanced Reactors: Characterization of Radionuclide Transport and Retention in a Sodium Cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-04-17

    A vital component of the U.S. reactor licensing process is an integrated safety analysis in which a source term representing the release of radionuclides during normal operation and accident sequences is analyzed. Historically, source term analyses have utilized bounding, deterministic assumptions regarding radionuclide release. However, advancements in technical capabilities and the knowledge state have enabled the development of more realistic and best-estimate retention and release models such that a mechanistic source term assessment can be expected to be a required component of future licensing of advanced reactors. Recently, as part of a Regulatory Technology Development Plan effort for sodium cooled fast reactors (SFRs), Argonne National Laboratory has investigated the current state of knowledge of potential source terms in an SFR via an extensive review of previous domestic experiments, accidents, and operation. As part of this work, the significant sources and transport processes of radionuclides in an SFR have been identified and characterized. This effort examines all stages of release and source term evolution, beginning with release from the fuel pin and ending with retention in containment. Radionuclide sources considered in this effort include releases originating both in-vessel (e.g. in-core fuel, primary sodium, cover gas cleanup system, etc.) and ex-vessel (e.g. spent fuel storage, handling, and movement). Releases resulting from a primary sodium fire are also considered as a potential source. For each release group, dominant transport phenomena are identified and qualitatively discussed. The key product of this effort was the development of concise, inclusive diagrams that illustrate the release and retention mechanisms at a high level, where unique schematics have been developed for in-vessel, ex-vessel and sodium fire releases. This review effort has also found that despite the substantial range of phenomena affecting radionuclide release, the

  18. Reactors and neutron-scattering instruments in Western Europe -an update on continuous neutron sources

    Science.gov (United States)

    Bauer, G. S.; Thamm, G.

    1991-10-01

    Research reactors as sources of continuous neutron beams have been operational in Western Europe since the late fifties and have triggered a series of successful developments in reactor technology and in particular in neutron-scattering instrumentation. The culmination point so far was the construction and operation of the high-flux reactor at the ILL in Grenoble. Although only one new reactor has been built since then, there were-and still are-efforts going on especially in Germany to provide a modern substitute for the older multi-purpose reactors still in operation or shut down already. Substantial effort also went into the modernization of the scattering instruments and experiment infrastructure (cold neutron sources, neutron guides, etc.) at several locations, showing that the scientific merits of neutron scattering are well recognized also on the political level. The ongoing construction of a cw-spallation neutron source in Switzerland might constitute an important step on the way to a new generation of high-flux sources.

  19. Representative Source Terms and the Influence of Reactor Attributes on Functional Containment in Modular High-Temperature Gas-Cooled Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Petti, D. A.; Hobbins, R. R.; Lowry, P.; Gougar, H.

    2013-11-01

    Modular high-temperature gas-cooled reactors (MHTGRs) offer a high degree of passive safety. The low power density of the reactor and the high heat capacity of the graphite core result in slow transients that do not challenge the integrity of the robust TRISO fuel. Another benefit of this fuel form and the surrounding graphite is their superior ability to retain fission products under all anticipated normal and off-normal conditions, which limits reactor accident source terms to very low values. In this paper, we develop estimates of the source term for a generic MHTGR to illustrate the performance of the radionuclide barriers that comprise the MHTGR functional containment. We also examine the influence of initial fuel quality, fuel performance/failure, reactor outlet temperature, and retention outside of the reactor core on the resultant source term to the environment.

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

    Energy Technology Data Exchange (ETDEWEB)

    Grabaskas, D.; Brunett, A.; Passerini, S.; Grelle, A.; Bucknor, M.

    2017-06-26

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

  1. Combustion flame-plasma hybrid reactor systems, and chemical reactant sources

    Science.gov (United States)

    Kong, Peter C

    2013-11-26

    Combustion flame-plasma hybrid reactor systems, chemical reactant sources, and related methods are disclosed. In one embodiment, a combustion flame-plasma hybrid reactor system comprising a reaction chamber, a combustion torch positioned to direct a flame into the reaction chamber, and one or more reactant feed assemblies configured to electrically energize at least one electrically conductive solid reactant structure to form a plasma and feed each electrically conductive solid reactant structure into the plasma to form at least one product is disclosed. In an additional embodiment, a chemical reactant source for a combustion flame-plasma hybrid reactor comprising an elongated electrically conductive reactant structure consisting essentially of at least one chemical reactant is disclosed. In further embodiments, methods of forming a chemical reactant source and methods of chemically converting at least one reactant into at least one product are disclosed.

  2. UCN sources at external beams of thermal neutrons. An example of PIK reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lychagin, E.V., E-mail: lychag@nf.jinr.ru [Joint Institute for Nuclear Research, 6 Joliot-Curie, Dubna 141980 (Russian Federation); Mityukhlyaev, V.A., E-mail: victim@pnpi.spb.ru [Petersburg Nuclear Physics Institute, Orlova Roscha, Gatchina 188300 (Russian Federation); Muzychka, A.Yu., E-mail: muz@nf.jinr.ru [Joint Institute for Nuclear Research, 6 Joliot-Curie, Dubna 141980 (Russian Federation); Nekhaev, G.V., E-mail: grigorijnekhaev@yandex.ru [Joint Institute for Nuclear Research, 6 Joliot-Curie, Dubna 141980 (Russian Federation); Nesvizhevsky, V.V., E-mail: nesvizhevsky@ill.eu [Institut Max von Laue – Paul Langevin, 71 Avenue des Martyrs, Grenoble 38042 (France); Onegin, M.S., E-mail: oneginm@gmail.com [Petersburg Nuclear Physics Institute, Orlova Roscha, Gatchina 188300 (Russian Federation); Sharapov, E.I., E-mail: sharapov@nf.jinr.ru [Joint Institute for Nuclear Research, 6 Joliot-Curie, Dubna 141980 (Russian Federation); Strelkov, A.V., E-mail: str@jinr.ru [Joint Institute for Nuclear Research, 6 Joliot-Curie, Dubna 141980 (Russian Federation)

    2016-07-01

    We consider ultracold neutron (UCN) sources based on a new method of UCN production in superfluid helium ({sup 4}He). The PIK reactor is chosen as a perspective example of application of this idea, which consists of installing {sup 4}He UCN source in the beam of thermal or cold neutrons and surrounding the source with moderator-reflector, which plays the role of cold neutron (CN) source feeding the UCN source. CN flux in the source can be several times larger than the incident flux, due to multiple neutron reflections from the moderator–reflector. We show that such a source at the PIK reactor would provide an order of magnitude larger density and production rate than an analogous source at the ILL reactor. We estimate parameters of {sup 4}He source with solid methane (CH{sub 4}) or/and liquid deuterium (D{sub 2}) moderator–reflector. We show that such a source with CH{sub 4} moderator–reflector at the PIK reactor would provide the UCN density of ~1·10{sup 5} cm{sup −3}, and the UCN production rate of ~2·10{sup 7} s{sup −1}. These values are respectively 1000 and 20 times larger than those for the most intense UCN user source. The UCN density in a source with D{sub 2} moderator-reflector would reach the value of ~2·10{sup 5} cm{sup −3}, and the UCN production rate would be equal ~8·10{sup 7} s{sup −1}. Installation of such a source in a beam of CNs would slightly increase the density and production rate.

  3. Declassification of radioactive water from a pool type reactor after nuclear facility dismantling

    Science.gov (United States)

    Arnal, J. M.; Sancho, M.; García-Fayos, B.; Verdú, G.; Serrano, C.; Ruiz-Martínez, J. T.

    2017-09-01

    This work is aimed to the treatment of the radioactive water from a dismantled nuclear facility with an experimental pool type reactor. The main objective of the treatment is to declassify the maximum volume of water and thus decrease the volume of radioactive liquid waste to be managed. In a preliminary stage, simulation of treatment by the combination of reverse osmosis (RO) and evaporation have been performed. Predicted results showed that the combination of membrane and evaporation technologies would result in a volume reduction factor higher than 600. The estimated time to complete the treatment was around 650 h (25-30 days). For different economical and organizational reasons which are explained in this paper, the final treatment of the real waste had to be reduced and only evaporation was applied. The volume reduction factor achieved in the real treatment was around 170, and the time spent for treatment was 194 days.

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

    Energy Technology Data Exchange (ETDEWEB)

    Steele, Robert; Mata, Angelica; Dunzik-Gougar, Mary Lou; van Rooyen, Isabella

    2016-06-01

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

  5. Design of sample carrier for neutron irradiation facility at TRIGA MARK II nuclear reactor

    Science.gov (United States)

    Abdullah, Y.; Hamid, N. A.; Mansor, M. A.; Ahmad, M. H. A. R. M.; Yusof, M. R.; Yazid, H.; Mohamed, A. A.

    2013-06-01

    The objective of this work is to design a sample carrier for neutron irradiation experiment at beam ports of research nuclear reactor, the Reaktor TRIGA PUSPATI (RTP). The sample carrier was designed so that irradiation experiment can be performed safely by researchers. This development will resolve the transferring of sample issues faced by the researchers at the facility when performing neutron irradiation studies. The function of sample carrier is to ensure the sample for the irradiation process can be transferred into and out from the beam port of the reactor safely and effectively. The design model used was House of Quality Method (HOQ) which is usually used for developing specifications for product and develop numerical target to work towards and determining how well we can meet up to the needs. The chosen sample carrier (product) consists of cylindrical casing shape with hydraulic cylinders transportation method. The sample placing can be done manually, locomotion was by wheel while shielding used was made of boron materials. The sample carrier design can shield thermal neutron during irradiation of sample so that only low fluencies fast neutron irradiates the sample.

  6. A study on source term assessment and waste disposal requirement of decontamination and decommissioning for the TRIGA research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Whang, Joo Ho; Lee, Kyung JIn; Lee, Jae Min; Choi, Gyu Seup; Shin, Byoung Sun [Kyunghee Univ., Seoul (Korea, Republic of)

    1999-08-15

    The objective and necessity of the project : TRIGA is the first nuclear facility that decide to decommission and decontamination in our nation. As we estimate the expected life of nuclear power generation at 30 or 40 years, the decommissioning business should be conducted around 2010, and the development of regulatory technique supporting it should be developed previously. From a view of decommissioning and decontamination, the research reactor is just small in scale but it include all decommissioning and decontamination conditions. So, the rules by regulatory authority with decommissioning will be a guide for nuclear power plant in the future. The basis of regulatory technique required when decommissioning the research reactor are the radiological safety security and the data for it. The source term is very important condition not only for security of worker but for evaluating how we dispose the waste is appropriate for conducting the middle store and the procedure after it when the final disposal is considered. The content and the scope in this report contain the procedure of conducting the assessment of the source term which is most important in understanding the general concept of the decommissioning procedure of the decommissioning and decontamination of TRIGA research reactor. That is, the sampling and measuring method is presented as how to measure the volume of the radioactivity of the nuclear facilities. And also, the criterion of classifying the waste occurred in other countries and the site release criteria which is the final step of decommissioning and decontamination presented through MARSSIM. Finally, the program to be applicable through comparing the methods of our nation and other countries ones is presented as plan for disposal of the waste in the decommissioning.

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

    Energy Technology Data Exchange (ETDEWEB)

    Erika N. Bailey

    2011-10-10

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

  8. A Mechanistic Source Term Calculation for a Metal Fuel Sodium Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Grabaskas, David; Bucknor, Matthew; Jerden, James

    2017-06-26

    A mechanistic source term (MST) calculation attempts to realistically assess the transport and release of radionuclides from a reactor system to the environment during a specific accident sequence. The U.S. Nuclear Regulatory Commission (NRC) has repeatedly stated its expectation that advanced reactor vendors will utilize an MST during the U.S. reactor licensing process. As part of a project to examine possible impediments to sodium fast reactor (SFR) licensing in the U.S., an analysis was conducted regarding the current capabilities to perform an MST for a metal fuel SFR. The purpose of the project was to identify and prioritize any gaps in current computational tools, and the associated database, for the accurate assessment of an MST. The results of the study demonstrate that an SFR MST is possible with current tools and data, but several gaps exist that may lead to possibly unacceptable levels of uncertainty, depending on the goals of the MST analysis.

  9. Measuring and monitoring KIPT Neutron Source Facility Reactivity

    Energy Technology Data Exchange (ETDEWEB)

    Cao, Yan [Argonne National Lab. (ANL), Argonne, IL (United States); Gohar, Yousry [Argonne National Lab. (ANL), Argonne, IL (United States); Zhong, Zhaopeng [Argonne National Lab. (ANL), Argonne, IL (United States)

    2015-08-01

    Argonne National Laboratory (ANL) of USA and Kharkov Institute of Physics and Technology (KIPT) of Ukraine have been collaborating on developing and constructing a neutron source facility at Kharkov, Ukraine. The facility consists of an accelerator-driven subcritical system. The accelerator has a 100 kW electron beam using 100 MeV electrons. The subcritical assembly has keff less than 0.98. To ensure the safe operation of this neutron source facility, the reactivity of the subcritical core has to be accurately determined and continuously monitored. A technique which combines the area-ratio method and the flux-to-current ratio method is purposed to determine the reactivity of the KIPT subcritical assembly at various conditions. In particular, the area-ratio method can determine the absolute reactivity of the subcritical assembly in units of dollars by performing pulsed-neutron experiments. It provides reference reactivities for the flux-to-current ratio method to track and monitor the reactivity deviations from the reference state while the facility is at other operation modes. Monte Carlo simulations are performed to simulate both methods using the numerical model of the KIPT subcritical assembly. It is found that the reactivities obtained from both the area-ratio method and the flux-to-current ratio method are spatially dependent on the neutron detector locations and types. Numerical simulations also suggest optimal neutron detector locations to minimize the spatial effects in the flux-to-current ratio method. The spatial correction factors are calculated using Monte Carlo methods for both measuring methods at the selected neutron detector locations. Monte Carlo simulations are also performed to verify the accuracy of the flux-to-current ratio method in monitoring the reactivity swing during a fuel burnup cycle.

  10. Investigation of Isfahan miniature neutron source reactor (MNSR for boron neutron capture therapy by MCNP simulation

    Directory of Open Access Journals (Sweden)

    S.Z Kalantari

    2015-01-01

    Full Text Available One of the important neutron sources for Boron Neutron Capture Therapy (BNCT is a nuclear reactor. It needs a high flux of epithermal neutrons. The optimum conditions of the neutron spectra for BNCT are provided by the International Atomic Energy Agency (IAEA. In this paper, Miniature Neutron Source Reactor (MNSR as a neutron source for BNCT was investigated. For this purpose, we designed a Beam Shaping Assembly (BSA for the reactor and the neutron transport from the core of the reactor to the output windows of BSA was simulated by MCNPX code. To optimize the BSA performance, two sets of parameters should be evaluated, in-air and in-phantom parameters. For evaluating in-phantom parameters, a Snyder head phantom was used and biological dose rate and dose-depth curve were calculated in brain normal and tumor tissues. Our calculations showed that the neutron flux of the MNSR reactor can be used for BNCT, and the designed BSA in optimum conditions had a good therapeutic characteristic for BNCT.

  11. Scaling Studies for High Temperature Test Facility and Modular High Temperature Gas-Cooled Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Richard R. Schult; Paul D. Bayless; Richard W. Johnson; James R. Wolf; Brian Woods

    2012-02-01

    The Oregon State University (OSU) High Temperature Test Facility (HTTF) is an integral experimental facility that will be constructed on the OSU campus in Corvallis, Oregon. The HTTF project was initiated, by the U.S. Nuclear Regulatory Commission (NRC), on September 5, 2008 as Task 4 of the 5-year High Temperature Gas Reactor Cooperative Agreement via NRC Contract 04-08-138. Until August, 2010, when a DOE contract was initiated to fund additional capabilities for the HTTF project, all of the funding support for the HTTF was provided by the NRC via their cooperative agreement. The U.S. Department of Energy (DOE) began their involvement with the HTTF project in late 2009 via the Next Generation Nuclear Plant (NGNP) project. Because the NRC's interests in HTTF experiments were only centered on the depressurized conduction cooldown (DCC) scenario, NGNP involvement focused on expanding the experimental envelope of the HTTF to include steady-state operations and also the pressurized conduction cooldown (PCC).

  12. Calculation of fast neutron flux in reactor pressure tubes and experimental facilities

    Energy Technology Data Exchange (ETDEWEB)

    Barnett, P.C. [Canadian General Electric (Canada)

    1968-07-15

    The computer program EPITHET was used to calculate the fast neutron flux (>1 MeV) in several reactor pressure tubes and experimental facilities in order to compare the fast neutron flux in the different cases and to provide a self-consistent set of flux values which may be used to relate creep strain to fast neutron flux . The facilities considered are shown below together with the calculated fast neutron flux (>1 MeV). Fast flux 10{sup 13} n/cm{sup 2}s: NPD 1.14, Douglas Point 2.66, Pickering 2.89, Gentilly 2.35, SGHWR 3.65, NRU U-1 and U-2 3.25'' pressure tube - 19 element fuel 3.05, NRU U-1 and U-2 4.07'' pressure tube - 28 element fuel 3.18, NRU U-1 and U-2 4.07'' pressure tube - 18 element fuel 2.90, NRX X-5 0.88, PRTR Mk I fuel 2.81, PRTR HPD fuel 3.52, WR-1 2.73, Mk IV creep machine (NRX) 0.85, Mk VI creep machine (NRU) 2.04, Biaxial creep insert (NRU U-49) 2.61.

  13. Location of an electric source facility and local area promotion

    Energy Technology Data Exchange (ETDEWEB)

    Shimohirao, Isao [Fukushima Univ. (Japan)

    1999-10-01

    Here were described on energy demand and supply, energy policy and local area promotion policy for basic problems important on location of electric source facilities. At present, co-existence business between electricity business and electric source location area is lacking in its activity. It seems to be necessary to enforce some systems to intend to promote it earnestly, and to effort to promote industry promotions such as introduction of some national projects, induction of electricity cost reduction for a means of business invitation, and so forth. And it is necessary to promote them under cooperations with electricity businesses, governments, universities and communities for the industrial promotion and fixation of the youth at local areas. In order to realize such necessities, further larger efforts are expected for national and local governments. (G.K.)

  14. In-Source Laser Resonance Ionization at ISOL Facilities

    CERN Document Server

    Marsh, Bruce; Feddosseev, Valentin

    Resonance ionization laser ion source development has been carried out at two radioactive ion beam facilities: ISOLDE (CERN, Switzerland) and the IGISOL facility (Jyvaskyla, Finland). The scope of the Resonance Ionization Laser Ion Source has been extended to 27 elements with the development of new three-step ionization schemes for Sb, Sc, Dy, Y and Au. The efficiencies were determined to be in the range of 2 - 20 %. Additionally, a new two-step ionization scheme has been developed for bismuth in an off-line atomic beam unit. The scheme relies on ionization via a strong and broad auto-ionizing resonance at an energy of 63196.79 cm$^{−1}$. This scheme may offer an improvement over the existing RILIS efficiency and will be more convenient for use during resonance ionization spectroscopy of Bi isotopes. The RILIS can be used as a spectroscopic tool to probe features such as the hyperfine structures and the isotope-shifts of radioisotopes with low production rates. By coupling a laser scanning process that dire...

  15. X-ray Optics for BES Light Source Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Mills, Dennis [Argonne National Lab. (ANL), Argonne, IL (United States); Padmore, Howard [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Lessner, Eliane [Dept. of Energy (DOE), Washington DC (United States). Office of Science

    2013-03-27

    potentially revolutionary science involves soft excitations such as magnons and phonons; in general, these are well below the resolution that can be probed by today’s optical systems. The study of these low-energy excitations will only move forward if advances are made in high-resolution gratings for the soft X-ray energy region, and higher-resolution crystal analyzers for the hard X-ray region. In almost all the forefront areas of X-ray science today, the main limitation is our ability to focus, monochromate, and manipulate X-rays at the level required for these advanced measurements. To address these issues, the U.S. Department of Energy (DOE) Office of Basic Energy Sciences (BES) sponsored a workshop, X-ray Optics for BES Light Source Facilities, which was held March 27–29, 2013, near Washington, D.C. The workshop addressed a wide range of technical and organizational issues. Eleven working groups were formed in advance of the meeting and sought over several months to define the most pressing problems and emerging opportunities and to propose the best routes forward for a focused R&D program to solve these problems. The workshop participants identified eight principal research directions (PRDs), as follows: Development of advanced grating lithography and manufacturing for high-energy resolution techniques such as soft X-ray inelastic scattering. Development of higher-precision mirrors for brightness preservation through the use of advanced metrology in manufacturing, improvements in manufacturing techniques, and in mechanical mounting and cooling. Development of higher-accuracy optical metrology that can be used in manufacturing, verification, and testing of optomechanical systems, as well as at wavelength metrology that can be used for quantification of individual optics and alignment and testing of beamlines. Development of an integrated optical modeling and design framework that is designed and maintained specifically for X-ray optics. Development of

  16. Monte Carlo simulation of NSE at reactor and spallation sources

    Energy Technology Data Exchange (ETDEWEB)

    Zsigmond, G.; Wechsler, D.; Mezei, F. [Hahn-Meitner-Institut Berlin, Berlin (Germany)

    2001-03-01

    A MC (Monte Carlo) computation study of NSE (Neutron Spin Echo) has been performed by means of VITESS investigating the classic and TOF-NSE options at spallation sources. The use of white beams in TOF-NSE makes the flipper efficiency in function of the neutron wavelength an important issue. The emphasis was put on exact evaluation of flipper efficiencies for wide wavelength-band instruments. (author)

  17. Modeling of Heat Transfer in the Helical-Coil Heat Exchanger for the Reactor Facility "UNITERM"

    Directory of Open Access Journals (Sweden)

    V. I. Solonin

    2014-01-01

    Full Text Available Circuit heat sink plays an important role in the reactor system. Therefore it imposes high requirements for quality of determining thermal-hydraulic parameters. This article is aimed at modeling of heat exchange process of the helical-coil heat exchanger, which is part of the heat sink circuit of the reactor facility "UNITERM."The simulation was performed using hydro-gas-dynamic software package ANSYS CFX. Computational fluid dynamics of this package allows us to perform calculations in a threedimensional setting, giving an idea of the fluid flow nature. The purpose of the simulation was to determine the parameters of the helical-coil heat exchanger (temperature, velocity at the outlet of the pipe and inter-tubular space, pressure drop, and the nature of the fluid flow of primary and intermediate coolants. Geometric parameters of the model were determined using the preliminary calculations performed by the criterion equations. In calculations Turbulence models k-ε RNG, Shear Stress Transport (SST are used. The article describes selected turbulence models, and considers relationship with wall function.The calculation results allow us to give the values obtained for thermal-hydraulic parameters, to compare selected turbulence models, as well as to show distribution patterns of the coolant temperature, pressure, and velocity at the outlet of the intermediate cooler.Calculations have shown that:- maximum values of primary coolant temperature at the outlet of the heat exchanger surface are encountered in the space between the helical-coil tubes;- higher temperatures of intermediate coolant at the outlet of the coils (in space of helicalcoil tubes are observed for the peripheral row;- primary coolant movement in the inter-tubular space of helical-coil surface is formed as a spiral flow, rather than as a in-line tube bank cross flow.

  18. Preliminary Analysis of the Transient Reactor Test Facility (TREAT) with PROTEUS

    Energy Technology Data Exchange (ETDEWEB)

    Connaway, H. M. [Argonne National Lab. (ANL), Argonne, IL (United States); Lee, C. H. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2015-11-30

    The neutron transport code PROTEUS has been used to perform preliminary simulations of the Transient Reactor Test Facility (TREAT). TREAT is an experimental reactor designed for the testing of nuclear fuels and other materials under transient conditions. It operated from 1959 to 1994, when it was placed on non-operational standby. The restart of TREAT to support the U.S. Department of Energy’s resumption of transient testing is currently underway. Both single assembly and assembly-homogenized full core models have been evaluated. Simulations were performed using a historic set of WIMS-ANL-generated cross-sections as well as a new set of Serpent-generated cross-sections. To support this work, further analyses were also performed using additional codes in order to investigate particular aspects of TREAT modeling. DIF3D and the Monte-Carlo codes MCNP and Serpent were utilized in these studies. MCNP and Serpent were used to evaluate the effect of geometry homogenization on the simulation results and to support code-to-code comparisons. New meshes for the PROTEUS simulations were created using the CUBIT toolkit, with additional meshes generated via conversion of selected DIF3D models to support code-to-code verifications. All current analyses have focused on code-to-code verifications, with additional verification and validation studies planned. The analysis of TREAT with PROTEUS-SN is an ongoing project. This report documents the studies that have been performed thus far, and highlights key challenges to address in future work.

  19. Creep-Fatigue Damage Evaluation of a Model Reactor Vessel and Reactor Internals of Sodium Test Facility according to ASME-NH and RCC-MRx Codes

    Energy Technology Data Exchange (ETDEWEB)

    Lim, Dong-Won; Lee, Hyeong-Yeon; Eoh, Jae-Hyuk; Son, Seok-Kwon; Kim, Jong-Bum; Jeong, Ji-Young [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    The objective of the STELLA-2 is to support the specific design approval for PGSFR by synthetic reviews of key safety issues and code validations through the integral effect tests. Due to its high temperature operation in SFRs (and in a testing facility) up to 550 °C, thermally induced creep-fatigue damage is very likely in components including a reactor vessel, reactor internals (interior structures), heat exchangers, pipelines, etc. In this study, structural integrity of the components such as reactor vessel and internals in STELLA-2 has been evaluated against creep-fatigue failures at a concept-design step. As 2D analysis yields far conservative results, a realistic 3D simulation is performed by a commercial software. A design integrity guarding against a creep-fatigue damage failure operating at high temperature was evaluated for the reactor vessel with its internal structure of the STELLA-2. Both the high temperature design codes were used for the evaluation, and results were compared. All the results showed the vessel as a whole is safely designed at the given operating conditions, while the ASME-NH gives a conservative evaluation.

  20. Structural biology facilities at Brookhaven National Laboratory`s high flux beam reactor

    Energy Technology Data Exchange (ETDEWEB)

    Korszun, Z.R.; Saxena, A.M.; Schneider, D.K. [Brookhaven National Laboratory, Upton, NY (United States)

    1994-12-31

    The techniques for determining the structure of biological molecules and larger biological assemblies depend on the extent of order in the particular system. At the High Flux Beam Reactor at the Brookhaven National Laboratory, the Biology Department operates three beam lines dedicated to biological structure studies. These beam lines span the resolution range from approximately 700{Angstrom} to approximately 1.5{Angstrom} and are designed to perform structural studies on a wide range of biological systems. Beam line H3A is dedicated to single crystal diffraction studies of macromolecules, while beam line H3B is designed to study diffraction from partially ordered systems such as biological membranes. Beam line H9B is located on the cold source and is designed for small angle scattering experiments on oligomeric biological systems.

  1. Completion Summary for Well NRF-16 near the Naval Reactors Facility, Idaho National Laboratory, Idaho

    Science.gov (United States)

    Twining, Brian V.; Fisher, Jason C.; Bartholomay, Roy C.

    2010-01-01

    In 2009, the U.S. Geological Survey in cooperation with the U.S. Department of Energy's Naval Reactors Laboratory Field Office, Idaho Branch Office cored and completed well NRF-16 for monitoring the eastern Snake River Plain (SRP) aquifer. The borehole was initially cored to a depth of 425 feet below land surface and water samples and geophysical data were collected and analyzed to determine if well NRF-16 would meet criteria requested by Naval Reactors Facility (NRF) for a new upgradient well. Final construction continued after initial water samples and geophysical data indicated that NRF-16 would produce chemical concentrations representative of upgradient aquifer water not influenced by NRF facility disposal, and that the well was capable of producing sustainable discharge for ongoing monitoring. The borehole was reamed and constructed as a Comprehensive Environmental Response Compensation and Liability Act monitoring well complete with screen and dedicated pump. Geophysical and borehole video logs were collected after coring and final completion of the monitoring well. Geophysical logs were examined in conjunction with the borehole core to identify primary flow paths for groundwater, which are believed to occur in the intervals of fractured and vesicular basalt and to describe borehole lithology in detail. Geophysical data also were examined to look for evidence of perched water and the extent of the annular seal after cement grouting the casing in place. Borehole videos were collected to confirm that no perched water was present and to examine the borehole before and after setting the screen in well NRF-16. Two consecutive single-well aquifer tests to define hydraulic characteristics for well NRF-16 were conducted in the eastern SRP aquifer. Transmissivity and hydraulic conductivity averaged from the aquifer tests were 4.8 x 103 ft2/d and 9.9 ft/d, respectively. The transmissivity for well NRF-16 was within the range of values determined from past aquifer

  2. IDLING AND NATURAL POWER TRANSFER LOSS SAVING BY MEANS OF SHUNT REACTORS AND REACTIVE POWER SOURCES

    Directory of Open Access Journals (Sweden)

    Patsiuk V.I

    2009-04-01

    Full Text Available The closed formulas for definition of the steady–state values of voltages, currents, active and reactive power in a line with the distributed and lumped constants are shown. The influence of the shunt reactors and reactive power sources in the form of capacitor banks on losses of idling with various wave lengths is investigated. For the half-wave transmissions line the optimal parameters (which allow increasing of the output during the natural-power transfer of the shunt reactors were obtained.

  3. Reactor units for power supply to the Russian Arctic regions: Priority assessment of nuclear energy sources

    Directory of Open Access Journals (Sweden)

    Mel'nikov N. N.

    2017-03-01

    Full Text Available Under conditions of competitiveness of small nuclear power plants (SNPP and feasibility of their use to supply power to remote and inaccessible regions the competition occurs between nuclear energy sources, which is caused by a wide range of proposals for solving the problem of power supply to different consumers in the decentralized area of the Russian Arctic power complex. The paper suggests a methodological approach for expert assessment of the priority of small power reactor units based on the application of the point system. The priority types of the reactor units have been determined based on evaluation of the unit's conformity to the following criteria: the level of referentiality and readiness degree of reactor units to implementation; duration of the fuel cycle, which largely determines an autonomy level of the nuclear energy source; the possibility of creating a modular block structure of SNPP; the maximum weight of a transported single equipment for the reactor unit; service life of the main equipment. Within the proposed methodological approach the authors have performed a preliminary ranking of the reactor units according to various criteria, which allows quantitatively determining relative difference and priority of the small nuclear power plants projects aimed at energy supply to the Russian Arctic. To assess the sensitivity of the ranking results to the parameters of the point system the authors have observed the five-point and ten-point scales under variations of importance (weights of different criteria. The paper presents the results of preliminary ranking, which have allowed distinguishing the following types of the reactor units in order of their priority: ABV-6E (ABV-6M, "Uniterm" and SVBR-10 in the energy range up to 20 MW; RITM-200 (RITM-200M, KLT-40S and SVBR-100 in the energy range above 20 MW.

  4. Simulation of Collimator for Neutron Imaging Facility of TRIGA MARK II PUSPATI Reactor

    Science.gov (United States)

    Zin, Muhammad Rawi Mohamed; Jamro, Rafhayudi; Yazid, Khairiah; Hussain, Hishamuddin; Yazid, Hafizal; Ahmad, Megat Harun Al Rashid Megat; Azman, Azraf; Mohamad, Glam Hadzir Patai; Hamzah, Nai'im Syaugi; Abu, Mohamad Puad

    Neutron Radiography facility in TRIGA MARK II PUSPATI reactor is being upgraded to obtain better image resolution as well as reducing exposure time. Collimator and exposure room are the main components have been designed for fabrication. This article focuses on the simulation part that was carried out to obtain the profile of collimated neutron beam by utilizing the neutron transport protocol code in the Monte Carlo N-Particle (MCNP) software. Particular interest is in the selection of materials for inlet section of the collimator. Results from the simulation indicates that a combination of Bismuth and Sapphire, each of which has 5.0 cm length that can significantly filter both the gamma radiation and the fast neutrons. An aperture made of Cadmium with 1.0 cm opening diameter provides thermal neutron flux about 1.8 x108 ncm-2s-1 at the inlet, but reduces to 2.7 x106 ncm-2s-1 at the sample plane. Still the flux obtained is expected to reduces exposure time as well as gaining better image resolution.

  5. Plutonium Consumption Program, CANDU Reactor Project: Feasibility of BNFP Site as MOX Fuel Supply Facility. Final report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-06-30

    An evaluation was made of the technical feasibility, cost, and schedule for converting the existing unused Barnwell Nuclear Fuel Facility (BNFP) into a Mixed Oxide (MOX) CANDU fuel fabrication plant for disposition of excess weapons plutonium. This MOX fuel would be transported to Ontario where it would generate electricity in the Bruce CANDU reactors. Because CANDU MOX fuel operates at lower thermal load than natural uranium fuel, the MOX program can be licensed by AECB within 4.5 years, and actual Pu disposition in the Bruce reactors can begin in 2001. Ontario Hydro will have to be involved in the entire program. Cost is compared between BNFP and FMEF at Hanford for converting to a CANDU MOX facility.

  6. Regulatory Technology Development Plan - Sodium Fast Reactor: Mechanistic Source Term – Trial Calculation

    Energy Technology Data Exchange (ETDEWEB)

    Grabaskas, David [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Bucknor, Matthew [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Jerden, James [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Brunett, Acacia J. [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Denman, Matthew [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States). Nuclear Engineering Division; Clark, Andrew [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States). Nuclear Engineering Division; Denning, Richard S. [Consultant, Columbus, OH (United States)

    2016-10-01

    The potential release of radioactive material during a plant incident, referred to as the source term, is a vital design metric and will be a major focus of advanced reactor licensing. The U.S. Nuclear Regulatory Commission has stated an expectation for advanced reactor vendors to present a mechanistic assessment of the potential source term in their license applications. The mechanistic source term presents an opportunity for vendors to realistically assess the radiological consequences of an incident, and may allow reduced emergency planning zones and smaller plant sites. However, the development of a mechanistic source term for advanced reactors is not without challenges, as there are often numerous phenomena impacting the transportation and retention of radionuclides. This project sought to evaluate U.S. capabilities regarding the mechanistic assessment of radionuclide release from core damage incidents at metal fueled, pool-type sodium fast reactors (SFRs). The purpose of the analysis was to identify, and prioritize, any gaps regarding computational tools or data necessary for the modeling of radionuclide transport and retention phenomena. To accomplish this task, a parallel-path analysis approach was utilized. One path, led by Argonne and Sandia National Laboratories, sought to perform a mechanistic source term assessment using available codes, data, and models, with the goal to identify gaps in the current knowledge base. The second path, performed by an independent contractor, performed sensitivity analyses to determine the importance of particular radionuclides and transport phenomena in regards to offsite consequences. The results of the two pathways were combined to prioritize gaps in current capabilities.

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

    Science.gov (United States)

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

    2002-01-01

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

  8. Recovery of spent high intensity neutron sources in Atalante Facility

    Energy Technology Data Exchange (ETDEWEB)

    Bros, P.; Boyer Deslys, V.; Millet, A.; Solinhac, I.; Donnet, L.; Maillard, C.; Paillard, S.; Ranchoux, M.

    2005-01-01

    The Atalante facility is required by law to recover both neutron and gamma sources with activity levels exceeding 300 mCi. Most of the neutron sources consist of mixtures of alpha-emitters (238Pu, 239Pu, 241Am or 244Cm) and beryllium. Several processes now under consideration are based on routine chemical separation techniques (selective precipitation, extraction chromatography, ion exchange). The treatment produces an actinide oxide (which is used later for R and D studies) and solid beryllium nitrate, which is considered as a waste and transferred to a surface interim storage site if the overall activity of the package after 300 years is less than 50 MBq (ANDRA specifications). The Material Analysis and Metrology Laboratory of Atalante validate the residual alpha activity in the waste. The techniques used include alpha spectrometry and L-line X-ray fluorescence for alpha emitters, and plasma torch measurements (ICP-AES and ICP-MS) for beryllium analysis. Specific equipment for transport (B type cask), storage and treatment (hot shielded cells) are used for this activity. (Author)

  9. Electron Accelerator Shielding Design of KIPT Neutron Source Facility

    Directory of Open Access Journals (Sweden)

    Zhaopeng Zhong

    2016-06-01

    Full Text Available The Argonne National Laboratory of the United States and the Kharkov Institute of Physics and Technology of the Ukraine have been collaborating on the design, development and construction of a neutron source facility at Kharkov Institute of Physics and Technology utilizing an electron-accelerator-driven subcritical assembly. The electron beam power is 100 kW using 100-MeV electrons. The facility was designed to perform basic and applied nuclear research, produce medical isotopes, and train nuclear specialists. The biological shield of the accelerator building was designed to reduce the biological dose to less than 5.0e-03 mSv/h during operation. The main source of the biological dose for the accelerator building is the photons and neutrons generated from different interactions of leaked electrons from the electron gun and the accelerator sections with the surrounding components and materials. The Monte Carlo N-particle extended code (MCNPX was used for the shielding calculations because of its capability to perform electron-, photon-, and neutron-coupled transport simulations. The photon dose was tallied using the MCNPX calculation, starting with the leaked electrons. However, it is difficult to accurately tally the neutron dose directly from the leaked electrons. The neutron yield per electron from the interactions with the surrounding components is very small, ∼0.01 neutron for 100-MeV electron and even smaller for lower-energy electrons. This causes difficulties for the Monte Carlo analyses and consumes tremendous computation resources for tallying the neutron dose outside the shield boundary with an acceptable accuracy. To avoid these difficulties, the SOURCE and TALLYX user subroutines of MCNPX were utilized for this study. The generated neutrons were banked, together with all related parameters, for a subsequent MCNPX calculation to obtain the neutron dose. The weight windows variance reduction technique was also utilized for both

  10. Nuclear-fuel-cycle risk assessment: descriptions of representative non-reactor facilities. Sections 1-14

    Energy Technology Data Exchange (ETDEWEB)

    Schneider, K.J.

    1982-09-01

    The Fuel Cycle Risk Assessment Program was initiated to provide risk assessment methods for assistance in the regulatory process for nuclear fuel cycle facilities other than reactors. This report, the first from the program, defines and describes fuel cycle elements that are being considered in the program. One type of facility (and in some cases two) is described that is representative of each element of the fuel cycle. The descriptions are based on real industrial-scale facilities that are current state-of-the-art, or on conceptual facilities where none now exist. Each representative fuel cycle facility is assumed to be located on the appropriate one of four hypothetical but representative sites described. The fuel cycles considered are for Light Water Reactors with once-through flow of spent fuel, and with plutonium and uranium recycle. Representative facilities for the following fuel cycle elements are described for uranium (or uranium plus plutonium where appropriate): mining, milling, conversion, enrichment, fuel fabrication, mixed-oxide fuel refabrication, fuel reprocessing, spent fuel storage, high-level waste storage, transuranic waste storage, spent fuel and high-level and transuranic waste disposal, low-level and intermediate-level waste disposal, and transportation. For each representative facility the description includes: mainline process, effluent processing and waste management, facility and hardware description, safety-related information and potential alternative concepts for that fuel cycle element. The emphasis of the descriptive material is on safety-related information. This includes: operating and maintenance requirements, input/output of major materials, identification and inventories of hazardous materials (particularly radioactive materials), unit operations involved, potential accident driving forces, containment and shielding, and degree of hands-on operation.

  11. Monitoring of dry anaerobic fermentation in experimental facility with use of biofilm reactor

    Directory of Open Access Journals (Sweden)

    Milan Šinkora

    2011-01-01

    Full Text Available Anaerobic fermentation is a process in which almost any organic mass may be transformed into an energetically rich biogas and a fermentation residue. Only strictly anaerobic microorganisms enter into the process; thus the process may take place only in a hermetically sealed environment. With regard to the world wide situation, where the increase in the proportion of energy from sustainable sources is in demand, anaerobic fermentation offers the possibility of transforming farm waste, farm products and municipality waste of biological character into electricity. This electricity may subsequently become an interesting source of income. The system may be proposed to agricultural companies as well as to municipality corporations. The process of fermentation may be carried out as dry fermentation or as liquid fermentation. Dry fermentation, working with materials where the percentage of dry matter exceeds 15 %, is the topic of this paper. This method has been frequently discussed as a method of processing organic material without waste water and thus the volume of material as well as the size of the biogas plant considerably decreases. To enable progress in the process, it is necessary to use a biologically active liquid solution containing the essential micro-organisms, often termed “percolate”. To activate a fresh substrate, fermented material adulterant containing cultivated microorganisms from previous processes is used; the ratio in which it is used is approximately one third to one fifth. “Percolate strategy” is another phrase used for sustaining the anaerobic fermentation; material is sprinkled on the percolate in the precisely defined cycles. In addition, the biologically active liquid solution contains organic substances washed out from the fermented material. With regard to its amount, this paper has become an impulse for the research in the amount of biogas which may be subsequently produced from the percolate in the so

  12. Probing light sterile neutrino signatures at reactor and Spallation Neutron Source neutrino experiments

    Science.gov (United States)

    Kosmas, T. S.; Papoulias, D. K.; Tórtola, M.; Valle, J. W. F.

    2017-09-01

    We investigate the impact of a fourth sterile neutrino at reactor and Spallation Neutron Source neutrino detectors. Specifically, we explore the discovery potential of the TEXONO and COHERENT experiments to subleading sterile neutrino effects through the measurement of the coherent elastic neutrino-nucleus scattering event rate. Our dedicated χ2-sensitivity analysis employs realistic nuclear structure calculations adequate for high purity sub-keV threshold Germanium detectors.

  13. Evaluation of the Initial Isothermal Physics Measurements at the Fast Flux Test Facility, a Prototypic Liquid Metal Fast Breeder Reactor

    Energy Technology Data Exchange (ETDEWEB)

    John D. Bess

    2010-03-01

    The Fast Flux Test Facility (FFTF) was a 400-MWt, sodium-cooled, low-pressure, high-temperature, fast-neutron flux, nuclear fission reactor plant designed for the irradiation testing of nuclear reactor fuels and materials for the development of liquid metal fast breeder reactors (LMFBRs). The FFTF was fueled with plutonium-uranium mixed oxide (MOX) and reflected by Inconel-600. Westinghouse Hanford Company operated the FFTF as part of the Hanford Engineering Development Laboratory (HEDL) for the U.S. Department of Energy on the Hanford Site near Richland, Washington. Although the FFTF was a testing facility not specifically designed to breed fuel or produce electricity, it did provide valuable information for LMFBR projects and base technology programs in the areas of plant system and component design, component fabrication, prototype testing, and site construction. The major objectives of the FFTF were to provide a strong, disciplined engineering base for the LMFBR program, provide fast flux testing for other U.S. programs, and contribute to the development of a viable self-sustaining competitive U.S. LMFBR industry. During its ten years of operation, the FFTF acted as a national research facility to test advanced nuclear fuels, materials, components, systems, nuclear power plant operating and maintenance procedures, and active and passive reactor safety technologies; it also produced a large number of isotopes for medical and industrial users, generated tritium for the U.S. fusion research program, and participated in cooperative, international research work. Prior to the implementation of the reactor characterization program, a series of isothermal physics measurements were performed; this acceptance testing program consisted of a series of control rod worths, critical rod positions, subcriticality measurements, maximum reactivity addition rates, shutdown margins, excess reactivity, and isothermal temperature coefficient reactivity. The results of these

  14. STAR: The Secure Transportable Autonomous Reactor System - Encapsulated Fission Heat Source

    Energy Technology Data Exchange (ETDEWEB)

    Ehud Greenspan

    2003-10-31

    OAK-B135 The Encapsulated Nuclear Heat Source (ENHS) is a novel 125 MWth fast spectrum reactor concept that was selected by the 1999 DOE NERI program as a candidate ''Generation-IV'' reactor. It uses Pb-Bi or other liquid-metal coolant and is intended to be factory manufactured in large numbers to be economically competitive. It is anticipated to be most useful to developing countries. The US team studying the feasibility of the ENHS reactor concept consisted of the University of California, Berkeley, Argonne National Laboratory (ANL), Lawrence Livermore National Laboratory (LLNL) and Westinghouse. Collaborating with the US team were three Korean organizations: Korean Atomic Energy Research Institute (KAERI), Korean Advanced Institute for Science and Technology (KAIST) and the University of Seoul, as well as the Central Research Institute of the Electrical Power Industry (CRIEPI) of Japan. Unique features of the ENHS include at least 20 years of operation without refueling; no fuel handling in the host country; no pumps and valves; excess reactivity does not exceed 1$; fully passive removal of the decay heat; very small probability of core damaging accidents; autonomous operation and capability of load-following over a wide range; very long plant life. In addition it offers a close match between demand and supply, large tolerance to human errors, is likely to get public acceptance via demonstration of superb safety, lack of need for offsite response, and very good proliferation resistance. The ENHS reactor is designed to meet the requirements of Generation IV reactors including sustainable energy supply, low waste, high level of proliferation resistance, high level of safety and reliability, acceptable risk to capital and, hopefully, also competitive busbar cost of electricity.

  15. Design and R&D Progress of China Lead-Based Reactor for ADS Research Facility

    Directory of Open Access Journals (Sweden)

    Yican Wu

    2016-03-01

    Full Text Available In 2011, the Chinese Academy of Sciences launched an engineering project to develop an accelerator-driven subcritical system (ADS for nuclear waste transmutation. The China Lead-based Reactor (CLEAR, proposed by the Institute of Nuclear Energy Safety Technology, was selected as the reference reactor for ADS development, as well as for the technology development of the Generation IV lead-cooled fast reactor. The conceptual design of CLEAR-I with 10 MW thermal power has been completed. KYLIN series lead-bismuth eutectic experimental loops have been constructed to investigate the technologies of the coolant, key components, structural materials, fuel assembly, operation, and control. In order to validate and test the key components and integrated operating technology of the lead-based reactor, the lead alloy-cooled non-nuclear reactor CLEAR-S, the lead-based zero-power nuclear reactor CLEAR-0, and the lead-based virtual reactor CLEAR-V are under realization.

  16. Review of nuclear data improvement needs for nuclear radiation measurement techniques used at the CEA experimental reactor facilities

    Directory of Open Access Journals (Sweden)

    Destouches Christophe

    2016-01-01

    Full Text Available The constant improvement of the neutron and gamma calculation codes used in experimental nuclear reactors goes hand in hand with that of the associated nuclear data libraries. The validation of these calculation schemes always requires the confrontation with integral experiments performed in experimental reactors to be completed. Nuclear data of interest, straight as cross sections, or elaborated ones such as reactivity, are always derived from a reaction rate measurement which is the only measurable parameter in a nuclear sensor. So, in order to derive physical parameters from the electric signal of the sensor, one needs specific nuclear data libraries. This paper presents successively the main features of the measurement techniques used in the CEA experimental reactor facilities for the on-line and offline neutron/gamma flux characterizations: reactor dosimetry, neutron flux measurements with miniature fission chambers and Self Power Neutron Detector (SPND and gamma flux measurements with chamber ionization and TLD. For each technique, the nuclear data necessary for their interpretation will be presented, the main identified needs for improvement identified and an analysis of their impact on the quality of the measurement. Finally, a synthesis of the study will be done.

  17. Investigation of TASS/SMR Capability to Predict a Natural Circulation in the Test Facility for an Integral Reactor

    Directory of Open Access Journals (Sweden)

    Young-Jong Chung

    2014-01-01

    Full Text Available System-integrated modular advanced reactor (SMART is a small-sized advanced integral type pressurized water reactor (PWR with a rated thermal power of 330 MW. It can produce 100 MW of electricity or 90 MW of electricity and 40,000 ton of desalinated water concurrently, which is sufficient for 100,000 residents. The design features contributing to safety enhancement are basically inherent safety improvement and passive safety features. TASS/SMR code was developed for an analysis of design based events and accidents in an integral type reactor reflecting the characteristics of the SMART design. The main purpose of the code is to analyze all relevant phenomena and processes. The code should be validated using experimental data in order to confirm prediction capability. TASS/SMR predicts well the overall thermal-hydraulic behavior under various natural circulation conditions at the experimental test facility for an integral reactor. A pressure loss should be provided a function of Reynolds number at low velocity conditions in order to simulate the mass flow rate well under natural circulations.

  18. Source term derivation and radiological safety analysis for the TRICO II research reactor in Kinshasa

    Energy Technology Data Exchange (ETDEWEB)

    Muswema, J.L., E-mail: jeremie.muswem@unikin.ac.cd [Faculty of Science, University of Kinshasa, P.O. Box 190, KIN XI (Congo, The Democratic Republic of the); Ekoko, G.B. [Faculty of Science, University of Kinshasa, P.O. Box 190, KIN XI (Congo, The Democratic Republic of the); Lukanda, V.M. [Faculty of Science, University of Kinshasa, P.O. Box 190, KIN XI (Congo, The Democratic Republic of the); Democratic Republic of the Congo' s General Atomic Energy Commission, P.O. Box AE1 (Congo, The Democratic Republic of the); Lobo, J.K.-K. [Faculty of Science, University of Kinshasa, P.O. Box 190, KIN XI (Congo, The Democratic Republic of the); Darko, E.O. [Radiation Protection Institute, Ghana Atomic Energy Commission, P.O. Box LG 80, Legon, Accra (Ghana); Boafo, E.K. [University of Ontario Institute of Technology, 2000 Simcoe St. North, Oshawa, ONL1 H7K4 (Canada)

    2015-01-15

    Highlights: • Atmospheric dispersion modeling for two credible accidents of the TRIGA Mark II research reactor in Kinshasa (TRICO II) was performed. • Radiological safety analysis after the postulated initiating events (PIE) was also carried out. • The Karlsruhe KORIGEN and the HotSpot Health Physics codes were used to achieve the objectives of this study. • All the values of effective dose obtained following the accident scenarios were below the regulatory limits for reactor staff members and the public, respectively. - Abstract: The source term from the 1 MW TRIGA Mark II research reactor core of the Democratic Republic of the Congo was derived in this study. An atmospheric dispersion modeling followed by radiation dose calculation were performed based on two possible postulated accident scenarios. This derivation was made from an inventory of peak radioisotope activities released in the core by using the Karlsruhe version of isotope generation code KORIGEN. The atmospheric dispersion modeling was performed with HotSpot code, and its application yielded to radiation dose profile around the site using meteorological parameters specific to the area under study. The two accident scenarios were picked from possible accident analyses for TRIGA and TRIGA-fueled reactors, involving the case of destruction of the fuel element with highest activity release and a plane crash on the reactor building as the worst case scenario. Deterministic effects of these scenarios are used to update the Safety Analysis Report (SAR) of the reactor, and for its current version, these scenarios are not yet incorporated. Site-specific meteorological conditions were collected from two meteorological stations: one installed within the Atomic Energy Commission and another at the National Meteorological Agency (METTELSAT), which is not far from the site. Results show that in both accident scenarios, radiation doses remain within the limits, far below the recommended maximum effective

  19. High-density ultracold neutron sources for the WWR-M and PIK reactors

    Energy Technology Data Exchange (ETDEWEB)

    Serebrov, A. P., E-mail: serebrov@pnpi.spb.ru; Fomin, A. K.; Kharitonov, A. G.; Lyamkin, V. A.; Prudnikov, D. V.; Ivanov, S. A.; Erykalov, A. N.; Onegin, M. S. [National Research Centre “Kurchatov Institute”, Petersburg Nuclear Physics Institute (Russian Federation); Gridnev, K. A. [St. Petersburg State University (Russian Federation)

    2016-01-15

    It is proposed to equip the PIK and WWR-M research reactors at the Petersburg Nuclear Physics Institute (PNPI) with high-density ultracold neutron (UCN) sources, where UCNs will be obtained based on the effect of their accumulation in superfluid helium (due to the specific features of this quantum fluid). The maximum UCN storage time in superfluid helium is obtained at temperatures on the order of 1 K. These sources are expected to yield UCN densities of 10{sup 3}–10{sup 4} cm{sup –3}, i.e., approximately three orders of magnitude higher than the density from existing UCN sources throughout the world. The development of highest intensity UCN sources will make PNPI an international center of fundamental UCN research.

  20. LMFBR source term experiments in the Fuel Aerosol Simulant Test (FAST) facility

    Energy Technology Data Exchange (ETDEWEB)

    Petrykowski, J.C.; Longest, A.W.

    1985-01-01

    The transport of uranium dioxide (UO/sub 2/) aerosol through liquid sodium was studied in a series of ten experiments in the Fuel Aerosol Simulant Test (FAST) facility at Oak Ridge National Laboratory (ORNL). The experiments were designed to provide a mechanistic basis for evaluating the radiological source term associated with a postulated, energetic core disruptive accident (CDA) in a liquid metal fast breeder reactor (LMFBR). Aerosol was generated by capacitor discharge vaporization of UO/sub 2/ pellets which were submerged in a sodium pool under an argon cover gas. Measurements of the pool and cover gas pressures were used to study the transport of aerosol contained by vapor bubbles within the pool. Samples of cover gas were filtered to determine the quantity of aerosol released from the pool. The depth at which the aerosol was generated was found to be the most critical parameter affecting release. The largest release was observed in the baseline experiment where the sample was vaporized above the sodium pool. In the nine ''undersodium'' experiments aerosol was generated beneath the surface of the pool at depths varying from 30 to 1060 mm. The mass of aerosol released from the pool was found to be a very small fraction of the original specimen. It appears that the bulk of aerosol was contained by bubbles which collapsed within the pool. 18 refs., 11 figs., 4 tabs.

  1. Topical report: Natural convection shutdown heat removal test facility (NSTF) evaluation for generating additional reactor cavity cooling system (RCCS) data.

    Energy Technology Data Exchange (ETDEWEB)

    Farmer, M. T.; Kilsdonk, D. J.; Tzanos, C.P.; Lomperski, S.; Aeschlimann, R.W.; Pointer, D.; Nuclear Engineering Division

    2005-09-01

    As part of the Department of Energy (DOE) Generation IV roadmapping activity, the Very High Temperature gas cooled Reactor (VHTR) has been selected as the principal concept for hydrogen production and other process-heat applications such as district heating and potable water production. On this basis, the DOE has selected the VHTR for additional R&D with the ultimate goal of demonstrating emission-free electricity and hydrogen production with this advanced reactor concept. One of the key passive safety features of the VHTR is the potential for decay heat removal by natural circulation of air in a Reactor Cavity Cooling System (RCCS). The air-cooled RCCS concept is notably similar to the Reactor Vessel Auxiliary Cooling System (RVACS) that was developed for the General Electric PRISM sodium-cooled fast reactor. As part of the DOE R&D program that supported the development of this fast reactor concept, the Natural Convection Shutdown Heat Removal Test Facility (NSTF) was developed at ANL to provide proof-of-concept data for the RVACS under prototypic natural convection flow, temperature, and heat flux conditions. Due to the similarity between RVACS and the RCCS, current VHTR R&D plans call for the utilization of the NSTF to provide RCCS model development and validation data, in addition to supporting design validation and optimization activities. Both air-cooled and water-cooled RCCS designs are to be included. In support of this effort, ANL has been tasked with the development of an engineering plan for mechanical and instrumentation modifications to NSTF to ensure that sufficiently detailed temperature, heat flux, velocity and turbulence profiles are obtained to adequately qualify the codes under the expected range of air-cooled RCCS flow conditions. Next year, similar work will be carried out for the alternative option of a water-cooled RCCS design. Analysis activities carried out in support of this experiment planning task have shown that: (a) in the RCCS, strong

  2. Effect of hypoiodous acid volatility on the iodine source term in reactor accidents

    Energy Technology Data Exchange (ETDEWEB)

    Routamo, T. [Imatran Voima Oy, Vantaa (Finland)

    1996-12-01

    A FORTRAN code ACT WATCH has been developed to establish an improved understanding of essential radionuclide behaviour mechanisms, especially related to iodine chemistry, in reactor accidents. The accident scenarios calculated in this paper are based on the Loss of Coolant accident at the Loviisa Nuclear Power Plant. The effect of different airborne species, especially HIO, on the iodine source term has been studied. The main cause of the high HIO release in the system modelled is the increase of I{sub 2} hydrolysis rate along with the temperature increase, which accelerates HIO production. Due to the high radiation level near the reactor core, I{sub 2} is produced from I{sup -}very rapidly. High temperature in the reactor coolant causes I{sub 2} to be transformed into HIO and through the boiling of the coolant volatile I{sub 2} and HIO are transferred efficiently into the gas phase. High filtration efficiency for particulate iodine causes I{sup -} release to be much lower than those of I{sub 2} and HIO. (author) 15 figs., 1 tab., refs.

  3. Regulatory Technology Development Plan - Sodium Fast Reactor. Mechanistic Source Term - Metal Fuel Radionuclide Release

    Energy Technology Data Exchange (ETDEWEB)

    Grabaskas, David [Argonne National Lab. (ANL), Argonne, IL (United States); Bucknor, Matthew [Argonne National Lab. (ANL), Argonne, IL (United States); Jerden, James [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-02-01

    The development of an accurate and defensible mechanistic source term will be vital for the future licensing efforts of metal fuel, pool-type sodium fast reactors. To assist in the creation of a comprehensive mechanistic source term, the current effort sought to estimate the release fraction of radionuclides from metal fuel pins to the primary sodium coolant during fuel pin failures at a variety of temperature conditions. These release estimates were based on the findings of an extensive literature search, which reviewed past experimentation and reactor fuel damage accidents. Data sources for each radionuclide of interest were reviewed to establish release fractions, along with possible release dependencies, and the corresponding uncertainty levels. Although the current knowledge base is substantial, and radionuclide release fractions were established for the elements deemed important for the determination of offsite consequences following a reactor accident, gaps were found pertaining to several radionuclides. First, there is uncertainty regarding the transport behavior of several radionuclides (iodine, barium, strontium, tellurium, and europium) during metal fuel irradiation to high burnup levels. The migration of these radionuclides within the fuel matrix and bond sodium region can greatly affect their release during pin failure incidents. Post-irradiation examination of existing high burnup metal fuel can likely resolve this knowledge gap. Second, data regarding the radionuclide release from molten high burnup metal fuel in sodium is sparse, which makes the assessment of radionuclide release from fuel melting accidents at high fuel burnup levels difficult. This gap could be addressed through fuel melting experimentation with samples from the existing high burnup metal fuel inventory.

  4. CHARACTERIZATION OF A PRECIPITATE REACTOR FEED TANK (PRFT) SAMPLE FROM THE DEFENSE WASTE PROCESSING FACILITY (DWPF)

    Energy Technology Data Exchange (ETDEWEB)

    Crawford, C.; Bannochie, C.

    2014-05-12

    A sample of from the Defense Waste Processing Facility (DWPF) Precipitate Reactor Feed Tank (PRFT) was pulled and sent to the Savannah River National Laboratory (SRNL) in June of 2013. The PRFT in DWPF receives Actinide Removal Process (ARP)/ Monosodium Titanate (MST) material from the 512-S Facility via the 511-S Facility. This 2.2 L sample was to be used in small-scale DWPF chemical process cell testing in the Shielded Cells Facility of SRNL. A 1L sub-sample portion was characterized to determine the physical properties such as weight percent solids, density, particle size distribution and crystalline phase identification. Further chemical analysis of the PRFT filtrate and dissolved slurry included metals and anions as well as carbon and base analysis. This technical report describes the characterization and analysis of the PRFT sample from DWPF. At SRNL, the 2.2 L PRFT sample was composited from eleven separate samples received from DWPF. The visible solids were observed to be relatively quick settling which allowed for the rinsing of the original shipping vials with PRFT supernate on the same day as compositing. Most analyses were performed in triplicate except for particle size distribution (PSD), X-ray diffraction (XRD), Scanning Electron Microscopy (SEM) and thermogravimetric analysis (TGA). PRFT slurry samples were dissolved using a mixed HNO3/HF acid for subsequent Inductively Coupled Plasma Atomic Emission Spectroscopy (ICPAES) and Inductively Coupled Plasma Mass Spectroscopy (ICP-MS) analyses performed by SRNL Analytical Development (AD). Per the task request for this work, analysis of the PRFT slurry and filtrate for metals, anions, carbon and base were primarily performed to support the planned chemical process cell testing and to provide additional component concentrations in addition to the limited data available from DWPF. Analysis of the insoluble solids portion of the PRFT slurry was aimed at detailed characterization of these solids (TGA, PSD

  5. Reactor target from metal chromium for "pure" high-intensive artificial neutrino source

    Science.gov (United States)

    Gavrin, V. N.; Kozlova, Yu. P.; Veretenkin, E. P.; Logachev, A. V.; Logacheva, A. I.; Lednev, I. S.; Okunkova, A. A.

    2017-01-01

    The paper presents the first results of development of manufacturing technology of metallic chromium targets from highly enriched isotope 50Cr for irradiation in a high flux nuclear reactor to obtain a compact high intensity neutrino source with low content of radionuclide impurities and minimum losses of enriched isotope. The main technological stages are the hydrolysis of chromyl fluoride, the electrochemical reduction of metallic chromium, the hot isostatic pressing of chromium powder and the electrical discharge machining of chromium bars. The technological stages of hot isostatic pressing of chromium powder and of electrical discharge machining of Cr rods have been tested.

  6. Reactor Physics

    Energy Technology Data Exchange (ETDEWEB)

    Ait Abderrahim, A

    2001-04-01

    The Reactor Physics and MYRRHA Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis of reactor fuel. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2000 are summarised.

  7. Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Martens, Frederick H. [Argonne National Laboratory; Jacobson, Norman H.

    1968-09-01

    This booklet discusses research reactors - reactors designed to provide a source of neutrons and/or gamma radiation for research, or to aid in the investigation of the effects of radiation on any type of material.

  8. C-188 Co-60 sources installation and source rack loading optimization processes in a gamma irradiation facility

    Energy Technology Data Exchange (ETDEWEB)

    Santos, Paulo de S.; Vasquez, Pablo A.S., E-mail: psantos@ipen.br, E-mail: pavsalva@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2017-11-01

    Since 2004, the Multipurpose Gamma Facility at the Nuclear and Energy Research Institute has been providing services on radiation processing for disinfection and sterilization of health care and disposable medical products as well to support research studies on modification of physical, chemical and biological properties of several materials. Recently, there was an increment in irradiation of the Cultural Heritages. This facility uses C-188 double-encapsulated radioactive Cobalt-60 sources known as pencils from manufactures outside of country. The activity of the cobalt sources decays into a stable nickel isotope with a half-life around 5.27 years, which means a loss of 12.3% annually. Then, additional pencils of Cobalt-60 are added periodically to the source rack to maintain the required capacity or installed activity of the facility. The manufacturer makes shipping of the radioactive sources inside a high density container type B(U) , by sea. This one involves many administrative, transport and radiation safety procedures. Once in the facility, the container is opened inside a deep pool water to remove the pencils. The required source geometry of the facility is obtained by loading these source pencils into predetermined diagram or positions in source modules and distributing these modules over the source rack of the facility. The dose variation can be reduced placing the higher activity source pencils near the periphery of the source rack. In this work are presented the procedures for perform the boiling leaching tests applied to the container, the Cobalt-60 sources installation, the loading processes and the source rack loading optimization. (author)

  9. Radiological survey support activities for the decommissioning of the Ames Laboratory Research Reactor Facility, Ames, Iowa

    Energy Technology Data Exchange (ETDEWEB)

    Wynveen, R.A.; Smith, W.H.; Sholeen, C.M.; Justus, A.L.; Flynn, K.F.

    1984-09-01

    At the request of the Engineering Support Division of the US Department of Energy-Chicago Operations Office and in accordance with the programmatic overview/certification responsibilities of the Department of Energy Environmental and Safety Engineering Division, the Argonne National Laboratory Radiological Survey Group conducted a series of radiological measurements and tests at the Ames Laboratory Research Reactor located in Ames, Iowa. These measurements and tests were conducted during 1980 and 1981 while the reactor building was being decontaminated and decommissioned for the purpose of returning the building to general use. The results of these evaluations are included in this report. Although the surface contamination within the reactor building could presumably be reduced to negligible levels, the potential for airborne contamination from tritiated water vapor remains. This vapor emmanates from contamination within the concrete of the building and should be monitored until such time as it is reduced to background levels. 2 references, 8 figures, 6 tables.

  10. Large Object Irradiation Facility In The Tangential Channel Of The JSI TRIGA Reactor

    CERN Document Server

    Radulovic, Vladimir; Kaiba, Tanja; Kavsek, Darko; Cindro, Vladimir; Mikuz, Marko; Snoj, Luka

    2017-01-01

    This paper presents the design and installation of a new irradiation device in the Tangential Channel of the JSI TRIGA reactor in Ljubljana, Slovenia. The purpose of the device is to enable on-line irradiation testing of electronic components considerably larger in size (of lateral dimensions of at least 12 cm) than currently possible in the irradiation channels located in the reactor core, in a relatively high neutron flux (exceeding 10^12 n cm^-2 s^-1) and to provide adequate neutron and gamma radiation shielding.

  11. Development of Pneumatic Transfer Irradiation Facility (PTS no.3) for Neutron Activation Analysis at HANARO Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Y. S.; Moon, J. H.; Kim, S. H.; Sun, G. M.; Baek, S. Y.; Kim, H. R.; Kim, Y. J

    2008-04-15

    A pneumatic transfer system (PTS) is one of the most important facilities used during neutron irradiation of a target material for instrumental neutron activation analysis (INAA) in a research reactor. In particular, a fast pneumatic transfer system is essential for the measurement of a short half-life nuclide. The pneumatic transfer irradiation system (PTS no.3) involving a manual system and an semi-automatic system were reconstructed with new designs of a functional improvement at the HANARO research reactor and NAA laboratory of RI building in 2006. In this technical report, the design, operation and control of these system (PTS no.3) was described. Also the experimental results and the characteristic parameters measured from a functional operation test and an irradiation test of these systems, such as the transfer time of irradiation capsule, the different neutron flux, the temperature of the irradiation position with an irradiation time, the radiation dose rate when the rabbit is returned, etc. are reported to provide a user information as well as a reactor's management and safety.

  12. Development of Pneumatic Transfer Irradiation Facility (PTS no.1) for Neutron Activation Analysis at HANARO Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Y. S.; Moon, J. H.; Kim, S. H.; Sun, G. M.; Baek, S. Y.; Kim, H. R.; Kim, Y. J

    2008-03-15

    A pneumatic transfer system (PTS) is one of the most important facilities used during neutron irradiation of a target material for instrumental neutron activation analysis (INAA) in a research reactor. In particular, a fast pneumatic transfer system is essential for the measurement of a short half-life nuclide and a delayed neutron counting system. The pneumatic transfer system (PTS no.1) involving a manual system and an semiautomatic system were reconstructed with new designs of a functional improvement at the HANARO research reactor in 2006. In this technical report, the conception, design, operation and control of these system (PTS no.1) was described. Also the experimental results and the characteristic parameters measured by a mock-up test, a functional operation test and an irradiation test of these systems, such as the transfer time of irradiation capsule, the different neutron flux, the temperature of the irradiation position with an irradiation time, the radiation dose rate when the rabbit is returned, etc. are reported to provide a user information as well as a reactor's management and safety.

  13. Development of Pneumatic Transfer Irradiation Facility (PTS no.2) for Neutron Activation Analysis at HANARO Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Y. S.; Moon, J. H.; Kim, S. H.; Sun, G. M.; Baek, S. Y.; Kim, H. R.; Kim, Y. J

    2008-03-15

    A pneumatic transfer irradiation system (PTS) is one of the most important facilities used during neutron irradiation of a target material for instrumental neutron activation analysis (INAA) in a research reactor. In particular, a fast pneumatic transfer system is essential for the measurement of a short half-life nuclide and a delayed neutron counting system. The pneumatic transfer irradiation system (PTS no.2) involving a manual system and an automatic system for delayed neutron activation analysis (DNAA) were reconstructed with new designs of a functional improvement at the HANARO research reactor in 2006. In this technical report, the conception, design, operation and control of PTS no.2 was described. Also the experimental results and the characteristic parameters measured by a mock-up test, a functional operation test and an irradiation test of these systems, such as the transfer time of irradiation capsule, automatic operation control by personal computer, delayed neutron counting system, the different neutron flux, the temperature of the irradiation position with an irradiation time, the radiation dose rate when the rabbit is returned, etc. are reported to provide a user information as well as a reactor's management and safety.

  14. Synthetic fuel production via carbon neutral cycles with high temperature nuclear reactors as a power source

    Energy Technology Data Exchange (ETDEWEB)

    Konarek, E.; Coulas, B.; Sarvinis, J. [Hatch Ltd., Mississauga, Ontario (Canada)

    2016-06-15

    This paper analyzes a number of carbon neutral cycles, which could be used to produce synthetic hydrocarbon fuels. Synthetic hydrocarbons are produced via the synthesis of Carbon Monoxide and Hydrogen. The . cycles considered will either utilize Gasification processes, or carbon capture as a source of feed material. In addition the cycles will be coupled to a small modular Nuclear Reactor (SMR) as a power and heat source. The goal of this analysis is to reduce or eliminate the need to transport diesel and other fossil fuels to remote regions and to provide a carbon neutral, locally produced hydrocarbon fuel for remote communities. The technical advantages as well as the economic case are discussed for each of the cycles presented. (author)

  15. SOURCE AND PATHWAY DETERMINATION FOR BERYLLIUM FOUND IN BECHTEL NEVADA NORTH LAS VEGAS FACILITIES

    Energy Technology Data Exchange (ETDEWEB)

    BECHTEL NEVADA

    2004-07-01

    In response to the report ''Investigation of Beryllium Exposure Cases Discovered at the North Las Vegas Facility of the National Nuclear Security Administration'', published by the U.S. Department of Energy (DOE), National Nuclear Security Administration (NNSA) in August 2003, Bechtel Nevada (BN) President and General Manager Dr. F. A. Tarantino appointed the Beryllium Investigation & Assessment Team (BIAT) to identify both the source and pathway for the beryllium found in the North Las Vegas (NLV) B-Complex. From September 8 to December 18, 2003, the BIAT investigated the pathway for beryllium and determined that a number of locations existed at the Nevada Test Site (NTS) which could have contained sufficient quantities of beryllium to result in contamination if transported. Operations performed in the B-1 Building as a result of characterization activities at the Engine Maintenance, Assembly, and Disassembly (EMAD); Reactor Maintenance, Assembly, and Disassembly (RMAD); Test Cells A and C; and the Central Support Facility in Area 25 had the greatest opportunity for transport of beryllium. Investigative monitoring and sampling was performed at these sites with subsequent transport of sample materials, equipment, and personnel from the NTS to the B-1 Building. The timeline established by the BIAT for potential transport of the beryllium contamination into the B-1 Building was from September 1997 through November 2002. Based on results of recently completed swipe sampling, no evidence of transport of beryllium from test areas has been confirmed. Results less than the DOE beryllium action level of 0.2 ???g/100 cm2 were noted for work support facilities located in Area 25. All of the identified sites in Area 25 worked within the B-1 tenant's residency timeline have been remediated. Legacy contaminants have either been disposed of or capped with clean borrow material. As such, no current opportunity exists for release or spread of beryllium

  16. Measurements Methods for the analysis of Nuclear Reactors Thermal Hydraulic in Water Scaled Facilities

    Science.gov (United States)

    Spaccapaniccia, C.; Planquart, P.; Buchlin, J. M. AB(; ), AC(; )

    2018-01-01

    The Belgian nuclear research institute (SCK•CEN) is developing MYRRHA. MYRRHA is a flexible fast spectrum research reactor, conceived as an accelerator driven system (ADS). The configuration of the primary loop is pool-type: the primary coolant and all the primary system components (core and heat exchangers) are contained within the reactor vessel, while the secondary fluid is circulating in the heat exchangers. The primary coolant is Lead Bismuth Eutectic (LBE). The recent nuclear accident of Fukushima in 2011 changed the requirements for the design of new reactors, which should include the possibility to remove the residual decay heat through passive primary and secondary systems, i.e. natural convection (NC). After the reactor shut down, in the unlucky event of propeller failures, the primary and secondary loops should be able to remove the decay heat in passive way (Natural Convection). The present study analyses the flow and the temperature distribution in the upper plenum by applying laser imaging techniques in a laboratory scaled water model. A parametric study is proposed to study stratification mitigation strategies by varying the geometry of the buffer tank simulating the upper plenum.

  17. Experimental Breeder Reactor II (EBR-II) Fuel-Performance Test Facility (FPTF)

    Energy Technology Data Exchange (ETDEWEB)

    Pardini, J.A.; Brubaker, R.C.; Veith, D.J.; Giorgis, G.C.; Walker, D.E.; Seim, O.S.

    1982-01-01

    The Fuel-Performance Test Facility (FPTF) is the latest in a series of special EBR-II instrumented in-core test facilities. A flow control valve in the facility is programmed to vary the coolant flow, and thus the temperature, in an experimental-irradiation subassembly beneath it and coupled to it. In this way, thermal transients can be simulated in that subassembly without changing the temperatures in surrounding subassemblies. The FPTF also monitors sodium flow and temperature, and detects delayed neutrons in the sodium effluent from the experimental-irradiation subassembly beneath it. This facility also has an acoustical detector (high-temperature microphone) for detecting sodium boiling.

  18. ECR ion source based low energy ion beam facility

    Indian Academy of Sciences (India)

    Mass analyzed highly charged ion beams of energy ranging from a few keV to a few MeV plays an important role in various aspects of research in modern physics. In this paper a unique low energy ion beam facility (LEIBF) set up at Nuclear Science Centre (NSC) for providing low and medium energy multiply charged ion ...

  19. Occupational radiation exposure at commercial nuclear power reactors and other facilities 1992; Twenty-fifth annual report, Volume 14

    Energy Technology Data Exchange (ETDEWEB)

    Raddatz, C.T. [US Nuclear Regulatory Commission, Washington, DC (United States). Division of Regulatory Applications; Hagemeyer, D. [Science Applications International Corp., Oak Ridge, TN (United States)

    1993-12-01

    This report summarizes the occupational radiation exposure information that has been reported to the NRC`s Radiation Exposure Information Reporting System (REIRS) by nuclear power facilities and certain other categories of NRC licensees during the years 1969 through 1992. The bulk of the data presented in the report was obtained from annual radiation exposure reports submitted in accordance with the requirements of 10CFR20.407 and the technical specifications of nuclear power plants. Data on workers terminating their employment at certain NRC licensed facilities were obtained from reports submitted pursuant to 10CFR20.408. The 1992 annual reports submitted by about 364 licensees indicated that approximately 204,365 individuals were monitored, 183,927 of whom were monitored by nuclear power facilities. They incurred an average individual dose of 0.16 rem (cSv) and an average measurable dose of about 0.30 (cSv). Termination radiation exposure reports were analyzed to reveal that about 74,566 individuals completed their employment with one or more of the 364 covered licensees during 1992. Some 71,846 of these individuals terminated from power reactor facilities, and about 9,724 of them were considered to be transient workers who received an average dose of 0.50 rem (cSv).

  20. ECR ion source based low energy ion beam facility

    Indian Academy of Sciences (India)

    A nanogan type of ECR source based on a fully permanent magnet design was chosen for this purpose [3]. The ECR ion source along with all its peripheral electron- ics and vacuum components placed on a 200 kV high voltage platform provides multiply charged positive ions in a widely varying energy range from a few ...

  1. Pulsed neutron source and instruments at neutron facility

    Energy Technology Data Exchange (ETDEWEB)

    Teshigawara, Makoto; Aizawa, Kazuya; Suzuki, Jun-ichi; Morii, Yukio; Watanabe, Noboru [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1997-11-01

    We report the results of design studies on the optimal target shape, target - moderator coupling, optimal layout of moderators, and neutron instruments for a next generation pulsed spallation source in JAERI. The source utilizes a projected high-intensity proton accelerator (linac: 1.5 GeV, {approx}8 MW in total beam power, compressor ring: {approx}5 MW). We discuss the target neutronics, moderators and their layout. The sources is designed to have at least 30 beam lines equipped with more than 40 instruments, which are selected tentatively to the present knowledge. (author)

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

    Science.gov (United States)

    Ambrožič, K; Žerovnik, G; Snoj, L

    2017-12-01

    The JSI TRIGA Mark II, IJS research reactor is equipped with numerous irradiation positions, where samples can be irradiated by neutrons and γ-rays. Irradiation position selection is based on its properties, such as physical size and accessibility, as well as neutron and γ-ray spectra, flux and dose intensities. This paper presents an overview on the neutron and γ-ray fluxes, spectra and dose intensities calculations using Monte Carlo MCNP software and ENDF/B-VII.0 nuclear data libraries. The dose-rates are presented in terms of ambient dose equivalents, air kerma, and silicon dose equivalent. At full reactor power the neutron ambient dose equivalent ranges from 5.5×103Svh-1 to 6×106Svh-1, silicon dose equivalent from 6×102Gy/hsi to 3×105Gy/hsi, and neutron air kerma from 4.3×103Gyh-1 to 2×105Gyh-1. Ratio of fast (1MeVreactor power from 3.4×103Svh-1 to 3.6×105Svh-1 and γ air kerma range 3.1×103Gyh-1 to 2.9×105Gyh-1. Copyright © 2017 Elsevier Ltd. All rights reserved.

  3. External neutron source anomalies analysis using Hurst's exponent for the Myrrha reactor

    Energy Technology Data Exchange (ETDEWEB)

    Henrice Junior, Edson; Goncalves, Alessandro C., E-mail: ejunior@con.ufrj.br [Coordenacao dos Cursos de Pos-Graduacao em Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear; Palma, Daniel A.P., E-mail: dapalma@cnen.gov.br [Comissao Nacional de Energia Nuclear (CNEN), Rio de Janeir, RJ (Brazil)

    2015-07-01

    Anomalous diffusion is usually marked by the non-linear growth of the variance in time, that is, the diffusion will be considered as anomalous if there is a deviation in the behaviour described before. This paper aims to identify anomalies in the neutron flux during the operation of an ADS (Accelerator Driven System) nuclear reactor as a result of a trip that originates in the proton accelerator as per project bases, from two different calculation methods for Hurst exponents. These methods are the R/S Method and the Detrended Fluctuation Analysis (DFA) Method. For that the Myrrha Reactor will be simulated using the SERPENT code and the neutron source will be subjected to a production peak at a given instant. The Hurst exponent has a direct application on determining the order of derivatives in fractional point-kinetics equations and the estimate for the fractional derivative can be related as being twice that of Hurst's exponent, according to the co-variance function in the Gauss' processes. After getting the Hurst's exponent a numerically solution is proposed. This subject being a theme very much in focus nowadays. (author)

  4. Intense positron source at the Munich research reactor FRM-II

    CERN Document Server

    Hugenschmidt, C; Schreckenbach, K; Strasser, B; Koegel, G; Sperr, P; Triftshaeuser, W

    2002-01-01

    The principle and the design of the in-pile positron source at the new Munich research reactor FRM-II are presented. Absorption of high-energy prompt gamma-rays from thermal neutron capture in sup 1 sup 1 sup 3 Cd generates positrons by pair production. For this purpose, a cadmium cap is placed inside the tip of the inclined beam tube SR11 in the neutron field of the reactor, where an undisturbed thermal neutron flux up to 2 x 10 sup 1 sup 4 n cm sup - sup 2 s sup - sup 1 is expected. At this position the flux ratio of thermal to fast neutrons will be better than 10 sup 4. Monte Carlo calculations showed that a mean capture rate in cadmium between 4.5 and 6.0 x 10 sup 1 sup 3 n cm sup - sup 2 s sup - sup 1 can be expected. Inside the cadmium cap a structure of platinum foils is placed for converting gamma-radiation into positron-electron pairs. The heated foils also act as positron moderators to generate monoenergetic positrons. After acceleration to 5 keV a positron beam is formed by electric lenses and guid...

  5. Use of Sources and SNM in COE Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Archuleta, Jeffrey Christopher [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-06-07

    This procedure describes how to use Special Nuclear Material (SNM) and sources in order to perform measurements in a safe manner. The hazards and controls associated with these activities are addressed in this document.

  6. An iterated tabu search heuristic for the Single Source Capacitated Facility Location Problem

    DEFF Research Database (Denmark)

    Ho, Sin C.

    2015-01-01

    This paper discusses the Single Source Capacitated Facility Location Problem (SSCFLP) where the problem consists in determining a subset of capacitated facilities to be opened in order to satisfy the customers’ demands such that total costs are minimized. The paper presents an iterated tabu searc...... competitive with other metaheuristic approaches for solving the SSCFLP....

  7. Fission Product Transport and Source Terms in HTRs: Experience from AVR Pebble Bed Reactor

    Directory of Open Access Journals (Sweden)

    Rainer Moormann

    2008-01-01

    Full Text Available Fission products deposited in the coolant circuit outside of the active core play a dominant role in source term estimations for advanced small pebble bed HTRs, particularly in design basis accidents (DBA. The deposited fission products may be released in depressurization accidents because present pebble bed HTR concepts abstain from a gas tight containment. Contamination of the circuit also hinders maintenance work. Experiments, performed from 1972 to 88 on the AVR, an experimental pebble bed HTR, allow for a deeper insight into fission product transport behavior. The activity deposition per coolant pass was lower than expected and was influenced by fission product chemistry and by presence of carbonaceous dust. The latter lead also to inconsistencies between Cs plate out experiments in laboratory and in AVR. The deposition behavior of Ag was in line with present models. Dust as activity carrier is of safety relevance because of its mobility and of its sorption capability for fission products. All metal surfaces in pebble bed reactors were covered by a carbonaceous dust layer. Dust in AVR was produced by abrasion in amounts of about 5 kg/y. Additional dust sources in AVR were ours oil ingress and peeling of fuel element surfaces due to an air ingress. Dust has a size of about 1  m, consists mainly of graphite, is partly remobilized by flow perturbations, and deposits with time constants of 1 to 2 hours. In future reactors, an efficient filtering via a gas tight containment is required because accidents with fast depressurizations induce dust mobilization. Enhanced core temperatures in normal operation as in AVR and broken fuel pebbles have to be considered, as inflammable dust concentrations in the gas phase.

  8. Development of Electrical Capacitance Sensors for Accident Tolerant Fuel (ATF) Testing at the Transient Reactor Test (TREAT) Facility

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Maolong; Ryals, Matthew; Ali, Amir; Blandford, Edward; Jensen, Colby; Condie, Keith; Svoboda, John; O' Brien, Robert

    2016-08-01

    A variety of instruments are being developed and qualified to support the Accident Tolerant Fuels (ATF) program and future transient irradiations at the Transient Reactor Test (TREAT) facility at Idaho National Laboratory (INL). The University of New Mexico (UNM) is working with INL to develop capacitance-based void sensors for determining the timing of critical boiling phenomena in static capsule fuel testing and the volume-averaged void fraction in flow-boiling in-pile water loop fuel testing. The static capsule sensor developed at INL is a plate-type configuration, while UNM is utilizing a ring-type capacitance sensor. Each sensor design has been theoretically and experimentally investigated at INL and UNM. Experiments are being performed at INL in an autoclave to investigate the performance of these sensors under representative Pressurized Water Reactor (PWR) conditions in a static capsule. Experiments have been performed at UNM using air-water two-phase flow to determine the sensitivity and time response of the capacitance sensor under a flow boiling configuration. Initial measurements from the capacitance sensor have demonstrated the validity of the concept to enable real-time measurement of void fraction. The next steps include designing the cabling interface with the flow loop at UNM for Reactivity Initiated Accident (RIA) ATF testing at TREAT and further characterization of the measurement response for each sensor under varying conditions by experiments and modeling.

  9. Efficient removal of antibiotics in a fluidized bed reactor by facile fabricated magnetic powdered activated carbon.

    Science.gov (United States)

    Ma, Jianqing; Yang, Qunfeng; Xu, Dongmei; Zeng, Xiaomei; Wen, Yuezhong; Liu, Weiping

    2017-02-01

    Powdered activated carbons (PACs) with micrometer size are showing great potential for enabling and improving technologies in water treatment. The critical problem in achieving practical application of PAC involves simple, effective fabrication of magnetic PAC and the design of a feasible reactor that can remove pollutants and recover the adsorbent efficiently. Herein, we show that such materials can be fabricated by the combination of PAC and magnetic Fe3O4 with chitosan-Fe hydrogel through a simple co-precipitation method. According to the characterization results, CS-Fe/Fe3O4/PAC with different micrometers in size exhibited excellent magnetic properties. The adsorption of tetracycline was fast and efficient, and 99.9% removal was achieved in 30 min. It also possesses good usability and stability to co-existing ions, organics, and different pH values due to its dispersive interaction nature. Finally, the prepared CS-Fe/Fe3O4/PAC also performed well in the fluidized bed reactor with electromagnetic separation function. It could be easily separated by applying a magnetic field and was effectively in situ regenerated, indicating a potential of practical application for the removal of pollutants from water.

  10. Irradiation facilities at the spallation neutron source SINQ

    Energy Technology Data Exchange (ETDEWEB)

    Lehmann, E.; Ledermann, J.; Aebersold, H.; Kuehne, G.; Kohlik, K. [Paul Scherrer Inst. (PSI), Villigen (Switzerland)

    1997-09-01

    Four independent experiments for sample irradiation are under construction and in preparation for operational tests at the spallation source SINQ. Three of them are located inside a thermal beam port with end positions inside or near the moderator tank. The other experiment will be established at the end position of a super mirror lined neutron guide for applications with cold neutrons. (author) 3 figs., 1 tab., 6 refs.

  11. Dosimetry and radiobiology at the new RA-3 reactor boron neutron capture therapy (BNCT) facility: Application to the treatment of experimental oral cancer

    Energy Technology Data Exchange (ETDEWEB)

    Pozzi, E. [Research and Production Reactors, National Atomic Energy Commission, Ezeiza Atomic Center (Argentina); Department of Radiobiology, National Atomic Energy Commission, Constituyentes Atomic Center (Argentina)], E-mail: epozzi@cnea.gov.ar; Nigg, D.W. [Idaho National Laboratory, Idaho Falls (United States); Miller, M.; Thorp, S.I. [Instrumentation and Control Department, National Atomic Energy Commission, Ezeiza Atomic Center (Argentina); Heber, E.M. [Department of Radiobiology, National Atomic Energy Commission, Constituyentes Atomic Center (Argentina); Zarza, L.; Estryk, G. [Research and Production Reactors, National Atomic Energy Commission, Ezeiza Atomic Center (Argentina); Monti Hughes, A.; Molinari, A.J.; Garabalino, M. [Department of Radiobiology, National Atomic Energy Commission, Constituyentes Atomic Center (Argentina); Itoiz, M.E. [Department of Radiobiology, National Atomic Energy Commission, Constituyentes Atomic Center (Argentina); Department of Oral Pathology, Faculty of Dentistry, University of Buenos Aires (Argentina); Aromando, R.F. [Department of Oral Pathology, Faculty of Dentistry, University of Buenos Aires (Argentina); Quintana, J. [Research and Production Reactors, National Atomic Energy Commission, Ezeiza Atomic Center (Argentina); Trivillin, V.A.; Schwint, A.E. [Department of Radiobiology, National Atomic Energy Commission, Constituyentes Atomic Center (Argentina)

    2009-07-15

    The National Atomic Energy Commission of Argentina (CNEA) constructed a novel thermal neutron source for use in boron neutron capture therapy (BNCT) applications at the RA-3 research reactor facility located in Buenos Aires. The aim of the present study was to perform a dosimetric characterization of the facility and undertake radiobiological studies of BNCT in an experimental model of oral cancer in the hamster cheek pouch. The free-field thermal flux was 7.1x10{sup 9} n cm{sup -2} s{sup -1} and the fast neutron flux was 2.5x10{sup 6} n cm{sup -2} s{sup -1}, indicating a very well-thermalized neutron field with negligible fast neutron dose. For radiobiological studies it was necessary to shield the body of the hamster from the neutron flux while exposing the everted cheek pouch bearing the tumors. To that end we developed a lithium (enriched to 95% in {sup 6}Li) carbonate enclosure. Groups of tumor-bearing hamsters were submitted to BPA-BNCT, GB-10-BNCT, (GB-10+BPA)-BNCT or beam only treatments. Normal (non-cancerized) hamsters were treated similarly to evaluate normal tissue radiotoxicity. The total physical dose delivered to tumor with the BNCT treatments ranged from 6 to 8.5 Gy. Tumor control at 30 days ranged from 73% to 85%, with no normal tissue radiotoxicity. Significant but reversible mucositis in precancerous tissue surrounding tumors was associated to BPA-BNCT. The therapeutic success of different BNCT protocols in treating experimental oral cancer at this novel facility was unequivocally demonstrated.

  12. Dosimetry and radiobiology at the new RA-3 reactor boron neutron capture therapy (BNCT) facility: application to the treatment of experimental oral cancer.

    Science.gov (United States)

    Pozzi, E; Nigg, D W; Miller, M; Thorp, S I; Heber, E M; Zarza, L; Estryk, G; Monti Hughes, A; Molinari, A J; Garabalino, M; Itoiz, M E; Aromando, R F; Quintana, J; Trivillin, V A; Schwint, A E

    2009-07-01

    The National Atomic Energy Commission of Argentina (CNEA) constructed a novel thermal neutron source for use in boron neutron capture therapy (BNCT) applications at the RA-3 research reactor facility located in Buenos Aires. The aim of the present study was to perform a dosimetric characterization of the facility and undertake radiobiological studies of BNCT in an experimental model of oral cancer in the hamster cheek pouch. The free-field thermal flux was 7.1 x 10(9) n cm(-2)s(-1) and the fast neutron flux was 2.5 x 10(6) n cm(-2)s(-1), indicating a very well-thermalized neutron field with negligible fast neutron dose. For radiobiological studies it was necessary to shield the body of the hamster from the neutron flux while exposing the everted cheek pouch bearing the tumors. To that end we developed a lithium (enriched to 95% in (6)Li) carbonate enclosure. Groups of tumor-bearing hamsters were submitted to BPA-BNCT, GB-10-BNCT, (GB-10+BPA)-BNCT or beam only treatments. Normal (non-cancerized) hamsters were treated similarly to evaluate normal tissue radiotoxicity. The total physical dose delivered to tumor with the BNCT treatments ranged from 6 to 8.5 Gy. Tumor control at 30 days ranged from 73% to 85%, with no normal tissue radiotoxicity. Significant but reversible mucositis in precancerous tissue surrounding tumors was associated to BPA-BNCT. The therapeutic success of different BNCT protocols in treating experimental oral cancer at this novel facility was unequivocally demonstrated.

  13. Productivity of a nuclear chemical reactor with gamma radioisotopic sources; Rendimiento de un reactor quimico-nuclear con fuentes radioisotopicas gamma

    Energy Technology Data Exchange (ETDEWEB)

    Anguis T, C

    1975-07-01

    According to an established mathematical model of successive Compton interaction processes the made calculations for major distances are extended checking the acceptability of the spheric geometry model for the experimental data for radioisotopic sources of Co-60 and Cs-137. Parameters such as the increasing factor and the absorbed dose served as comparative base. calculations for the case of a punctual source succession inside a determined volume cylinder are made to obtain the total dose, the deposited energy by each photons energetic group and the total absorbed energy inside the reactor. Varying adequately the height/radius relation for different cylinders, the distinct energy depositions are compared in each one of them once a time standardized toward a standard value of energy emitted by the reactor volume. A relation between the quantity of deposited energy in each point of the reactor and the conversion values of chemical species is established. They are induced by electromagnetic radiation and that are reported as ''G'' in the scientific literature (number of molecules formed or disappeared by each 100 e.v. of energy). Once obtained the molecular performance inside the reactor for each type of geometry, it is optimized the height/radius relation according to the maximum production of molecules by unity of time. It is completed a bibliographical review of ''G'' values reported by different types of aqueous solutions with the purpose to determine the maximum performance of molecular hydrogen as a function of pH of the solution and of the used type of solute among other factors. Calculations for the ethyl bromide production as an example of one of the industrial processes which actually work using the gamma radiation as reactions inductor are realized. (Author)

  14. Upgrade of the ultracold neutron source at the pulsed reactor TRIGA Mainz

    Science.gov (United States)

    Kahlenberg, J.; Ries, D.; Ross, K. U.; Siemensen, C.; Beck, M.; Geppert, C.; Heil, W.; Hild, N.; Karch, J.; Karpuk, S.; Kories, F.; Kretschmer, M.; Lauss, B.; Reich, T.; Sobolev, Y.; Trautmann, N.

    2017-11-01

    The performance of the upgraded solid deuterium ultracold neutron source at the pulsed reactor TRIGA Mainz is described. The current configuration stage comprises the installation of a He liquefier to run UCN experiments over long-term periods, the use of stainless steel neutron guides with improved transmission as well as sputter-coated non-magnetic 58NiMo alloy at the inside walls of the thermal bridge and the converter cup. The UCN yield was measured in a "standard" UCN storage bottle (stainless steel) with a volume of 32 litres outside the biological shield at the experimental area yielding UCN densities of 8.5/cm3; an increase by a factor of 3.5 compared to the former setup. The measured UCN storage curve is in good agreement with the predictions from a Monte Carlo simulation developed to model the source. The growth and formation of the solid deuterium converter during freeze-out are affected by the ortho/para ratio of the H2 premoderator.

  15. Source term model evaluations for the low-level waste facility performance assessment

    Energy Technology Data Exchange (ETDEWEB)

    Yim, M.S.; Su, S.I. [North Carolina State Univ., Raleigh, NC (United States)

    1995-12-31

    The estimation of release of radionuclides from various waste forms to the bottom boundary of the waste disposal facility (source term) is one of the most important aspects of LLW facility performance assessment. In this work, several currently used source term models are comparatively evaluated for the release of carbon-14 based on a test case problem. The models compared include PRESTO-EPA-CPG, IMPACTS, DUST and NEFTRAN-II. Major differences in assumptions and approaches between the models are described and key parameters are identified through sensitivity analysis. The source term results from different models are compared and other concerns or suggestions are discussed.

  16. Advanced Electron Beam Ion Sources (EBIS) for 2-nd generation carbon radiotherapy facilities

    CERN Document Server

    Shornikov, A.

    2016-01-01

    In this work we analyze how advanced Electron Beam Ion Sources (EBIS) can facilitate the progress of carbon therapy facilities. We will demonstrate that advanced ion sources enable operation of 2-nd generation ion beam therapy (IBT) accelerators. These new accelerator concepts with designs dedicated to IBT provide beams better suited for therapy and, are more cost efficient than contemporary IBT facilities. We will give a sort overview of the existing new IBT concepts and focus on those where ion source technology is the limiting factor. We will analyse whether this limitation can be overcome in the near future thanks to ongoing EBIS development.

  17. Electrostatic levitation facility optimized for neutron diffraction studies of high temperature liquids at a spallation neutron source

    Energy Technology Data Exchange (ETDEWEB)

    Mauro, N. A., E-mail: namauro@noctrl.edu [Department of Physics, North Central College, Naperville, Illinois 60540 (United States); Vogt, A. J. [Instrument and Source Division, Oak Ridge National Laboratory, Oak Ridge, Tennessee 37831 (United States); Derendorf, K. S. [Mechanical Engineering and Materials Science, Washington University, St. Louis, Missouri 63130 (United States); Johnson, M. L.; Kelton, K. F. [Department of Physics and Institute of Materials Science and Engineering, Washington University, 1 Brookings Drive, St. Louis, Missouri 63130 (United States); Rustan, G. E.; Quirinale, D. G.; Goldman, A. I. [Department of Physics and Astronomy, Iowa State University, Ames, Iowa 50011 (United States); Kreyssig, A. [Department of Physics and Astronomy, Iowa State University, Ames, Iowa 50011 (United States); Division of Materials Sciences and Engineering, Ames Laboratory, Ames, Iowa 50011 (United States); Lokshin, K. A. [Department of Materials Science and Engineering, University of Tennessee, Knoxville, Tennessee 37996 (United States); Quantum Condensed Matter Division, Oak Ridge National Laboratory, Oak Ridge, Tennessee 37831 (United States); Neuefeind, J. C.; An, Ke [Chemical and Engineering Materials Division, Oak Ridge National Laboratory, Oak Ridge, Tennessee 37831 (United States); Wang, Xun-Li [Department of Physics and Materials Science, City University of Hong Kong, 83 Tat Chee Ave., Kowloon (Hong Kong); Egami, T. [Department of Materials Science and Engineering, University of Tennessee, Knoxville, Tennessee 37996 (United States); Department of Physics and Astronomy, Joint Institute for Neutron Sciences, University of Tennessee, Knoxville, Tennessee 37996 (United States)

    2016-01-15

    Neutron diffraction studies of metallic liquids provide valuable information about inherent topological and chemical ordering on multiple length scales as well as insight into dynamical processes at the level of a few atoms. However, there exist very few facilities in the world that allow such studies to be made of reactive metallic liquids in a containerless environment, and these are designed for use at reactor-based neutron sources. We present an electrostatic levitation facility, NESL (for Neutron ElectroStatic Levitator), which takes advantage of the enhanced capabilities and increased neutron flux available at spallation neutron sources (SNSs). NESL enables high quality elastic and inelastic neutron scattering experiments to be made of reactive metallic and other liquids in the equilibrium and supercooled temperature regime. The apparatus is comprised of a high vacuum chamber, external and internal neutron collimation optics, and a sample exchange mechanism that allows up to 30 samples to be processed between chamber openings. Two heating lasers allow excellent sample temperature homogeneity, even for samples approaching 500 mg, and an automated temperature control system allows isothermal measurements to be conducted for times approaching 2 h in the liquid state, with variations in the average sample temperature of less than 0.5%. To demonstrate the capabilities of the facility for elastic scattering studies of liquids, a high quality total structure factor for Zr{sub 64}Ni{sub 36} measured slightly above the liquidus temperature is presented from experiments conducted on the nanoscale-ordered materials diffractometer (NOMAD) beam line at the SNS after only 30 min of acquisition time for a small sample (∼100 mg)

  18. Development of a test facility for analyzing transients in supercritical water-cooled reactors by fractional scaling analysis

    Energy Technology Data Exchange (ETDEWEB)

    Roberto, Thiago D., E-mail: thiagodbtr@gmail.com [Instituto de Engenharia Nuclear (IEN/CNEN—RJ), Rua Hélio de Almeida, 75 21941-972, Rio de Janeiro Caixa-Postal: 68550, RJ (Brazil); Silva, Mário A. B. da, E-mail: mabs500@gmail.com [Departamento de Energia Nuclear (CTG/UFPE), Av. Professor Luiz Freire, 1000, Recife 50740-540, PE (Brazil); Lapa, Celso M.F., E-mail: lapa@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN—RJ), Rua Hélio de Almeida, 75 21941-972, Rio de Janeiro Caixa-Postal: 68550, RJ (Brazil)

    2016-01-15

    The feasibility of performing experiments using water under supercritical conditions is limited by technical and financial difficulties. These difficulties can be overcome by using model fluids that are characterized by feasible supercritical conditions, that is, lower critical pressure and critical temperature. Experimental investigations are normally used to determine the conditions under which model fluids reliably represent supercritical fluids under steady-state conditions. A fluid-to-fluid scaling approach has been proposed to determine the model fluids that represent supercritical fluids in a transient state. Recently, a similar technique known as fractional scaling analysis was developed to establish the conditions under which experiments can be performed using models that represent transients in prototypes. This paper presents a fractional scaling analysis application to determine parameters for a test facility in which transient conditions in supercritical water-cooled reactors are simulated by using carbon dioxide as a model fluid, whose critical point conditions are more feasible than those of water. Similarity is obtained between water (prototype) and carbon dioxide (model) by depressurization in a simple vessel. The main parameters required for the construction of a future test facility are obtained using the proposed method.

  19. Specific factors influencing information system/information and communication technology sourcing strategies in healthcare facilities.

    Science.gov (United States)

    Potančok, Martin; Voříšek, Jiří

    2016-09-01

    Healthcare facilities use a number of information system/information and communication technologies. Each healthcare facility faces a need to choose sourcing strategies most suitable to ensure provision of information system/information and communication technology services, processes and resources. Currently, it is possible to observe an expansion of sourcing possibilities in healthcare informatics, which creates new requirements for sourcing strategies. Thus, the aim of this article is to identify factors influencing information system/information and communication technology sourcing strategies in healthcare facilities. The identification was based on qualitative research, namely, a case study. This study provides a set of internal and external factors with their impact levels. The findings also show that not enough attention is paid to these factors during decision-making. © The Author(s) 2015.

  20. Risk-Informing Safety Reviews for Non-Reactor Nuclear Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Mubayi, V.; Azarm, A.; Yue, M.; Mukaddam, W.; Good, G.; Gonzalez, F.; Bari, R.A.

    2011-03-13

    This paper describes a methodology used to model potential accidents in fuel cycle facilities that employ chemical processes to separate and purify nuclear materials. The methodology is illustrated with an example that uses event and fault trees to estimate the frequency of a specific energetic reaction that can occur in nuclear material processing facilities. The methodology used probabilistic risk assessment (PRA)-related tools as well as information about the chemical reaction characteristics, information on plant design and operational features, and generic data about component failure rates and human error rates. The accident frequency estimates for the specific reaction help to risk-inform the safety review process and assess compliance with regulatory requirements.

  1. SBLOCA AND LOFW EXPERIMENTS IN A SCALED-DOWN IET FACILITY OF REX-10 REACTOR

    Directory of Open Access Journals (Sweden)

    YEON-GUN LEE

    2013-06-01

    Full Text Available This paper presents an experimental investigation of the small-break loss-of-coolant accident (SBLOCA and the loss-of-feedwater accident (LOFW in a scaled integral test facility of REX-10. REX-10 is a small integral-type PWR in which the coolant flow is driven by natural circulation, and the RCS is pressurized by the steam-gas pressurizer. The postulated accidents of REX-10 include the system depressurization initiated by the break of a nitrogen injection line connected to the steam-gas pressurizer and the complete loss of normal feedwater flow by the malfunction of control systems. The integral effect tests on SBLOCA and LOFW are conducted at the REX-10 Test Facility (RTF, a full-height full-pressure facility with reduced power by 1/50. The SBLOCA experiment is initiated by opening a flow passage out of the pressurizer vessel, and the LOFW experiment begins with the termination of the feedwater supply into the helical-coil steam generator. The experimental results reveal that the RTF can assure sufficient cooldown capability with the simulated PRHRS flow during these DBAs. In particular, the RTF exhibits faster pressurization during the LOFW test when employing the steam-gas pressurizer than the steam pressurizer. This experimental study can provide unique data to validate the thermal-hydraulic analysis code for REX-10.

  2. Derivation of the source term, dose results and associated radiological consequences for the Greek Research Reactor – 1

    Energy Technology Data Exchange (ETDEWEB)

    Pappas, Charalampos, E-mail: chpappas@ipta.demokritos.gr; Ikonomopoulos, Andreas; Sfetsos, Athanasios; Andronopoulos, Spyros; Varvayanni, Melpomeni; Catsaros, Nicolas

    2014-07-01

    Highlights: • Source term derivation of postulated accident sequences in a research reactor. • Various containment ventilation scenarios considered for source term calculations. • Source term parametric analysis performed in case of lack of ventilation. • JRODOS employed for dose calculations under eighteen modeled scenarios. • Estimation of radiological consequences during typical and adverse weather scenarios. - Abstract: The estimated source term, dose results and radiological consequences of selected accident sequences in the Greek Research Reactor – 1 are presented and discussed. A systematic approach has been adopted to perform the necessary calculations in accordance with the latest computational developments and IAEA recommendations. Loss-of-coolant, reactivity insertion and fuel channel blockage accident sequences have been selected to derive the associated source terms under three distinct containment ventilation scenarios. Core damage has been conservatively assessed for each accident sequence while the ventilation has been assumed to function within the efficiency limits defined at the Safety Analysis Report. In case of lack of ventilation a parametric analysis is also performed to examine the dependency of the source term on the containment leakage rate. A typical as well as an adverse meteorological scenario have been defined in the JRODOS computational platform in order to predict the effective, lung and thyroid doses within a region defined by a 15 km radius downwind from the reactor building. The radiological consequences of the eighteen scenarios associated with the accident sequences are presented and discussed.

  3. Isotope Mixes, Corresponding Nuclear Properties and Reactor Design Implications of Naturally Occurring Lead Sources

    Science.gov (United States)

    2013-06-01

    then becomes plutonium (Pu239). This plutonium isotope can be reprocessed and used as more reactor fuel.” [19] Lead and sodium are potential...coolants for use in fast reactors. The ability to convert the non-fissile (238) isotope of uranium into a fissile isotope of plutonium dramatically...used to perform simulations of the reactor model to create meaningful results. A. GEOCHEMISTRY Geochemistry is the science of using chemistry to

  4. Biological phosphorus removal in anoxic-aerobic sequencing batch reactor with starch as sole carbon source.

    Science.gov (United States)

    Luo, Dacheng; Yuan, Linjiang; Liu, Lun; Chai, Lu; Wang, Xin

    2017-01-01

    In traditional biological phosphorus removal (BPR), phosphorus release in anaerobic stage is the prerequisite of phosphorus excessive uptake in aerobic conditions. Moreover, when low molecular weight of the organic substance such as volatile fatty acids (VFAs) is scarce in bulk liquid or anaerobic condition does not exist, phosphate accumulating organisms (PAOs) have difficulty removing phosphorus. However, in this work, phosphorus removal in two anoxic-aerobic sequencing batch reactors (SBRs) was observed when starch was supplied as a sole carbon source. The relations of the BPR with idle period were investigated in the two identical SBRs; the idle times were set to 0.5 hr (R1) and 4 hr (R2), respectively. Results of the study showed that, in the two SBRs, phosphorus concentrations of 0.26-3.11 mg/L in effluent were obtained after aeration when phosphorus concentration in influent was about 8 mg/L. Moreover, lower accumulations/transformations of polyhydroxyalkanoates (PHAs) and higher transformation of glycogen occurred in the SBRs, indicating that glycogen was the main energy source that was different from the traditional mechanism of BPR. Under the different idle time, the phosphorus removal was a little different. In R2, which had a longer idle period, phosphorus release was very obvious just as occurs in a anaerobic-aerobic regime, but there was a special phenomenon of chemical oxygen demand increase, while VFAs had no notable change. It is speculated that PAOs can assimilate organic compounds in the mixed liquor, which were generated from glycolysis by fermentative organisms, coupled with phosphorus release. In R1, which had a very short idle period, anaerobic condition did not exist; phosphorus removal rate reached 63%. It is implied that a new metabolic pathway can occur even without anaerobic phosphorus release when starch is supplied as the sole carbon source.

  5. A hybrid algorithm for stochastic single-source capacitated facility location problem with service level requirements

    Directory of Open Access Journals (Sweden)

    Hosseinali Salemi

    2016-04-01

    Full Text Available Facility location models are observed in many diverse areas such as communication networks, transportation, and distribution systems planning. They play significant role in supply chain and operations management and are one of the main well-known topics in strategic agenda of contemporary manufacturing and service companies accompanied by long-lasting effects. We define a new approach for solving stochastic single source capacitated facility location problem (SSSCFLP. Customers with stochastic demand are assigned to set of capacitated facilities that are selected to serve them. It is demonstrated that problem can be transformed to deterministic Single Source Capacitated Facility Location Problem (SSCFLP for Poisson demand distribution. A hybrid algorithm which combines Lagrangian heuristic with adjusted mixture of Ant colony and Genetic optimization is proposed to find lower and upper bounds for this problem. Computational results of various instances with distinct properties indicate that proposed solving approach is efficient.

  6. NASA's Nuclear Frontier: The Plum Brook Reactor Facility, 1941-2002

    Science.gov (United States)

    Bowles, Mark D.; Arrighi, Robert S.

    2004-01-01

    In 1953, President Eisenhower delivered a speech called "Atoms for Peace" to the United Nations General Assembly. He described the emergence of the atomic age and the weapons of mass destruction that were piling up in the storehouses of the American and Soviet nations. Although neither side was aiming for global destruction, Eisenhower wanted to "move out of the dark chambers of horrors into the light, to find a way by which the minds of men, the hopes of men, the souls of men everywhere, can move towards peace and happiness and well-being." One way Eisenhower hoped this could happen was by transforming the atom from a weapon of war into a useful tool for civilization. Many people believed that there were unprecedented opportunities for peaceful nuclear applications. These included hopeful visions of atomic-powered cities, cars, airplanes, and rockets. Nuclear power might also serve as an efficient way to generate electricity in space to support life and machines. Eisenhower wanted to provide scientists and engineers with "adequate amounts of fission- able material with which to test and develop their ideas." But, in attempting to devise ways to use atomic power for peaceful purposes, scientists realized how little they knew about the nature and effects of radiation. As a result, the United States began constructing nuclear test reactors to enable scientists to conduct research by producing neutrons.

  7. Report on emergency electrical power supply systems for nuclear fuel cycle and reactor facilities security systems

    Energy Technology Data Exchange (ETDEWEB)

    1977-01-01

    The report includes information that will be useful to those responsible for the planning, design and implementation of emergency electric power systems for physical security and special nuclear materials accountability systems. Basic considerations for establishing the system requirements for emergency electric power for security and accountability operations are presented. Methods of supplying emergency power that are available at present and methods predicted to be available in the future are discussed. The characteristics of capacity, cost, safety, reliability and environmental and physical facility considerations of emergency electric power techniques are presented. The report includes basic considerations for the development of a system concept and the preparation of a detailed system design.

  8. A new method to determine in situ the transmission of a neutron-guide system at a reactor source

    CERN Document Server

    Haan, V O D; Gommers, R M; Labohm, F; Well, A A V; De Leege, P F A; Schebetov, A; Pusenkov, V

    2002-01-01

    In this paper, a description of a new method to determine the transmission of neutron guides after they are installed in a beam-tube at a reactor source is given. The method is based on activation measurements of gold foils at the entrance of the beam-tube and at the exit of the neutron guides compared to Monte-Carlo calculations. In this method, a quality factor is defined as the ratio between the actual transmission and the theoretical maximum attainable transmission. This method is used to determine the quality of an optimised neutron-guide system developed for beam-tube R2 of the HOR. The HOR is a pool-type nuclear research reactor at the Interfaculty Reactor Institute of the Delft University of Technology. It is shown that the quality factors of the newly installed neutron guides are between 0.49 and 0.63.

  9. Traditional blood irradiation facilities based on radioactive sources are phased out; Tradisjonelle blodbestraalingsanlegg basert paa radioaktive kilder fases ut

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-03-01

    Blood irradiation facilities containing radioactive sources are among the most powerful sources of radiation. As government we want to phase out this type of facility for the benefit of virtually risk-free blood irradiation facility based on X-ray technology.(eb)

  10. Safety considerations in the disposal of disused sealed radioactive sources in borehole facilities

    CERN Document Server

    International Atomic Energ Agency. Vienna

    2003-01-01

    Sealed radioactive sources are used in medicine, industry and research for a wide range of purposes. They can contain different radionuclides in greatly varying amounts. At the end of their useful lives, they are termed 'disused sources' but their activity levels can still be quite high. They are, for all practical purposes, another type of radioactive waste that needs to be disposed of safely. Disused sealed radioactive sources can represent a significant hazard to people if not managed properly. Many countries have no special facilities for the management or disposal of radioactive waste, as they have no nuclear power programmes requiring such facilities. Even in countries with developed nuclear programmes, disused sealed sources present problems as they often fall outside the common categories of radioactive waste for which disposal options have been identified. As a result, many disused sealed sources are kept in storage. Depending on the nature of the storage arrangements, this situation may represent a ...

  11. Effects of Feedstock Sources on Inoculant Acclimatization: Start-up Strategies and Reactor Performance.

    Science.gov (United States)

    Yang, Qian; Wei, Liang-Huan; Li, Wei-Zun; Chen, Yu; Ju, Mei-Ting

    2017-11-01

    Different inoculum sources and acclimatization methods result in different substrate adaptation and biodegradability. To increase straw degradation rate, shorten the digester start-up time, and enhance the biogas production, we domesticated anaerobic sludge by adding microcrystalline cellulose (MCC). During acclimatization, the start-up strategies and reactor performance were investigated to analyze changes in feedstock adaption, biodegradability, and methanogen activity. The effect of the domesticated inoculum was evaluated by testing batch un-pretreated corn stover with a dewatered sludge (DS)-domesticated inoculum as a control. The results showed that (1) using MCC as a substrate rapidly improved microorganism biodegradability and adaptation. (2) MCC as domesticated substrate has relatively stable system and high mass conversion, but with low buffer capacity. (3) Macro- and micronutrients should be added for improving the activity of methanogenic and system's buffer capacity. (4) Using the domesticated inoculums and batch tests to anaerobically digest untreated corn stover yielded rapid biogas production of 292 mL, with an early peak value on the first day. The results indicated that cultivating directional inoculum can efficiently and quickly start-up digester. These investigated results to promote anaerobic digestion of straw for producing biogas speed up the transformation of achievements of biomass solid waste utilization have a positive promoting significance.

  12. Occupational Radiation Exposure at Commercial Nuclear Power Reactors and Other Facilities 2008

    Energy Technology Data Exchange (ETDEWEB)

    U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research

    2009-12-01

    This report summarizes the occupational exposure data that are maintained in the U.S. Nuclear Regulatory Commission (NRC) Radiation Exposure Information and Reporting System (REIRS). The bulk of the information contained in the report was compiled from the 2008 annual reports submitted by five of the seven categories1 of NRC licensees subject to the reporting requirements of 10 CFR 20.2206. The annual reports submitted by these licensees consist of radiation exposure records for each monitored individual. These records are analyzed for trends and presented in this report in terms of collective dose and the distribution of dose among the monitored individuals. Because there are no geologic repositories for high-level waste currently licensed and no low-level waste disposal facilities in operation, only five categories will be considered in this report.

  13. GEANT4 used for neutron beam design of a neutron imaging facility at TRIGA reactor in Morocco

    Science.gov (United States)

    Ouardi, A.; Machmach, A.; Alami, R.; Bensitel, A.; Hommada, A.

    2011-09-01

    Neutron imaging has a broad scope of applications and has played a pivotal role in visualizing and quantifying hydrogenous masses in metallic matrices. The field continues to expand into new applications with the installation of new neutron imaging facilities. In this scope, a neutron imaging facility for computed tomography and real-time neutron radiography is currently being developed around 2.0MW TRIGA MARK-II reactor at Maamora Nuclear Research Center in Morocco (Reuscher et al., 1990 [1]; de Menezes et al., 2003 [2]; Deinert et al., 2005 [3]). The neutron imaging facility consists of neutron collimator, real-time neutron imaging system and imaging process systems. In order to reduce the gamma-ray content in the neutron beam, the tangential channel was selected. For power of 250 kW, the corresponding thermal neutron flux measured at the inlet of the tangential channel is around 3×10 11 ncm 2/s. This facility will be based on a conical neutron collimator with two circular diaphragms with diameters of 4 and 2 cm corresponding to L/D-ratio of 165 and 325, respectively. These diaphragms' sizes allow reaching a compromise between good flux and efficient L/D-ratio. Convergent-divergent collimator geometry has been adopted. The beam line consists of a gamma filter, fast neutrons filter, neutron moderator, neutron and gamma shutters, biological shielding around the collimator and several stages of neutron collimator. Monte Carlo calculations by a fully 3D numerical code GEANT4 were used to design the neutron beam line ( http://www.info.cern.ch/asd/geant4/geant4.html[4]). To enhance the neutron thermal beam in terms of quality, several materials, mainly bismuth (Bi) and sapphire (Al 2O 3) were examined as gamma and neutron filters respectively. The GEANT4 simulations showed that the gamma and epithermal and fast neutron could be filtered using the bismuth (Bi) and sapphire (Al 2O 3) filters, respectively. To get a good cadmium ratio, GEANT 4 simulations were used to

  14. 76 FR 11291 - University of New Mexico AGN-201M Reactor Notice of Issuance of Renewed Facility Operating...

    Science.gov (United States)

    2011-03-01

    ..., New Mexico. The UNMR is a solid homogeneous core research reactor licensed to operate at a steady... Nuclear Regulatory Commission. Jessie Quichocho, Chief, Research and Test Reactors Licensing Branch, Division of Policy and Rulemaking, Office of Nuclear Reactor Regulation. BILLING CODE 7590-01-P ...

  15. The Effect of Initial Inoculum Source on the Microbial Community Structure and Dynamics in Laboratory-Scale Sequencing Batch Reactors

    KAUST Repository

    Hernandez, Susana

    2011-07-01

    Understanding the factors that shapes the microbial community assembly in activated sludge wastewater treatment processes provide a conceptual foundation for improving process performance. The aim of this study was to compare two major theories (deterministic theory and neutral theory) regarding the assembly of microorganisms in activated sludge: Six lab-scale activated sludge sequencing batch reactors were inoculated with activated sludge collected from three different sources (domestic, industrial, and sugar industry WWTP). Additionally, two reactors were seeded with equal proportion of sludge from the three WWTPs. Duplicate reactors were used for each sludge source (i.e. domestic, industrial, sugar and mix). Reactors were operated in parallel for 11 weeks under identical conditions. Bacterial diversity and community structure in the eight SBRs were assessed by 16S rRNA gene pyrosequencing. The 16S rRNA gene sequences were analyzed using taxonomic and clustering analysis and by measuring diversity indices (Shannon-weaver and Chao1 indices). Cluster analysis revealed that the microbial community structure was dynamic and that replicate reactors evolved differently. Also the microbial community structure in the SBRs seeded with a different sludge did not converge after 11 weeks of operation under identical conditions. These results suggest that history and distribution of taxa in the source inoculum were stronger regulating factors in shaping bacterial community structure than environmental factors. This supports the neutral theory which states that the assembly of the local microbial community from the metacommunity is random and is regulated by the size and diversity of the metacommunity. Furthermore, sludge performance, measured by COD and ammonia removal, confirmed that broad-scale functions (e.g. COD removal) are not influenced by dynamics in the microbial composition, while specific functions (e.g. nitrification) are more susceptible to these changes.

  16. Effect of different carbon sources on the biological phosphorus removal by a sequencing batch reactor using pressurized pure oxygen.

    Science.gov (United States)

    Wei, Jie; Imai, Tsuyoshi; Higuchi, Takaya; Arfarita, Novi; Yamamoto, Koichi; Sekine, Masahiko; Kanno, Ariyo

    2014-05-04

    The effect of different carbon source on the efficiency of enhanced biological phosphorus removal (EBPR) from synthetic wastewater with acetate and two ratios of acetate/starch as a carbon source was investigated. Three pressurized pure oxygen sequencing batch reactor (POSBR) experiments were operated. The reactors (POSBR1, POSBR2 and POSBR3) were developed and studied at different carbon source ratios of 100% acetate, 75% acetate plus 25% starch and 50% acetate plus 50% starch, respectively. The results showed that POSBR1 had a higher phosphate release-to-uptake ratio and, respectively, in a much higher phosphorus removal efficiency (93.8%) than POSBR2 (84.7%) and POSBR3 (77.3%) within 30 days of operation. This indicated that the phosphorus removal efficiency decreased the higher the starch concentration was. It was also found that POSBR1 produced more polyhydroxyalkanoates (PHAs) than the other reactors. Based on the effect of the carbon source on the PHA concentration and consumption, the conditions of POSBR1 were favourable for the growth of polyphosphate-accumulating organisms and therefore, beneficial for the biological phosphorus removal process.

  17. The development of a realistic source term for sodium-cooled fast reactors : assessment of current status and future needs.

    Energy Technology Data Exchange (ETDEWEB)

    LaChance, Jeffrey L.; Phillips, Jesse; Parma, Edward J., Jr.; Olivier, Tara Jean; Middleton, Bobby D.

    2011-06-01

    Sodium-cooled fast reactors (SFRs) continue to be proposed and designed throughout the United States and the world. Although the number of SFRs actually operating has declined substantially since the 1980s, a significant interest in advancing these types of reactor systems remains. Of the many issues associated with the development and deployment of SFRs, one of high regulatory importance is the source term to be used in the siting of the reactor. A substantial amount of modeling and experimental work has been performed over the past four decades on accident analysis, sodium coolant behavior, and radionuclide release for SFRs. The objective of this report is to aid in determining the gaps and issues related to the development of a realistic, mechanistically derived source term for SFRs. This report will allow the reader to become familiar with the severe accident source term concept and gain a broad understanding of the current status of the models and experimental work. Further, this report will allow insight into future work, in terms of both model development and experimental validation, which is necessary in order to develop a realistic source term for SFRs.

  18. Large-scale experimental facility for emergency condition investigation of a new generation NPP WWER-640 reactor with passive safety systems

    Energy Technology Data Exchange (ETDEWEB)

    Aniskevich, Y.N.; Vasilenko, V.A.; Zasukha, V.K.; Migrov, Y.A.; Khabensky, V.B. [Research Inst. of Technology NITI (Russian Federation)

    1997-12-31

    The creation of the large-scale integral experimental facility (KMS) is specified by the programme of the experimental investigations to justify the engineering decisions on the safety of the design of the new generation NPP with the reactor WWER-640. The construction of KMS in a full volume will allow to conduct experimental investigations of all physical phenomena and processes, practically, occurring during the accidents on the NPPs with the reactor of WWER type and including the heat - mass exchange processes with low rates of the coolant, which is typical during the utilization of the passive safety systems, process during the accidents with a large leak, and also the complex intercommunicated processes in the reactor unit, passive safety systems and in the containment with the condition of long-term heat removal to the final absorber. KMS is being constructed at the Research Institute of Technology (NITI), Sosnovy Bor, Leningrad region, Russia. (orig.). 5 refs.

  19. Development of Coupled Interface System between the FADAS Code and a Source-term Evaluation Code XSOR for CANDU Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Son, Han Seong; Song, Deok Yong [ENESYS, Taejon (Korea, Republic of); Kim, Ma Woong; Shin, Hyeong Ki; Lee, Sang Kyu; Kim, Hyun Koon [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of)

    2006-07-01

    An accident prevention system is essential to the industrial security of nuclear industry. Thus, the more effective accident prevention system will be helpful to promote safety culture as well as to acquire public acceptance for nuclear power industry. The FADAS(Following Accident Dose Assessment System) which is a part of the Computerized Advisory System for a Radiological Emergency (CARE) system in KINS is used for the prevention against nuclear accident. In order to enhance the FADAS system more effective for CANDU reactors, it is necessary to develop the various accident scenarios and reliable database of source terms. This study introduces the construction of the coupled interface system between the FADAS and the source-term evaluation code aimed to improve the applicability of the CANDU Integrated Safety Analysis System (CISAS) for CANDU reactors.

  20. New research possibilities at the Budapest research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hargitai, T.; Vidovszky, I. [KFKI Atomic Energy Research Institute, H-1525 Budapest (Hungary)

    2001-07-01

    The Budapest Research Reactor is the first nuclear facility of Hungary. It was commissioned in 1959, reconstructed and upgraded in 1967 and 1986-92. The main purpose of the reactor is to serve neutron research. The reactor was extended by a liquid hydrogen type cold neutron source in 2000. The research possibilities are much improved by the CNS both in neutron scattering and neutron activation. (author)

  1. Technical specifications, Hanford production reactors

    Energy Technology Data Exchange (ETDEWEB)

    Gilbert, W.D. [comp.

    1962-06-25

    These technical specifications are applicable to the eight operating production reactor facilities, B, C, D, DR, F, H, KE, and KW. Covered are operating and performance restrictions and administrative procedures. Areas covered by the operating and performance restrictions are reactivity, reactor control and safety elements, power level, temperature and heat flux, reactor fuel loadings, reactor coolant systems, reactor confinement, test facilities, code compliance, and reactor scram set points. Administrative procedures include process control procedures, training programs, audits and inspections, and reports and records.

  2. Aurora Flash X-Ray Facility as a Source-Region EMP Simulator,

    Science.gov (United States)

    1981-01-01

    displacemeni- current can be neglected, but consideration of the two extre-mes is useful for a cualitative understanding.) When "local effects dominate...Use of the AURORA Machine as a Source-Region Simulator an Antenna CouL Lng Analysis Facility, Proceedings of NATO D_-fense Rer- ch Groui Seminar

  3. An improved cut-and-solve algorithm for the single-source capacitated facility location problem

    DEFF Research Database (Denmark)

    Gadegaard, Sune Lauth; Klose, Andreas; Nielsen, Lars Relund

    2017-01-01

    In this paper, we present an improved cut-and-solve algorithm for the single-source capacitated facility location problem. The algorithm consists of three phases. The first phase strengthens the integer program by a cutting plane algorithm to obtain a tight lower bound. The second phase uses a two...

  4. User`s Manual for the SOURCE1 and SOURCE2 Computer Codes: Models for Evaluating Low-Level Radioactive Waste Disposal Facility Source Terms (Version 2.0)

    Energy Technology Data Exchange (ETDEWEB)

    Icenhour, A.S.; Tharp, M.L.

    1996-08-01

    The SOURCE1 and SOURCE2 computer codes calculate source terms (i.e. radionuclide release rates) for performance assessments of low-level radioactive waste (LLW) disposal facilities. SOURCE1 is used to simulate radionuclide releases from tumulus-type facilities. SOURCE2 is used to simulate releases from silo-, well-, well-in-silo-, and trench-type disposal facilities. The SOURCE codes (a) simulate the degradation of engineered barriers and (b) provide an estimate of the source term for LLW disposal facilities. This manual summarizes the major changes that have been effected since the codes were originally developed.

  5. Preliminary Feasibility, Design, and Hazard Analysis of a Boiling Water Test Loop Within the Idaho National Laboratory Advanced Test Reactor National Scientific User Facility

    Energy Technology Data Exchange (ETDEWEB)

    Douglas M. Gerstner

    2009-05-01

    The Advanced Test Reactor (ATR) is a pressurized light-water reactor with a design thermal power of 250 MW. The principal function of the ATR is to provide a high neutron flux for testing reactor fuels and other materials. The ATR and its support facilities are located at the Idaho National Laboratory (INL). A Boiling Water Test Loop (BWTL) is being designed for one of the irradiation test positions within the. The objective of the new loop will be to simulate boiling water reactor (BWR) conditions to support clad corrosion and related reactor material testing. Further it will accommodate power ramping tests of candidate high burn-up fuels and fuel pins/rods for the commercial BWR utilities. The BWTL will be much like the pressurized water loops already in service in 5 of the 9 “flux traps” (region of enhanced neutron flux) in the ATR. The loop coolant will be isolated from the primary coolant system so that the loop’s temperature, pressure, flow rate, and water chemistry can be independently controlled. This paper presents the proposed general design of the in-core and auxiliary BWTL systems; the preliminary results of the neutronics and thermal hydraulics analyses; and the preliminary hazard analysis for safe normal and transient BWTL and ATR operation.

  6. Using Soluble Reactive Phosphorus and Ammonia to Identify Point Source Discharge from Large Livestock Facilities

    Science.gov (United States)

    Borrello, M. C.; Scribner, M.; Chessin, K.

    2013-12-01

    A growing body of research draws attention to the negative environmental impacts on surface water from large livestock facilities. These impacts are mostly in the form of excessive nutrient loading resulting in significantly decreased oxygen levels. Over-application of animal waste on fields as well as direct discharge into surface water from facilities themselves has been identified as the main contributor to the development of hypoxic zones in Lake Erie, Chesapeake Bay and the Gulf of Mexico. Some regulators claim enforcement of water quality laws is problematic because of the nature and pervasiveness of non-point source impacts. Any direct discharge by a facility is a violation of permits governed by the Clean Water Act, unless the facility has special dispensation for discharge. Previous research by the principal author and others has shown runoff and underdrain transport are the main mechanisms by which nutrients enter surface water. This study utilized previous work to determine if the effects of non-point source discharge can be distinguished from direct (point-source) discharge using simple nutrient analysis and dissolved oxygen (DO) parameters. Nutrient and DO parameters were measured from three sites: 1. A stream adjacent to a field receiving manure, upstream of a large livestock facility with a history of direct discharge, 2. The same stream downstream of the facility and 3. A stream in an area relatively unimpacted by large-scale agriculture (control site). Results show that calculating a simple Pearson correlation coefficient (r) of soluble reactive phosphorus (SRP) and ammonia over time as well as temperature and DO, distinguishes non-point source from point source discharge into surface water. The r value for SRP and ammonia for the upstream site was 0.01 while the r value for the downstream site was 0.92. The control site had an r value of 0.20. Likewise, r values were calculated on temperature and DO for each site. High negative correlations

  7. Target-ion-source system for RNB facility at KEK Tanashi

    CERN Document Server

    Jeong, S C; Kawakami, H; Miyatake, H; Ishida, Y; Tojyo, E; Oyaizu, M; Mizutani, S; Yoshikawa, N; Tanaka, M H; Kawashima, H; Fuchi, Y; Wada, M; Katayama, I; Nomura, T

    2002-01-01

    For the radioactive nuclear beam facility at KEK Tanashi based on an isotope separator on-line (ISOL), two different types of ion sources are currently employed: a single stage 6.4 GHz ECR and a surface ionization-type ion sources. Several proton-rich radioactive nuclear beams were developed for astrophysical interests. LiF powder target in a water-cooled Cu container is used for the production of Ne isotopes and CaF sub 2 mixed with fine graphite powder for Na isotopes. The ion beams are bunched at the exit of the ion source in order to reduce the beam loss in post acceleration.

  8. Screensaver: an open source lab information management system (LIMS) for high throughput screening facilities.

    Science.gov (United States)

    Tolopko, Andrew N; Sullivan, John P; Erickson, Sean D; Wrobel, David; Chiang, Su L; Rudnicki, Katrina; Rudnicki, Stewart; Nale, Jennifer; Selfors, Laura M; Greenhouse, Dara; Muhlich, Jeremy L; Shamu, Caroline E

    2010-05-18

    Shared-usage high throughput screening (HTS) facilities are becoming more common in academe as large-scale small molecule and genome-scale RNAi screening strategies are adopted for basic research purposes. These shared facilities require a unique informatics infrastructure that must not only provide access to and analysis of screening data, but must also manage the administrative and technical challenges associated with conducting numerous, interleaved screening efforts run by multiple independent research groups. We have developed Screensaver, a free, open source, web-based lab information management system (LIMS), to address the informatics needs of our small molecule and RNAi screening facility. Screensaver supports the storage and comparison of screening data sets, as well as the management of information about screens, screeners, libraries, and laboratory work requests. To our knowledge, Screensaver is one of the first applications to support the storage and analysis of data from both genome-scale RNAi screening projects and small molecule screening projects. The informatics and administrative needs of an HTS facility may be best managed by a single, integrated, web-accessible application such as Screensaver. Screensaver has proven useful in meeting the requirements of the ICCB-Longwood/NSRB Screening Facility at Harvard Medical School, and has provided similar benefits to other HTS facilities.

  9. Screensaver: an open source lab information management system (LIMS for high throughput screening facilities

    Directory of Open Access Journals (Sweden)

    Nale Jennifer

    2010-05-01

    Full Text Available Abstract Background Shared-usage high throughput screening (HTS facilities are becoming more common in academe as large-scale small molecule and genome-scale RNAi screening strategies are adopted for basic research purposes. These shared facilities require a unique informatics infrastructure that must not only provide access to and analysis of screening data, but must also manage the administrative and technical challenges associated with conducting numerous, interleaved screening efforts run by multiple independent research groups. Results We have developed Screensaver, a free, open source, web-based lab information management system (LIMS, to address the informatics needs of our small molecule and RNAi screening facility. Screensaver supports the storage and comparison of screening data sets, as well as the management of information about screens, screeners, libraries, and laboratory work requests. To our knowledge, Screensaver is one of the first applications to support the storage and analysis of data from both genome-scale RNAi screening projects and small molecule screening projects. Conclusions The informatics and administrative needs of an HTS facility may be best managed by a single, integrated, web-accessible application such as Screensaver. Screensaver has proven useful in meeting the requirements of the ICCB-Longwood/NSRB Screening Facility at Harvard Medical School, and has provided similar benefits to other HTS facilities.

  10. TMI-2 Reactor Building source term measurements: surfaces and basement water and sediment

    Energy Technology Data Exchange (ETDEWEB)

    McIsaac, C V; Keefer, D G

    1984-10-01

    Presented in this report are the results of radiochemical and elemental analyses performed on samples collected from the Three Mile Island Unit 2 Reactor Building from August 1979 to December 1983. The quantities of fission products and core materials that were measured on the external surfaces in the Reactor Building or in the water and sediment in its basement are summarized. Recent analysis results for access panels removed from the air cooling assembly and for liquid and particulate samples collected from the Reactor Building sump and reactor coolant drain tank are included in the report. Measurements show that 59% of the /sup 3/H, 2.7% of the /sup 90/Sr, 15% of the /sup 129/I, 20% of the /sup 131/I, and 42% of the /sup 137/Cs originally in the core at the time of the accident could be accounted for outside the core in the Reactor Building. With the exceptions of /sup 90/Sr and /sup 144/Ce, the vast majority of each radionuclide released was found dispersed in the water and sediment in the basement.

  11. The energy release and temperature field in the ultracold neutron source of the WWR-M reactor at the Petersburg Nuclear Physics Institute

    Energy Technology Data Exchange (ETDEWEB)

    Serebrov, A. P., E-mail: serebrov@pnpi.spb.ru; Kislitsin, B. V.; Onegin, M. S.; Lyamkin, V. A.; Prudnikov, D. V.; Ilatovskiy, V. A.; Orlov, S. P.; Kirsanov, G. A.; Fomin, A. K.; Filchenkova, D. V. [National Research Center Kurchatov Institute, Petersburg Nuclear Physics Institute (Russian Federation)

    2016-12-15

    Results of calculations of energy releases and temperature fields in the ultracold neutron source under design at the WWR-M reactor are presented. It is shown that, with the reactor power of 18 MW, the power of energy release in the 40-L volume of the source with superfluid helium will amount to 28.5 W, while 356 W will be released in a liquid-deuterium premoderator. The lead shield between the reactor core and the source reduces the radiative heat release by an order of magnitude. A thermal power of 22 kW is released in it, which is removed by passage of water. The distribution of temperatures in all components of the vacuum structure is presented, and the temperature does not exceed 100°C at full reactor power. The calculations performed make it possible to go to design of the source.

  12. Environmental impacts of nonfusion power systems. [Data on environmental effects of all power sources that may be competitive with fusion reactor power plants

    Energy Technology Data Exchange (ETDEWEB)

    Brouns, R.J.

    1976-09-01

    Data were collected on the environmental effects of power sources that may be competitive with future fusion reactor power plants. Data are included on nuclear power plants using HTGR, LMBR, GCFR, LMFBR, and molten salt reactors; fossil-fuel electric power plants; geothermal power plants; solar energy power plants, including satellite-based solar systems; wind energy power plants; ocean thermal gradient power plants; tidal energy power plants; and power plants using hydrogen and other synthetic fuels as energy sources.

  13. Upgrade of the BATMAN test facility for H{sup −} source development

    Energy Technology Data Exchange (ETDEWEB)

    Heinemann, B., E-mail: bernd.heinemann@ipp.mpg.de; Fröschle, M.; Falter, H.-D.; Fantz, U.; Franzen, P.; Kraus, W.; Nocentini, R.; Riedl, R.; Ruf, B. [Max-Planck-Institut für Plasmaphysik, Boltzmannstr. 2, 85748 Garching (Germany)

    2015-04-08

    The development of a radio frequency (RF) driven source for negative hydrogen ions for the neutral beam heating devices of fusion experiments has been successfully carried out at IPP since 1996 on the test facility BATMAN. The required ITER parameters have been achieved with the prototype source consisting of a cylindrical driver on the back side of a racetrack like expansion chamber. The extraction system, called “Large Area Grid” (LAG) was derived from a positive ion accelerator from ASDEX Upgrade (AUG) using its aperture size (ø 8 mm) and pattern but replacing the first two electrodes and masking down the extraction area to 70 cm2. BATMAN is a well diagnosed and highly flexible test facility which will be kept operational in parallel to the half size ITER source test facility ELISE for further developments to improve the RF efficiency and the beam properties. It is therefore planned to upgrade BATMAN with a new ITER-like grid system (ILG) representing almost one ITER beamlet group, namely 5 × 14 apertures (ø 14 mm). Additionally to the standard three grid extraction system a repeller electrode upstream of the grounded grid can optionally be installed which is positively charged against it by 2 kV. This is designated to affect the onset of the space charge compensation downstream of the grounded grid and to reduce the backstreaming of positive ions from the drift space backwards into the ion source. For magnetic filter field studies a plasma grid current up to 3 kA will be available as well as permanent magnets embedded into a diagnostic flange or in an external magnet frame. Furthermore different source vessels and source configurations are under discussion for BATMAN, e.g. using the AUG type racetrack RF source as driver instead of the circular one or modifying the expansion chamber for a more flexible position of the external magnet frame.

  14. Rotating-bed reactor as a power source for EM gun applications

    Energy Technology Data Exchange (ETDEWEB)

    Powell, J.; Botts, T.; Stickley, C.M.; Meth, S.

    1980-01-01

    Electromagnetic gun applications of the Rotating Bed Reactor (RBR) are examined. The RBR is a compact (approx. 1 m/sup 3/), (up to several thousand MW(th)), high-power reactor concept, capable of producing a high-temperature (up to approx. 300/sup 0/K) gas stream with a MHD generator coupled to it, the RBR can generate electric power (up to approx. 1000 MW(e)) in the pulsed or cw modes. Three EM gun applications are investigated: a rail gun thruster for orbit transfer, a rapid-fire EM gun for point defense, and a direct ground-to-space launch. The RBR appears suitable for all applications.

  15. Informal proposal for an Atomic Physics Facility at the National Synchrotron Light Source

    Energy Technology Data Exchange (ETDEWEB)

    Jones, K.W.; Johnson, B.M.; Meron, M.

    1986-01-01

    An Atomic Physics Facility (APF) for experiments that will use radiation from a superconducting wiggler on the NSLS X-13 port is described. The scientific justification for the APF is given and the elements of the facility are discussed. It is shown that it will be possible to conduct a uniquely varied set of experiments that can probe most aspects of atomic physics. A major component of the proposal is a heavy-ion storage ring capable of containing ions with energies of about 10 MeV/nucleon. The ring can be filled with heavy ions produced at the BNL MP Tandem Laboratory or from independent ion-source systems. A preliminary cost estimate for the facility is presented.

  16. Contamination Sources Effects Analysis (CSEA) - A Tool to Balance Cost/Schedule While Managing Facility Availability

    Science.gov (United States)

    Wilcox, Margaret

    2008-01-01

    A CSEA is similar to a Failure Modes Effects Analysis (FMEA). A CSEA tracks risk, deterrence, and occurrence of sources of contamination and their mitigation plans. Documentation is provided spanning mechanical and electrical assembly, precision cleaning, thermal vacuum bake-out, and thermal vacuum testing. These facilities all may play a role in contamination budgeting and reduction ultimately affecting test and flight. With a CSEA, visibility can be given to availability of these facilities, test sequencing and trade-offs. A cross-functional team including specialty engineering, contamination control, electrostatic dissipation, manufacturing, testing, and material engineering participate in an exercise that identifies contaminants and minimizes the complexity of scheduling these facilities considering their volatile schedules. Care can be taken in an efficient manner to insure correct cleaning processes are employed. The result is reduction in cycle time ("schedule hits"), reduced cost due to rework, reduced risk and improved communication and quality while achieving adherence to the Contamination Control Plan.

  17. A Methodology for the Integration of a Mechanistic Source Term Analysis in a Probabilistic Framework for Advanced Reactors

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-06-26

    GE Hitachi Nuclear Energy (GEH) and Argonne National Laboratory are currently engaged in a joint effort to modernize and develop probabilistic risk assessment (PRA) techniques for advanced non-light water reactors. At a high level, the primary outcome of this project will be the development of next-generation PRA methodologies that will enable risk-informed prioritization of safety- and reliability-focused research and development, while also identifying gaps that may be resolved through additional research. A subset of this effort is the development of PRA methodologies to conduct a mechanistic source term (MST) analysis for event sequences that could result in the release of radionuclides. The MST analysis seeks to realistically model and assess the transport, retention, and release of radionuclides from the reactor to the environment. The MST methods developed during this project seek to satisfy the requirements of the Mechanistic Source Term element of the ASME/ANS Non-LWR PRA standard. The MST methodology consists of separate analysis approaches for risk-significant and non-risk significant event sequences that may result in the release of radionuclides from the reactor. For risk-significant event sequences, the methodology focuses on a detailed assessment, using mechanistic models, of radionuclide release from the fuel, transport through and release from the primary system, transport in the containment, and finally release to the environment. The analysis approach for non-risk significant event sequences examines the possibility of large radionuclide releases due to events such as re-criticality or the complete loss of radionuclide barriers. This paper provides details on the MST methodology, including the interface between the MST analysis and other elements of the PRA, and provides a simplified example MST calculation for a sodium fast reactor.

  18. A multicharge ion source (Supernanogan) for the OLIS facility at ISAC/TRIUMF.

    Science.gov (United States)

    Jayamanna, K; Wight, G; Gallop, D; Dube, R; Jovicic, V; Laforge, C; Marchetto, M; Leross, M; Louie, D; Laplante, R; Laxdal, R; McDonald, M; Wiebe, G J; Wang, V; Yan, F

    2010-02-01

    The Off-Line Ion Source (OLIS) [K. Jayamanna, D. Yuan, T. Kuo, M. MacDonald, P. Schmor, and G. Dutto, Rev. Sci. Instrum. 67, 1061 (1996); K. Jayamanna, Rev. Sci. Instrum. 79, 02711 (2008)] facility consists of a high voltage terminal containing a microwave cusp ion source, either a surface ion source or a hybrid surface-arc discharge ion source [K. Jayamanna and C. Vockenhuber, Rev. Sci. Instrum. 79, 02C712 (2008)], and an electrostatic switch that allows the selection of any one of the sources without mechanical intervention. These sources provide a variety of +1 beams up to mass 30 for Isotope Separator and ACcelerator (ISAC) [R. E. Laxdal, Nucl. Instrum. Methods Phys. Res. B 204, 400 (2003)] experiments, commissioning the accelerators, setting up the radioactive experiments, and for tuning the beam lines. The radio frequency quadrupole (RFQ) [M. Marchetto, Z. T. Ang, K. Jayamanna, R. E. Laxdal, A. Mitra, and V. Zvyagintsev, Eur. Phys. J. Spec. Top. 150, 241 (2005)] injector accelerator is a constant velocity machine designed to accept only 2 keV/u and the source extraction energy is limited to 60 kV. Further stripping is then needed downstream of the RFQ to inject the beam into the drift tube linac [M. Marchetto, Z. T. Ang, K. Jayamanna, R. E. Laxdal, A. Mitra, and V. Zvyagintsev, Eur. Phys. J. Spec. Top. 150, 241 (2005)] accelerator that requires A/q up to 6. Base on this constraints a multicharge ion source capable to deliver beams above mass 30 with A/q up to 6 was needed in order to reach full capability of the ISAC facility. A Supernanogan [C. Bieth et al., Nucleonika 48, S93 (2003)] multicharge ion source was then purchased from Pantechnik and was installed in the OLIS terminal. Commissioning and performance of the Supernanogan with some results such as emittance dependence of the charge states as well as charge state efficiencies are presented.

  19. X-ray source development for EXAFS measurements on the National Ignition Facility

    Science.gov (United States)

    Coppari, F.; Thorn, D. B.; Kemp, G. E.; Craxton, R. S.; Garcia, E. M.; Ping, Y.; Eggert, J. H.; Schneider, M. B.

    2017-08-01

    Extended X-ray absorption Fine Structure (EXAFS) measurements require a bright, spectrally smooth, and broad-band x-ray source. In a laser facility, such an x-ray source can be generated by a laser-driven capsule implosion. In order to optimize the x-ray emission, different capsule types and laser irradiations have been tested at the National Ignition Facility (NIF). A crystal spectrometer is used to disperse the x-rays and high efficiency image plate detectors are used to measure the absorption spectra in transmission geometry. EXAFS measurements at the K-edge of iron at ambient conditions have been obtained for the first time on the NIF laser, and the requirements for optimization have been established.

  20. Low power microwave tests on RF gun prototype of the Iranian Light Source Facility

    Directory of Open Access Journals (Sweden)

    A Sadeghipanah

    2017-08-01

    Full Text Available In this paper, we introduce RF electron gun of Iranian Light Source Facility (ILSF pre-injection system. Design, fabrication and low-power microwave tests results of the prototype RF electron gun have been described in detail. This paper also explains the tuning procedure of the prototype RF electron gun to the desired resonant frequency. The outcomes of this project brighten the path to the fabrication of the RF electron gun by the local industries  

  1. Proposal on experience learning of a nuclear reactor for children in future. A basic concept on a nuclear reactor facility for demonstration and education

    Energy Technology Data Exchange (ETDEWEB)

    Murata, Takashi [Kyoto Univ., Graduate School of Energy Science, Kyoto (Japan); Yoshiki, Nobuya; Kinehara, Yoshiki; Nakagawa, Haruo

    2001-12-01

    The Science Council of Japan indicates in a proposal on R and D on nuclear energy forward the 21st Century that it is important to expand the educational object on nuclear energy from colleges and gradual schools to elementary, middle high schools. And, the Committee of Japan Nuclear Energy Industries also proposed that as an effort forward security of reliability and popularization of knowledge, completeness of learning chance on energy and nuclear energy in education such as usage of general learning time, concept on establishment of educational reactor for demonstration and experience, is essential. Here was described on a concept on establishment of nuclear reactor for demonstration and experience at objectives of common national peoples, which was based on results of searches and investigations carried out by authors and aimed to supply to a field to grow up a literary adequately and widely capable of judging various information on the peoples by focusing to effectiveness of empirical learning as a method of promoting corrective understanding of common citizens on high class technical system and by establishment of the reactor aiming at general education on nuclear energy at a place easily accessible by common citizens, such as large city. (G.K.)

  2. Space and Terrestrial Power System Integration Optimization Code BRMAPS for Gas Turbine Space Power Plants With Nuclear Reactor Heat Sources

    Science.gov (United States)

    Juhasz, Albert J.

    2007-01-01

    In view of the difficult times the US and global economies are experiencing today, funds for the development of advanced fission reactors nuclear power systems for space propulsion and planetary surface applications are currently not available. However, according to the Energy Policy Act of 2005 the U.S. needs to invest in developing fission reactor technology for ground based terrestrial power plants. Such plants would make a significant contribution toward drastic reduction of worldwide greenhouse gas emissions and associated global warming. To accomplish this goal the Next Generation Nuclear Plant Project (NGNP) has been established by DOE under the Generation IV Nuclear Systems Initiative. Idaho National Laboratory (INL) was designated as the lead in the development of VHTR (Very High Temperature Reactor) and HTGR (High Temperature Gas Reactor) technology to be integrated with MMW (multi-megawatt) helium gas turbine driven electric power AC generators. However, the advantages of transmitting power in high voltage DC form over large distances are also explored in the seminar lecture series. As an attractive alternate heat source the Liquid Fluoride Reactor (LFR), pioneered at ORNL (Oak Ridge National Laboratory) in the mid 1960's, would offer much higher energy yields than current nuclear plants by using an inherently safe energy conversion scheme based on the Thorium --> U233 fuel cycle and a fission process with a negative temperature coefficient of reactivity. The power plants are to be sized to meet electric power demand during peak periods and also for providing thermal energy for hydrogen (H2) production during "off peak" periods. This approach will both supply electric power by using environmentally clean nuclear heat which does not generate green house gases, and also provide a clean fuel H2 for the future, when, due to increased global demand and the decline in discovering new deposits, our supply of liquid fossil fuels will have been used up. This is

  3. Uncertainty Analysis of non-point source pollution control facilities design techniques in Korea

    Science.gov (United States)

    Lee, J.; Okjeong, L.; Gyeong, C. B.; Park, M. W.; Kim, S.

    2015-12-01

    The design of non-point sources control facilities in Korea is divided largely by the stormwater capture ratio, the stormwater load capture ratio, and the pollutant reduction efficiency of the facility. The stormwater capture ratio is given by a design formula as a function of the water quality treatment capacity, the greater the capacity, the more the amount of stormwater intercepted by the facility. The stormwater load capture ratio is defined as the ratio of the load entering the facility of the total pollutant load generated in the target catchment, and is given as a design formula represented by a function of the stormwater capture ratio. In order to estimate the stormwater capture ratio and load capture ratio, a lot of quantitative analysis of hydrologic processes acted in pollutant emission is required, but these formulas have been applied without any verification. Since systematic monitoring programs were insufficient, verification of these formulas was fundamentally impossible. However, recently the Korean ministry of Environment has conducted an long-term systematic monitoring project, and thus the verification of the formulas became possible. In this presentation, the stormwater capture ratio and load capture ratio are re-estimated using actual TP data obtained from long-term monitoring program at Noksan industrial complex located in Busan, Korea. Through the re-estimated process, the uncertainty included in the design process that has been applied until now will be shown in a quantitative extent. In addition, each uncertainty included in the stormwater capture ratio estimation and in the stormwater load capture ratio estimation will be expressed to quantify the relative impact on the overall non-point pollutant control facilities design process. Finally, the SWMM-Matlab interlocking module for model parameters estimation will be introduced. Acknowledgement This subject is supported by Korea Ministry of Environment as "The Eco Innovation Project : Non

  4. 10 CFR 171.16 - Annual fees: Materials licensees, holders of certificates of compliance, holders of sealed source...

    Science.gov (United States)

    2010-01-01

    ... Heap Leach facilities 31,200 (b) Basic In Situ Recovery facilities 29,700 (c) Expanded In Situ Recovery..., heap-leaching, ore buying stations, ion exchange facilities and in-processing of ores containing source... than power reactors and spent fuel storage/reactor decommissioning, the in-situ leach rulemaking, and...

  5. A facile and efficient method of enzyme immobilization on silica particles via Michael acceptor film coatings: immobilized catalase in a plug flow reactor.

    Science.gov (United States)

    Bayramoglu, Gulay; Arica, M Yakup; Genc, Aysenur; Ozalp, V Cengiz; Ince, Ahmet; Bicak, Niyazi

    2016-06-01

    A novel method was developed for facile immobilization of enzymes on silica surfaces. Herein, we describe a single-step strategy for generating of reactive double bonds capable of Michael addition on the surfaces of silica particles. This method was based on reactive thin film generation on the surfaces by heating of impregnated self-curable polymer, alpha-morpholine substituted poly(vinyl methyl ketone) p(VMK). The generated double bonds were demonstrated to be an efficient way for rapid incorporation of enzymes via Michael addition. Catalase was used as model enzyme in order to test the effect of immobilization methodology by the reactive film surface through Michael addition reaction. Finally, a plug flow type immobilized enzyme reactor was employed to estimate decomposition rate of hydrogen peroxide. The highly stable enzyme reactor could operate continuously for 120 h at 30 °C with only a loss of about 36 % of its initial activity.

  6. Characterization and quantification of an in-core neutron irradiation facility at a TRIGA II research reactor

    Science.gov (United States)

    Aghara, Sukesh; Charlton, William

    2006-07-01

    Experiments have been performed to characterize the neutron environment at an in-core TRIGA type nuclear research reactor. Steady-state thermal and epithermal neutron environment testing is important for many applications including, materials, electronics and biological cells. A well characterized neutron environment at a research reactor, including energy spectrum and spatial distribution, can be useful to many research communities and for educational research. This paper describes the characterization process and an application of exposing electronics to high neutron fluence.

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

    Energy Technology Data Exchange (ETDEWEB)

    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 is 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).

  8. Review of ion-source developments for radioactive ion-beam facilities

    CERN Document Server

    Lettry, Jacques

    1999-01-01

    The ion-sources dedicated to the production of radioactive ion beams (RIB) shall be highly efficient, selective and fast. This efficiency is mandatory since only limited amounts of radionuclides are produced. Chemical selectivity is needed to confine other elements near to the production site and to suppress isobaric contaminants. Eventually, the ion-source shall only decay the radioisotopes by a fraction of their half-life to reduce decay losses. The world wide spread RIB facilities came up with a large variety of solutions to meet part or all of these requirements such as: ion traps, surface, plasma, sputtering, electron cyclotron resonance and laser ion- sources. In this review, the latest developments are presented and their applications to charge states breeder systems proposed for post-acceleration are discussed. (59 refs).

  9. Occupational radiation exposure at commercial nuclear power reactors and other facilities 1994. Twenty-seventh annual report

    Energy Technology Data Exchange (ETDEWEB)

    Thomas, M.L.; Hagemeyer, D. [Science Applications International Corporation, Oak Ridge, TN (United States)

    1996-01-01

    This report summarizes the occupational exposure data that are maintained in the U.S. Nuclear Regulatory Commission`s (NRC) Radiation Exposure Information and Reporting System (REIRS). Annual reports for 1994 were received from a total of 303 NRC licensees, of which 109 were operators of nuclear power reactors in commercial operation. Compilations of the reports submitted by the 303 licensees indicated that 152,028 individuals were monitored, 79,780 of whom received a measurable dose. The collective dose incurred by these individuals was 24,740 person-cSv (person-rem){sup 2} which represents a 15% decrease from the 1993 value. The number of workers receiving a measurable dose also decreased, resulting in the average measurable dose of 0.31 cSv (rem) for 1994. The average measurable dose is defined to be the total collective dose (TEDE) divided by the number of workers receiving a measurable dose. These figures have been adjusted to account for transient reactor workers. In 1994, the annual collective dose per reactor for light water reactor licensees (LWRs) was 198 person-cSv (person-rem). This represents a 18% decrease from the 1993 value of 242 person-cSv (person-rem). The annual collective dose per reactor for boiling water reactors (BWRs) was 327 person-cSv (person-rem) and, for pressurized water reactors (PWRs), it was 131 person-cSv (person-rem). Analyses of transient worker data indicate that 18,178 individuals completed work assignments at two or more licensees during the monitoring year. The dose distributions are adjusted each year to account for the duplicate reporting of transient workers by multiple licensees. In 1994, the average measurable dose calculated from reported data was 0.28 cSv (rem). The corrected dose distribution resulted in an average measurable dose of 0.31 cSv (rem).

  10. Advanced Neutron Source Cross Section Libraries (ANSL-V): ENDF/B-V based multigroup cross-section libraries for advanced neutron source (ANS) reactor studies

    Energy Technology Data Exchange (ETDEWEB)

    Ford, W.E. III; Arwood, J.W.; Greene, N.M.; Moses, D.L.; Petrie, L.M.; Primm, R.T. III; Slater, C.O.; Westfall, R.M.; Wright, R.Q.

    1990-09-01

    Pseudo-problem-independent, multigroup cross-section libraries were generated to support Advanced Neutron Source (ANS) Reactor design studies. The ANS is a proposed reactor which would be fueled with highly enriched uranium and cooled with heavy water. The libraries, designated ANSL-V (Advanced Neutron Source Cross Section Libraries based on ENDF/B-V), are data bases in AMPX master format for subsequent generation of problem-dependent cross-sections for use with codes such as KENO, ANISN, XSDRNPM, VENTURE, DOT, DORT, TORT, and MORSE. Included in ANSL-V are 99-group and 39-group neutron, 39-neutron-group 44-gamma-ray-group secondary gamma-ray production (SGRP), 44-group gamma-ray interaction (GRI), and coupled, 39-neutron group 44-gamma-ray group (CNG) cross-section libraries. The neutron and SGRP libraries were generated primarily from ENDF/B-V data; the GRI library was generated from DLC-99/HUGO data, which is recognized as the ENDF/B-V photon interaction data. Modules from the AMPX and NJOY systems were used to process the multigroup data. Validity of selected data from the fine- and broad-group neutron libraries was satisfactorily tested in performance parameter calculations.

  11. Occupational radiation exposure at commercial nuclear power reactors and other facilities 1995: Twenty-eighth annual report. Volume 17

    Energy Technology Data Exchange (ETDEWEB)

    Thomas, M.L. [Nuclear Regulatory Commission, Washington, DC (United States). Div. of Regulatory Applications; Hagemeyer, D. [Science Applications International Corp., Oak Ridge, TN (United States)

    1997-01-01

    This report summarizes the occupational exposure data that are maintained in the US Nuclear Regulatory Commission`s (NRC) Radiation Exposure Information and Reporting System (REIRS). The bulk of the information contained in the report was compiled from the 1995 annual reports submitted by six of the seven categories of NRC licensees subject to the reporting requirements of 10 CFR 20.2206. Since there are no geologic repositories for high-level waste currently licensed, only six categories will be considered in this report. In 1995, the annual collective dose per reactor for light water reactor licensees (LWRs) was 199 person-cSv (person-rem). This is the same value that was reported for 1994. The annual collective dose per reactor for boiling water reactors (BWRs) was 256 person-cSv (person-rem) and, for pressurized water reactors (PWRs), it was 170 person-cSv (person-rem). Analyses of transient worker data indicate that 17,153 individuals completed work assignments at two or more licensees during the monitoring year. The dose distributions are adjusted each year to account for the duplicate reporting of transient workers by multiple licensees. In 1995, the average measurable dose calculated from reported data was 0.26 cSv (rem). The corrected dose distribution resulted in an average measurable dose of 0.32 cSv (rem).

  12. Post-closure safety assessment of near surface disposal facilities for disused sealed radioactive sources

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Seunghee; Kim, Juyoul, E-mail: gracemi@fnctech.com

    2017-03-15

    Highlights: • Post-closure safety assessment of near surface disposal facility for DSRS was performed. • Engineered vault and rock-cavern type were considered for normal and well scenario. • {sup 14}C, {sup 226}Ra, {sup 241}Am were primary nuclides contributing large portion of exposure dose. • Near surface disposal of DSRSs containing {sup 14}C, {sup 226}Ra and {sup 241}Am should be restricted. - Abstract: Great attention has been recently paid to the post-closure safety assessment of low- and intermediate-level radioactive waste (LILW) disposal facility for disused sealed radioactive sources (DSRSs) around the world. Although the amount of volume of DSRSs generated from industry, medicine and research and education organization was relatively small compared with radioactive wastes from commercial nuclear power plants, some DSRSs can pose a significant hazard to human health due to their high activities and long half-lives, if not appropriately managed and disposed. In this study, post-closure safety assessment was carried out for DSRSs generated from 1991 to 2014 in Korea in order to ensure long-term safety of near surface disposal facilities. Two kinds of disposal options were considered, i.e., engineered vault type disposal facility and rock-cavern type disposal facility. Rock-cavern type disposal facility has been under operation in Gyeongju city, republic of Korea since August 2015 and engineered vault type disposal facility will be constructed until December 2020 in the vicinity of rock-cavern disposal facility. Assessment endpoint was individual dose to the member of critical group, which was modeled by GoldSim, which has been widely used as probabilistic risk analysis software based on Monte Carlo simulation in the area of safety assessment of radioactive waste facilities. In normal groundwater scenario, the maximum exposure dose was extremely low, approximately 1 × 10{sup −7} mSv/yr, for both disposal options and satisfied the regulatory limit

  13. The current status of Kartini research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tri Wulan Tjiptono; Syarip

    1998-10-01

    The Kartini reactor reached the first criticality on January 25, 1979. In the first three years, the reactor power is limited up to 50 kW thermal power and on July 1, 1982 has been increased to 100 kW. It has been used as experiments facility by researcher of Atomic Energy National Agency and students of the Universities. Three beam tubes used as experiments facilities, the first, is used as a neutron source for H{sub 2}O-Natural Uranium Subcritical Assembly, the second, is developed for neutron radiography facility and the third, is used for gamma radiography facility. The other facilities are rotary rack and two pneumatic transfer systems, one for delayed neutron counting system and the other for the new Neutron Activation Analysis (NAA) facility. The rotary rack used for isotope production for NAA purpose (for long time irradiation), the delayed neutron counting system used for analysis the Uranium contents of the ores and the new NAA is provided for short live elements analysis. In the last three years the Reactor Division has a joint use program with the Nuclear Component and Engineering Center in research reactor instrumentation and control development. (author)

  14. Bacterial community dynamics in a biodenitrification reactor packed with polylactic acid/poly (3-hydroxybutyrate-co-3-hydroxyvalerate) blend as the carbon source and biofilm carrier.

    Science.gov (United States)

    Qiu, Tianlei; Xu, Ying; Gao, Min; Han, Meilin; Wang, Xuming

    2017-05-01

    While heterotrophic denitrification has been widely used for treating such nitrogen-rich wastewater, it requires the use of additional carbon sources. With fluctuations in the nitrate concentration in the influent, controlling the C/N ratio to avoid carbon breakthrough becomes difficult. To overcome this obstacle, solid-phase denitrification (SPD) using biodegradable polymers has been used, where denitrification and carbon source biodegradation depend on microorganisms growing within the reactor. However, the microbial community dynamics in continuous-flow SPD reactors have not been fully elucidated yet. Here, we aimed to study bacterial community dynamics in a biodenitrification reactor packed with a polylactic acid/poly (3-hydroxybutyrate-co-3-hydroxyvalerate) (PLA/PHBV) blend as the carbon source and biofilm carrier. A lab-scale denitrifying reactor filled with a PLA/PHBV blend was used. With 85 mg/L of influent NO3-N concentration and a hydraulic retention time (HRT) of 2.5 h, more than 92% of the nitrate was removed. The bacterial community of inoculated activated sludge had the highest species richness in all samples. Bacterial species diversity in the reactor first decreased and then increased to a stable level. Diaphorobacter species were predominant in the reactor after day 24. In total, 178 clones were retrieved from the 16S rRNA gene clone library constructed from the biofilm samples in the reactor at 62 days of operation, and 80.9% of the clones were affiliated with Betaproteobacteria. Of these, 97.2% were classified into phylotypes corresponding to Diaphorobacter nitroreducens strain NA10B with 99% sequence similarity. Diaphorobacter, Rhizobium, Acidovorax, Rubrivivax, Azospira, Thermomonas, and Acidaminobacter constituted the biofilm microflora in the stably running reactor. Copyright © 2016 The Society for Biotechnology, Japan. Published by Elsevier B.V. All rights reserved.

  15. Magnetic measurement of Iranian Light Source Facility quadrupole storage ring prototype

    Directory of Open Access Journals (Sweden)

    Y Radkhorrami

    2017-08-01

    Full Text Available Magnetic Measurement Lab is one of the most significant divisions of Research and Development (R&D Lab of Iranian Light Source Facility. The main duty of this lab is to measure and check qualification of the accelerator magnets, including permanent and electromagnets, being applied in Iran for the fisrt time. The ILSF measurement lab consists of precise measurement equipment, in proportion  to synchrotron needs, such as Hall Effect probe measurement bench, rotating coil and Helmholtz coil. Recently, the lab has been provided with Hall probe measurement bench and uncompensated rotating coil and has made it possible to measure prototype magnets. In this article, the results of measuring quadrupole prototype are studied using Hall probe and rotating coil, to determine and compare errors in measuring multipole magnets and their sources

  16. An Experimental Facility to Validate Ground Source Heat Pump Optimisation Models for the Australian Climate

    Directory of Open Access Journals (Sweden)

    Yuanshen Lu

    2017-01-01

    Full Text Available Ground source heat pumps (GSHPs are one of the most widespread forms of geothermal energy technology. They utilise the near-constant temperature of the ground below the frost line to achieve energy-efficiencies two or three times that of conventional air-conditioners, consequently allowing a significant offset in electricity demand for space heating and cooling. Relatively mature GSHP markets are established in Europe and North America. GSHP implementation in Australia, however, is limited, due to high capital price, uncertainties regarding optimum designs for the Australian climate, and limited consumer confidence in the technology. Existing GSHP design standards developed in the Northern Hemisphere are likely to lead to suboptimal performance in Australia where demand might be much more cooling-dominated. There is an urgent need to develop Australia’s own GSHP system optimisation principles on top of the industry standards to provide confidence to bring the GSHP market out of its infancy. To assist in this, the Queensland Geothermal Energy Centre of Excellence (QGECE has commissioned a fully instrumented GSHP experimental facility in Gatton, Australia, as a publically-accessible demonstration of the technology and a platform for systematic studies of GSHPs, including optimisation of design and operations. This paper presents a brief review on current GSHP use in Australia, the technical details of the Gatton GSHP facility, and an analysis on the observed cooling performance of this facility to date.

  17. Development of BWR regional stability experimental facility SIRIUS-F, which simulates thermal-hydraulics-neutronics coupling in reactor core, and stability evaluation of ABWR

    Energy Technology Data Exchange (ETDEWEB)

    Masahiro Furuya; Fumio Inada [Central Research Institute of Electric Power Industry (CRIEPI) 2-11-1 Iwado-kita, Komae, Tokyo 201-8511 (Japan); Takanori Fukahori [Global Nuclear Fuel Japan (GNF-J) 2-3-1 Uchikawa, Yokosuka, Kanagawa 239-0836 (Japan); Shinya Mizokami [Tokyo Electric Power Company (TEPCO) 1-1-3 Uchisaiwai-cho, Chiyoda, Tokyo 100-0011 (Japan)

    2005-07-01

    Full text of publication follows: The SIRIUS-F facility was designed and constructed for highly accurate simulation of channel, core-wide and regional instabilities of an ABWR. A real-time simulation is performed for the modal-point kinetics of reactor neutronics and fuel-rod conduction on the basis of a measured void fraction in a reactor core section of the facility. A noise analysis method was performed to calculate decay ratios from dominant poles of transfer function on the basis of the AR method by applying time series of a core inlet flow rate. By utilizing this method, one can estimate stability at any specific operating point online without assuming excess conservative conditions. Channel and regional stability experiments were conducted for a wide range of operating conditions including maximum power points along the minimum pump speed line and the natural circulation line of the ABWR. The decay ratios and the resonance frequencies are in good agreement with those from the design analysis code, ODYSY. The SIRIUS-F experimental results demonstrated stability characteristics such as a stabilizing effect of the power, and reviled a sufficiently large stability margin even under hypothetical conditions of power enlargement. (authors)

  18. Dependence of the source performance on plasma parameters at the BATMAN test facility

    Science.gov (United States)

    Wimmer, C.; Fantz, U.

    2015-04-01

    The investigation of the dependence of the source performance (high jH-, low je) for optimum Cs conditions on the plasma parameters at the BATMAN (Bavarian Test MAchine for Negative hydrogen ions) test facility is desirable in order to find key parameters for the operation of the source as well as to deepen the physical understanding. The most relevant source physics takes place in the extended boundary layer, which is the plasma layer with a thickness of several cm in front of the plasma grid: the production of H-, its transport through the plasma and its extraction, inevitably accompanied by the co-extraction of electrons. Hence, a link of the source performance with the plasma parameters in the extended boundary layer is expected. In order to characterize electron and negative hydrogen ion fluxes in the extended boundary layer, Cavity Ring-Down Spectroscopy and Langmuir probes have been applied for the measurement of the H- density and the determination of the plasma density, the plasma potential and the electron temperature, respectively. The plasma potential is of particular importance as it determines the sheath potential profile at the plasma grid: depending on the plasma grid bias relative to the plasma potential, a transition in the plasma sheath from an electron repelling to an electron attracting sheath takes place, influencing strongly the electron fraction of the bias current and thus the amount of co-extracted electrons. Dependencies of the source performance on the determined plasma parameters are presented for the comparison of two source pressures (0.6 Pa, 0.45 Pa) in hydrogen operation. The higher source pressure of 0.6 Pa is a standard point of operation at BATMAN with external magnets, whereas the lower pressure of 0.45 Pa is closer to the ITER requirements (p ≤ 0.3 Pa).

  19. Dependence of the source performance on plasma parameters at the BATMAN test facility

    Energy Technology Data Exchange (ETDEWEB)

    Wimmer, C.; Fantz, U. [Max-Planck-Institut für Plasmaphysik, Boltzmannstr. 2, 85748 Garching (Germany)

    2015-04-08

    The investigation of the dependence of the source performance (high j{sub H{sup −}}, low j{sub e}) for optimum Cs conditions on the plasma parameters at the BATMAN (Bavarian Test MAchine for Negative hydrogen ions) test facility is desirable in order to find key parameters for the operation of the source as well as to deepen the physical understanding. The most relevant source physics takes place in the extended boundary layer, which is the plasma layer with a thickness of several cm in front of the plasma grid: the production of H{sup −}, its transport through the plasma and its extraction, inevitably accompanied by the co-extraction of electrons. Hence, a link of the source performance with the plasma parameters in the extended boundary layer is expected. In order to characterize electron and negative hydrogen ion fluxes in the extended boundary layer, Cavity Ring-Down Spectroscopy and Langmuir probes have been applied for the measurement of the H{sup −} density and the determination of the plasma density, the plasma potential and the electron temperature, respectively. The plasma potential is of particular importance as it determines the sheath potential profile at the plasma grid: depending on the plasma grid bias relative to the plasma potential, a transition in the plasma sheath from an electron repelling to an electron attracting sheath takes place, influencing strongly the electron fraction of the bias current and thus the amount of co-extracted electrons. Dependencies of the source performance on the determined plasma parameters are presented for the comparison of two source pressures (0.6 Pa, 0.45 Pa) in hydrogen operation. The higher source pressure of 0.6 Pa is a standard point of operation at BATMAN with external magnets, whereas the lower pressure of 0.45 Pa is closer to the ITER requirements (p ≤ 0.3 Pa)

  20. The New Cold Neutron Radiography Facility (CNRF) at the Mianyang Research Reactor of the China Academy of Engineering Physics

    Science.gov (United States)

    Bin, Tang; Heyong, Huo; Ke, Tang; Rogers, John; Haste, Martin; Christodoulou, Marios

    A new cold neutron radiography beamline has been designed and constructed for the Mianyang reactor at the Institute of Nuclear Physics and Chemistry of the China Academy of Engineering Physics. This paper describes the components of the system and demonstrates the achievable image resolution.

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

    Science.gov (United States)

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

    1987-01-01

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

  2. Accident source terms for pressurized water reactors with high-burnup cores calculated using MELCOR 1.8.5.

    Energy Technology Data Exchange (ETDEWEB)

    Gauntt, Randall O.; Powers, Dana Auburn; Ashbaugh, Scott G.; Leonard, Mark Thomas; Longmire, Pamela

    2010-04-01

    In this study, risk-significant pressurized-water reactor severe accident sequences are examined using MELCOR 1.8.5 to explore the range of fission product releases to the reactor containment building. Advances in the understanding of fission product release and transport behavior and severe accident progression are used to render best estimate analyses of selected accident sequences. Particular emphasis is placed on estimating the effects of high fuel burnup in contrast with low burnup on fission product releases to the containment. Supporting this emphasis, recent data available on fission product release from high-burnup (HBU) fuel from the French VERCOR project are used in this study. The results of these analyses are treated as samples from a population of accident sequences in order to employ approximate order statistics characterization of the results. These trends and tendencies are then compared to the NUREG-1465 alternative source term prescription used today for regulatory applications. In general, greater differences are observed between the state-of-the-art calculations for either HBU or low-burnup (LBU) fuel and the NUREG-1465 containment release fractions than exist between HBU and LBU release fractions. Current analyses suggest that retention of fission products within the vessel and the reactor coolant system (RCS) are greater than contemplated in the NUREG-1465 prescription, and that, overall, release fractions to the containment are therefore lower across the board in the present analyses than suggested in NUREG-1465. The decreased volatility of Cs2MoO4 compared to CsI or CsOH increases the predicted RCS retention of cesium, and as a result, cesium and iodine do not follow identical behaviors with respect to distribution among vessel, RCS, and containment. With respect to the regulatory alternative source term, greater differences are observed between the NUREG-1465 prescription and both HBU and LBU predictions than exist between HBU and LBU

  3. Using single-chamber microbial fuel cells as renewable power sources of electro-Fenton reactors for organic pollutant treatment

    KAUST Repository

    Zhu, Xiuping

    2013-05-01

    Electro-Fenton reactions can be very effective for organic pollutant degradation, but they typically require non-sustainable electrical power to produce hydrogen peroxide. Two-chamber microbial fuel cells (MFCs) have been proposed for pollutant treatment using Fenton-based reactions, but these types of MFCs have low power densities and require expensive membranes. Here, more efficient dual reactor systems were developed using a single-chamber MFC as a low-voltage power source to simultaneously accomplish H2O2 generation and Fe2+ release for the Fenton reaction. In tests using phenol, 75±2% of the total organic carbon (TOC) was removed in the electro-Fenton reactor in one cycle (22h), and phenol was completely degraded to simple and readily biodegradable organic acids. Compared to previously developed systems based on two-chamber MFCs, the degradation efficiency of organic pollutants was substantially improved. These results demonstrate that this system is an energy-efficient and cost-effective approach for industrial wastewater treatment of certain pollutants. © 2013 Elsevier B.V.

  4. 40 CFR 63.5692 - How do I know if my boat manufacturing facility is a new source or an existing source?

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 12 2010-07-01 2010-07-01 true How do I know if my boat manufacturing... AIR POLLUTANTS FOR SOURCE CATEGORIES National Emission Standards for Hazardous Air Pollutants for Boat Manufacturing What the Subpart Covers § 63.5692 How do I know if my boat manufacturing facility is a new source...

  5. Occupational Radiation Exposure at Commercial Nuclear Power Reactors and Other Facilities 2010, Prepared for the Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, May 2012

    Energy Technology Data Exchange (ETDEWEB)

    D. E. Lewis D. A. Hagemeyer Y. U. McCormick

    2012-07-07

    This report summarizes the occupational exposure data that are maintained in the U.S. Nuclear Regulatory Commission’s (NRC) Radiation Exposure Information and Reporting System (REIRS). The bulk of the information contained in the report was compiled from the 2010 annual reports submitted by five of the seven categories of NRC licensees subject to the reporting requirements of 10 CFR 20.2206. Because there are no geologic repositories for high-level waste currently licensed and no NRC-licensed low-level waste disposal facilities currently in operation, only five categories will be considered in this report. The annual reports submitted by these licensees consist of radiation exposure records for each monitored individual. These records are analyzed for trends and presented in this report in terms of collective dose and the distribution of dose among the monitored individuals. Annual reports for 2010 were received from a total of 190 NRC licensees. The summation of reports submitted by the 190 licensees indicated that 192,424 individuals were monitored, 81,961 of whom received a measurable dose. When adjusted for transient workers who worked at more than one licensee during the year, there were actually 142,471 monitored individuals and 62,782 who received a measurable dose. The collective dose incurred by these individuals was 10,617 person-rem, which represents a 12% decrease from the 2009 value. This decrease was primarily due to the decrease in collective dose at commercial nuclear power reactors, as well as a decrease in the collective dose for most of the other categories of NRC licensees. The number of individuals receiving a measurable dose also decreased, resulting in an average measurable dose of 0.13 rem for 2010. The average measurable dose is defined as the total effective dose equivalent (TEDE) divided by the number of individuals receiving a measurable dose. In calendar year 2010, the average annual collective dose per reactor for light water reactor

  6. Advanced Neutron Source (ANS) Project progress report

    Energy Technology Data Exchange (ETDEWEB)

    McBee, M.R.; Chance, C.M. (eds.) (Oak Ridge National Lab., TN (USA)); Selby, D.L.; Harrington, R.M.; Peretz, F.J. (Oak Ridge National Lab., TN (USA))

    1990-04-01

    This report discusses the following topics on the advanced neutron source: quality assurance (QA) program; reactor core development; fuel element specification; corrosion loop tests and analyses; thermal-hydraulic loop tests; reactor control concepts; critical and subcritical experiments; material data, structural tests, and analysis; cold source development; beam tube, guide, and instrument development; hot source development; neutron transport and shielding; I C research and development; facility concepts; design; and safety.

  7. Studies Related to the Oregon State University High Temperature Test Facility: Scaling, the Validation Matrix, and Similarities to the Modular High Temperature Gas-Cooled Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Richard R. Schultz; Paul D. Bayless; Richard W. Johnson; William T. Taitano; James R. Wolf; Glenn E. McCreery

    2010-09-01

    The Oregon State University (OSU) High Temperature Test Facility (HTTF) is an integral experimental facility that will be constructed on the OSU campus in Corvallis, Oregon. The HTTF project was initiated, by the U.S. Nuclear Regulatory Commission (NRC), on September 5, 2008 as Task 4 of the 5 year High Temperature Gas Reactor Cooperative Agreement via NRC Contract 04-08-138. Until August, 2010, when a DOE contract was initiated to fund additional capabilities for the HTTF project, all of the funding support for the HTTF was provided by the NRC via their cooperative agreement. The U.S. Department of Energy (DOE) began their involvement with the HTTF project in late 2009 via the Next Generation Nuclear Plant project. Because the NRC interests in HTTF experiments were only centered on the depressurized conduction cooldown (DCC) scenario, NGNP involvement focused on expanding the experimental envelope of the HTTF to include steady-state operations and also the pressurized conduction cooldown (PCC). Since DOE has incorporated the HTTF as an ingredient in the NGNP thermal-fluids validation program, several important outcomes should be noted: 1. The reference prismatic reactor design, that serves as the basis for scaling the HTTF, became the modular high temperature gas-cooled reactor (MHTGR). The MHTGR has also been chosen as the reference design for all of the other NGNP thermal-fluid experiments. 2. The NGNP validation matrix is being planned using the same scaling strategy that has been implemented to design the HTTF, i.e., the hierarchical two-tiered scaling methodology developed by Zuber in 1991. Using this approach a preliminary validation matrix has been designed that integrates the HTTF experiments with the other experiments planned for the NGNP thermal-fluids verification and validation project. 3. Initial analyses showed that the inherent power capability of the OSU infrastructure, which only allowed a total operational facility power capability of 0.6 MW, is

  8. Accident source terms for boiling water reactors with high burnup cores.

    Energy Technology Data Exchange (ETDEWEB)

    Gauntt, Randall O.; Powers, Dana Auburn; Leonard, Mark Thomas

    2007-11-01

    The primary objective of this report is to provide the technical basis for development of recommendations for updates to the NUREG-1465 Source Term for BWRs that will extend its applicability to accidents involving high burnup (HBU) cores. However, a secondary objective is to re-examine the fundamental characteristics of the prescription for fission product release to containment described by NUREG-1465. This secondary objective is motivated by an interest to understand the extent to which research into the release and behaviors of radionuclides under accident conditions has altered best-estimate calculations of the integral response of BWRs to severe core damage sequences and the resulting radiological source terms to containment. This report, therefore, documents specific results of fission product source term analyses that will form the basis for the HBU supplement to NUREG-1465. However, commentary is also provided on observed differences between the composite results of the source term calculations performed here and those reflected NUREG-1465 itself.

  9. Evaluation of marine algae as a source of biogas in a two-stage anaerobic reactor system

    Energy Technology Data Exchange (ETDEWEB)

    Vergara-Fernandez, Alberto; Vargas, Gisela [Escuela de Ingenieria Ambiental, Facultad de Ingenieria, Universidad Catolica de Temuco, Manuel Montt 56, Casilla 15-D, Temuco (Chile); Alarcon, Nelson [Departamento de Ingenieria Quimica, Facultad de Ingenieria y Ciencias Geologicas, Universidad Catolica del Norte (Chile); Velasco, Antonio [Centro Nacional de Investigacion y Capacitacion Ambiental del Instituto Nacional de Ecologia (CENICA-INE), Av. San Rafael Atlixco 186, Col. Vicentina, Del. Iztapalapa, 09340, Mexico, DF (Mexico)

    2008-04-15

    The marine algae are considered an important biomass source; however, their utilization as energy source is still low around the world. The technical feasibility of marine algae utilization as a source of renewable energy was studied to laboratory scale. The anaerobic digestion of Macrocystis pyrifera, Durvillea antarctica and their blend 1:1 (w/w) was evaluated in a two-phase anaerobic digestion system, which consisted of an anaerobic sequencing batch reactor (ASBR) and an upflow anaerobic filter (UAF). The results show that 70% of the total biogas produced in the system was generated in the UAF, and both algae species have similar biogas productions of 180.4({+-}1.5) mL g{sup -1} dry algae d{sup -1}, with a methane concentration around 65%. The same methane content was observed in biogas yield of algae blend; however, a lower biogas yield was obtained. In conclusion, either algae species or their blend can be utilized to produce methane gas in a two-phase digestion system. (author)

  10. First production of ultracold neutrons with a solid deuterium source at the pulsed reactor TRIGA Mainz⋆

    Science.gov (United States)

    Frei, A.; Sobolev, Yu.; Altarev, I.; Eberhardt, K.; Gschrey, A.; Gutsmiedl, E.; Hackl, R.; Hampel, G.; Hartmann, F. J.; Heil, W.; Kratz, J. V.; Lauer, Th.; Liźon Aguilar, A.; Müller, A. R.; Paul, S.; Pokotilovski, Yu.; Schmid, W.; Tassini, L.; Tortorella, D.; Trautmann, N.; Trinks, U.; Wiehl, N.

    2007-11-01

    The production rates of ultracold neutrons (UCN) with a solid deuterium converter have been measured at the pulsed reactor TRIGA Mainz. Exposed to a thermal neutron fluence of ensuremath ˜ 1\\cdot 10^{13} n·cm^-2·pulse^-1, the number of detected very cold and ultracold neutrons ranges up to 200 000 at 7mol of solid deuterium (sD2) in combination with a pre-moderator (mesitylene). About 50% of the measured neutrons can be assigned to UCN with energies E of ensuremath V_F(sD_2)≤ E ≤ V_F(guide) where V F( sD 2) = 105 neV and V F( guide) = 190 neV are the Fermi potentials of the sD2 converter and our stainless steel neutron guides, respectively. Thermal cycling of solid deuterium, which was frozen out from the gas phase, considerably improved the UCN yield, in particular at higher amounts of sD2.

  11. Development of a microsecond X-ray protein footprinting facility at the Advanced Light Source.

    Science.gov (United States)

    Gupta, Sayan; Celestre, Richard; Petzold, Christopher J; Chance, Mark R; Ralston, Corie

    2014-07-01

    X-ray footprinting (XF) is an important structural biology tool used to determine macromolecular conformations and dynamics of both nucleic acids and proteins in solution on a wide range of timescales. With the impending shut-down of the National Synchrotron Light Source, it is ever more important that this tool continues to be developed at other synchrotron facilities to accommodate XF users. Toward this end, a collaborative XF program has been initiated at the Advanced Light Source using the white-light bending-magnet beamlines 5.3.1 and 3.2.1. Accessibility of the microsecond time regime for protein footprinting is demonstrated at beamline 5.3.1 using the high flux density provided by a focusing mirror in combination with a micro-capillary flow cell. It is further reported that, by saturating samples with nitrous oxide, the radiolytic labeling efficiency is increased and the imprints of bound versus bulk water can be distinguished. These results both demonstrate the suitability of the Advanced Light Source as a second home for the XF experiment, and pave the way for obtaining high-quality structural data on complex protein samples and dynamics information on the microsecond timescale.

  12. Bright x-ray stainless steel K-shell source development at the National Ignition Facility

    Energy Technology Data Exchange (ETDEWEB)

    May, M. J.; Fournier, K. B.; Colvin, J. D.; Barrios, M. A.; Dewald, E. L.; Moody, J.; Patterson, J. R.; Schneider, M.; Widmann, K. [Lawrence Livermore National Laboratory, P.O. Box 808 L170, Livermore, California 94551 (United States); Hohenberger, M.; Regan, S. P. [Laboratory for Laser Energetics, University of Rochester, Rochester, New York 14623 (United States)

    2015-06-15

    High x-ray conversion efficiency (XRCE) K-shell sources are being developed for high energy density experiments for use as backlighters and for the testing of materials exposed to high x-ray fluxes and fluences. Recently, sources with high XRCE in the K-shell x-ray energy range of iron and nickel were investigated at the National Ignition Facility (NIF). The x-ray conversion efficiency in the 5–9 keV spectral range was determined to be 6.8% ± 0.3%. These targets were 4.1 mm diameter, 4 mm tall hollow epoxy tubes having a 50 μm thick wall supporting a tube of 3 to 3.5 μm thick stainless steel. The NIF laser deposited ∼460 kJ of 3ω light into the target in a 140 TW, 3.3 ns square pulse. The absolute x-ray emission of the source was measured by two calibrated Dante x-ray spectrometers. Time resolved images filtered for the Fe K-shell were recorded to follow the heating of the target. Time integrated high-resolution spectra were recorded in the K-shell range.

  13. Reactor Safety Planning for Prometheus Project, for Naval Reactors Information

    Energy Technology Data Exchange (ETDEWEB)

    P. Delmolino

    2005-05-06

    The purpose of this letter is to submit to Naval Reactors the initial plan for the Prometheus project Reactor Safety work. The Prometheus project is currently developing plans for cold physics experiments and reactor prototype tests. These tests and facilities may require safety analysis and siting support. In addition to the ground facilities, the flight reactor units will require unique analyses to evaluate the risk to the public from normal operations and credible accident conditions. This letter outlines major safety documents that will be submitted with estimated deliverable dates. Included in this planning is the reactor servicing documentation and shipping analysis that will be submitted to Naval Reactors.

  14. Low-Energy Microfocus X-Ray Source for Enhanced Testing Capability in the Stray Light Facility

    Science.gov (United States)

    Gaskin, Jessica; O'Dell, Stephen; Kolodziejczak, Jeff

    2015-01-01

    Research toward high-resolution, soft x-ray optics (mirrors and gratings) necessary for the next generation large x-ray observatories requires x-ray testing using a low-energy x-ray source with fine angular size (energy microfocus (approximately 0.1 mm spot) x-ray source from TruFocus Corporation that mates directly to the Stray Light Facility (SLF). MSFC X-ray Astronomy team members are internationally recognized for their expertise in the development, fabrication, and testing of grazing-incidence optics for x-ray telescopes. One of the key MSFC facilities for testing novel x-ray instrumentation is the SLF. This facility is an approximately 100-m-long beam line equipped with multiple x-ray sources and detectors. This new source adds to the already robust compliment of instrumentation, allowing MSFC to support additional internal and community x-ray testing needs.

  15. Pharmaceutical Formulation Facilities as Sources of Opioids and Other Pharmaceuticals to Wastewater Treatment Plant Effluents

    Science.gov (United States)

    2010-01-01

    Facilities involved in the manufacture of pharmaceutical products are an under-investigated source of pharmaceuticals to the environment. Between 2004 and 2009, 35 to 38 effluent samples were collected from each of three wastewater treatment plants (WWTPs) in New York and analyzed for seven pharmaceuticals including opioids and muscle relaxants. Two WWTPs (NY2 and NY3) receive substantial flows (>20% of plant flow) from pharmaceutical formulation facilities (PFF) and one (NY1) receives no PFF flow. Samples of effluents from 23 WWTPs across the United States were analyzed once for these pharmaceuticals as part of a national survey. Maximum pharmaceutical effluent concentrations for the national survey and NY1 effluent samples were generally 400 μg/L. Maximum concentrations of oxycodone (1700 μg/L) and metaxalone (3800 μg/L) in samples from NY3 effluent exceeded 1000 μg/L. Three pharmaceuticals (butalbital, carisoprodol, and oxycodone) in samples of NY2 effluent had median concentrations ranging from 2 to 11 μg/L. These findings suggest that current manufacturing practices at these PFFs can result in pharmaceuticals concentrations from 10 to 1000 times higher than those typically found in WWTP effluents. PMID:20521847

  16. Emergency preparedness source term development for the Office of Nuclear Material Safety and Safeguards-Licensed Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Sutter, S.L.; Mishima, J.; Ballinger, M.Y.; Lindsey, C.G.

    1984-08-01

    In order to establish requirements for emergency preparedness plans at facilities licensed by the Office of Nuclear Materials Safety and Safeguards, the Nuclear Regulatory Commission (NRC) needs to develop source terms (the amount of material made airborne) in accidents. These source terms are used to estimate the potential public doses from the events, which, in turn, will be used to judge whether emergency preparedness plans are needed for a particular type of facility. Pacific Northwest Laboratory is providing the NRC with source terms by developing several accident scenarios for eleven types of fuel cycle and by-product operations. Several scenarios are developed for each operation, leading to the identification of the maximum release considered for emergency preparedness planning (MREPP) scenario. The MREPP scenarios postulated were of three types: fire, tornado, and criticality. Fire was significant at oxide fuel fabrication, UF/sub 6/ production, radiopharmaceutical manufacturing, radiopharmacy, sealed source manufacturing, waste warehousing, and university research and development facilities. Tornadoes were MREPP events for uranium mills and plutonium contaminated facilities, and criticalities were significant at nonoxide fuel fabrication and nuclear research and development facilities. Techniques for adjusting the MREPP release to different facilities are also described.

  17. The Advanced Photon Source: A national synchrotron radiation research facility at Argonne National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-10-01

    The vision of the APS sprang from prospective users, whose unflagging support the project has enjoyed throughout the decade it has taken to make this facility a reality. Perhaps the most extraordinary aspect of synchrotron radiation research, is the extensive and diverse scientific makeup of the user community. From this primordial soup of scientists exchanging ideas and information, come the collaborative and interdisciplinary accomplishments that no individual alone could produce. So, unlike the solitary Roentgen, scientists are engaged in a collective and dynamic enterprise with the potential to see and understand the structures of the most complex materials that nature or man can produce--and which underlie virtually all modern technologies. This booklet provides scientists and laymen alike with a sense of both the extraordinary history of x-rays and the knowledge they have produced, as well as the potential for future discovery contained in the APS--a source a million million times brighter than the Roentgen tube.

  18. Design and construction of a pre-injector for the Iranian Light Source Facility

    Directory of Open Access Journals (Sweden)

    A Sadeghipanah

    2015-09-01

    Full Text Available Every synchrotron accelerator requires a pre-injector for primary injection of the electrons into the booster ring. The Iranian Light Source Facility (ILSF pre-injector is a 150 MeV S-band linear accelerator with a thermionic cathode RF gun. The design of the pre-injector lattice and its beam dynamics calculation results together with the design of RF gun, alpha magnet, quadrupole magnets and linear accelerator structures are described in this article. The measurement results of the RF gun prototype fabricated in Iran demonstrate a dimension error less than 20 μm and a surface roughness of less than 0.8 μm

  19. Numerical analysis of the beam position monitor pickup for the Iranian light source facility

    Science.gov (United States)

    Shafiee, M.; Feghhi, S. A. H.; Rahighi, J.

    2017-03-01

    In this paper, we describe the design of a button type Beam Position Monitor (BPM) for the low emittance storage ring of the Iranian Light Source Facility (ILSF). First, we calculate sensitivities, induced power and intrinsic resolution based on solving Laplace equation numerically by finite element method (FEM), in order to find the potential at each point of BPM's electrode surface. After the optimization of the designed BPM, trapped high order modes (HOM), wakefield and thermal loss effects are calculated. Finally, after fabrication of BPM, it is experimentally tested by using a test-stand. The results depict that the designed BPM has a linear response in the area of 2×4 mm2 inside the beam pipe and the sensitivity of 0.080 and 0.087 mm-1 in horizontal and vertical directions. Experimental results also depict that they are in a good agreement with numerical analysis.

  20. Numerical analysis of the beam position monitor pickup for the Iranian light source facility

    Energy Technology Data Exchange (ETDEWEB)

    Shafiee, M., E-mail: mehdish@ipm.ir [Radiation Applications Department, Shahid Beheshti University, G. C., Tehran (Iran, Islamic Republic of); Feghhi, S.A.H. [Radiation Applications Department, Shahid Beheshti University, G. C., Tehran (Iran, Islamic Republic of); Rahighi, J. [Iranian Light Source Facility (ILSF), Institute for Research in Fundamental Sciences (IPM), Tehran (Iran, Islamic Republic of)

    2017-03-01

    In this paper, we describe the design of a button type Beam Position Monitor (BPM) for the low emittance storage ring of the Iranian Light Source Facility (ILSF). First, we calculate sensitivities, induced power and intrinsic resolution based on solving Laplace equation numerically by finite element method (FEM), in order to find the potential at each point of BPM's electrode surface. After the optimization of the designed BPM, trapped high order modes (HOM), wakefield and thermal loss effects are calculated. Finally, after fabrication of BPM, it is experimentally tested by using a test-stand. The results depict that the designed BPM has a linear response in the area of 2×4 mm{sup 2} inside the beam pipe and the sensitivity of 0.080 and 0.087 mm{sup −1} in horizontal and vertical directions. Experimental results also depict that they are in a good agreement with numerical analysis.

  1. Demonstration of a long pulse X-ray source at the National Ignition Facility

    Science.gov (United States)

    May, M. J.; Opachich, Y. P.; Kemp, G. E.; Colvin, J. D.; Barrios, M. A.; Widmann, K. W.; Fournier, K. B.; Hohenberger, M.; Albert, F.; Regan, S. P.

    2017-04-01

    A long duration high fluence x-ray source has been developed at the National Ignition Facility (NIF). The target was a 14.4 mm tall, 4.1 mm diameter, epoxy walled, gas filled pipe. Approximately 1.34 MJ from the NIF laser was used to heat the mixture of (55:45) Kr:Xe at 1.2 atm (˜5.59 mg/cm3) to emit in a fairly isotropic radiant intensity of 400-600 GW/sr from the Ephoton = 3-7 keV spectral range for a duration of ≈ 14 ns. The HYDRA simulated radiant intensities were in reasonable agreement with experiments but deviated at late times.

  2. General design of the International Fusion Materials Irradiation Facility deuteron injector: Source and beam linea)

    Science.gov (United States)

    Gobin, R.; Blideanu, V.; Bogard, D.; Bourdelle, G.; Chauvin, N.; Delferrière, O.; Girardot, P.; Jannin, J. L.; Langlois, S.; Loiseau, D.; Pottin, B.; Rousse, J.-Y.; Senée, F.

    2010-02-01

    In the framework of the International Fusion Materials Irradiation Facility-Engineering Validation and Engineering Design Activities (IFMIF-EVEDA) project, CEA/IRFU is in charge of the design and realization of the 140 mA cw deuteron Injector. The electron cyclotron resonance ion source operates at 2.45 GHz and a 4 electrode extraction system has been chosen. A 2 solenoid beam line, together with a high space charge compensation have been optimized for a proper beam injection in the 175 MHz radio frequency quadrupole. The injector will be tested with proton and deuteron beam production either in pulsed mode or in cw mode on the CEA-Saclay site before to be shipped to Japan. Special attention was paid to neutron emission due to (d,D) reaction. In this paper, the general IFMIF Injector design is reported, pointing out beam dynamics, radioprotection, diagnostics, and mechanical aspects.

  3. Impact of the High Flux Isotope Reactor HEU to LEU Fuel Conversion on Cold Source Nuclear Heat Generation Rates

    Energy Technology Data Exchange (ETDEWEB)

    Chandler, David [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2014-03-01

    Under the sponsorship of the US Department of Energy National Nuclear Security Administration, staff members at the Oak Ridge National Laboratory have been conducting studies to determine whether the High Flux Isotope Reactor (HFIR) can be converted from high enriched uranium (HEU) fuel to low enriched uranium (LEU) fuel. As part of these ongoing studies, an assessment of the impact that the HEU to LEU fuel conversion has on the nuclear heat generation rates in regions of the HFIR cold source system and its moderator vessel was performed and is documented in this report. Silicon production rates in the cold source aluminum regions and few-group neutron fluxes in the cold source moderator were also estimated. Neutronics calculations were performed with the Monte Carlo N-Particle code to determine the nuclear heat generation rates in regions of the HFIR cold source and its vessel for the HEU core operating at a full reactor power (FP) of 85 MW(t) and the reference LEU core operating at an FP of 100 MW(t). Calculations were performed with beginning-of-cycle (BOC) and end-of-cycle (EOC) conditions to bound typical irradiation conditions. Average specific BOC heat generation rates of 12.76 and 12.92 W/g, respectively, were calculated for the hemispherical region of the cold source liquid hydrogen (LH2) for the HEU and LEU cores, and EOC heat generation rates of 13.25 and 12.86 W/g, respectively, were calculated for the HEU and LEU cores. Thus, the greatest heat generation rates were calculated for the EOC HEU core, and it is concluded that the conversion from HEU to LEU fuel and the resulting increase of FP from 85 MW to 100 MW will not impact the ability of the heat removal equipment to remove the heat deposited in the cold source system. Silicon production rates in the cold source aluminum regions are estimated to be about 12.0% greater at BOC and 2.7% greater at EOC for the LEU core in comparison to the HEU core. Silicon is aluminum s major transmutation product and

  4. The technological safety in facilities that manage radioactive sources; La seguridad tecnologica en instalaciones que manejan fuentes radiactivas

    Energy Technology Data Exchange (ETDEWEB)

    Lizcano, D., E-mail: david.lizcano@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2014-10-15

    The sealed radioactive sources are used inside a wide range of applications in the medicine, industry and investigation around the world. These sources can contain a great radionuclides variety, exhibiting a wide spectrum of activities and radiological half lives. This way, we can find pattern sources of radionuclides as Americium-241, Plutonium-238, Plutonium-239, Thorium-228 and Thorium-230, etc., with some activity of kBq in research laboratories, Iridium-192 and Cesium-137 sources used in brachytherapy with GBq activities, until sources with P Bq activities in industrial irradiators of Cobalt-60 and Cesium-137. This document approach the physical safety that entities like the IAEA recommends for the facilities that contain sealed sources, especially the measures that are taking in the Instituto Nacional de Investigaciones Nucleares (ININ) and others government facilities. (Author)

  5. Co-digestion and model simulations of source separated municipal organic waste with cattle manure under batch and continuously stirred tank reactors

    DEFF Research Database (Denmark)

    Tsapekos, Panagiotis; Kougias, Panagiotis; Kuthiala, Sidhant

    2018-01-01

    This study investigates the co-digestion of source separated municipal organic waste (SSMOW), pretreated using a biopulper, and cattle manure both in batch and continuous stirred tank reactors. The optimum co-digestion feeding mixture was consisted of 90% SSMOW and 10% cattle manure on organic...

  6. NUCLEAR REACTOR

    Science.gov (United States)

    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.

  7. Partial nitritation of stored source-separated urine by granular activated sludge in a sequencing batch reactor.

    Science.gov (United States)

    Chen, Liping; Yang, Xiaoxiao; Tian, Xiujun; Yao, Song; Li, Jiuyi; Wang, Aimin; Yao, Qian; Peng, Dangcong

    2017-12-01

    The combination of partial nitritation (PN) and anaerobic ammonium oxidation (anammox) has been proposed as an ideal process for nitrogen removal from source-separated urine, while the high organic matters in urine cause instability of single-stage PN-anammox process. This study aims to remove the organic matters and partially nitrify the nitrogen in urine, producing an ammonium/nitrite solution suitable for anammox. The organic matters in stored urine were used as the electron donors to achieve 40% total nitrogen removal in nitritation-denitrification process in a sequencing batch reactor (SBR). Granular aggregates were observed and high mixed liquor suspended solids (9.5 g/L) were maintained in the SBR. Around 70-75% ammonium was oxidized to nitrite under the volumetric loading rates of 3.23 kg chemical oxygen demand (COD)/(m(3) d) and 1.86 kg N/(m(3) d), respectively. The SBR produced an ammonium/nitrite solution free of biodegradable organic matters, with a NO2(-)-N:NH4(+)-N of 1.24 ± 0.13. Fluorescence in situ hybridization images showed that Nitrosomonas-like ammonium-oxidizing bacteria, accounting for 7.2% of total bacteria, located in the outer layer (25 μm), while heterotrophs distributed homogeneously throughout the granular aggregates. High concentrations of free ammonia and nitrous acids in the reactor severely inhibited the growth of nitrite-oxidizing bacteria, resulting in their absence in the granular sludge. The microbial diversity analysis indicated Proteobacteria was the predominant phylum, in which Pseudomonas was the most abundant genus.

  8. Chemical and Radiochemical Constituents in Water from Wells in the Vicinity of the Naval Reactors Facility, Idaho National Engineering and Environmental Laboratory, Idaho, 1996

    Energy Technology Data Exchange (ETDEWEB)

    L. L. Knobel; R. C. Bartholomay; B. J. Tucker; L. M. Williams (USGS)

    1999-10-01

    The U.S. Geological Survey, in response to a request from the U.S. Department of Energy's Pittsburgh Naval Reactors Office, Idaho Branch Office (IBO), samples water from 13 wells during 1996 as part of a long-term project to monitor water quality to the Snake River Plain aquifer in the vicinity of the Naval Reactors Facility (NRF), Idaho National Engineering and Environmental Laboratory, Idaho. The IBO requires information about the mobility of radionuclide- and chemical-waste constituents in the Snake River Plain aquifer. Waste-constituent mobility is determined principally by (1) the rate and direction of ground-water flow; (2) the locations, quantities, and methods of waste disposal; (3) waste-constituents chemistry; and (4) the geochemical processes taking place in the aquifer. The purpose of the data-collection program is to provide IBO with water-chemistry data to evaluate the effect of NRF activities on the water quality of the Snake River Plain aquifer. Water samples were analyzed for naturally occurring constituents and man-made contaminants.

  9. Chemical constituents in water from wells in the vicinity of the Naval Reactors Facility, Idaho National Engineering Laboratory, Idaho, 1991--93

    Energy Technology Data Exchange (ETDEWEB)

    Tucker, B.J.; Knobel, L.L.; Bartholomay, R.C.

    1995-11-01

    The US Geological Survey, in response to a request from the US Department of Energy`s Pittsburgh Naval Reactors Office, Idaho Branch Office, sampled 14 wells during 1991--93 as part of a long-term project to monitor water quality of the Snake River Plain aquifer in the vicinity of the Naval Reactors Facility, Idaho National Engineering Laboratory, Idaho. Water samples were analyzed for manmade contaminants and naturally occurring constituents. One hundred sixty-one samples were collected from 10 ground-water monitoring wells and 4 production wells. Twenty-one quality-assurance samples also were collected and analyzed; 2 were blank samples and 19 were replicate samples. The two blank samples contained concentrations of six inorganic constituents that were slightly greater than the laboratory reporting levels (the smallest measured concentration of a constituent that can be reported using a given analytical method). Concentrations of other constituents in the blank samples were less than their respective reporting levels. The 19 replicate samples and their respective primary samples generated 614 pairs of analytical results for a variety of chemical and radiochemical constituents. Of the 614 data pairs, 588 were statistically equivalent at the 95% confidence level; about 96% of the analytical results were in agreement. Two pairs of turbidity measurements were not evaluated because of insufficient information and one primary sample collected in January 1992 contained tentatively identified organic compounds when the replicate sample did not.

  10. Neutron Flux Monitoring Based on Blind Source Separation Algorithms in Moroccan TRIGA MARK II Reactor

    Directory of Open Access Journals (Sweden)

    Hanane Arahmane

    2017-01-01

    Full Text Available We present an overview of fission chamber’s functioning modes, theoretical aspects of the nonnegative matrix factorization methods, and the opportunities that offer neutron data processing in order to achieve neutron flux monitoring tasks. Indeed, it is a part of research project that aimed at applying Blind Source Separation methods for in-core and ex-core neutron flux monitoring while analyzing the outputs of fission chamber. The latter could be used as a key issue for control, fuel management, safety concerns, and material irradiation experiments. The Blind Source Separation methods had been used in many scientific fields such as biomedical engineering and telecommunications. Recently, they were used for gamma spectrometry data processing. The originality of this research work is to apply these powerful methods to process the fission chamber output signals. We illustrated the effectiveness of this tool using simulated fission chamber signals.

  11. Fission Product Transport and Source Terms in HTRs: Experience from AVR Pebble Bed Reactor

    OpenAIRE

    Rainer Moormann

    2008-01-01

    Fission products deposited in the coolant circuit outside of the active core play a dominant role in source term estimations for advanced small pebble bed HTRs, particularly in design basis accidents (DBA). The deposited fission products may be released in depressurization accidents because present pebble bed HTR concepts abstain from a gas tight containment. Contamination of the circuit also hinders maintenance work. Experiments, performed from 1972 to 88 on the AVR, an experimental ...

  12. RF power source for the compact linear collider test facility (CTF3)

    CERN Document Server

    McMonagle, G; Brown, Peter; Carron, G; Hanni, R; Mourier, J; Rossat, G; Syratchev, I V; Tanner, L; Thorndahl, L

    2004-01-01

    The CERN CTF3 facility will test and demonstrate many vital components of CLIC (Compact Linear Collider). This paper describes the pulsed RF power source at 2998.55 MHz for the drive-beam accelerator (DBA), which produces a beam with an energy of 150 MeV and a current of 3.5 Amps. Where possible, existing equipment from the LEP preinjector, especially the modulators and klystrons, is being used and upgraded to achieve this goal. A high power RF pulse compression system is used at the output of each klystron, which requires sophisticated RF phase programming on the low level side to achieve the required RF pulse. In addition to the 3 GHz system two pulsed RF sources operating at 1.5 GHz are being built. The first is a wide-band, low power, travelling wave tube (TWT) for the subharmonic buncher (SHB) system that produces a train of "phase coded" subpulses as part of the injector scheme. The second is a high power narrow band system to produce 20 MW RF power to the 1.5 GHz RF deflectors in the delay loop situate...

  13. Reanalysis of the Gas-cooled fast reactor experiments at the zero power facility Proteus - Spectral indices

    Science.gov (United States)

    Perret, G.; Pattupara, R. M.; Girardin, G.; Chawla, R.

    2013-03-01

    PROTEUS is a zero power reactor at the Paul Scherrer Institute which has been employed during the 1970's to study experimentally the physics of the gas-cooled fast reactor. Reaction rate distributions, flux spectrum and reactivity effects have been measured in several configurations featuring PuO2/UO2 fuel, absorbers, large iron shields, and thorium oxide and thorium metal fuel either distributed quasihomogeneously in the reference PuO2/UO2 lattice or introduced in the form of radial and axial blanket zones. This papers focus on the spectral indices - including fission and capture in 232Th and 237Np - measured in the reference PuO2/UO2 lattices and their predictions with an MCNPX model specially developed for the PROTEUS-GCFR core. Predictions were obtained with JEFF-3.1 and -3.11, ENDF/B-VII.0 and VII.1, and JENDL-3.3 and -4.0. A general good agreement was demonstrated. The ratio of 232Th fission to 239Pu fission, however, was under-predicted by 8.7±2.1% and 6.5±2.1% using ENDF/B-VII.0 and VII.1, respectively. Finally, the capture rates in 237Np tended to be underpredicted by the JEFF and JENDL libraries, although the new cross section in JEFF-3.1.1 slightly improved the 237Np capture to 239Pu fission results (3.4±2.4%).

  14. On the possible use of the MASURCA reactor as a flexible, high-intensity, fast neutron beam facility

    Science.gov (United States)

    Dioni, Luca; Jacqmin, Robert; Sumini, Marco; Stout, Brian

    2017-09-01

    In recent work [1, 2], we have shown that the MASURCA research reactor could be used to deliver a fairly-intense continuous fast neutron beam to an experimental room located next to the reactor core. As a consequence of the MASURCA favorable characteristics and diverse material inventories, the neutron beam intensity and spectrum can be further tailored to meet the users' needs, which could be of interest for several applications. Monte Carlo simulations have been performed to characterize in detail the extracted neutron (and photon) beam entering the experimental room. These numerical simulations were done for two different bare cores: A uranium metallic core (˜30% 235U enriched) and a plutonium oxide core (˜25% Pu fraction, ˜78% 239Pu). The results show that the distinctive resonance energy structures of the two core leakage spectra are preserved at the channel exit. As the experimental room is large enough to house a dedicated set of neutron spectrometry instruments, we have investigated several candidate neutron spectrum measurement techniques, which could be implemented to guarantee well-defined, repeatable beam conditions to users. Our investigation also includes considerations regarding the gamma rays in the beams.

  15. Helium heater design for the helium direct cycle component test facility. [for gas-cooled nuclear reactor power plant

    Science.gov (United States)

    Larson, V. R.; Gunn, S. V.; Lee, J. C.

    1975-01-01

    The paper describes a helium heater to be used to conduct non-nuclear demonstration tests of the complete power conversion loop for a direct-cycle gas-cooled nuclear reactor power plant. Requirements for the heater include: heating the helium to a 1500 F temperature, operating at a 1000 psia helium pressure, providing a thermal response capability and helium volume similar to that of the nuclear reactor, and a total heater system helium pressure drop of not more than 15 psi. The unique compact heater system design proposed consists of 18 heater modules; air preheaters, compressors, and compressor drive systems; an integral control system; piping; and auxiliary equipment. The heater modules incorporate the dual-concentric-tube 'Variflux' heat exchanger design which provides a controlled heat flux along the entire length of the tube element. The heater design as proposed will meet all system requirements. The heater uses pressurized combustion (50 psia) to provide intensive heat transfer, and to minimize furnace volume and heat storage mass.

  16. Application of the Modified Source Multiplication (MSM) Technique to Subcritical Reactivity Worth Measurements in Thermal and Fast Reactor Systems

    Science.gov (United States)

    Blaise, Patrick; Mellier, Frédéric; Fougeras, Philippe

    2011-06-01

    The Amplified Source Multiplication (ASM) method and its improved Modified Source Multiplication (MSM) method have been widely used in the CEA's EOLE and MASURCA critical facilities over the past decades for the determination of reactivity worths by using fission chambers in subcritical configurations. The ASM methodology uses relatively simple relationships between count rates of efficient miniature fission chambers located in slightly subcritical reference and perturbed configurations. While this method works quite well for small reactivity variations, the raw results need to be corrected to take into account the flux perturbation at the fission chamber location. This is performed by applying to the measurement a correction factor called MSM. This paper describes in detail both methodologies, with their associated uncertainties. Applications on absorber cluster worth in the MISTRAL-4 full MOX mock-up core and the last core loaded in MASURCA show the importance of the MSM correction on raw ASM data.

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

    Energy Technology Data Exchange (ETDEWEB)

    De Bruyn, D.; Engelen, J. [Belgian Nuclear Research Centre SCK CEN, Boeretang 200, 2400 Mol (Belgium); Ortega, A.; Aguado, M. P. [Empresarios Agrupados A.I.E., Magallanes 3, 28015 Madrid (Spain)

    2012-07-01

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

  18. Extraction simulations and emittance measurements of a Holifield Radioactive Ion Beam Facility electron beam plasma source for radioactive ion beams

    Energy Technology Data Exchange (ETDEWEB)

    Mendez, A. J. II; Liu, Y. [Holifield Radioactive Ion Beam Facility, Oak Ridge National Laboratory, Oak Ridge, Tennessee 37831 (United States)

    2010-02-15

    The Holifield Radioactive Ion Beam Facility (HRIBF) at Oak Ridge National Laboratory has a variety of ion sources used to produce radioactive ion beams (RIBs). Of these, the workhorse is an electron beam plasma (EBP) ion source. The recent addition of a second RIB injector, the Injector for Radioactive Ion Species 2 (IRIS2), for the HRIBF tandem accelerator prompted new studies of the optics of the beam extraction from the EBP source. The source was modeled using SIMION V8.0, and results will be presented, including comparison of the emittances as predicted by simulation and as measured at the HRIBF offline ion source test facilities. Also presented will be the impact on phase space shape resulting from extraction optics modifications implemented at IRIS2.

  19. The Potential and Utilization of Unused Energy Sources for Large-Scale Horticulture Facility Applications under Korean Climatic Conditions

    Directory of Open Access Journals (Sweden)

    In Tak Hyun

    2014-07-01

    Full Text Available As the use of fossil fuel has increased, not only in construction, but also in agriculture due to the drastic industrial development in recent times, the problems of heating costs and global warming are getting worse. Therefore, introduction of more reliable and environmentally-friendly alternative energy sources has become urgent and the same trend is found in large-scale horticulture facilities. In this study, among many alternative energy sources, we investigated the reserves and the potential of various different unused energy sources which have infinite potential, but are nowadays wasted due to limitations in their utilization. In addition, we utilized available unused energy as a heat source for a heat pump in a large-scale horticulture facility and analyzed its feasibility through EnergyPlus simulation modeling. Accordingly, the discharge flow rate from the Fan Coil Unit (FCU in the horticulture facility, the discharge air temperature, and the return temperature were analyzed. The performance and heat consumption of each heat source were compared with those of conventional boilers. The result showed that the power load of the heat pump was decreased and thus the heat efficiency was increased as the temperature of the heat source was increased. Among the analyzed heat sources, power plant waste heat which had the highest heat source temperature consumed the least electric energy and showed the highest efficiency.

  20. Construction of a photochemical reactor combining a CCD spectrophotometer and a LED radiation source.

    Science.gov (United States)

    Gombár, Melinda; Józsa, Éva; Braun, Mihály; Ősz, Katalin

    2012-10-01

    An inexpensive photoreactor using LED light sources and a fibre-optic CCD spectrophotometer as a detector was built by designing a special cell holder for standard 1.000 cm cuvettes. The use of this device was demonstrated by studying the aqueous photochemical reaction of 2,5-dichloro-1,4-benzoquinone. The developed method combines the highly quantitative data collection of CCD spectrophotometers with the possibility of illuminating the sample independently of the detecting light beam, which is a substantial improvement of the method using diode array spectrophotometers as photoreactors.

  1. Experimental investigation of iodine removal and containment depressurization in containment spray system test facility of 700 MWe Indian pressurized heavy water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Jain, Manish [Department of Mechanical Engineering, I.I.T., Bombay, Powai, Mumbai (India); Kandar, T.K.; Vhora, S.F.; Mohan, Nalini [Directorate of Technology Development, Nuclear Power Corporation of India Limited, Mumbai (India); Iyer, K.N. [Department of Mechanical Engineering, I.I.T., Bombay, Powai, Mumbai (India); Prabhu, S.V., E-mail: svprabhu@iitb.ac.in [Department of Mechanical Engineering, I.I.T., Bombay, Powai, Mumbai (India)

    2017-05-15

    Highlights: • Depressurization rate in a scaled down vessel filled with air and steam is studied. • Iodine removal rate in a scaled down vessel filled with steam/air is investigated. • Effect of SMD and vessel pressure on depressurization rate is studied. • Depressurization rate decreases with the increase in the droplet size (590 μm – 1 mm) • Decrease in pressure and iodine concentration with time follow exponential trend. - Abstract: As an additional safety measure in the new 700 MWe Indian pressurized heavy water reactors, the first of a kind system called containment Spray System is introduced. The system is designed to cater/mitigate the conditions after design basis accidents i.e., loss of coolant accident and main steam line break. As a contribution to the safety analysis of condition following loss-of-coolant accidents, experiments are carried out to establish the performance of the system. The loss of coolant is simulated by injecting saturated steam and iodine vapors into the containment vessel in which air is enclosed at atmospheric and room temperature, and then the steam-air mixture is cooled by sprays of water. The effect of water spray on the containment vessel pressure and the iodine scrubbing in a scaled down facility is investigated for the containment spray system of Indian pressurized heavy water reactors. The experiments are carried out in the scaled down vessel of the diameter of 2.0 m and height of 3.5 m respectively. Experiments are conducted with water at room temperature as the spray medium. Two different initial vessel pressure i.e. 0.7 bar and 1.0 bar are chosen for the studies as they are nearing the loss of coolant accident & main steam line break pressures in Indian pressurized heavy water reactors. These pressures are chosen based on the containment resultant pressures after a design basis accident. The transient temperature and pressure distribution of the steam in the vessel are measured during the depressurization

  2. VIVO Open Source Software: Connecting Facilities to Promote Discovery and Further Research.

    Science.gov (United States)

    Gross, M. B.; Rowan, L. R.; Mayernik, M. S.; Daniels, M. D.; Stott, D.; Allison, J.; Maull, K. E.; Krafft, D. B.; Khan, H.

    2016-12-01

    EarthCollab (http://earthcube.org/group/earthcollab), a National Science Foundation (NSF) EarthCube Building Block project, has adapted an open source semantic web application, VIVO, for use within the earth science domain. EarthCollab is a partnership between UNAVCO, an NSF facility supporting research through geodetic services, the Earth Observing Laboratory (EOL) at the National Center for Atmospheric Research (NCAR), and Cornell University, where VIVO was created to highlight the scholarly output of researchers at universities. Two public sites have been released: Connect UNAVCO (connect.unavco.org) and Arctic Data Connects (vivo.eol.ucar.edu). The core VIVO software and ontology have been extended to work better with concepts necessary for capturing work within UNAVCO's and EOL's province such as principal investigators for continuous GPS/GNSS stations at UNAVCO and keywords describing cruise datasets at EOL. The sites increase discoverability of large and diverse data archives by linking data with people, research, and field projects. Disambiguation is a major challenge when using VIVO and open data when "anyone can say anything about anything." Concepts and controlled vocabularies help to build consistent and easily searchable connections within VIVO. We use aspects of subject heading services such as FAST and LOC, as well as AGU and GSA fields of research and subject areas to reveal connections, especially with VIVO instances at other institutions. VIVO works effectively with persistent IDs and the projects strive to utilize publication and data DOIs, ORCIDs for people, and ISNI and GRID for organizations. ORCID, an open source project, is very useful for disambiguation and unlike other identifier systems for people developed by publishers, makes public data available via an API. VIVO utilizes Solr and Freemarker, which are open source search engine and templating technologies, respectively. Additionally, a handful of popular open source libraries and

  3. Updates to the Generation of Physics Data Inputs for MAMMOTH Simulations of the Transient Reactor Test Facility - FY2016

    Energy Technology Data Exchange (ETDEWEB)

    Ortensi, Javier [Idaho National Lab. (INL), Idaho Falls, ID (United States); Baker, Benjamin Allen [Idaho National Lab. (INL), Idaho Falls, ID (United States); Schunert, Sebastian [Idaho National Lab. (INL), Idaho Falls, ID (United States); Wang, Yaqi [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gleicher, Frederick Nathan [Idaho National Lab. (INL), Idaho Falls, ID (United States); DeHart, Mark David [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-06-01

    The INL is currently evolving the modeling and simulation (M&S) capability that will enable improved core operation as well as design and analysis of TREAT experiments. This M&S capability primarily uses MAMMOTH, a reactor physics application being developed under Multi-physics Object Oriented Simulation Environment (MOOSE) framework. MAMMOTH allows the coupling of a number of other MOOSE-based applications. This second year of work has been devoted to the generation of a deterministic reference solution for the full core, the preparation of anisotropic diffusion coefficients, the testing of the SPH equivalence method, and the improvement of the control rod modeling. In addition, this report includes the progress made in the modeling of the M8 core configuration and experiment vehicle since January of this year.

  4. The economic and community impacts of closing Hanford's N Reactor and nuclear materials production facilities

    Energy Technology Data Exchange (ETDEWEB)

    Scott, M.J.; Belzer, D.B.; Nesse, R.J.; Schultz, R.W.; Stokowski, P.A.; Clark, D.C.

    1987-08-01

    This study discusses the negative economic impact on local cities and counties and the State of Washington of a permanent closure of nuclear materials production at the Hanford Site, located in the southeastern part of the state. The loss of nuclear materials production, the largest and most important of the five Department of Energy (DOE) missions at Hanford, could occur if Hanford's N Reactor is permanently closed and not replaced. The study provides estimates of statewide and local losses in jobs, income, and purchases from the private sector caused by such an event; it forecasts impacts on state and local government finances; and it describes certain local community and social impacts in the Tri-Cities (Richland, Kennewick, and Pasco) and surrounding communities. 33 refs., 8 figs., 22 tabs.

  5. Reanalysis of the Gas-cooled fast reactor experiments at the zero power facility Proteus – Spectral indices

    Directory of Open Access Journals (Sweden)

    Girardin G.

    2013-03-01

    Full Text Available PROTEUS is a zero power reactor at the Paul Scherrer Institute which has been employed during the 1970’s to study experimentally the physics of the gas-cooled fast reactor. Reaction rate distributions, flux spectrum and reactivity effects have been measured in several configurations featuring PuO2/UO2 fuel, absorbers, large iron shields, and thorium oxide and thorium metal fuel either distributed quasihomogeneously in the reference PuO2/UO2 lattice or introduced in the form of radial and axial blanket zones. This papers focus on the spectral indices – including fission and capture in 232Th and 237Np - measured in the reference PuO2/UO2 lattices and their predictions with an MCNPX model specially developed for the PROTEUS-GCFR core. Predictions were obtained with JEFF-3.1 and -3.11, ENDF/B-VII.0 and VII.1, and JENDL-3.3 and -4.0. A general good agreement was demonstrated. The ratio of 232Th fission to 239Pu fission, however, was under-predicted by 8.7±2.1% and 6.5±2.1% using ENDF/B-VII.0 and VII.1, respectively. Finally, the capture rates in 237Np tended to be underpredicted by the JEFF and JENDL libraries, although the new cross section in JEFF-3.1.1 slightly improved the 237Np capture to 239Pu fission results (3.4±2.4%.

  6. Occupational radiation exposure at commercial nuclear power reactors and other facilities 1996: Twenty-ninth annual report. Volume 18

    Energy Technology Data Exchange (ETDEWEB)

    Thomas, M.L. [Nuclear Regulatory Commission, Washington, DC (United States). Div. of Regulatory Applications; Hagemeyer, D. [Science Applications International Corp., Oak Ridge, TN (United States)

    1998-02-01

    This report summarizes the occupational exposure data that are maintained in the US Nuclear Regulatory Commission`s (NRC) Radiation Exposure Information and Reporting System (REIRS). The bulk of the information contained in the report was compiled from the 1996 annual reports submitted by six of the seven categories of NRC licensees subject to the reporting requirements of 10 CFR 20.2206. Since there are no geologic repositories for high level waste currently licensed, only six categories will be considered in this report. Annual reports for 1996 were received from a total of 300 NRC licensees, of which 109 were operators of nuclear power reactors in commercial operation. Compilations of the reports submitted by the 300 licensees indicated that 138,310 individuals were monitored, 75,139 of whom received a measurable dose. The collective dose incurred by these individuals was 21,755 person-cSv (person-rem){sup 2} which represents a 13% decrease from the 1995 value. The number of workers receiving a measurable dose also decreased, resulting in the average measurable dose of 0.29 cSv (rem) for 1996. The average measurable dose is defined to be the total collective dose (TEDE) divided by the number of workers receiving a measurable dose. These figures have been adjusted to account for transient reactor workers. Analyses of transient worker data indicate that 22,348 individuals completed work assignments at two or more licensees during the monitoring year. The dose distributions are adjusted each year to account for the duplicate reporting of transient workers by multiple licensees. In 1996, the average measurable dose calculated from reported was 0.24 cSv (rem). The corrected dose distribution resulted in an average measurable dose of 0.29 cSv (rem).

  7. Design of the 'half-size' ITER neutral beam source for the test facility ELISE

    Energy Technology Data Exchange (ETDEWEB)

    Heinemann, B. [Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, Postfach 1533, D-85740 Garching (Germany)], E-mail: bernd.heinemann@ipp.mpg.de; Falter, H.; Fantz, U.; Franzen, P.; Froeschle, M.; Gutser, R.; Kraus, W.; Nocentini, R.; Riedl, R.; Speth, E.; Staebler, A.; Wuenderlich, D. [Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, Postfach 1533, D-85740 Garching (Germany); Agostinetti, P. [Consorzio RFX, EURATOM Association, Corso Stati Uniti 4, I-35127 Padova (Italy); Jiang, T. [Southwestern Institute of Physics, ChengDu (China)

    2009-06-15

    In 2007 the radio frequency driven negative hydrogen ion source developed at IPP in Garching was chosen by the ITER board as the new reference source for the ITER neutral beam system. In order to support the design and the commissioning and operating phases of the ITER test facilities ISTF and NBTF in Padua, IPP is presently constructing a new test facility ELISE (Extraction from a Large Ion Source Experiment). ELISE will be operated with the so-called 'half-size ITER source' which is an intermediate step between the present small IPP RF sources (1/8 ITER size) and the full size ITER source. The source will have approximately the width but only half the height of the ITER source. The modular concept with 4 drivers will allow an easy extrapolation to the full ITER size with 8 drivers. Pulsed beam extraction and acceleration up to 60 kV (corresponding to pre-acceleration voltage of SINGAP) is foreseen. The aim of the design of the ELISE source and extraction system was to be as close as possible to the ITER design; it has however some modifications allowing a better diagnostic access as well as more flexibility for exploring open questions. Therefore one major difference compared to the source of ITER, NBTF or ISTF is the possible operation in air. Specific requirements for RF sources as found on IPP test facilities BATMAN and MANITU are implemented [A. Staebler, et al., Development of a RF-driven ion source for the ITER NBI system, SOFT Conference 2008, Fusion Engineering and Design, 84 (2009) 265-268].

  8. Termination of light-water reactor core-melt accidents with a chemical core catcher: the core-melt source reduction system (COMSORS)

    Energy Technology Data Exchange (ETDEWEB)

    Forsberg, C.W.; Parker, G.W.; Rudolph, J.C.; Osborne-Lee, I.W. [Oak Ridge National Lab., TN (United States); Kenton, M.A. [Dames and Moore, Westmont, IL (United States)

    1996-09-01

    The Core-Melt Source Reduction System (COMSORS) is a new approach to terminate light-water reactor core melt accidents and ensure containment integrity. A special dissolution glass is placed under the reactor vessel. If core debris is released onto the glass, the glass melts and the debris dissolves into the molten glass, thus creating a homogeneous molten glass. The molten glass, with dissolved core debris, spreads into a wide pool, distributing the heat for removal by radiation to the reactor cavity above or by transfer to water on top of the molten glass. Expected equilibrium glass temperatures are approximately 600 degrees C. The creation of a low-temperature, homogeneous molten glass with known geometry permits cooling of the glass without threatening containment integrity. This report describes the technology, initial experiments to measure key glass properties, and modeling of COMSORS operations.

  9. Operating manual for the Bulk Shielding Reactor

    Energy Technology Data Exchange (ETDEWEB)

    1987-03-01

    The BSR is a pool-type reactor. It has the capabilities of continuous operation at a power level of 2 MW or at any desired lower power level. This manual presents descriptive and operational information. The reactor and its auxiliary facilities are described from physical and operational viewpoints. Detailed operating procedures are included which are applicable from source-level startup to full-power operation. Also included are procedures relative to the safety of personnel and equipment in the areas of experiments, radiation and contamination control, emergency actions, and general safety. This manual supersedes all previous operating manuals for the BSR.

  10. Operating manual for the Bulk Shielding Reactor

    Energy Technology Data Exchange (ETDEWEB)

    1983-04-01

    The BSR is a pool-type reactor. It has the capabilities of continuous operation at a power level of 2 MW or at any desired lower power level. This manual presents descriptive and operational information. The reactor and its auxillary facilities are described from physical and operational viewpoints. Detailed operating procedures are included which are applicable from source-level startup to full-power operation. Also included are procedures relative to the safety of personnel and equipment in the areas of experiments, radiation and contamination control, emergency actions, and general safety. This manual supercedes all previous operating manuals for the BSR.

  11. Crystal and source characterization for the Crystal Backlighter Imager capability at the National Ignition Facility

    Science.gov (United States)

    Krauland, C. M.; Hall, G. N.; Buscho, J. G.; Hibbard, R.; McCarville, T. J.; Lowe-Webb, R.; Ayers, S. L.; Kalantar, D.; Kohut, T.; Kemp, G. E.; Bradley, D. K.; Bell, P.; Landen, O. L.; Brewster, T. N.; Piston, K.

    2017-10-01

    The Crystal Backlighter Imager (CBI) is a very narrow bandwidth ( 10 eV) x-ray radiography system that uses Bragg reflection from a spherically-curved crystal at near normal incidence. This diagnostic has the capability to image late in an ICF implosion because it only requires the brightness of the backlighter to be larger than the capsule self-emission in that narrow bandwidth. While the limited bandwidth is advantageous for this reason, it also requires that the effective energy of the backlighter atomic line is known to 1 eV accuracy for proper crystal alignment. Any Doppler shift in the line energy must be understood for the imaging system to work. The work presented details characterization experiments done at the Jupiter Laser Facility with a Si (8 6 2) crystal that will be used with a Selenium backlighter in the NIF CBI diagnostic. We used the spherically-bent crystals to image a small ( 200 µm) He α source generated by the Janus laser on a Se foil. Scanning Bragg angles over multiple shots allowed us to map out the spectral line intensity distribution for optimal alignment in NIF. A subsequent Doppler shift measurement using CBI on NIF will also be presented with complementary HYDRA modeling for both experiments. Prepared by LLNL under Contract DE-AC52-07NA27344 and by General Atomics under Contract DE-NA0001808.

  12. Multi-keV x-ray sources from HYBRID targets on GEKKO and OMEGA facilities

    Directory of Open Access Journals (Sweden)

    Primout M.

    2013-11-01

    Full Text Available The feasibility of efficient X-ray sources for radiography on the LMJ (Laser MégaJoule in the multi-kJ/ns range was demonstrated on the OMEGA laser facility (Univ. Rochester from 2002 to 2004 [1,2]. We significantly enhanced the conversion efficiency of titanium (4–6 keV, copper (8–10 keV and germanium (9–13 keV foils by using an optimized pre-pulse/pulse combination. Since higher X-ray energy and therefore electronic temperature need hydroconfinement, plastic cylindrical hohlraums internally coated with titanium, copper and germanium with various OMEGA beam configurations were successfully tested from 2005 to 2009 [3–5]. In addition, many shots with metal-doped aerogel (Ti, Fe, Ge were tested on OMEGA [6]. Recently we tested a new concept of “HYBRID sources” based on the combination of a thin titanium foil at the exit hole of a plastic cylinder filled with very low density SiO2 aerogel (2 and 5 mg/cc. The benefit of the underdense medium is, first, to transport the laser energy to the titanium foil after its conversion into a supersonic ionization front and, second, to prevent foil expansion and excessive kinetic energy losses by longitudinal hydroconfinement.

  13. Global shielding analysis for the three-element core advanced neutron source reactor under normal operating conditions

    Energy Technology Data Exchange (ETDEWEB)

    Slater, C.O.; Bucholz, J.A.

    1995-08-01

    Two-dimensional discrete ordinates radiation transport calculations were performed for a model of the three-element core Advanced Neutron Source reactor design under normal operating conditions. The core consists of two concentric upper elements and a lower element radially centered in the annulus between the upper elements. The initial radiation transport calculations were performed with the DORT two-dimensional discrete ordinates radiation transport code using the 39-neutron-group/44-gamma-ray-group ANSL-V cross-section library, an S{sub 6} quadrature, and a P{sub 1} Legendre polynomial expansion of the cross sections to determine the fission neutron source distribution in the core fuel elements. These calculations were limited to neutron groups only. The final radiation transport calculations, also performed with DORT using the 39-neutron-group/44-gamma-ray-group ANSL-V cross-section library, an S{sub l0} quadrature, and a P{sub 3} Legendre polynomial expansion of the cross sections, produced neutron and gamma-ray fluxes over the full extent of the geometry model. Responses (or activities) at various locations in the model were then obtained by folding the appropriate response functions with the fluxes at those locations. Some comparisons were made with VENTURE-calculated (diffusion theory) 20-group neutron fluxes that were summed into four broad groups. Tne results were in reasonably good agreement when the effects of photoneutrons were not included, thus verifying the physics model upon which the shielding model was based. Photoneutrons increased the fast-neutron flux levels deep within the D{sub 2}0 several orders of magnitude. Results are presented as tables of activity values for selected radial and axial traverses, plots of the radial and axial traverse data, and activity contours superimposed on the calculational geometry model.

  14. Design of a homogeneous subcritical nuclear reactor based on thorium with a source of californium 252; Diseno de un reactor nuclear subcritico homogeneo a base de Torio con una fuente de Californio 252

    Energy Technology Data Exchange (ETDEWEB)

    Delgado H, C. E.; Vega C, H. R. [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No. 10, Fracc. La Penuela, 98068 Zacatecas, Zac. (Mexico); Sajo B, L., E-mail: ce_delgado89@hotmail.com [Universidad Simon Bolivar, Laboratorio de Fisica Nuclear, Apdo. 89000, 1080A Caracas (Venezuela, Bolivarian Republic of)

    2015-10-15

    Full text: One of the energy alternatives to fossil fuels which do not produce greenhouse gases is the nuclear energy. One of the drawbacks of this alternative is the generation of radioactive wastes of long half-life and its relation to the generation of nuclear materials to produce weapons of mass destruction. An option to these drawbacks of nuclear energy is to use Thorium as part of the nuclear fuel which it becomes in U{sup 233} when capturing neutrons, that is a fissile material. In this paper Monte Carlo methods were used to design a homogeneous subcritical reactor based on thorium. As neutron reflector graphite was used. The reactor core is homogeneous and is formed of 70% light water as moderator, 12% of enriched uranium UO{sub 2}(NO{sub 3}){sub 4} and 18% of thorium Th(NO{sub 3}){sub 4} as fuel. To start the nuclear fission chain reaction an isotopic source of californium 252 was used with an intensity of 4.6 x 10{sup 7} s{sup -1}. In the design the value of the effective multiplication factor, whose value turned out k{sub eff} <1 was calculated. Also, the neutron spectra at different distances from the source and the total fluence were calculated, as well as the values of the ambient dose equivalent in the periphery of the reactor. (Author)

  15. Proposal to DOE Basic Energy Sciences: Ultrafast X-ray science facility at the Advanced Light Source

    Energy Technology Data Exchange (ETDEWEB)

    Schoenlein, Robert W.; Falcone, Roger W.; Abela, R.; Alivisatos, A.P.; Belkacem, A.; Berrah, N.; Bozek, J.; Bressler, C.; Cavalleri, A.; Chergui, M.; Glover, T.E.; Heimann, P.A.; Hepburn, J.; Larsson, J.; Lee, R.W.; McCusker, J.; Padmore, H.A.; Pattison, P.; Pratt, S.T.; Shank, C.V.; Wark, J.; Chang, Z.; Robin, D.W.; Schlueter, R.D.; Zholents, A.A.; Zolotorev, M.S.

    2001-12-12

    We propose to develop a true user facility for ultrafast x-ray science at the Advanced Light Source. This facility will be unique in the world, and will fill a critical need for the growing ultrafast x-ray research community. The development of this facility builds upon the expertise from long-standing research efforts in ultrafast x-ray spectroscopy and the development of femtosecond x-ray sources and techniques at both the Lawrence Berkeley National Laboratory and at U.C. Berkeley. In particular, the technical feasibility of a femtosecond x-ray beamline at the ALS has already been demonstrated, and existing ultrafast laser technology will enable such a beamline to operate near the practical limit for femtosecond x-ray flux and brightness from a 3rd generation synchrotron.

  16. Feasibility study for the implementation of an intense linac-based neutron source facility in Sao Paulo

    CERN Document Server

    Arruda-Neto, J D T

    1999-01-01

    The implementation of an intense linac-based neutron source is proposed here, using the forthcoming 30 MeV electron CW-Microtron under construction in Sao Paulo. A method to estimate photoneutron fluxes, based on thin target bremsstrahlung approximation, has been developed. The versatility and possible broad range of applications associated with such a facility are addressed.

  17. Implementation of ground source heat pumps in DON facilities - a feasibility study

    OpenAIRE

    Carson, John W. III

    2000-01-01

    CIVINS (Civilian Institutions) Thesis document Approved for public release, distribution unlimited The Civil Engineering Corps and the Naval Facilities Engineering Command has the distinct honor and challenge to oversee all facilities management functions from design and contract, to construction, to maintenance and repair and finally to demolition and disposal. In order to assist this monumental undertaking, the Naval Facilities Engineering Command (NAVFAC) is organized with Engneerin...

  18. Pressurizing new reactors

    Energy Technology Data Exchange (ETDEWEB)

    Neill, J.S.

    1956-01-30

    The Technical Division was asked recently to consider designs for new reactors that would add 8000 MW capacity to the Savannah River Plant. One modification of the existing SRP design that would enable a higher power rating, and therefore require fewer new reactors, is an increase in the maximum pressure in the D{sub 2}O system. The existing reactors at SRP are designed for a maximum pressure in the gas plenum of only 5 psig. Higher pressures enable higher D{sub 2} temperatures and higher sheath temperatures without local boiling or burnout. The requirements in reactor cooling facilities at any given power level would therefore be reduced by pressurizing.

  19. Reactor Engineering Department annual report (April 1, 1996 - March 31, 1997)

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-10-01

    This report summarizes the research and development activities in the Reactor Engineering Department of JAERI during the fiscal year of 1996 (April 1, 1996 - March 31, 1997). The major Department`s programs promoted in the year are the design activities of advanced reactor system and the development of a high power proton linear accelerator to construct an intense neutron source for innovative neutron science. Other Major tasks of the Department are various basics researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analysis, the fusion neutronics, the radiation shielding, the reactor instrumentation, the reactor control/diagnosis, the thermal hydraulics and the technology developments related to the reactor engineering facilities, the accelerator facilities and the thermal hydraulic facilities. The cooperative works to JAERI`s major projects such as the high temperature gas cooled reactor, the fusion reactor and PNC`s fast reactor project were also progressed. The 99 papers are indexed individually. (J.P.N.)

  20. Reactor operation safety information document

    Energy Technology Data Exchange (ETDEWEB)

    1990-01-01

    The report contains a reactor facility description which includes K, P, and L reactor sites, structures, operating systems, engineered safety systems, support systems, and process and effluent monitoring systems; an accident analysis section which includes cooling system anomalies, radioactive materials releases, and anticipated transients without scram; a summary of onsite doses from design basis accidents; severe accident analysis (reactor core disruption); a description of operating contractor organization and emergency planning; and a summary of reactor safety evolution. (MB)

  1. Central and Eastern United States (CEUS) Seismic Source Characterization (SSC) for Nuclear Facilities Project

    Energy Technology Data Exchange (ETDEWEB)

    Kevin J. Coppersmith; Lawrence A. Salomone; Chris W. Fuller; Laura L. Glaser; Kathryn L. Hanson; Ross D. Hartleb; William R. Lettis; Scott C. Lindvall; Stephen M. McDuffie; Robin K. McGuire; Gerry L. Stirewalt; Gabriel R. Toro; Robert R. Youngs; David L. Slayter; Serkan B. Bozkurt; Randolph J. Cumbest; Valentina Montaldo Falero; Roseanne C. Perman' Allison M. Shumway; Frank H. Syms; Martitia (Tish) P. Tuttle

    2012-01-31

    Seismic Hazard Analysis: Guidance on Uncertainty and Use of Experts. The model will be used to assess the present-day composite distribution for seismic sources along with their characterization in the CEUS and uncertainty. In addition, this model is in a form suitable for use in PSHA evaluations for regulatory activities, such as Early Site Permit (ESPs) and Combined Operating License Applications (COLAs). Applications, Values, and Use Development of a regional CEUS seismic source model will provide value to those who (1) have submitted an ESP or COLA for Nuclear Regulatory Commission (NRC) review before 2011; (2) will submit an ESP or COLA for NRC review after 2011; (3) must respond to safety issues resulting from NRC Generic Issue 199 (GI-199) for existing plants and (4) will prepare PSHAs to meet design and periodic review requirements for current and future nuclear facilities. This work replaces a previous study performed approximately 25 years ago. Since that study was completed, substantial work has been done to improve the understanding of seismic sources and their characterization in the CEUS. Thus, a new regional SSC model provides a consistent, stable basis for computing PSHA for a future time span. Use of a new SSC model reduces the risk of delays in new plant licensing due to more conservative interpretations in the existing and future literature. Perspective The purpose of this study, jointly sponsored by EPRI, the U.S. Department of Energy (DOE), and the NRC was to develop a new CEUS SSC model. The team assembled to accomplish this purpose was composed of distinguished subject matter experts from industry, government, and academia. The resulting model is unique, and because this project has solicited input from the present-day larger technical community, it is not likely that there will be a need for significant revision for a number of years. See also Sponsors Perspective for more details. The goal of this project was to implement the CEUS SSC work plan

  2. Effect of carbon sources on the removal of 1,1,2-trichloroethane and 1,1,2,2-tetrachloroethane in UASB reactor.

    Science.gov (United States)

    Basu, Debolina; Asolekar, Shyam R

    2012-01-01

    The effect of two carbon sources namely sodium acetate and ethanol was studied in bench-scale Upflow Anaerobic Sludge Blanket (UASB) reactors for the removal of chlorinated ethanes i.e., 1,1,2-Trichloroethane (TCA) and 1,1,2,2-Tetrachloroethane (TeCA) contained in the simulated wastewaters. The Hydraulic Retention Time (HRT) was maintained as 24 hours in all the reactors. The granular biomass in the test reactors R2 and R3 were acclimated to 40 mg/L of TCA and 20 mg/L of TeCA, respectively. The effluent TCA and TeCA concentrations were 0.03 mg/L and 0.18 mg/L, respectively, at the end of acclimation phase. Sodium acetate and ethanol both were found to be suitable as the primary substrates in the biodegradation of TCA and TeCA. However, lower concentrations of the toxic pollutants (TCA and TeCA) were obtained in the effluents with the use of sodium acetate. The COD removal efficiency in the test reactors (R2 and R3) varied in the range of 95 % to 98.2 % accompanied by the formation of 1,2-Dichloroethane (DCA) as the major intermediate.

  3. Use the results of measurements on KBR facility for testing of neutron data of main structural materials for fast reactors

    Science.gov (United States)

    Koscheev, Vladimir; Manturov, Gennady; Pronyaev, Vladimir; Rozhikhin, Evgeny; Semenov, Mikhail; Tsibulya, Anatoly

    2017-09-01

    Several k∞ experiments were performed on the KBR critical facility at the Institute of Physics and Power Engineering (IPPE), Obninsk, Russia during the 1970s and 80s for study of neutron absorption properties of Cr, Mn, Fe, Ni, Zr, and Mo. Calculations of these benchmarks with almost any modern evaluated nuclear data libraries demonstrate bad agreement with the experiment. Neutron capture cross sections of the odd isotopes of Cr, Mn, Fe, and Ni in the ROSFOND-2010 library have been reevaluated and another evaluation of the Zr nuclear data has been adopted. Use of the modified nuclear data for Cr, Mn, Fe, Ni, and Zr leads to significant improvement of the C/E ratio for the KBR assemblies. Also a significant improvement in agreement between calculated and evaluated values for benchmarks with Fe reflectors was observed. C/E results obtained with the modified ROSFOND library for complex benchmark models that are highly sensitive to the cross sections of structural materials are no worse than results obtained with other major evaluated data libraries. Possible improvement in results by decreasing the capture cross section for Zr and Mo at the energies above 1 keV is indicated.

  4. Use the results of measurements on KBR facility for testing of neutron data of main structural materials for fast reactors

    Directory of Open Access Journals (Sweden)

    Koscheev Vladimir

    2017-01-01

    Full Text Available Several k∞ experiments were performed on the KBR critical facility at the Institute of Physics and Power Engineering (IPPE, Obninsk, Russia during the 1970s and 80s for study of neutron absorption properties of Cr, Mn, Fe, Ni, Zr, and Mo. Calculations of these benchmarks with almost any modern evaluated nuclear data libraries demonstrate bad agreement with the experiment. Neutron capture cross sections of the odd isotopes of Cr, Mn, Fe, and Ni in the ROSFOND-2010 library have been reevaluated and another evaluation of the Zr nuclear data has been adopted. Use of the modified nuclear data for Cr, Mn, Fe, Ni, and Zr leads to significant improvement of the C/E ratio for the KBR assemblies. Also a significant improvement in agreement between calculated and evaluated values for benchmarks with Fe reflectors was observed. C/E results obtained with the modified ROSFOND library for complex benchmark models that are highly sensitive to the cross sections of structural materials are no worse than results obtained with other major evaluated data libraries. Possible improvement in results by decreasing the capture cross section for Zr and Mo at the energies above 1 keV is indicated.

  5. Novel test-bed facility for PSI issues in fusion reactor conditions on the base of next generation QSPA plasma accelerator

    Science.gov (United States)

    Garkusha, I. E.; Chebotarev, V. V.; Herashchenko, S. S.; Makhlaj, V. A.; Kulik, N. V.; Ladygina, M. S.; Marchenko, A. K.; Petrov, Yu. V.; Staltsov, V. V.; Shevchuk, P. V.; Solyakov, D. G.; Yelisyeyev, D. V.

    2017-11-01

    In this report a concept of a new generation QSPA with external B-field up to 2 T has been discussed. A novel test-bed facility, which was recently constructed in Kharkov IPP NSC KIPT, has been described. It allows for a new level of plasma stream parameters and its wide variation in new QSPA-M device, as well as possible combination of steady-state and pulsed plasma loads to the materials during the exposures. First plasma is recently obtained. Careful optimization of the operational regimes of the plasma accelerator’s functional components and plasma dynamics in the magnetic system of QSPA-M device has started approaching step by step the necessary level of plasma parameters and their effective variation. The relevant results on plasma stream characterization are presented. Energy density distributions in plasma stream have been measured with calorimetry. Spectroscopy and probe technique have also been applied for plasma parameters measurements. The obtained results demonstrate the ability of QSPA-M to reproduce the ELM impacts in fusion reactor, both in terms of heat load and particle flux to the surface.

  6. Two-color infrared FEL facility employing a 250-MeV linac injector of Saga synchrotron light source

    CERN Document Server

    Tomimasu, T; Koga, N; Hashiguchi, Y; Ochiai, Y; Ishibashi, M

    2001-01-01

    A two-color infrared free electron laser (FEL) facility is proposed. This FEL facility will employ a new 250-MeV linac injector of the Saga synchrotron light source (SLS). The linac has two operation modes: short macropulse mode of 1 mu s at 250 MeV is for injection to a 1.4-GeV storage ring and long macropulse mode of 13 mu s at 40 MeV is for the two-color FEL facility. The two-color FEL uses a single electron beam and simultaneously provides both infrared (IR) and far-IR laser pulses for pump-probe studies of quantum-well structures and studying vibrational relaxation of molecules. The Saga SLS will be operated in 2004 to promote material science, bio-medical and industrial applications in Kyushu.

  7. 40 CFR 63.5795 - How do I know if my reinforced plastic composites production facility is a new affected source or...

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 12 2010-07-01 2010-07-01 true How do I know if my reinforced plastic....5795 How do I know if my reinforced plastic composites production facility is a new affected source or an existing affected source? (a) A reinforced plastic composites production facility is a new...

  8. High Flux Isotope Reactor (HFIR)

    Data.gov (United States)

    Federal Laboratory Consortium — The HFIR at Oak Ridge National Laboratory is a light-water cooled and moderated reactor that is the United States’ highest flux reactor-based neutron source. HFIR...

  9. Beamlines of the biomedical imaging and therapy facility at the Canadian light source – part 3

    Energy Technology Data Exchange (ETDEWEB)

    Wysokinski, Tomasz W., E-mail: bmit@lightsource.ca [Canadian Light Source, Saskatoon, SK (Canada); Chapman, Dean [Anatomy and Cell Biology, University of Saskatchewan, Saskatoon, SK (Canada); Adams, Gregg [Western College of Veterinary Medicine, Saskatoon, SK (Canada); Renier, Michel [European Synchrotron Radiation Facility, Grenoble (France); Suortti, Pekka [Department of Physics, University of Helsinki (Finland); Thomlinson, William [Department of Physics, University of Saskatchewan, Saskatoon, SK (Canada)

    2015-03-01

    The BioMedical Imaging and Therapy (BMIT) facility provides synchrotron-specific imaging and radiation therapy capabilities [1–4]. We describe here the Insertion Device (ID) beamline 05ID-2 with the beam terminated in the SOE-1 (Secondary Optical Enclosure) experimental hutch. This endstation is designed for imaging and therapy research primarily in animals ranging in size from mice to humans to horses, as well as tissue specimens including plants. Core research programs include human and animal reproduction, cancer imaging and therapy, spinal cord injury and repair, cardiovascular and lung imaging and disease, bone and cartilage growth and deterioration, mammography, developmental biology, gene expression research as well as the introduction of new imaging methods. The source for the ID beamline is a multi-pole superconducting 4.3 T wiggler [5]. The high field gives a critical energy over 20 keV. The high critical energy presents shielding challenges and great care must be taken to assess shielding requirements [6–9]. The optics in the POE-1 and POE-3 hutches [4,10] prepare a monochromatic beam that is 22 cm wide in the last experimental hutch SOE-1. The double crystal bent-Laue or Bragg monochromator, or the single-crystal K-edge subtraction (KES) monochromator provide an energy range appropriate for imaging studies in animals (20–100+ keV). SOE-1 (excluding the basement structure 4 m below the experimental floor) is 6 m wide, 5 m tall and 10 m long with a removable back wall to accommodate installation and removal of the Large Animal Positioning System (LAPS) capable of positioning and manipulating animals as large as a horse [11]. This end-station also includes a unique detector positioner with a vertical travel range of 4.9 m which is required for the KES imaging angle range of +12.3° to –7.3°. The detector positioner also includes moveable shielding integrated with the safety shutters. An update on the status of the other two end-stations at BMIT

  10. Beamlines of the biomedical imaging and therapy facility at the Canadian light source - part 3

    Science.gov (United States)

    Wysokinski, Tomasz W.; Chapman, Dean; Adams, Gregg; Renier, Michel; Suortti, Pekka; Thomlinson, William

    2015-03-01

    The BioMedical Imaging and Therapy (BMIT) facility provides synchrotron-specific imaging and radiation therapy capabilities [1-4]. We describe here the Insertion Device (ID) beamline 05ID-2 with the beam terminated in the SOE-1 (Secondary Optical Enclosure) experimental hutch. This endstation is designed for imaging and therapy research primarily in animals ranging in size from mice to humans to horses, as well as tissue specimens including plants. Core research programs include human and animal reproduction, cancer imaging and therapy, spinal cord injury and repair, cardiovascular and lung imaging and disease, bone and cartilage growth and deterioration, mammography, developmental biology, gene expression research as well as the introduction of new imaging methods. The source for the ID beamline is a multi-pole superconducting 4.3 T wiggler [5]. The high field gives a critical energy over 20 keV. The high critical energy presents shielding challenges and great care must be taken to assess shielding requirements [6-9]. The optics in the POE-1 and POE-3 hutches [4,10] prepare a monochromatic beam that is 22 cm wide in the last experimental hutch SOE-1. The double crystal bent-Laue or Bragg monochromator, or the single-crystal K-edge subtraction (KES) monochromator provide an energy range appropriate for imaging studies in animals (20-100+ keV). SOE-1 (excluding the basement structure 4 m below the experimental floor) is 6 m wide, 5 m tall and 10 m long with a removable back wall to accommodate installation and removal of the Large Animal Positioning System (LAPS) capable of positioning and manipulating animals as large as a horse [11]. This end-station also includes a unique detector positioner with a vertical travel range of 4.9 m which is required for the KES imaging angle range of +12.3° to -7.3°. The detector positioner also includes moveable shielding integrated with the safety shutters. An update on the status of the other two end-stations at BMIT, described

  11. Photocatalytic reactor

    Science.gov (United States)

    Bischoff, Brian L.; Fain, Douglas E.; Stockdale, John A. D.

    1999-01-01

    A photocatalytic reactor for processing selected reactants from a fluid medium comprising at least one permeable photocatalytic membrane having a photocatalytic material. The material forms an area of chemically active sites when illuminated by light at selected wavelengths. When the fluid medium is passed through the illuminated membrane, the reactants are processed at these sites separating the processed fluid from the unprocessed fluid. A light source is provided and a light transmitting means, including an optical fiber, for transmitting light from the light source to the membrane.

  12. Closed Brayton Cycle power system with a high temperature pellet bed reactor heat source for NEP applications

    Science.gov (United States)

    Juhasz, Albert J.; El-Genk, Mohamed S.; Harper, William B., Jr.

    1992-01-01

    Capitalizing on past and future development of high temperature gas reactor (HTGR) technology, a low mass 15 MWe closed gas turbine cycle power system using a pellet bed reactor heating helium working fluid is proposed for Nuclear Electric Propulsion (NEP) applications. Although the design of this directly coupled system architecture, comprising the reactor/power system/space radiator subsystems, is presented in conceptual form, sufficient detail is included to permit an assessment of overall system performance and mass. Furthermore, an attempt is made to show how tailoring of the main subsystem design characteristics can be utilized to achieve synergistic system level advantages that can lead to improved reliability and enhanced system life while reducing the number of parasitic load driven peripheral subsystems.

  13. Discrimination of source reactor type by multivariate statistical analysis of uranium and plutonium isotopic concentrations in unknown irradiated nuclear fuel material.

    Science.gov (United States)

    Robel, Martin; Kristo, Michael J

    2008-11-01

    The problem of identifying the provenance of unknown nuclear material in the environment by multivariate statistical analysis of its uranium and/or plutonium isotopic composition is considered. Such material can be introduced into the environment as a result of nuclear accidents, inadvertent processing losses, illegal dumping of waste, or deliberate trafficking in nuclear materials. Various combinations of reactor type and fuel composition were analyzed using Principal Components Analysis (PCA) and Partial Least Squares Discriminant Analysis (PLSDA) of the concentrations of nine U and Pu isotopes in fuel as a function of burnup. Real-world variation in the concentrations of (234)U and (236)U in the fresh (unirradiated) fuel was incorporated. The U and Pu were also analyzed separately, with results that suggest that, even after reprocessing or environmental fractionation, Pu isotopes can be used to determine both the source reactor type and the initial fuel composition with good discrimination.

  14. A study on the application of CRUDTRAN code in primary systems of domestic pressurized heavy-water reactors for prediction of radiation source term

    Energy Technology Data Exchange (ETDEWEB)

    Song, Jong Soon; Cho, Hoon Jo; Jung, Min Young; Lee, Sang Heon [Dept. of Nuclear Engineering, Chosun University, Gwangju (Korea, Republic of)

    2017-04-15

    The importance of developing a source-term assessment technology has been emphasized owing to the decommissioning of Kori nuclear power plant (NPP) Unit 1 and the increase of deteriorated NPPs. We analyzed the behavioral mechanism of corrosion products in the primary system of a pressurized heavy-water reactor-type NPP. In addition, to check the possibility of applying the CRUDTRAN code to a Canadian Deuterium Uranium Reactor (CANDU)-type NPP, the type was assessed using collected domestic onsite data. With the assessment results, it was possible to predict trends according to operating cycles. Values estimated using the code were similar to the measured values. The results of this study are expected to be used to manage the radiation exposures of operators in high-radiation areas and to predict decommissioning processes in the primary system.

  15. Safety techniques in the change of nuclear systems. Radiation protection at spallation neutron sources and transmutation facilities; Sicherheitstechnik im Wandel Nuklearer Systeme. Strahlenschutz bei Spallationsneutronenquellen und Transmutationsanlagen

    Energy Technology Data Exchange (ETDEWEB)

    Nuenighoff, Kay

    2009-07-01

    To push the boundary towards higher neutron fluxes concepts based on spallation reactions have been discussed. Here neutrons are produced by bombarding a heavy metal target (e.g. mercury, tungsten, or tantalum) with high energetic protons. Up to now such facilities could not be realised because of the high power particle accelerators needed. Recent developments of the accelerator technology open the possibility of construction and operating proton accelerators in the MW region. This is demonstrated by construction and commissioning of two MW spallation neutron sources, namely SNS (Oak Ridge, Tennessee, USA) with a power of 1.4 MW and J-PARC (Japan) with 1 MW. The realisation of proton accelerators at this power level will open the way towards energy amplifiers, as proposed e.g. by Carlo Rubbia. Such a facility will not only produce electric power. Furthermore longliving radionuclides can be transmutated into shortlived or even stable nuclides by neutron induced nuclear reactions. A mitigation of the problem of nuclear waste disposal. The above discussed developments prove that accelerators are not only constructed for research, moreover application of these technology became state of the art. With the emergence of particle accelerators in the MW region, radiation protection is confronted with new kind of problems to be solved. Especially the higher kinetic energies of the primary beam particles requires modification and expansion of computer programs well known in nuclear engineering. In contrast to nuclear reactors with kinetic energies up to 2-3 MeV, in spallation reaction secondary particles up to the incident energy in the GeV region will be produced. Problems related to radiation protection have to be considered in an energy range three orders of magnitude higher than known from nuclear reactors. In this thesis existing computer codes are compared and validated with data from selected experiments. Questions concerning radiation protection covers a broad range

  16. Modeling Tool to Quantify Metal Sources in Stormwater Discharges at Naval Facilities (NESDI Project 455)

    Science.gov (United States)

    2014-06-01

    demonstration/validation project to assess the use of the urban stormwater model Windows Source Loading and Management Model (WinSLAMM) to characterize...the urban stormwater model Windows Source Loading and Management Model (WinSlamm) to characterize sources of copper and zinc in storm runoff at Navy...are ubiquitous contaminants found in stormwater discharges in urban and industrialized areas. These contaminants originate from a variety of sources

  17. Impact of a 2 MWth research reactor on radioactivity in sediments.

    Science.gov (United States)

    Bose, S R; Williamson, T G; Mulder, R U; Molla, M A

    1993-08-01

    The objective of this study was to determine the environmental impact caused by liquid effluent discharge from the University of Virginia's 2 MWth Research Reactor during the course of its first quarter-century of normal operation. Sediment samples were collected from the reactor pond (situated in a restricted area next to the Reactor Facility), the stream feeding it, and its exit stream. For a comparative study, sediment samples were taken from a nearby closed reference pond having no direct link with the reactor pond. Concentrations of long-lived alpha, beta and gamma emitting radionuclides, from natural and nuclear weapons fallout sources, were detected in pond and stream sediments. Low levels of activation product radioisotopes from the research reactor were detected in the reactor pond sediment. It was observed that both natural and artificial radionuclide concentrations were higher in the UVAR pond (with the exception of 54Mn and 65Zn) as compared to exist stream and reference sediments.

  18. Testing SLURM open source batch system for a Tierl/Tier2 HEP computing facility

    Science.gov (United States)

    Donvito, Giacinto; Salomoni, Davide; Italiano, Alessandro

    2014-06-01

    In this work the testing activities that were carried on to verify if the SLURM batch system could be used as the production batch system of a typical Tier1/Tier2 HEP computing center are shown. SLURM (Simple Linux Utility for Resource Management) is an Open Source batch system developed mainly by the Lawrence Livermore National Laboratory, SchedMD, Linux NetworX, Hewlett-Packard, and Groupe Bull. Testing was focused both on verifying the functionalities of the batch system and the performance that SLURM is able to offer. We first describe our initial set of requirements. Functionally, we started configuring SLURM so that it replicates all the scheduling policies already used in production in the computing centers involved in the test, i.e. INFN-Bari and the INFN-Tier1 at CNAF, Bologna. Currently, the INFN-Tier1 is using IBM LSF (Load Sharing Facility), while INFN-Bari, an LHC Tier2 for both CMS and Alice, is using Torque as resource manager and MAUI as scheduler. We show how we configured SLURM in order to enable several scheduling functionalities such as Hierarchical FairShare, Quality of Service, user-based and group-based priority, limits on the number of jobs per user/group/queue, job age scheduling, job size scheduling, and scheduling of consumable resources. We then show how different job typologies, like serial, MPI, multi-thread, whole-node and interactive jobs can be managed. Tests on the use of ACLs on queues or in general other resources are then described. A peculiar SLURM feature we also verified is triggers on event, useful to configure specific actions on each possible event in the batch system. We also tested highly available configurations for the master node. This feature is of paramount importance since a mandatory requirement in our scenarios is to have a working farm cluster even in case of hardware failure of the server(s) hosting the batch system. Among our requirements there is also the possibility to deal with pre-execution and post

  19. Nuclear reactors built, being built, or planned, 1994

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-07-01

    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, 1994. 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: a commercial reactor locator map and tables of the characteristic and statistical data that follow; a table of abbreviations; tables of data for reactors operating, being built, or planned; and 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).

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

    Energy Technology Data Exchange (ETDEWEB)

    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 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. 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).

  1. Comparative study on nutrient removal of agricultural non-point source pollution for three filter media filling schemes in eco-soil reactors.

    Science.gov (United States)

    Du, Fuyi; Xie, Qingjie; Fang, Longxiang; Su, Hang

    2016-08-01

    Nutrients (nitrogen and phosphorus) from agricultural non-point source (NPS) pollution have been increasingly recognized as a major contributor to the deterioration of water quality in recent years. The purpose of this article is to investigate the discrepancies in interception of nutrients in agricultural NPS pollution for eco-soil reactors using different filling schemes. Parallel eco-soil reactors of laboratory scale were created and filled with filter media, such as grit, zeolite, limestone, and gravel. Three filling schemes were adopted: increasing-sized filling (I-filling), decreasing-sized filling (D-filling), and blend-sized filling (B-filling). The systems were intermittent operations via simulated rainstorm runoff. The nutrient removal efficiency, biomass accumulation and vertical dissolved oxygen (DO) distribution were defined to assess the performance of eco-soil. The results showed that B-filling reactor presented an ideal DO for partial nitrification-denitrification across the eco-soil, and B-filling was the most stable in the change of bio-film accumulation trends with depth in the three fillings. Simultaneous and highest removals of NH4(+)-N (57.74-70.52%), total nitrogen (43.69-54.50%), and total phosphorus (42.50-55.00%) were obtained in the B-filling, demonstrating the efficiency of the blend filling schemes of eco-soil for oxygen transfer and biomass accumulation to cope with agricultural NPS pollution.

  2. H Reactor

    Data.gov (United States)

    Federal Laboratory Consortium — The H Reactor was the first reactor to be built at Hanford after World War II.It became operational in October of 1949, and represented the fourth nuclear reactor on...

  3. Mixed Waste Management Facility (MWMF) Old Burial Ground (OBG) source control technology and inventory study

    Energy Technology Data Exchange (ETDEWEB)

    Flach, G.P.; Rehder, T.E.; Kanzleiter, J.P.

    1996-10-02

    This report has been developed to support information needs for wastes buried in the Burial Ground Complex. Information discussed is presented in a total of four individual attachments. The general focus of this report is to collect information on estimated source inventories, leaching studies, source control technologies, and to provide information on modeling parameters and associated data deficiencies.

  4. TWAC facility and the use of the laser ion source for production of intense heavy ion beams

    CERN Document Server

    Sharkov, B Yu; Shumshurov, A V; Meshcheryakov, N D; Rudskoy, I; Homenko, S; Makarov, K; Rörich, V; Stepanov, A; Satov, Yu A; Haseroth, H; Kugler, H; Lisi, N; Scrivens, R

    1999-01-01

    Current activities on upgrading of the ITEP heavy ion accelerator complex in the framework of the ITEP-TWAC project are reported. The project being in progress since 1997 is aiming at production of intense (100 kJ/100 ns) heavy ion beams. The basic idea of the project is the application of the non-Liouvillian technique in an existing accelerator facility based on a heavy ion synchrotron for its adaptation to heavy ion fusion related experiments. Special attention is paid to the results on generation of highly charged medium mass and heavy ions in the laser produced plasma. Development of key elements of the laser ion source based on the use of a 100 J repetition rate CO/sub 2/-laser for filling of ITEP and CERN accelerator facilities in the single turn injection mode is presented. (4 refs).

  5. Extraction and low energy beam transport from a surface ion source at the TRIUMF-ISAC facility

    Science.gov (United States)

    Sen, A.; Ames, F.; Bricault, P.; Lassen, J.; Laxdal, A.; Mjos, A.

    2016-06-01

    A large fraction of radioactive beams produced and delivered at TRIUMF's isotope separator and accelerator facility, ISAC, are using either a surface ion source or a resonant ionization laser ion source, which share a common design. To characterize the operation of the ion sources, simulations were performed to determine the ion beam optics and beam envelope properties of the extracted beam. Furthermore ion-optics calculations were performed to determine the transmission parameters through the mass separator magnet. Emittances are measured in the ISAC low energy beam line right after the mass separator. The recent addition of a channeltron to the Allison emittance meter scanner now allows us to measure emittances for ion beams with intensities as low as 105 ions/s. This is particularly useful for establishing high resolution, high throughput mass separator tunes for radioactive isotope beams. This paper discusses emittance measurements of low intensity beams, typical emittance scans for the surface ion source and the resonant laser ionized source for different source parameters. The observed results are compared to the simulations and discussed.

  6. Extraction and low energy beam transport from a surface ion source at the TRIUMF-ISAC facility

    Energy Technology Data Exchange (ETDEWEB)

    Sen, A., E-mail: asen.kvi@gmail.com; Ames, F.; Bricault, P.; Lassen, J.; Laxdal, A.; Mjos, A.

    2016-06-01

    A large fraction of radioactive beams produced and delivered at TRIUMF’s isotope separator and accelerator facility, ISAC, are using either a surface ion source or a resonant ionization laser ion source, which share a common design. To characterize the operation of the ion sources, simulations were performed to determine the ion beam optics and beam envelope properties of the extracted beam. Furthermore ion-optics calculations were performed to determine the transmission parameters through the mass separator magnet. Emittances are measured in the ISAC low energy beam line right after the mass separator. The recent addition of a channeltron to the Allison emittance meter scanner now allows us to measure emittances for ion beams with intensities as low as 10{sup 5} ions/s. This is particularly useful for establishing high resolution, high throughput mass separator tunes for radioactive isotope beams. This paper discusses emittance measurements of low intensity beams, typical emittance scans for the surface ion source and the resonant laser ionized source for different source parameters. The observed results are compared to the simulations and discussed.

  7. Energetic and Economic Assessment of Pipe Network Effects on Unused Energy Source System Performance in Large-Scale Horticulture Facilities

    Directory of Open Access Journals (Sweden)

    Jae Ho Lee

    2015-04-01

    Full Text Available As the use of fossil fuel has increased, not only in construction, but also in agriculture due to the drastic industrial development in recent times, the problems of heating costs and global warming are getting worse. Therefore, the introduction of more reliable and environmentally-friendly alternative energy sources has become urgent and the same trend is found in large-scale horticulture facilities. In this study, among many alternative energy sources, we investigated the reserves and the potential of various different unused energy sources which have infinite potential, but are nowadays wasted due to limitations in their utilization. This study investigated the effects of the distance between the greenhouse and the actual heat source by taking into account the heat transfer taking place inside the pipe network. This study considered CO2 emissions and economic aspects to determine the optimal heat source. Payback period analysis against initial investment cost shows that a heat pump based on a power plant’s waste heat has the shortest payback period of 7.69 years at a distance of 0 km. On the other hand, the payback period of a heat pump based on geothermal heat showed the shortest payback period of 10.17 year at the distance of 5 km, indicating that heat pumps utilizing geothermal heat were the most effective model if the heat transfer inside the pipe network between the greenhouse and the actual heat source is taken into account.

  8. Reactor Safety Research Programs

    Energy Technology Data Exchange (ETDEWEB)

    Edler, S. K.

    1981-07-01

    This document summarizes the work performed by Pacific Northwest Laboratory (PNL) from January 1 through March 31, 1981, for the Division of Reactor Safety Research within the U.S. Nuclear Regulatory Commission (NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining the strength of structural graphite, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the integrity of pressurized water reactor (PWR) steam generator tubes where service-induced degradation has been indicated. Experimental data and analytical models are being provided to aid in decision-making regarding pipeto- pipe impacts following postulated breaks in high-energy fluid system piping. Core thermal models are being developed to provide better digital codes to compute the behavior of full-scale reactor systems under postulated accident conditions. Fuel assemblies and analytical support are being provided for experimental programs at other facilities. These programs include loss-ofcoolant accident (LOCA) simulation tests at the NRU reactor, Chalk River, Canada; fuel rod deformation, severe fuel damage, and postaccident coolability tests for the ESSOR reactor Super Sara Test Program, Ispra, Italy; the instrumented fuel assembly irradiation program at Halden, Norway; and experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory (INEL). These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  9. Federal Technology Alert: Ground-Source Heat Pumps Applied to Federal Facilities--Second Edition

    Energy Technology Data Exchange (ETDEWEB)

    Hadley, Donald L.

    2001-03-01

    This Federal Technology Alert, which was sponsored by the U.S. Department of Energy's Office of Federal Energy Management Programs, provides the detailed information and procedures that a Federal energy manager needs to evaluate most ground-source heat pump applications. This report updates an earlier report on ground-source heat pumps that was published in September 1995. In the current report, general benefits of this technology to the Federal sector are described, as are ground-source heat pump operation, system types, design variations, energy savings, and other benefits. In addition, information on current manufacturers, technology users, and references for further reading are provided.

  10. Preliminary assessment of the nuclide migration from the activation zone around the proposed Spallation Neutron Source facility

    Energy Technology Data Exchange (ETDEWEB)

    Dole, L.R.

    1998-09-01

    The purpose of this study is to investigate the potential impacts of migrating radionuclides from the activation zone around the proposed Spallation Neutron Source (SNS). Using conservatively high estimates of the potential inventory of radioactive activation products that could form in the proposed compacted-soil shield berm around an SNS facility on the Oak Ridge Reservation (ORR), a conservative, simplified transport model was used to estimate the potential worst-case concentrations of the 12 long-lived isotopes in the groundwater under a site with the hydrologic characteristics of the ORR.

  11. Modeling of the charge-state separation at ITEP experimental facility for material science based on a Bernas ion source.

    Science.gov (United States)

    Barminova, H Y; Saratovskyh, M S

    2016-02-01

    The experiment automation system is supposed to be developed for experimental facility for material science at ITEP, based on a Bernas ion source. The program CAMFT is assumed to be involved into the program of the experiment automation. CAMFT is developed to simulate the intense charged particle bunch motion in the external magnetic fields with arbitrary geometry by means of the accurate solution of the particle motion equation. Program allows the consideration of the bunch intensity up to 10(10) ppb. Preliminary calculations are performed at ITEP supercomputer. The results of the simulation of the beam pre-acceleration and following turn in magnetic field are presented for different initial conditions.

  12. Modeling Geometric Arrangements of TiO2-Based Catalyst Substrates and Isotropic Light Sources to Enhance the Efficiency of a Photocatalystic Oxidation (PCO) Reactor

    Science.gov (United States)

    Richards, Jeffrey T.; Levine, Lanfang H.; Husk, Geoffrey K.

    2011-01-01

    The closed confined environments of the ISS, as well as in future spacecraft for exploration beyond LEO, provide many challenges to crew health. One such challenge is the availability of a robust, energy efficient, and re-generable air revitalization system that controls trace volatile organic contaminants (VOCs) to levels below a specified spacecraft maximum allowable concentration (SMAC). Photocatalytic oxidation (PCO), which is capable of mineralizing VOCs at room temperature and of accommodating a high volumetric flow, is being evaluated as an alternative trace contaminant control technology. In an architecture of a combined air and water management system, placing a PCO unit before a condensing heat exchanger for humidity control will greatly reduce the organic load into the humidity condensate loop ofthe water processing assembly (WPA) thereby enhancing the life cycle economics ofthe WPA. This targeted application dictates a single pass efficiency of greater than 90% for polar VOCs. Although this target was met in laboratory bench-scaled reactors, no commercial or SBIR-developed prototype PCO units examined to date have achieved this goal. Furthermore, the formation of partial oxidation products (e.g., acetaldehyde) was not eliminated. It is known that single pass efficiency and partial oxidation are strongly dependent upon the contact time and catalyst illumination, hence the requirement for an efficient reactor design. The objective of this study is to maximize the apparent contact time and illuminated catalyst surface area at a given reactor volume and volumetric flow. In this study, a Ti02-based photocatalyst is assumed to be immobilized on porous substrate panels and illumination derived from linear isotropic light sources. Mathematical modeling using computational fluid dynamics (CFD) analyses were performed to investigate the effect of: 1) the geometry and configuration of catalyst-coated substrate panels, 2) porosity of the supporting substrate, and 3

  13. Development of the radiation models of a BWR type reactor and it facility in the SUN-RAH; Desarrollo de modelos de radiacion de un reactor tipo BWR y su instalacion en el SUN-RAH

    Energy Technology Data Exchange (ETDEWEB)

    Barron A, I. [Facultad de Ingenieria, UNAM, 04510 Mexico D.F. (Mexico)]. e-mail: isbarron@yahoo.com.mx

    2005-07-01

    This work about generation models, transport in processes and radioactive contamination of areas of a BWR central, is an amplification to the project developed in the UNAM to have a support tool in subjects or electric generation courses. It is planned about the implementation of models of radiation generation in a BWR type reactor for complement the functions developed in the University Simulator of Nucleo electric- Boiling water reactor (SUN-RAH) which it has been implemented in Simulink of MatLab and it has a model for the dynamics of one nucleo electric central that presents the main characteristics of the reactor vessel, the recirculation system, steam lines, turbines, generator, condensers and feeding water, defined by the main processes that intervene in the generation of energy of these plants. By this way the radiation monitoring systems for area and process, operate simultaneously with the processes of energy generation, with that is possible to observe the changes that present with respect to the operation conditions of the plant, and likewise to appreciate the radiation transport process through the components of the reactor, steam lines and turbines, for different operation conditions and possible faults that they could be presented during the reactor operation. (Author)

  14. UV Lamp as a Facile Ozone Source for Structural Analysis of Unsaturated Lipids Via Electrospray Ionization-Mass Spectrometry

    Science.gov (United States)

    Stinson, Craig A.; Zhang, Wenpeng; Xia, Yu

    2017-12-01

    Ozonolysis of alkene functional groups is a type of highly specific and effective chemical reaction, which has found increasing applications in structural analysis of unsaturated lipids via coupling with mass spectrometry (MS). In this work, we utilized a low-pressure mercury lamp (6 W) to initiate ozonolysis inside electrospray ionization (ESI) sources. By placing the lamp near a nanoESI emitter that partially transmits 185 nm ultraviolet (UV) emission from the lamp, dissolved dioxygen in the spray solution was converted into ozone, which subsequently cleaved the double bonds within fatty acyls of lipids. Solvent conditions, such as presence of water and acid solution pH, were found to be critical in optimizing ozonolysis yields. Fast (on seconds time scale) and efficient (50%-100% yield) ozonolysis was achieved for model unsaturated phospholipids and fatty acids with UV lamp-induced ozonolysis incorporated on a static and an infusion nanoESI source. The method was able to differentiate double bond location isomers and identify the geometry of the double bond based on yield. The analytical utility of UV lamp-induced ozonolysis was further demonstrated by implementation on a liquid chromatography (LC)-MS platform. Ozonolysis was effected in a flow microreactor that was made from ozone permeable tubing, so that ambient ozone produced by the lamp irradiation could diffuse into the reactor and induce online ozonolysis post-LC separation and before ESI-MS.

  15. Comparison of three small-break loss-of-coolant accident tests with different break locations using the system-integrated modular advanced reactor-integral test loop facility to estimate the safety of the smart design

    Directory of Open Access Journals (Sweden)

    Hwang Bae

    2017-08-01

    Full Text Available Three small-break loss-of-coolant accident (SBLOCA tests with safety injection pumps were carried out using the integral-effect test loop for SMART (System-integrated Modular Advanced ReacTor, i.e., the SMART-ITL facility. The types of break are a safety injection system line break, shutdown cooling system line break, and pressurizer safety valve line break. The thermal–hydraulic phenomena show a traditional behavior to decrease the temperature and pressure whereas the local phenomena are slightly different during the early stage of the transient after a break simulation. A safety injection using a high-pressure pump effectively cools down and recovers the inventory of a reactor coolant system. The global trends show reproducible results for an SBLOCA scenario with three different break locations. It was confirmed that the safety injection system is robustly safe enough to protect from a core uncovery.

  16. Comparison of three small-break loss-of-coolant accident tests with different break locations using the system-integrated modular advanced reactor-integral test loop facility to estimate the safety of the smart design

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Hwang; Ryu, Sung Uk; Yi, Sung Jae; Park, Hyun Sik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kim, Dong Eok [Dept. of Precision Mechanical Engineering, Kyungpook National University, Sangju (Korea, Republic of)

    2017-08-15

    Three small-break loss-of-coolant accident (SBLOCA) tests with safety injection pumps were carried out using the integral-effect test loop for SMART (System-integrated Modular Advanced ReacTor), i.e., the SMART-ITL facility. The types of break are a safety injection system line break, shutdown cooling system line break, and pressurizer safety valve line break. The thermal–hydraulic phenomena show a traditional behavior to decrease the temperature and pressure whereas the local phenomena are slightly different during the early stage of the transient after a break simulation. A safety injection using a high-pressure pump effectively cools down and recovers the inventory of a reactor coolant system. The global trends show reproducible results for an SBLOCA scenario with three different break locations. It was confirmed that the safety injection system is robustly safe enough to protect from a core uncovery.

  17. The spallation neutron source SINQ. A new large facility for research at PSI

    Energy Technology Data Exchange (ETDEWEB)

    Bauer, G.S.; Crawford, J.F.

    1994-12-31

    This document is intended to familiarize the non-specialist with the principles of neutron scattering and some of its applications. It presents an overview of the foundations of neutron scattering, the basic types of instruments used, and their principles of operation. The design concept and some technical details of the spallation neutron source are described for the benefit of the scientifically or technically interested reader. In future this source will form the heart of the instruments available to PSI`s wide community of neutron scattering researchers. (author) 32 figs., 1 tab.

  18. Implementation of the k{sub 0} technique using multi-detectors on diverse irradiation facilities of TRIGA Reactor; Implementacion de la tecnica k{sub 0} usando multidetectores en diferentes instalaciones de irradiacion del Reactor TRIGA

    Energy Technology Data Exchange (ETDEWEB)

    Caldera C, M. de G.

    2013-07-01

    The k{sub 0} method with the technique of neutron activation analysis allows obtaining important characteristics parameters that describe a nuclear reactor. Among these parameters are the form factor of epithermal neutron flux, α and the ratio of thermal neutron flux with respect to the epithermal neutron flux, f. These parameters were obtained by irradiation of two different monitors, one of Au-Zr and the other of Au-Mo-Cr, where the last one was made and implemented for the first time. Both monitors were irradiated in different positions in the TRIGA Mark III Reactor at the National Institute of Nuclear Research. (Author)

  19. Access to water source, latrine facilities and other risk factors of active trachoma in Ankober, Ethiopia.

    Directory of Open Access Journals (Sweden)

    Ilya Golovaty

    2009-08-01

    Full Text Available This study aims to determine the prevalence and correlates of active trachoma in Ankober, Ethiopia.A cross-sectional community-based study was conducted during July 2007. A total of 507 children (ages 1-9 years, from 232 households were included in the study. All children were examined for trachoma by ophthalmic nurses using the WHO simplified clinical grading system. Interviews and observations were used to assess risk factors. Logistic regression procedures were used to determine associations between potential risk factors and signs of active trachoma.Overall, the prevalence of active trachoma was found to be 53.9% (95%CI 49.6%-58.2%. Presence of fly-eye (fly contact with the eyelid margin during eye examination (Odds Ratio (OR = 4.03 95% CI 1.40-11.59, absence of facial cleanliness (OR = 7.59; 95%CI 4.60-12.52, an illiterate mother (OR = 5.88; 95%CI 2.10-15.95, lack of access to piped water (OR = 2.19; 95%CI 1.14-6.08, and lack of access to latrine facilities (OR = 4.36; 95%CI 1.49-12.74 were statistically significantly associated with increased risk of active trachoma.Active trachoma among children 1-9 years of age in Ankober is highly prevalent and significantly associated with a number of risk factors including access to water and latrine facilities. Trachoma prevention programs that include improved access to water and sanitation, active fly control, and hygiene education are recommended to lower the burden of trachoma in Ankober, Ethiopia.

  20. A new 2.5 MeV injector and beam test facility for the spallation neutron source

    Science.gov (United States)

    Welton, R. F.; Aleksandrov, A.; Han, B. X.; Kang, Y. W.; Middendorf, M. M.; Murray, S. N.; Piller, M.; Pennisi, T. R.; Peplov, V.; Saethre, R.; Santana, M.; Stinson, C.; Stockli, M. P.

    2017-08-01

    The U.S. Spallation Neutron Source (SNS) now operates with 1.2 MW of beam power on target with the near-term goal of delivering 1.4 MW and a longer-term goal of delivering >2 MW to support a planned second target station. Presently, H- beam pulses (50-60 mA, 1 ms, 60 Hz) from an RF-driven, Cs-enhanced, multi-cusp ion source are first accelerated to 2.5 MeV by a Radio Frequency Quadrupole (RFQ) accelerator, injected into a ˜1 GeV linac, compressed to transmission, the initial applications of the BTF will be to conduct 6D beam dynamic studies, develop & demonstrate ion sources capable of meeting the current and future requirements of the SNS, and contribute to neutron moderator development. This report provides a facility update, description of the BTF ion source systems as well as a discussion of the first LEBT and RFQ beam current measurements performed at the BTF.

  1. Reactors as a Source of Antineutrinos: Effects of Fuel Loading and Burnup for Mixed-Oxide Fuels

    Science.gov (United States)

    Bernstein, Adam; Bowden, Nathaniel S.; Erickson, Anna S.

    2018-01-01

    In a conventional light-water reactor loaded with a range of uranium and plutonium-based fuel mixtures, the variation in antineutrino production over the cycle reflects both the initial core fissile inventory and its evolution. Under an assumption of constant thermal power, we calculate the rate at which antineutrinos are emitted from variously fueled cores, and the evolution of that rate as measured by a representative ton-scale antineutrino detector. We find that antineutrino flux decreases with burnup for low-enriched uranium cores, increases for full mixed-oxide (MOX) cores, and does not appreciably change for cores with a MOX fraction of approximately 75%. Accounting for uncertainties in the fission yields in the emitted antineutrino spectra and the detector response function, we show that the difference in corewide MOX fractions at least as small as 8% can be distinguished using a hypothesis test. The test compares the evolution of the antineutrino rate relative to an initial value over part or all of the cycle. The use of relative rates reduces the sensitivity of the test to an independent thermal power measurement, making the result more robust against possible countermeasures. This rate-only approach also offers the potential advantage of reducing the cost and complexity of the antineutrino detectors used to verify the diversion, compared to methods that depend on the use of the antineutrino spectrum. A possible application is the verification of the disposition of surplus plutonium in nuclear reactors.

  2. CER. Research reactors in France

    Energy Technology Data Exchange (ETDEWEB)

    Estrade, Jerome [CEA, DEN, DER, Saint-Paul-lez-Durance (France). Jules Horowitz Reactor (JHR)

    2012-10-15

    Networking and the establishment of coalitions between research reactors are important to guarantee a high technical quality of the facility, to assure well educated and trained personnel, to harmonize the codes of standards and the know-ledge of the personnel as well as to enhance research reactor utilization. In addition to the European co-operation, country-specific working groups have been established for many years, such as the French research reactor Club d'Exploitants des Reacteurs (CER). It is the association of French research reactors representing all types of research reactors from zero power up to high flux reactors. CER was founded in 1990 and today a number of 14 research reactors meet twice a year for an exchange of experience. (orig.)

  3. Compact muon source with electron accelerator for a mobile muSR facility

    Energy Technology Data Exchange (ETDEWEB)

    Nagamine, K., E-mail: kanetada.nagamine@ucr.ed [Department of Physics and Astronomy, University of California, Riverside, CA 92521 (United States); Muon Science Laboratory, IMSS, KEK, Oho, Tsukuba, Ibaraki 305-0801 (Japan); Atomic Physics Laboratory, RIKEN, Wako, Saitama 351-0191 (Japan); Miyadera, H.; Jason, A. [AOT-ABS, Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Seki, R. [Department of Physics, State University of California, Northridge, CA 91330-8268 (United States)

    2009-04-15

    In order to increase accessibility to the muSR spectroscopy for people in various fields of science and engineering, a conceptual design study was made to realize a compact and inexpensive muon source by using 300 MeV electron microtron and a large-acceptance muon-capture. Advanced radiography imaging with muon spin probes will become possible for bio-medical studies, inspection of re-enforced architectures, etc.

  4. Ageing management and refurbishment of Ghana Research Reactor-1 (GHARR-1)

    Energy Technology Data Exchange (ETDEWEB)

    Amponsahabu, Edward Oscar; Gbadago, Joseph Korbla; Addo, Moses Ankamah; Sogbadji, Robert Bright Mawuko; Odoi, Henry Cecil; Gyamfi, Kwame; Ampong, Atta Gyekye; Opate, Nicholas Sackitey [Ghana Atomic Energy Commission, Accra (Ghana)

    2013-07-01

    Ageing management is an essential component of the routine practices at the Ghana Research Reactor-1 (GHARR-1) Facility. The reactor is Miniature Neutron Source Reactor with a rated power of 30 kW. GHARR-1 was installed and attained criticality on December 17, 1994 and commissioned on 8th March, 1995. It has since been in operation. The routine practices and operational procedures have been set out with clear emphasis on ageing policy at the facility. Some electronic components are changed regularly during maintenance sessions and keeping to regular purification of the reactor and pool water to mitigate against corrosion. This paper outlines the ageing management programme, mitigation practices, strategies for ageing management, periodic safety reviews, consideration of ageing during designing, design features for components and unit replacement, top beryllium shim addition, and succession planning. Information sharing with other operating organization is one of the means considered by GHARR-1 to attain excellence.

  5. Effect of dissolved oxygen on nitrate removal using polycaprolactone as an organic carbon source and biofilm carrier in fixed-film denitrifying reactors.

    Science.gov (United States)

    Luo, Guozhi; Xu, Guimei; Gao, Jinfang; Tan, Hongxin

    2016-05-01

    Nitrate-nitrogen (NO3(-)-N) always accumulates in commercial recirculating aquaculture systems (RASs) with aerobic nitrification units. The ability to reduce NO3(-)-N consistently and confidently could help RASs to become more sustainable. The rich dissolved oxygen (DO) content and sensitive organisms stocked in RASs increase the difficulty of denitrifying technology. A denitrifying process using biologically degradable polymers as an organic carbon source and biofilm carrier was proposed because of its space-efficient nature and strong ability to remove NO3(-)-N from RASs. The effect of dissolved oxygen (DO) levels on heterotrophic denitrification in fixed-film reactors filled with polycaprolactone (PCL) was explored in the current experiment. DO conditions in the influent of the denitrifying reactors were set up as follows: the anoxic treatment group (Group A, average DO concentration of 0.28±0.05mg/L), the low-oxygen treatment DO group (Group B, average DO concentration of 2.50±0.24mg/L) and the aerated treatment group (Group C, average DO concentration of 5.63±0.57mg/L). Feeding with 200mg/L of NO3(-)-N, the NO3(-)-N removal rates were 1.53, 1.60 and 1.42kg/m(3) PCL/day in Groups A, B and C, respectively. No significant difference in NO3(-)-N removal rates was observed among the three treatments. It was concluded that the inhibitory effects of DO concentrations lower than 6mg/L on heterotrophic denitrification in the fixed-film reactors filled with PCL can be mitigated. Copyright © 2015. Published by Elsevier B.V.

  6. Beamlines of the biomedical imaging and therapy facility at the Canadian light source-Part 1

    Energy Technology Data Exchange (ETDEWEB)

    Wysokinski, Tomasz W. [Canadian Light Source, Saskatoon, SK (Canada)], E-mail: tomasz.wysokinski@lightsource.ca; Chapman, Dean [Anatomy and Cell Biology, University of Saskatchewan, Saskatoon, SK (Canada); Adams, Gregg [Western College of Veterinary Medicine, Saskatoon, SK (Canada); Renier, Michel [European Synchrotron Radiation Facility, Grenoble (France); Suortti, Pekka [Department of Physical Sciences, University of Helsinki (Finland); Thomlinson, William [Canadian Light Source, Saskatoon, SK (Canada)

    2007-11-11

    The BioMedical Imaging and Therapy (BMIT) Facility will provide synchrotron-specific imaging and therapy capabilities. This paper describes one of the BMIT beamlines: the bend magnet (BM) beamline 05B1-1. It plays a complementary role to the insertion device (ID) beamline 051D-2 and allows either monochromatic or filtered white beam to be used in the experimental hutch. The monochromatic spectral range will span 8-40 keV, and the beam is more than 200 mm wide in the experimental hutch for imaging studies of small and medium-size animals (up to sheep size). The experimental hutch will have a positioning system that will allow imaging (computed tomography and planar imaging) as well as radiation therapy applications with both filtered white and monochromatic X-ray beams and will handle subjects up to 120 kg. Several different focal plane detectors (cameras) will be available with resolutions ranging from 10 to 150 {mu}m.

  7. Source Definition Issue for KN Power - Front Range Energy Associates, LLC/PSCo Generating Facility

    Science.gov (United States)

    This document may be of assistance in applying the New Source Review (NSR) air permitting regulations including the Prevention of Significant Deterioration (PSD) requirements. This document is part of the NSR Policy and Guidance Database. Some documents in the database are a scanned or retyped version of a paper photocopy of the original. Although we have taken considerable effort to quality assure the documents, some may contain typographical errors. Contact the office that issued the document if you need a copy of the original.

  8. Implementation and qualification of neutronic calculation methodology in subcritical reactors driven by external neutron sources and applications

    OpenAIRE

    Thiago Carluccio

    2011-01-01

    O trabalho teve como objetivo a investigação de Metodologias de Cálculo dos Reatores Subcríticos acionados por fonte externa de nêutrons, tais como, \\"Accelerator Driven Subcritical Reactor\\" (ADSR) e \\"Fusion Driven Subcritical Reator\\" (FDSR) , que são reatores nucleares subcríticos com uma fonte externa de nêutrons. Tais nêutrons são produzidos, no caso do ADSR, através da interação de partículas aceleradas (prótons, deutério) com um alvo (Pb, Bi, etc) ou através das reações de fusão, no c...

  9. RELAP5 model for advanced neutron source reactor thermal-hydraulic transients, three-element-core design

    Energy Technology Data Exchange (ETDEWEB)

    Chen, N.C.J.; Wendel, M.W.; Yoder, G.L.

    1996-02-01

    In order to utilize reduced enrichment fuel, the three-element-core design has been proposed. The proposed core configuration consists of inner, middle, and outer elements, with the middle element offset axially beneath the inner and outer elements, which are axially aligned. The three-element-core RELAP5 model assumes that the reactor hardware is changed only within the core region, so that the loop piping, heat exchangers, and pumps remain as assumed for the two-element-core configuration. However, the total flow rate through the core is greater and the pressure drop across the core is less so that the primary coolant pumps and heat exchangers are operating at a different point in their performance curves. This report describes the new RELAP5 input for the core components.

  10. Preliminary analysis of the postulated changes needed to achieve rail cask handling capabilities at selected light water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Konzek, G.J.

    1986-02-01

    Reactor-specific railroad and crane information for all LWRs in the US was extracted from current sources of information. Based on this information, reactors were separated into two basic groups consisting of reactors with existing, usable rail cask capabilities and those without these capabilities. The latter group is the main focus of this study. The group of reactors without present rail cask handling capabilities was further separated into two subgroups consisting of reactors considered essentially incapable of handling a large rail cask of about 100 tons and reactors where postulated facility changes could result in rail cask handling capabilities. Based on a selected population of 127 reactors, the results of this assessment indicate that usable rail cask capabilities exist at 83 (65%) of the reactors. Twelve (27%) of the remaining 44 reactors are deemed incapable of handling a large rail cask without major changes, and 32 reactors are considered likely candidates for potentially achieving rail cask handling capabilities. In the latter group, facility changes were postulated that would conceptually enable these reactors to handle large rail casks. The estimated cost per plant of required facility changes varied widely from a high of about $35 million to a low of <$0.3 million. Only 11 of the 32 plants would require crane upgrades. Spur track and right-of-way costs would apparently vary widely among sites. These results are based on preliminary analyses using available generic cost data. They represent lower bound values that are useful for developing an initial assessment of the viability of the postulated changes on a system-wide basis, but are not intended to be absolute values for specific reactors or sites.

  11. A New 14 GHz ECR Ion Source for the Heavy Ion Accelerator Facility ATLAS

    Science.gov (United States)

    Schlapp, M.; Vondrasek, R. C.; Szczech, J.; Billquist, P. J.; Pardo, R. C.; Xie, Z. Q.; Harkewicz, R.

    1997-05-01

    A new 14 GHz ECRIS has been designed and built over the last two years. The source design incorporates the latest results from ECR developments to produce intense beams of highly charged ions, i.e. an improved magnetic electron confinement. The aluminum plasma chamber and extraction electrode as well as a biased disk on axis at the microwave injection side donate additional electrons to the plasma, making use of the large secondary electron yield from aluminum oxide. The source will be capable of ECR plasma heating using two different frequencies simultaneously to increase the electron energy gain for the production of high charge states. To be able to deliver usable intensities of the heaviest ion beams the design will also allow axial access for metal evaporation ovens and solid material samples. The main design goal is to produce several eμA of at least ^238U^34+ in order to accelerate the beam to coulomb-barrier energies without further stripping. That should also significantly improve the beam quality over beams requiring stripping for acceleration. This work was supported by US D.O.E. Nuclear Physics Division under contract W-31-109-ENG-38.

  12. Mirror reactor surface study

    Energy Technology Data Exchange (ETDEWEB)

    Hunt, A. L.; Damm, C. C.; Futch, A. H.; Hiskes, J. R.; Meisenheimer, R. G.; Moir, R. W.; Simonen, T. C.; Stallard, B. W.; Taylor, C. E.

    1976-09-01

    A general survey is presented of surface-related phenomena associated with the following mirror reactor elements: plasma first wall, ion sources, neutral beams, director converters, vacuum systems, and plasma diagnostics. A discussion of surface phenomena in possible abnormal reactor operation is included. Several studies which appear to merit immediate attention and which are essential to the development of mirror reactors are abstracted from the list of recommended areas for surface work. The appendix contains a discussion of the fundamentals of particle/surface interactions. The interactions surveyed are backscattering, thermal desorption, sputtering, diffusion, particle ranges in solids, and surface spectroscopic methods. A bibliography lists references in a number of categories pertinent to mirror reactors. Several complete published and unpublished reports on surface aspects of current mirror plasma experiments and reactor developments are also included.

  13. Reactor safeguards

    CERN Document Server

    Russell, Charles R

    1962-01-01

    Reactor Safeguards provides information for all who are interested in the subject of reactor safeguards. Much of the material is descriptive although some sections are written for the engineer or physicist directly concerned with hazards analysis or site selection problems. The book opens with an introductory chapter on radiation hazards, the construction of nuclear reactors, safety issues, and the operation of nuclear reactors. This is followed by separate chapters that discuss radioactive materials, reactor kinetics, control and safety systems, containment, safety features for water reactor

  14. Reactor operation

    CERN Document Server

    Shaw, J

    2013-01-01

    Reactor Operation covers the theoretical aspects and design information of nuclear reactors. This book is composed of nine chapters that also consider their control, calibration, and experimentation.The opening chapters present the general problems of reactor operation and the principles of reactor control and operation. The succeeding chapters deal with the instrumentation, start-up, pre-commissioning, and physical experiments of nuclear reactors. The remaining chapters are devoted to the control rod calibrations and temperature coefficient measurements in the reactor. These chapters also exp

  15. A time resolved microfocus XEOL facility at the Diamond Light Source

    Science.gov (United States)

    Mosselmans, J. F. W.; Taylor, R. P.; Quinn, P. D.; Finch, A. A.; Cibin, G.; Gianolio, D.; Sapelkin, A. V.

    2013-03-01

    We have constructed a Time-Resolved X-ray Excited Optical Luminescence (TR-XEOL) detection system at the Microfocus Spectroscopy beamline I18 at the Diamond Light Source. Using the synchrotron in "hybrid bunch mode", the data collection is triggered by the RF clock, and we are able to record XEOL photons with a time resolution of 6.1 ps during the 230 ns gap between the hybrid bunch and the main train of electron bunches. We can detect photons over the range 180-850 nm using a bespoke optical fibre, with X-ray excitation energies between 2 and 20 keV. We have used the system to study a range of feldspars. The detector is portable and has also been used on beamline B18 to collect Optically Determined X-ray Absorption Spectroscopy (OD-XAS) in QEXAFS mode.

  16. Multimodal hard x-ray nanoprobe facility by nested Montel mirrors aimed for 40nm resolution at Taiwan Photon Source

    Energy Technology Data Exchange (ETDEWEB)

    Yin, Gung-Chian, E-mail: gcyin@nsrrc.org.tw; Chang, Shi-Hung; Chen, Bo-Yi; Chen, Huang-Yeh; Lin, Bi-Hsuan; Tseng, Shao-Chin; Lee, Chian-Yao; Tang, Mau-Tsu [National Synchrotron Radiation Research Center, Hsinchu 30076, Taiwan (China); Wu, Shao-Yun [National Tsing-Hua University, Hsinchu 30076, Taiwan (China)

    2016-01-28

    The hard X-ray nanoprobe facility at Taiwan Photon Source (TPS) provides multimodal X-ray detections, including XRF, XAS, XEOL, projection microscope, CDI, etc. Resulting from the large numerical aperture obtained by utilizing nested Montel mirrors, the beamline with a moderate length 75 meters can conduct similar performance with those beamlines longer than 100 meters. The mirrors are symmetrically placed with a 45 degrees cut. The beamline optics is thus designed to take the advantage of the symmetry of mirrors such that a round focal spot is accomplished. The size and the divergence of the focus spot are simulated around 40 nm and 6.29 mrad, respectively. The whole facility including the beamline and the stations will be operated under vacuum to preserve the photon coherence as well as to prevent the system from unnecessary environmental interference. A SEM in close cooperation with laser interferometers is equipped to precisely locate the position of the sample. This endstation is scheduled to be commissioned in the fall of 2016.

  17. 3-dimensional shielding design for a spallation neutron source facility in the high-intensity proton accelerator project

    Energy Technology Data Exchange (ETDEWEB)

    Tamura, Masaya; Maekawa, Fujio [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2003-03-01

    Evaluation of shielding performance for a 1 MW spallation neutron source facility in the Materials and Life Science Facility being constructed in the High-Intensity Proton Accelerator Project (J-PARC) is important from a viewpoint of radiation safety and optimization of arrangement of components. This report describes evaluated results for the shielding performance with modeling three-dimensionally whole structural components including gaps between them in detail. A Monte Carlo calculation method with MCNPX2.2.6 code and LA-150 library was adopted. Streaming and void effects, optimization of shield for cost reduction and optimization of arrangement of structures such as shutters were investigated. The streaming effects were investigated quantitatively by changing the detailed structure of components and gap widths built into the calculation model. Horizontal required shield thicknesses were ranged from about 6.5 m to 7.5 m as a function of neutron beam line angles. A shutter mechanism for a horizontal neutron reflectometer that was directed downward was devised, and it was shown that the shielding performance of the shutter was acceptable. An optimal biological shield configuration was finally determined according to the calculated results. (author)

  18. THERMAL NEUTRON FLUX MAPPING ON A TARGET CAPSULE AT RABBIT FACILITY OF RSG-GAS REACTOR FOR USE IN k0-INAA

    Directory of Open Access Journals (Sweden)

    Sutisna Sutisna

    2015-03-01

    Full Text Available Instrumental neutron activation analysis based on the k0 method (k0-INAA requires the availability of the accurate reactor parameter data, in particular a thermal neutron flux that interact with a targets inside the target capsule. This research aims to determine and map the thermal neutron flux inside the capsule and irradiation channels used for the elemental quantification using the k0-AANI. Mapping of the thermal neutron flux (фth on two type of irradiation capsule have been done for RS01 and RS02 facilities of RSG-GAS reactor. Thermal neutron flux determined using Al-0,1%Au alloy through 197Au(n,g 198Au nuclear reaction, while the flux mapping done using statistics R. Thermal neutron flux are calculated using k0-IAEA software provided by IAEA. The results showed the average thermal neutron flux is (5.6±0.3×10+13 n.cm-2.s-1; (5.6±0.4×10+13 n.cm-2.s-1; (5.2±0.4×10+13 n.cm-2.s-1 and (5.3±0.4×10+13 n.cm-2.s-1 for Polyethylene capsule of 1st , 2nd, 3rd and 4th layer respectively. In the case of Aluminum capsule, the thermal neutron flux was lower compared to that on Polyethylene capsule. There were (3.0±0.2×10+13 n.cm-2.s-1; (2.8±0.1×10+13 n.cm-2.s-1; (3.2±0.3×10+13 n.cm-2.s-1 for 1st, 2nd and 3rd layers respectively. For each layer in the capsule, the thermal neutron flux is not uniform and it was no degradation flux in the axial direction, both for polyethylene and aluminum capsules. Contour map of eight layer on polyethylene capsule and six layers on aluminum capsule for RS01 and RS02 irradiation channels had a similar pattern with a small diversity for all type of the irradiation capsule. Keywords: thermal neutron, flux, capsule, NAA   Analisis aktivasi neutron instrumental berbasis metode k0 (k0-AANI memerlukan ketersediaan data parameter reaktor yang akurat, khususnya data fluks neutron termal yang berinteraksi dengan inti sasaran di dalam kapsul target. Penelitian ini bertujuan menentukan dan memetakan fluks neutron termal

  19. The neutronic design and performance of the Indiana University Cyclotron Facility (IUCF) Low Energy Neutron Source (LENS)

    Science.gov (United States)

    Lavelle, Christopher M.

    Neutron scattering research is performed primarily at large-scale facilities. However, history has shown that smaller scale neutron scattering facilities can play a useful role in education and innovation while performing valuable materials research. This dissertation details the design and experimental validation of the LENS TMR as an example for a small scale accelerator driven neutron source. LENS achieves competitive long wavelength neutron intensities by employing a novel long pulse mode of operation, where the neutron production target is irradiated on a time scale comparable to the emission time of neutrons from the system. Monte Carlo methods have been employed to develop a design for optimal production of long wavelength neutrons from the 9Be(p,n) reaction at proton energies ranging from 7 to 13 MeV proton energy. The neutron spectrum was experimentally measured using time of flight, where it is found that the impact of the long pulse mode on energy resolution can be eliminated at sub-eV neutron energies if the emission time distribution of neutron from the system is known. The emission time distribution from the TMR system is measured using a time focussed crystal analyzer. Emission time of the fundamental cold neutron mode is found to be consistent with Monte Carlo results. The measured thermal neutron spectrum from the water reflector is found to be in agreement with Monte Carlo predictions if the scattering kernels employed are well established. It was found that the scattering kernels currently employed for cryogenic methane are inadequate for accurate prediction of the cold neutron intensity from the system. The TMR and neutronic modeling have been well characterized and the source design is flexible, such that it is possible for LENS to serve as an effective test bed for future work in neutronic development. Suggestions for improvements to the design that would allow increased neutron flux into the instruments are provided.

  20. The Tower Shielding Facility: Its glorious past

    Energy Technology Data Exchange (ETDEWEB)

    Muckenthaler, F.J.

    1997-05-07

    The Tower Shielding Facility (TSF) is the only reactor facility in the US that was designed and built for radiation-shielding studies in which both the reactor source and shield samples could be raised into the air to allow measurements to be made without interference from ground scattering or other spurious effects. The TSF proved its usefulness as many different programs were successfully completed. It became active in work for the Defense Atomic Support Agency (DASA) Space Nuclear Auxiliary Power, Defense Nuclear Agency, Liquid Metal Fast Breeder Reactor Program, the Gas-Cooled and High-Temperature Gas-Cooled Reactor programs, and the Japanese-American Shielding Program of Experimental Research, just to mention a few of the more extensive ones. The history of the TSF as presented in this report describes the various experiments that were performed using the different reactors. The experiments are categorized as to the programs which they supported and placed in corresponding chapters. The experiments are described in modest detail, along with their purpose when appropriate. Discussion of the results is minimal, but references are given to more extensive topical reports.

  1. Acclimation of denitrifying activated sludge to a single vs. complex external carbon source during a start-up of sequencing batch reactors treating ammonium-rich anaerobic sludge digester liquors.

    Science.gov (United States)

    Czerwionka, Krzysztof; Luczkiewicz, Aneta; Majtacz, Joanna; Kowal, Przemyslaw; Jankowska, Katarzyna; Ciesielski, Slawomir; Pagilla, Krishna; Makinia, Jacek

    2014-11-01

    In this study, denitrification of ammonium-reach anaerobic sludge digester liquor was investigated during start-up periods of two laboratory-scale "fill-and-draw" reactors. One reactor was fed with a single carbon source (ethanol), whereas the other reactor was fed with a complex carbon source (fusel oil). During two acclimation experiments, the structure of microbial community involved in denitrification was analyzed using 16S rDNA polymerase chain reaction-denaturing gradient gel electrophoresis fingerprints and fluorescent in situ hybridization. The characteristics of the mixed liquor were additionally supported by regular measurements of nitrate uptake rates. The addition of fusel oil and ethanol resulted in a significant enhancement of the denitrification rate and efficiency combined with the increasing volumetric addition of sludge digester liquor up to 15 % of the reactor volume. The microbiological analyses revealed that the addition of sludge digester liquor as well as both external carbon sources (fusel oil and ethanol) did not affect the structure of microbial communities in a severe way. In both reactors, Curvibacter sp. and Azoarcus sp. were found as the most abundant representatives of denitrifiers.

  2. Source inventory for Department of Energy solid low-level radioactive waste disposal facilities: What it means and how to get one of your own

    Energy Technology Data Exchange (ETDEWEB)

    Smith, M.A. [Science Applications International Corp., Oak Ridge, TN (United States). Environmental Compliance Group

    1991-12-31

    In conducting a performance assessment for a low-level waste (LLW) disposal facility, one of the important considerations for determining the source term, which is defined as the amount of radioactivity being released from the facility, is the quantity of radioactive material present. This quantity, which will be referred to as the source inventory, is generally estimated through a review of historical records and waste tracking systems at the LLW facility. In theory, estimating the total source inventory for Department of Energy (DOE) LLW disposal facilities should be possible by reviewing the national data base maintained for LLW operations, the Solid Waste Information Management System (SWIMS), or through the annual report that summarizes the SWIMS data, the Integrated Data Base (IDB) report. However, in practice, there are some difficulties in making this estimate. This is not unexpected, since the SWIMS and the IDB were not developed with the goal of developing a performance assessment source term in mind. The practical shortcomings using the existing data to develop a source term for DOE facilities will be discussed in this paper.

  3. Biological nutrient removal by a sequencing batch reactor (SBR) using an internal organic carbon source in digested piggery wastewater.

    Science.gov (United States)

    Obaja, D; Macé, S; Mata-Alvarez, J

    2005-01-01

    Experiments in a lab-scale SBR were conducted to demonstrate the feasibility of using an internal carbon source (non-digested pig manure) for biological nitrogen and phosphorus removal in digested piggery wastewater. The internal C-source used for denitrification had similar effects to acetate. 99.8% of nitrogen and 97.8% of phosphate were removed in the SBR, from an initial content in the feed of 900 mg/l ammonia and 90 mg/l phosphate.

  4. Reactor flux calculations

    Energy Technology Data Exchange (ETDEWEB)

    Lhuillier, D. [Commissariat à l' Énergie Atomique et aux Énergies Alternatives, Centre de Saclay, IRFU/SPhN, 91191 Gif-sur-Yvette (France)

    2013-02-15

    The status of the prediction of reactor anti-neutrino spectra is presented. The most accurate method is still the conversion of total β spectra of fissionning isotopes as measured at research reactors. Recent re-evaluations of the conversion process led to an increased predicted flux by few percent and were at the origin of the so-called reactor anomaly. The up to date predictions are presented with their main sources of error. Perspectives are given on the complementary ab-initio predictions and upcoming experimental cross-checks of the predicted spectrum shape.

  5. Development of safety analysis technology for LMR; development of safety analysis technology for LMR/ development of radioactive source terms in liquid metal reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kamg, Chang Sun; Song, Jae Hyuk; Cho, Young Ho; Go, Hyun Seok; Lee, Young Wook; Jang, Mee [Seoul National Univ., Seoul (Korea)

    2002-05-01

    PRISM source term is reviewed that had much influence on development of KALIMER. A series of experiments and simulations made in many countries are studied and source terms for liquid metal reactors except for PRISM are also reviewed. Thus, KALIMER HCDA source term is determined reasonably and conservatively. Sodium pool fire and sodium spray fire are selected as HCDA scenarios for performance analysis for KALIMER containment dome. Performance analysis for KALIMER containment dome was carried out using CONTAIN-LMR code. Comparing code calculation results with containment design parameters, we determined whether KALIMER containment dome would fail or not. The major parameters are peak pressure and peak temperature. Then, using CONTAIN-LMR code and MACCS code, radiation dose at site boundary was calculated. By comparing code calculation results with PAG guideline and 10 CFR limit, radiological consequences for HCDA was evaluated. The performance analysis showed that KALIMER containment could maintain its integrity and achieve its purpose to mitigate accident consequences and prevent release of radioactive materials in case of HCDA. Sodium pool fire caused higher radiation doses than sodium spray fire. But, dose values evaluated for HCDA were much lower than dose limit values for both sodium pool fire and sodium spray fire. 23 refs., 55 figs., 21 tabs. (Author)

  6. Operating limits Hanford Production Reactors. Revision 2

    Energy Technology Data Exchange (ETDEWEB)

    Owsley, G.F. [comp.

    1963-05-20

    This report is applicable to the eight operating production reactors, B, C, D, DR, F, H, KE, and KW. It covers the following: operating parameter limitations; reactivity limitations; control and safety systems; reactor fuel loading; coolant requirements with irradiated fuel in reactor; reactor confinement; test facilities; code compliance; safety instrumentation and set points; and control criteria. Also discussed are administrative procedures for process control, training, audits and inspection, and reports and records.

  7. Study on bulk shielding for a spallation neutron source facility in the high-intensity proton accelerator project

    CERN Document Server

    Maekawa, F; Takada, H; Teshigawara, M; Watanabe, N

    2002-01-01

    Under the JAERI-KEK High-Intensity Proton Accelerator Project, a spallation neutron source driven by a 3 GeV-1 MW proton beam is planed to be constructed in a main part of the Materials and Life Science Facility. This report describes results of a study on bulk shielding performance of a biological shield for the spallation neutron source by means of a Monte Carlo calculation method, that is important in terms of radiation safety and cost reduction. A shielding configuration was determined as a reference case by considering preliminary studies and interaction with other components, then shielding thickness that was required to achieve a target dose rate of 1 mu Sv/h was derived. Effects of calculation conditions such as shielding materials and dimensions on the shielding performance was investigated by changing those parameters. By taking all the results and design margins into account, a shielding configuration that was identified as the most appropriate was finally determined as follows. An iron shield regi...

  8. Development of a research nuclear reactor simulator using LABVIEW®

    Energy Technology Data Exchange (ETDEWEB)

    Lage, Aldo Marcio Fonseca; Mesquita, Amir Zacarias; Pinto, Antonio Juscelino; Souza, Luiz Claudio Andrade [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2015-07-01

    The International Atomic Energy Agency recommends the use of safety and friendly interfaces for monitoring and controlling the operational parameters of the nuclear reactors. The most important variable in the nuclear reactors control is the power released by fission of the fuel in the core which is directly proportional to neutron flux. It was developed a digital system to simulate the neutron evolution flux and monitoring their interaction on the other operational parameters. The control objective is to bring the reactor power from its source level (mW) to a few W. It is intended for education of basic reactor neutronic principles such as the multiplication factor, criticality, reactivity, period, delayed neutron and control by rods. The 250 kW IPR-R1 TRIGA research reactor at Nuclear Technology Development Center - CDTN (Belo Horizonte/Brazil) was used as reference. TRIGA reactors, developed by General Atomics (GA), are the most widely used research reactor in the world. They are cooled by light water under natural convection and are characterized by being inherently safety. The simulation system was developed using the LabVIEW® (Laboratory Virtual Instruments Engineering Workbench) software, considering the modern concept of virtual instruments (VI's). The main purpose of the system is to provide to analyze the behavior, and the tendency of some processes that occur in the reactor using a user-friendly operator interface. The TRIGA simulator system will allow the study of parameters, which affect the reactor operation, without the necessity of using the facility.(author)

  9. Facility Registry Service (FRS)

    Data.gov (United States)

    U.S. Environmental Protection Agency — The Facility Registry Service (FRS) provides an integrated source of comprehensive (air, water, and waste) environmental information about facilities across EPA,...

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

    Science.gov (United States)

    2010-01-01

    ... 10 Energy 1 2010-01-01 2010-01-01 false Office of Nuclear Reactor Regulation. 1.43 Section 1.43... Program Offices § 1.43 Office of Nuclear Reactor Regulation. The Office of Nuclear Reactor Regulation— (a... health and safety, the environment, or the safeguarding of nuclear reactor facilities; (c) Assesses and...

  11. CONVECTION REACTOR

    Science.gov (United States)

    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.

  12. Long-term operation of a pilot-scale reactor for phosphorus recovery as struvite from source-separated urine

    NARCIS (Netherlands)

    Zamora, Patricia; Georgieva, Tanya; Salcedo, Inmaculada; Elzinga, Nico; Kuntke, Philipp; Buisman, Cees J.N.

    2017-01-01

    BACKGROUND: The treatment of source separated urine allows for a more sustainable approach to nutrients recovery in actual wastewater treatment plants. Struvite precipitation from urine yields a slow-release fertilizer (struvite) with a high marketable value for agricultural use. Extensive

  13. Process and device for the regulation of operating conditions of a catalytic cracking reactor using an ionizing radiation source

    Energy Technology Data Exchange (ETDEWEB)

    Beretta, A.; Boute, B.; Pontier, R.

    1991-05-31

    An ionizing radiation source and a radiation detector allow, through a radiation absorption technique, the measurement of the in-motion solid mean density, perpendicularly to the charge migration axis. The density measurement allows the measure and control of the solid quantity and/or the fluid characteristics and/or the hydrocarbon charge.

  14. Natural convection reactor

    Energy Technology Data Exchange (ETDEWEB)

    Babcock, D.F.; Bernath, L.; Menegus, R.L.; Ring, H.F.

    1956-05-01

    A previous report described the conceptual design of a plutonium producing reactor that may be characterized as follows: Power output (2000 MW); cooling - (natural convection of light water through the reactor, up through a draft tube to an evaporative cooling pond, then back to the reactor, and fuel (400 to 500 tons of uranium enriched to 1.2% U-235). Because this reactor would be cooled by the natural convection of light water, it is believed that the construction costs would be significantly less than for a Savannah or Hanford type reactor. Such expensive items as water treatment and water pumping facilities would be eliminated entirely. The inventory of 500 tons of slightly enriched uranium, however, is an unattractive feature. It represents not only a large dollar investment but also makes the reactor less attractive for construction during periods of national emergency because of the almost certain scarcity of even slightly enriched uranium at that time. The Atomic Energy Commission asked that the design be reviewed with the objective of reducing the inventory of uranium, The results of this review are given in this report.

  15. Analytical solution of neutron transport equation in an annular reactor with a rotating pulsed source; Resolucao analitica da equacao de transporte de neutrons em um reator anelar com fonte pulsada rotativa

    Energy Technology Data Exchange (ETDEWEB)

    Teixeira, Paulo Cleber Mendonca

    2002-12-01

    In this study, an analytical solution of the neutron transport equation in an annular reactor is presented with a short and rotating neutron source of the type S(x) {delta} (x- Vt), where V is the speed of annular pulsed reactor. The study is an extension of a previous study by Williams [12] carried out with a pulsed source of the type S(x) {delta} (t). In the new concept of annular pulsed reactor designed to produce continuous high flux, the core consists of a subcritical annular geometry pulsed by a rotating modulator, producing local super prompt critical condition, thereby giving origin to a rotating neutron pulse. An analytical solution is obtained by opening up of the annular geometry and applying one energy group transport theory in one dimension using applied mathematical techniques of Laplace transform and Complex Variables. The general solution for the flux consists of a fundamental mode, a finite number of harmonics and a transient integral. A condition which limits the number of harmonics depending upon the circumference of the annular geometry has been obtained. Inverse Laplace transform technique is used to analyse instability condition in annular reactor core. A regenerator parameter in conjunction with perimeter of the ring and nuclear properties is used to obtain stable and unstable harmonics and to verify if these exist. It is found that the solution does not present instability in the conditions stated in the new concept of annular pulsed reactor. (author)

  16. Sensitivity analysis of the spectra of the core neutronic source in the calculation of radiation damage in internal of PWR reactor vessel. Internal; Analisis de sensibilidad a los espectros de la fuente neutronica del nucleo en el calculo del dano por irradiacion en los internos de la vasija de un reactor PWR

    Energy Technology Data Exchange (ETDEWEB)

    Cadenas Mendicoa, A. M.; Benito Hernandez, M.; Barrerira Pereira, P.

    2012-07-01

    This study is to analyze the sensitivity to the expected differences in the energy spectra characterizing the neutron source that radiates the vessel internals of a commercial PWR reactor, in order to quantify their influence in the quantities that determine the damage in materials metal.

  17. International workshop on cold neutron sources

    Energy Technology Data Exchange (ETDEWEB)

    Russell, G.J.; West, C.D. (comps.) (Los Alamos National Lab., NM (United States))

    1991-08-01

    The first meeting devoted to cold neutron sources was held at the Los Alamos National Laboratory on March 5--8, 1990. Cosponsored by Los Alamos and Oak Ridge National Laboratories, the meeting was organized as an International Workshop on Cold Neutron Sources and brought together experts in the field of cold-neutron-source design for reactors and spallation sources. Eighty-four people from seven countries attended. Because the meeting was the first of its kind in over forty years, much time was spent acquainting participants with past and planned activities at reactor and spallation facilities worldwide. As a result, the meeting had more of a conference flavor than one of a workshop. The general topics covered at the workshop included: Criteria for cold source design; neutronic predictions and performance; energy deposition and removal; engineering design, fabrication, and operation; material properties; radiation damage; instrumentation; safety; existing cold sources; and future cold sources.

  18. Source identification of nitrous oxide emission pathways from a single-stage nitritation-anammox granular reactor

    KAUST Repository

    Ali, Muhammad

    2016-06-16

    Nitrous oxide (N2O) production pathway in a signal-stage nitritation-anammox sequencing batch reactor (SBR) was investigated based on a multilateral approach including real-time N2O monitoring, N2O isotopic composition analysis, and in-situ analyses of spatial distribution of N2O production rate and microbial populations in granular biomass. N2O emission rate was high in the initial phase of the operation cycle and gradually decreased with decreasing NH4+ concentration. The average emission of N2O was 0.98 ± 0.42% and 1.35 ± 0.72% of the incoming nitrogen load and removed nitrogen, respectively. The N2O isotopic composition analysis revealed that N2O was produced via NH2OH oxidation and NO2− reduction pathways equally, although there is an unknown influence from N2O reduction and/or anammox N2O production. However, the N2O isotopomer analysis could not discriminate the relative contribution of nitrifier denitrification and heterotrophic denitrification in the NO2− reduction pathway. Various in-situ techniques (e.g. microsensor measurements and FISH (fluorescent in-situ hybridization) analysis) were therefore applied to further identify N2O producers. Microsensor measurements revealed that approximately 70% of N2O was produced in the oxic surface zone, where nitrifiers were predominantly localized. Thus, NH2OH oxidation and NO2 reduction by nitrifiers (nitrifier-denitrification) could be responsible for the N2O production in the oxic zone. The rest of N2O (ca. 30%) was produced in the anammox bacteria-dominated anoxic zone, probably suggesting that NO2− reduction by coexisting putative heterotrophic denitrifiers and some other unknown pathway(s) including the possibility of anammox process account for the anaerobic N2O production. Further study is required to identify the anaerobic N2O production pathways. Our multilateral approach can be useful to quantitatively examine the relative contributions of N2O production pathways. Good understanding of the key N2O

  19. Enhanced nitrogen and phosphorus removal from municipal wastewater in an anaerobic-aerobic-anoxic sequencing batch reactor with sludge fermentation products as carbon source.

    Science.gov (United States)

    Liu, Jinjin; Yuan, Yue; Li, Baikun; Zhang, Qiong; Wu, Lei; Li, Xiyao; Peng, Yongzhen

    2017-11-01

    An anaerobic-aerobic-anoxic sequencing batch reactor (AOA-SBR) using sludge fermentation products as carbon source was developed to enhance nitrogen and phosphorus removal in municipal wastewater with low C/N ratio (<4) and reduce sludge production. The AOA-SBR achieved simultaneous partial nitrification and denitrification (SND), aerobic phosphorus uptake and anoxic denitrification through the real-time control and the addition of sludge fermentation products. The average removal efficiencies of total nitrogen (TN), phosphorus (PO4(3-)-P) and chemical oxygen demand (COD) after 145-day operation were 88.8%, 99.3% and 81.2%, respectively. Nitrite accumulation ratio (NAR) reached 99.1% and sludge reduction rate reached 44.1-52.1%. Specifically, 34.4% of the TN removal was carried out by SND and 57.5% by denitrification. Illumina MiSeq sequencing indicated that ammonium-oxidizing bacteria (Nitrosomonas) were enriched and nitrite-oxidizing bacteria (Nitrospira) did not exist in AOA-SBR. The system demonstrated potential to solve the dual problem of insufficient carbon source and sludge reduction. Copyright © 2017 Elsevier Ltd. All rights reserved.

  20. Short Baseline Reactor Antineutrino-Electron Scattering Experiments and Non-Standard Neutrino Interactions at Source and Detector

    OpenAIRE

    Khan, Amir N.; McKay, Douglas W.; Tahir, F

    2014-01-01

    We investigate non-standard interaction effects in antineutrino-electron scattering experiments with baselines short enough to ignore standard oscillation phenomena. The setup is free of ambiguities from the interference between new physics and oscillation effects and is sensitive to both semileptonic new physics at the source and purely leptonic new physics in the weak interaction scattering at the detector. We draw on the TEXONO experiment as the model system, extending its analysis of non-...

  1. 9 CFR 71.20 - Approval of livestock facilities.

    Science.gov (United States)

    2010-01-01

    ... facility's sale days, which shall indicate the types of animals that will be handled at the facility on... facility will handle equine infectious anemia (EIA) reactors: —This facility will not handle horses known...

  2. The physics of accelerator driven sub-critical reactors

    Indian Academy of Sciences (India)

    Keywords. Accelerator driven systems; nuclear waste transmutation; computer codes; reactor physics; reactor noise; kinetics; burnup; transport theory; Monte Carlo; thorium utilization; neutron multiplication; sub-criticality; sub-critical facilities.

  3. Background information on sources of low-level radionuclide emissions to air

    Energy Technology Data Exchange (ETDEWEB)

    Corbit, C.D.; Herrington, W.N.; Higby, D.P.; Stout, L.A.; Corley, J.P.

    1983-09-01

    This report provides a general description and reported emissions for eight low-level radioactive source categories, including facilties that are licensed by the Nuclear Regulatory Commission (NRC) and Agreement States, and non-Department of Energy (DOE) federal facilities. The eight categories of low-level radioactive source facilities covered by this report are: research and test reactors, accelerators, the radiopharmaceutical industry, source manufacturers, medical facilities, laboratories, naval shipyards, and low-level commercial waste disposal sites. Under each category five elements are addressed: a general description, a facility and process description, the emission control systems, a site description, and the radionuclides released to air (from routine operations).

  4. Preliminary Assessment of the Nuclide Migration from the Activation Zone Around the Proposed Spallation Neutron Source Facility

    Energy Technology Data Exchange (ETDEWEB)

    Dole, L.R.

    1998-09-01

    The purpose of this study is to investigate the potential impacts of migrating radionuclides from the activation zone around the proposed Spallation Neutron Source (SNS). Using conservatively high estimates of the potential inventory of radioactive activation products that could form in the proposed compacted-soil shield berm around an SNS facility on the Oak Ridge Reservation (ORR), a conservative, simplified transport model was used to estimate the potential worst-case concentrations of the 12 long-lived isotopes in the groundwater under a site with the hydrologic characteristics of the ORR. Of the 12, only 3 isotopes showed any potential to exceed the U.S. Nuclear Regulatory Commission (NRC) 10 Code of Federal Regulations (CFR) Part 20 Drinking Water Limits (DWLs). These isotopes were 14C, 22Na, and 54Mn. The latter two activation products have very short half-lives of 2.6 years and 0.854 year, respectively. Therefore, these will decay before reaching an off-site receptor, and they cannot pose off-site hazards. However, for this extremely conservative model, which overestimates the mobility of the contaminant, 14C, which has a 5,730-year half-life, was shown to represent a potential concern in the context of this study's conservative assumptions. This study examines alternative modifications to the SNS shield berm and makes recommendations.

  5. Study of the background in the measuring station at the n_TOF facility at CERN: sources and solutions

    CERN Document Server

    Zanini, L; Aerts, G; Andriamonje, Samuel A; Andrzejewski, J; Angelopoulous, A; Assimakopoulos, Panayiotis; Bacri, C-O; Badurek, G; Berthoumieux, E; Baumann, P; Beer, H; Benlliure, J; Berthier, B; Bondarenko, I; Borcea, C; Bos, A J J; Boscolo-Marchi, E; Bustreo, N; Calviño, F; Cano-Ott, D; Capote, R; Carlson, P; Charpak, Georges; Chauvin, N; Cennini, P; Chepel, V; Colonna, N; Cortés, G; Cortina-Gil, D; Corvi, F; Cusmano, A; Dababneh, S; Dahlfors, M; Damianoglou, D; David, S; Dimovasili, E; Domingo, C; Doroshenko, A; Duran-Escribano, I; Eleftheriadis, C; Embid, M; Ferrant, L; Ferrari, A; Ferreira-Marques, R; Frais-Kölbl, H; Furman, W; Fursov, B; Garzón, J A; Giomataris, Ioanis; Gledenov, Y; Gonzalez-Romero, E; Goverdovski, A; Gramegna, F; Griesmayer, E; Gunsing, F; Haefner, P; Haight, R; Heil, M; Herrera-martinez, A; Hollander, P; Ioannou, P; Isaev, S; Jericha, E; Kadi, Y; Kappeler, F; Karadimos, D; Karamanis, D; Kayukova, A; Kazakov, L; Kelic, A; Ketlerov, V; Kitis, G; Köhler, P E; Kopach, Y; Kossionides, E; Kroshkina, I; Lacoste, V; Lamboudis, C; Leeb, H; Leprêtre, A; Lopes, M; Lozano, M; Marrone, S; Martínez-Val, J M; Mastinu, P; Mengoni, A; Meunier, R; Mezentsev, A J; Milazzo, P; Minguez, E; Mitrofanov, V; Moreau, C; Müller, A; Nicolis, N; Nikolenkov, V; Oberhummer, Heinz; Pakou, A; Pancin, J; Papadopoulous, K; Papaevangelou, T; Paradela, C; Paradelis, T; Pavlik, A; Pavlopoulos, P; Perrez-Parra, A; Perriale, L; Perlado, J M; Peskov, Vladimir; Piksaikin, V; Plag, R; Plompen, A; Plukis, A; Poch, A; Policarpo, Armando; Popov, A; Popov, Y P; Pretel, C; Quesada, J M; Radermacher, E; Rapp, W; Rauscher, T; Reifarth, R; Rejmund, F; Rubbia, Carlo; Rudolf, G; Rullhusen, P; Sakelliou, L; Saldaña, F; Samylin, B; Savvidis, I; Savvidis, S; Sedyshev, P; Stéphan, C; Szalanski, P; Tagliente, G; Taín, J L; Tapia, C; Tassan-Got, L; Terchychnyi, R; Tsabaris, C; Tsangas, N; van Eijk, C W E; Vannini, G; Ventura, A; Villamarin, A; Vlachoudis, V; Vlastou, R; Voinov, A; Voss, F; Wendler, H; Wiescher, M; Wisshak, K; Zanini, L; Zeinalov, S; Zhuravlev, B; CERN. Geneva. SPS and LHC Division

    2001-01-01

    A background roughly two orders of magnitude higher than tolerable was found in the n_TOF facility at CERN during the first measurements [1]. This note describes a series of additional measurements performed in the n_TOF experimental area to study the origin and the characteristics of the background. The program of these measurements was determined taking into account the results from the simulations carried out by the EET group [2]. A first phase of measurements confirmed what was expected from the simulations, namely that the dominant source of background was due to neutrons generated by negative muon capture. Actions to reduce the background were taken according to the results from both measurements and simulations. An iron shielding wall 3.2 m thick was then placed in between the sweeping magnet and the second collimator, with the purpose of stopping most of the muons. In a second phase of measurements, results showed that the additional shielding reduced the main component of the background by about a fa...

  6. Safety Assessment Of The Centralized Storage Facility For Disused Sealed Radioactive Sources In United Republic Of Tanzania

    Energy Technology Data Exchange (ETDEWEB)

    Abel, Vitus [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of); Lee, JaeSeong [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2014-10-15

    SRS are no longer in use, they are declared as disused, and they are transferred to the central radioactive management facility (CRMF) belonging to Tanzania Atomic Energy Commission (regulatory body) and managed as radioactive waste. In order to reduce the risk associated with disused sealed radioactive sources (DSRS), the first priority would be to bring them to appropriate controls under the regulatory body. When DSRS are safely managed, regulatory body need to make assessment of the likelihood and potential impact of incidents, accidents and hazards for proper management plan. The paper applies Analytical Hierarchy Process (AHP) for assessing and allocating weights and priorities for solving the problem of mult criteria consideration for management plan. Using pairwise comparisons, the relative importance of one criterion over another can be expressed. The method allows decision makers to provide judgments about the relative importance of each criterion and to estimate radiological risk by using expert's judgments or probability of occurrence. AHP is the step by step manner where the resulting priorities are shown and the possible inconsistencies are determined. The Information provided by experts helps to rank hazards according to probability of occurrence, potential impact and mitigation cost. The strength of the AHP method lies in its ability to incorporate both qualitative and quantitative data in decision making. AHP present a powerful tool for weighting and prioritizing hazards in terms of occurrence probability. However, AHP also has some weak points. AHP requires data based on experience, knowledge and judgment which are subjective for each decision-maker.

  7. Modelling and simulation the radioactive source-term of fission products in PWR type reactors; Modelagem e simulacao do termo-fonte radioativo de produtos de fissao em reatores nucleares do tipo PWR

    Energy Technology Data Exchange (ETDEWEB)

    Porfirio, Rogilson Nazare da Silva

    1996-07-01

    The source-term was defined with the purpose the quantify all radioactive nuclides released the nuclear reactor in the case of accidents. Nowadays the source-term is limited to the coolant of the primary circuit of reactors and may be measured or modelled with computer coders such as the TFP developed in this work. The calculational process is based on the linear chain techniques used in the CINDER-2 code. The TFP code considers forms of fission products release from the fuel pellet: Recoil, Knockout and Migration. The release from the gap to the coolant fluid is determined from the ratio between activity measured in the coolant and calculated activity in the gap. Considered the operational data of SURRY-1 reactor, the TFP code was run to obtain the source=term of this reactor. From the measured activities it was verified the reliability level of the model and the employed computational logic. The accuracy of the calculated quantities were compared to the measured data was considered satisfactory. (author)

  8. Comparison of diverse climatic sources for the simulation of facilities of solar refrigeration; Comparacion de diversas fuentes climaticas para la simulacion de instalaciones de refrigeracion solar

    Energy Technology Data Exchange (ETDEWEB)

    Bujedo, L. A.; Vicente, J.; Torre, C. de; Macia, A.; Rodriguez, J.

    2008-07-01

    With the present work they are sought to analyze the results based on simulations hourly of an installation of solar refrigeration by absorption, subjected to different sources of weather data. For it, the authors have modelled one of the facilities of the Technology Centre CARTIF located in Boecillo (Valladolid), under the environment TRNSYS. (Author)

  9. Work Smarter Not Harder: Utilizing an Environmental Management Information System to Meet Regulatory Compliance and Reporting Requirements for a Major Source Title V Facility

    Science.gov (United States)

    2011-05-10

    Environmental Management Information System to Meet Regulatory Compliance and Reporting Requirements for a Major Source Title V Facility. Tannis Danley...AND SUBTITLE Work Smarter Not Harder: Utilizing an Environmental Management Information System to Meet Regulatory Compliance and Reporting...Carson) – EMS (Hawaii Garrison, West Virginia National Guard) Environmental Management Information System (EMIS) National Defense Center for Energy and

  10. Integrating cut-and-solve and semi-Lagrangean based dual ascent for the single-source capacitated facility location problem

    DEFF Research Database (Denmark)

    Gadegaard, Sune Lauth

    This paper describes how the cut-and-solve framework and semi-Lagrangean based dual ascent algorithms can be integrated in two natural ways in order to solve the single source capacitated facility location problem. The first uses the cut-and-solve framework both as a heuristic and as an exact...

  11. REACTOR COOLING

    Science.gov (United States)

    Quackenbush, C.F.

    1959-09-29

    A nuclear reactor with provisions for selectively cooling the fuel elements is described. The reactor has a plurality of tubes extending throughout. Cylindrical fuel elements are disposed within the tubes and the coolant flows through the tubes and around the fuel elements. The fuel elements within the central portion of the reactor are provided with roughened surfaces of material. The fuel elements in the end portions of the tubes within the reactor are provlded with low conduction jackets and the fuel elements in the region between the central portion and the end portions are provided with smooth surfaces of high heat conduction material.

  12. Report on design and technical standard planning of vibration controlling structure on the buildings, in the Tokai Reprocessing Facility, Power Reactor and Nuclear Fuel Development Corporation

    Energy Technology Data Exchange (ETDEWEB)

    Uryu, Mitsuru; Terada, Shuji; Shinohara, Takaharu; Yamazaki, Toshihiko; Nakayama, Kazuhiko [Power Reactor and Nuclear Fuel Development Corp., Tokai, Ibaraki (Japan). Tokai Works; Kondo, Toshinari; Hosoya, Hisashi

    1997-10-01

    The Tokai reprocessing facility buildings are constituted by a lower foundation, vibration controlling layers, and upper structure. At the vibration controlling layer, a laminated rubber aiming support of the building load and extension of the eigenfrequency and a damper aiming absorption of earthquake energy are provided. Of course, the facility buildings are directly supported at the arenaceous shale (Taga Layer) of the Miocene in the Neogene confirmed to the stablest ground, as well the buildings with high vibration resistant importance in Japan. This report shows that when the vibration controlling structure is adopted for the reprocessing facility buildings where such high vibration resistance is required, reduction of input acceleration for equipments and pipings can be achieved and the earthquake resistant safety can also be maintained with sufficient tolerance and reliability. (G.K.)

  13. Determining initial enrichment, burnup, and cooling time of pressurized-water-reactor spent fuel assemblies by analyzing passive gamma spectra measured at the Clab interim-fuel storage facility in Sweden

    Science.gov (United States)

    Favalli, A.; Vo, D.; Grogan, B.; Jansson, P.; Liljenfeldt, H.; Mozin, V.; Schwalbach, P.; Sjöland, A.; Tobin, S. J.; Trellue, H.; Vaccaro, S.

    2016-06-01

    The purpose of the Next Generation Safeguards Initiative (NGSI)-Spent Fuel (SF) project is to strengthen the technical toolkit of safeguards inspectors and/or other interested parties. The NGSI-SF team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins; (3) estimate the plutonium mass [which is also a function of the variables in (1)]; (4) estimate the decay heat; and (5) determine the reactivity of spent fuel assemblies. Since August 2013, a set of measurement campaigns has been conducted at the Central Interim Storage Facility for Spent Nuclear Fuel (Clab), in collaboration with Swedish Nuclear Fuel and Waste Management Company (SKB). One purpose of the measurement campaigns was to acquire passive gamma spectra with high-purity germanium and lanthanum bromide scintillation detectors from Pressurized Water Reactor and Boiling Water Reactor spent fuel assemblies. The absolute 137Cs count rate and the 154Eu/137Cs, 134Cs/137Cs, 106Ru/137Cs, and 144Ce/137Cs isotopic ratios were extracted; these values were used to construct corresponding model functions (which describe each measured quantity's behavior over various combinations of burnup, cooling time, and initial enrichment) and then were used to determine those same quantities in each measured spent fuel assembly. The results obtained in comparison with the operator declared values, as well as the methodology developed, are discussed in detail in the paper.

  14. Determining initial enrichment, burnup, and cooling time of pressurized-water-reactor spent fuel assemblies by analyzing passive gamma spectra measured at the Clab interim-fuel storage facility in Sweden

    Energy Technology Data Exchange (ETDEWEB)

    Favalli, A., E-mail: afavalli@lanl.gov [Los Alamos National Laboratory, Los Alamos, NM (United States); Vo, D. [Los Alamos National Laboratory, Los Alamos, NM (United States); Grogan, B. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Jansson, P. [Uppsala University, Uppsala (Sweden); Liljenfeldt, H. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Mozin, V. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Schwalbach, P. [European Commission, DG Energy, Euratom Safeguards Luxemburg, Luxemburg (Luxembourg); Sjöland, A. [Swedish Nuclear Fuel and Waste Management Company, Stockholm (Sweden); Tobin, S.J.; Trellue, H. [Los Alamos National Laboratory, Los Alamos, NM (United States); Vaccaro, S. [European Commission, DG Energy, Euratom Safeguards Luxemburg, Luxemburg (Luxembourg)

    2016-06-01

    The purpose of the Next Generation Safeguards Initiative (NGSI)–Spent Fuel (SF) project is to strengthen the technical toolkit of safeguards inspectors and/or other interested parties. The NGSI–SF team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins; (3) estimate the plutonium mass [which is also a function of the variables in (1)]; (4) estimate the decay heat; and (5) determine the reactivity of spent fuel assemblies. Since August 2013, a set of measurement campaigns has been conducted at the Central Interim Storage Facility for Spent Nuclear Fuel (Clab), in collaboration with Swedish Nuclear Fuel and Waste Management Company (SKB). One purpose of the measurement campaigns was to acquire passive gamma spectra with high-purity germanium and lanthanum bromide scintillation detectors from Pressurized Water Reactor and Boiling Water Reactor spent fuel assemblies. The absolute {sup 137}Cs count rate and the {sup 154}Eu/{sup 137}Cs, {sup 134}Cs/{sup 137}Cs, {sup 106}Ru/{sup 137}Cs, and {sup 144}Ce/{sup 137}Cs isotopic ratios were extracted; these values were used to construct corresponding model functions (which describe each measured quantity’s behavior over various combinations of burnup, cooling time, and initial enrichment) and then were used to determine those same quantities in each measured spent fuel assembly. The results obtained in comparison with the operator declared values, as well as the methodology developed, are discussed in detail in the paper.

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

    Energy Technology Data Exchange (ETDEWEB)

    Magoulas, V. E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-07-28

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

  16. Household trends in access to improved water sources and sanitation facilities in Vietnam and associated factors: findings from the Multiple Indicator Cluster Surveys, 2000-2011.

    Science.gov (United States)

    Tuyet-Hanh, Tran Thi; Lee, Jong-Koo; Oh, Juhwan; Van Minh, Hoang; Ou Lee, Chul; Hoan, Le Thi; Nam, You-Seon; Long, Tran Khanh

    2016-01-01

    Despite progress made by the Millennium Development Goal (MDG) number 7.C, Vietnam still faces challenges with regard to the provision of access to safe drinking water and basic sanitation. This paper describes household trends in access to improved water sources and sanitation facilities separately, and analyses factors associated with access to improved water sources and sanitation facilities in combination. Secondary data from the Vietnam Multiple Indicator Cluster Survey in 2000, 2006, and 2011 were analyzed. Descriptive statistics and tests of significance describe trends over time in access to water and sanitation by location, demographic and socio-economic factors. Binary logistic regressions (2000, 2006, and 2011) describe associations between access to water and sanitation, and geographic, demographic, and socio-economic factors. There have been some outstanding developments in access to improved water sources and sanitation facilities from 2000 to 2011. In 2011, the proportion of households with access to improved water sources and sanitation facilities reached 90% and 77%, respectively, meeting the 2015 MDG targets for safe drinking water and basic sanitation set at 88% and 75%, respectively. However, despite these achievements, in 2011, only 74% of households overall had access to combined improved drinking water and sanitation facilities. There were also stark differences between regions. In 2011, only 47% of households had access to both improved water and sanitation facilities in the Mekong River Delta compared with 94% in the Red River Delta. In 2011, households in urban compared to rural areas were more than twice as likely (odds ratio [OR]: 2.2; 95% confidence interval [CI]: 1.9-2.5) to have access to improved water and sanitation facilities in combination, and households in the highest compared with the lowest wealth quintile were over 40 times more likely (OR: 42.3; 95% CI: 29.8-60.0). More efforts are required to increase household access to

  17. Reactor design and kinetics study of 4,4'-dichlorobiphenyl photodecay in surfactant solution by using a photosensitizer and hydrogen source.

    Science.gov (United States)

    Chu, W; Kwan, C Y

    2003-05-01

    The system design based on the photodegradation kinetics of 4,4'-dichlorobiphenyl (4,4'-DCB) in surfactant solution with the aid of solvents (acetone and/or squalane) has been studied. Organic solvents acetone and squalane were added as a photosensitizer and a hydrogen source, respectively, to achieve better photolysis performance. The quantum yield of 4,4'-DCB photodecay in Tween 80 micellar solution in the presence of added acetone was increased from 0.032 to 0.043 at its optimal condition. Acetone was shown to be an effective photosensitizer at low concentration, but an overdose would quench the reaction. Furthermore, the addition of squalane could further promote the photodechlorination of 4,4'-DCB in the Tween 80/acetone solutions for an additional 59% of rate improvement. However, an overdose of either solvent was found to cause UV light attenuation and to reduce the observed quantum yield. This effect has been justified and quantified in this study by a proposed light attenuation model, which has also been incorporated into the kinetic equation so that the resulted formula can be used to design the UV reactors for water and wastewater treatment works.

  18. 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

    Energy Technology Data Exchange (ETDEWEB)

    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 philosophy 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.

  19. Reactor Antineutrino Signals at Morton and Boulby

    CERN Document Server

    Dye, Steve

    2016-01-01

    Increasing the distance from which an antineutrino detector is capable of monitoring the operation of a registered reactor, or discovering a clandestine reactor, strengthens the Non-Proliferation of Nuclear Weapons Treaty. This report presents calculations of reactor antineutrino interactions, from quasi-elastic neutrino-proton scattering and elastic neutrino-electron scattering, in a water-based detector operated >10 km from a commercial power reactor. It separately calculates signal from the proximal reactor and background from all other registered reactors. The main results are interaction rates and kinetic energy distributions of charged leptons scattered from quasi-elastic and elastic processes. Comparing signal and background distributions evaluates reactor monitoring capability. Scaling the results to detectors of different sizes, target media, and standoff distances is straightforward. Calculations are for two examples of a commercial reactor (P_th~3 GW) operating nearby (L~20 km) an underground facil...

  20. What can Recycling in Thermal Reactors Accomplish?

    Energy Technology Data Exchange (ETDEWEB)

    Steven Piet; Gretchen E. Matthern; Jacob J. Jacobson

    2007-09-01

    Thermal recycle provides several potential benefits when used as stop-gap, mixed, or backup recycling to recycling in fast reactors. These three roles involve a mixture of thermal and fast recycling; fast reactors are required to some degree at some time. Stop-gap uses thermal reactors only until fast reactors are adequately deployed and until any thermal-recycle-only facilities have met their economic lifetime. Mixed uses thermal and fast reactors symbiotically for an extended period of time. Backup uses thermal reactors only if problems later develop in the fast reactor portion of a recycling system. Thermal recycle can also provide benefits when used as pure thermal recycling, with no intention to use fast reactors. However, long term, the pure thermal recycling approach is inadequate to meet several objectives.

  1. Neutrino Oscillation Studies with Reactors

    CERN Document Server

    Vogel, Petr; Zhang, Chao

    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 flavors 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 $\\theta_{13}$. In the near future, reactors will help to determine the neutrino mass hierarchy and to solve the puzzling issue of sterile neutrinos.

  2. The identification of autoionizing states of atomic chromium for the resonance ionization laser ion source of the ISOLDE radioactive ion beam facility

    CERN Document Server

    Goodacre, T Day; Fedorovc, D; Fedosseev, V N; Marsh, B A; Molkanov, P; Rossel, R E; Rothe, S; Seiffert, C

    2015-01-01

    The resonance ionization laser ion source (RILIS) is the principal ion source of the ISOLDE radioactive beam facility based at CERN. Using the method of in-source resonance ionization spectroscopy, an optimal three-step, three-resonance photo-ionization scheme has been developed for chromium. The scheme uses an ionizing transition to one of the 14 newly observed autoionizing states. This work increases the range of ISOLDE-RILIS ionized beams to 32 chemical elements. Details of the spectroscopic studies are described and the new ionization scheme is summarized. A link to the complete version of this document will be added here following publication:

  3. 77 FR 22808 - Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses...

    Science.gov (United States)

    2012-04-17

    ... the Code of Federal Regulations (10 CFR), 50.92, this means that operation of the facility in... symbol associated with the nominal Reactor Coolant System operating pressure (P'). The P' provided in TS...: (a) TS 3.4.6, ``Chemistry,'' (b) TS 3.7.5, ``Flood Protection,'' (c) TS 3.7.9, ``Sealed Source...

  4. Multifunctional reactors

    NARCIS (Netherlands)

    Westerterp, K.R.

    1992-01-01

    Multifunctional reactors are single pieces of equipment in which, besides the reaction, other functions are carried out simultaneously. The other functions can be a heat, mass or momentum transfer operation and even another reaction. Multifunctional reactors are not new, but they have received much

  5. 3D CAD model of the subcritical nuclear reactor of IPN; Modelo CAD 3D del reactor nuclear subcritico del IPN

    Energy Technology Data Exchange (ETDEWEB)

    Pahuamba V, F. de J.; Delfin L, A.; Gomez T, A. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Ibarra R, G.; Del Valle G, E.; Sanchez R, A., E-mail: narehc@hotmail.com [IPN, Escuela Superior de Fisica y Matematicas, Av. IPN, Edif. 9, Unidad Profesional Adolfo Lopez Mateos, San Pedro Zacatenco, 07738 Ciudad de Mexico (Mexico)

    2016-09-15

    The three-dimensional (3D) CAD model of the subcritical reactor Chicago model 9000 of Instituto Politecnico Nacional (IPN) allows obtaining a 3D view with the dimensions of each of its components, such as: natural uranium cylindrical rods, fuel elements, hexagonal reactor core arrangement, cylindrical stainless steel tank containing the core, fuel element support grids and reactor water cleaning system. As a starting point for the development of the model, the Chicago model 9000 subcritical reactor manual provided by the manufacturer was used, the measurement and verification of the components to adapt the geometric, physical and mechanical characteristics was carried out and materials standards were used to obtain a design that allows to elaborate a new manual according to the specifications. In addition, the 3D models of the building of the Advanced Physics Laboratory, neutron generator, cobalt source and the corridors connecting to the subcritical reactor facility were developed, allowing an animated ride, developed by computer-aided design software. The manual provided by the company Nuclear Chicago, dates from the year 1959 and presents diverse deviations in the design and dimensions of the reactor components. The model developed; in addition to supporting the development of the new manual represents a learning tool to visualize the reactor components. (Author)

  6. An updated overview of low and intermediate level waste disposal facilities around the world

    Energy Technology Data Exchange (ETDEWEB)

    Cuccia, Valeria; Uemura, George; Ferreira, Vinicius Verna M.; Tello, Cledola Cassia O. de, E-mail: vc@cdtn.br, E-mail: george@cdtn.br, E-mail: vvmf@cdtn.br, E-mail: tellocc@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil); Malta, Ricardo Scott V. [SEMC Engenharia e Consultoria Ltda., Belo Horizonte, MG (Brazil)

    2011-07-01

    Low and intermediate level radioactive waste should be disposed off in proper disposal facilities. Some countries already have these facilities and others are planning theirs. Information about disposal facilities around the world is useful and necessary; however, data on this matter are usually scattered in official reports per country. In order to allow an easier access to this information, this paper aims to provide an overview of disposal facilities for low and intermediate level radioactive waste around the world, as updated as possible. Also, characteristics of the facilities are provided, when possible. Considering that the main source of radioactive waste are the activities of nuclear reactors in research or power generation, the paper will also provide a summarized overview of these reactors around the world, updated until April, 2011. This data collection may be an important tool for researchers, and other professionals in this field. Also, it might provide an overview about the final disposal of radioactive waste. (author)

  7. Limited electricity access in health facilities of sub-Saharan Africa: a systematic review of data on electricity access, sources, and reliability

    Science.gov (United States)

    Adair-Rohani, Heather; Zukor, Karen; Bonjour, Sophie; Wilburn, Susan; Kuesel, Annette C; Hebert, Ryan; Fletcher, Elaine R

    2013-01-01

    ABSTRACT Background: Access to electricity is critical to health care delivery and to the overarching goal of universal health coverage. Data on electricity access in health care facilities are rarely collected and have never been reported systematically in a multi-country study. We conducted a systematic review of available national data on electricity access in health care facilities in sub-Saharan Africa. Methods: We identified publicly-available data from nationally representative facility surveys through a systematic review of articles in PubMed, as well as through websites of development agencies, ministries of health, and national statistics bureaus. To be included in our analysis, data sets had to be collected in or after 2000, be nationally representative of a sub-Saharan African country, cover both public and private health facilities, and include a clear definition of electricity access. Results: We identified 13 health facility surveys from 11 sub-Saharan African countries that met our inclusion criteria. On average, 26% of health facilities in the surveyed countries reported no access to electricity. Only 28% of health care facilities, on average, had reliable electricity among the 8 countries reporting data. Among 9 countries, an average of 7% of facilities relied solely on a generator. Electricity access in health care facilities increased by 1.5% annually in Kenya between 2004 and 2010, and by 4% annually in Rwanda between 2001 and 2007. Conclusions: Energy access for health care facilities in sub-Saharan African countries varies considerably. An urgent need exists to improve the geographic coverage, quality, and frequency of data collection on energy access in health care facilities. Standardized tools should be used to collect data on all sources of power and supply reliability. The United Nations Secretary-General's “Sustainable Energy for All” initiative provides an opportunity to comprehensively monitor energy access in health care

  8. A large-volume press facility at the Advanced Photon Source: diffraction and imaging studies on materials relevant to the cores of planetary bodies

    CERN Document Server

    Uchida, T; Rivers, M L; Sutton, S R; Weidner, D J; Vaughan, M T; Chen, J; Li, B; Secco, R A; Rutter, M D; Liu, H

    2002-01-01

    A new large-volume, high-pressure facility is being utilized and developed as part of GeoSoilEnviroCARS at a third-generation synchrotron, the Advanced Photon Source. This user facility consists of two large-volume presses (LVP), a 2.5 MN (250 ton) LVP installed at the bending magnet beamline, and a 10 MN (1000 ton) LVP at the insertion device beamline. Here we report some techniques currently being developed with the 10 MN LVP and the latest scientific results obtained using the 2.5 MN LVP.

  9. Thermal Reactor Safety

    Energy Technology Data Exchange (ETDEWEB)

    1980-06-01

    Information is presented concerning fire risk and protection; transient thermal-hydraulic analysis and experiments; class 9 accidents and containment; diagnostics and in-service inspection; risk and cost comparison of alternative electric energy sources; fuel behavior and experiments on core cooling in LOCAs; reactor event reporting analysis; equipment qualification; post facts analysis of the TMI-2 accident; and computational methods.

  10. Fast Reactor Fuel Type and Reactor Safety Performance

    Energy Technology Data Exchange (ETDEWEB)

    R. Wigeland; J. Cahalan

    2009-09-01

    Fast Reactor Fuel Type and Reactor Safety Performance R. Wigeland , Idaho National Laboratory J. Cahalan, Argonne National Laboratory The sodium-cooled fast neutron reactor is currently being evaluated for the efficient transmutation of the highly-hazardous, long-lived, transuranic elements that are present in spent nuclear fuel. One of the fundamental choices that will be made is the selection of the fuel type for the fast reactor, whether oxide, metal, carbide, nitride, etc. It is likely that a decision on the fuel type will need to be made before many of the related technologies and facilities can be selected, from fuel fabrication to spent fuel reprocessing. A decision on fuel type should consider all impacts on the fast reactor system, including safety. Past work has demonstrated that the choice of fuel type may have a significant impact on the severity of consequences arising from accidents, especially for severe accidents of low probability. In this paper, the response of sodium-cooled fast reactors is discussed for both oxide and metal fuel types, highlighting the similarities and differences in reactor response and accident consequences. Any fast reactor facility must be designed to be able to successfully prevent, mitigate, or accommodate all consequences of potential events, including accidents. This is typically accomplished by using multiple barriers to the release of radiation, including the cladding on the fuel, the intact primary cooling system, and most visibly the reactor containment building. More recently, this has also included the use of ‘inherent safety’ concepts to reduce or eliminate the potential for serious damage in some cases. Past experience with oxide and metal fuel has demonstrated that both fuel types are suitable for use as fuel in a sodium-cooled fast reactor. However, safety analyses for these two fuel types have also shown that there can be substantial differences in accident consequences due to the neutronic and

  11. Optimization of the irradiation beam in the BNCT research facility at IEA-R1 reactor; Otimizacao do feixe de irradiacao na instalacao para estudos em BNCT junto ao reator IEA-R1

    Energy Technology Data Exchange (ETDEWEB)

    Castro, Vinicius Alexandre de

    2014-07-01

    Boron Neutron Capture Therapy (BNCT) is a radiotherapeutic technique for the treatment of some types of cancer whose useful energy comes from a nuclear reaction that occurs when thermal neutron impinges upon a Boron-10 atom. In Brazil there is a research facility built along the beam hole number 3 of the IEA-R1 research reactor at IPEN, which was designed to perform BNCT research experiments. For a good performance of the technique, the irradiation beam should be mostly composed of thermal neutrons with a minimum as possible gamma and above thermal neutron components. This work aims to monitor and evaluate the irradiation beam on the sample irradiation position through the use of activation detectors (activation foils) and also to propose, through simulation using the radiation transport code, MCNP, new sets of moderators and filters which shall deliver better irradiation fields at the irradiation sample position In this work, a simulation methodology, based on a MCNP card, known as wwg (weight window generation) was studied, and the neutron energy spectrum has been experimentally discriminated at 5 energy ranges by using a new set o activation foils. It also has been concluded that the BNCT research facility has the required thermal neutron flux to perform studies in the area and it has a great potential for improvement for tailoring the irradiation field. (author)

  12. A spallation-based irradiation test facility for fusion and future fission materials

    CERN Document Server

    Samec, K; Kadi, Y; Luis, R; Romanets, Y; Behzad, M; Aleksan, R; Bousson, S

    2014-01-01

    The EU’s FP7 TIARA program for developing accelerator-based facilities has recently demonstrated the unique capabilities of a compact and powerful spallation source for irradiating advanced nuclear materials. The spectrum and intensity of the neutron flux produced in the proposed facility fulfils the requirements of the DEMO fusion reactor for ITER, ADS reactors and also Gen III / IV reactors. Test conditions can be modulated, covering temperature from 400 to 550°C, liquid metal corrosion, cyclical or static stress up to 500 MPa and neutron/proton irradiation damage of up to 25 DPA per annum. The entire “TMIF” facility fits inside a cube 2 metres on a side, and is dimensioned for an accelerator beam power of 100 kW, thus reducing costs and offering great versatility and flexibility.

  13. Reactor containment research and development

    Energy Technology Data Exchange (ETDEWEB)

    Weil, N. A.

    1963-06-15

    An outline is given of containment concepts, sources and release rates of energy, responses of containment structures, effects of projectiles, and leakage rates of radioisotopes, with particular regard to major reactor accidents. (T.F.H.)

  14. TREAT neutron-radiography facility

    Energy Technology Data Exchange (ETDEWEB)

    Harrison, L.J.

    1981-01-01

    The TREAT reactor was built as a transient irradiation test reactor. By taking advantage of built-in system features, it was possible to add a neutron-radiography facility. This facility has been used over the years to radiograph a wide variety and large number of preirradiated fuel pins in many different configurations. Eight different specimen handling casks weighing up to 54.4 t (60 T) can be accommodated. Thermal, epithermal, and track-etch radiographs have been taken. Neutron-radiography service can be provided for specimens from other reactor facilities, and the capacity for storing preirradiated specimens also exists.

  15. United States Domestic Research Reactor Infrastructure - TRIGA Reactor Fuel Support

    Energy Technology Data Exchange (ETDEWEB)

    Morrell, Douglas [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415 (United States)

    2008-10-29

    The purpose of the United State Domestic Research Reactor Infrastructure Program is to provide fresh nuclear reactor fuel to United States universities at no, or low, cost to the university. The title of the fuel remains with the United States government and when universities are finished with the fuel, the fuel is returned to the United States government. The program is funded by the United States Department of Energy - Nuclear Energy division, managed by Department of Energy - Idaho Field Office, and contracted to the Idaho National Laboratory's Management and Operations Contractor - Battelle Energy Alliance. Program has been at Idaho since 1977 and INL subcontracts with 26 United States domestic reactor facilities (13 TRIGA facilities, 9 plate fuel facilities, 2 AGN facilities, 1 Pulstar fuel facility, 1 Critical facility). University has not shipped fuel since 1968 and as such, we have no present procedures for shipping spent fuel. In addition: floor loading rate is unknown, many interferences must be removed to allow direct access to the reactor tank, floor space in the reactor cell is very limited, pavement ends inside our fence; some of the surface is not finished. The whole approach is narrow, curving and downhill. A truck large enough to transport the cask cannot pull into the lot and then back out (nearly impossible / refused by drivers); a large capacity (100 ton), long boom crane would have to be used due to loading dock obstructions. Access to the entrance door is on a sidewalk. The campus uses it as a road for construction equipment, deliveries and security response. Large trees are on both sides of sidewalk. Spent fuel shipments have never been done, no procedures approved or in place, no approved casks, no accident or safety analysis for spent fuel loading. Any cask assembly used in this facility will have to be removed from one crane, moved on the floor and then attached to another crane to get from the staging area to the reactor room

  16. INTEGRATION OF FACILITY MODELING CAPABILITIES FOR NUCLEAR NONPROLIFERATION ANALYSIS

    Energy Technology Data Exchange (ETDEWEB)

    Gorensek, M.; Hamm, L.; Garcia, H.; Burr, T.; Coles, G.; Edmunds, T.; Garrett, A.; Krebs, J.; Kress, R.; Lamberti, V.; Schoenwald, D.; Tzanos, C.; Ward, R.

    2011-07-18

    Developing automated methods for data collection and analysis that can facilitate nuclear nonproliferation assessment is an important research area with significant consequences for the effective global deployment of nuclear energy. Facility modeling that can integrate and interpret observations collected from monitored facilities in order to ascertain their functional details will be a critical element of these methods. Although improvements are continually sought, existing facility modeling tools can characterize all aspects of reactor operations and the majority of nuclear fuel cycle processing steps, and include algorithms for data processing and interpretation. Assessing nonproliferation status is challenging because observations can come from many sources, including local and remote sensors that monitor facility operations, as well as open sources that provide specific business information about the monitored facilities, and can be of many different types. Although many current facility models are capable of analyzing large amounts of information, they have not been integrated in an analyst-friendly manner. This paper addresses some of these facility modeling capabilities and illustrates how they could be integrated and utilized for nonproliferation analysis. The inverse problem of inferring facility conditions based on collected observations is described, along with a proposed architecture and computer framework for utilizing facility modeling tools. After considering a representative sampling of key facility modeling capabilities, the proposed integration framework is illustrated with several examples.

  17. Reference worldwide model for antineutrinos from reactors

    OpenAIRE

    Baldoncini, Marica; Callegari, Ivan; Fiorentini, Giovanni; Mantovani, Fabio; Ricci, Barbara; Strati, Virginia; Xhixha, Gerti

    2014-01-01

    Antineutrinos produced at nuclear reactors constitute a severe source of background for the detection of geoneutrinos, which bring to the Earth's surface information about natural radioactivity in the whole planet. In this framework we provide a reference worldwide model for antineutrinos from reactors, in view of reactors operational records yearly published by the International Atomic Energy Agency (IAEA). We evaluate the expected signal from commercial reactors for ongoing (KamLAND and Bor...

  18. Creating a clinical video-conferencing facility in a security-constrained environment using open-source AccessGrid software and consumer hardware.

    Science.gov (United States)

    Terrazas, Enrique; Hamill, Timothy R; Wang, Ye; Channing Rodgers, R P

    2007-10-11

    The Department of Laboratory Medicine at the University of California, San Francisco (UCSF) has been split into widely separated facilities, leading to much time being spent traveling between facilities for meetings. We installed an open-source AccessGrid multi-media-conferencing system using (largely) consumer-grade equipment, connecting 6 sites at 5 separate facilities. The system was accepted rapidly and enthusiastically, and was inexpensive compared to alternative approaches. Security was addressed by aspects of the AG software and by local network administrative practices. The chief obstacles to deployment arose from security restrictions imposed by multiple independent network administration regimes, requiring a drastically reduced list of network ports employed by AG components.

  19. Overview of Spallation Neutron Source Physics

    Science.gov (United States)

    Russell, G. J.; Pitcher, E. J.; Muhrer, G.; Mezei, F.; Ferguson, P. D.

    In December 1971 , the world's most advanced steady-state research reactor, the High Flux Reactor at the Institut Laue-Langevin (ILL) in Grenoble, France, reached full power operation. The reactor has recently undergone an extensive renovation, is equipped with hot and cold sources, and has a complement of word class instruments. As such, the ILL reactor is the worldwide center for neutron research at a reactor installation. With present technology, the constraints of heat removal and fuel cost place a limit on the available flux of a steadystate research reactor at levels not much higher than that of the ILL reactor. There has been extensive progress worldwide to realize new high-flux neutron facilities using the technology of spallation. When coupled with the spallation process in appropriate target materials, highpower accelerators can be used to produce large numbers of neutrons, thus providing an alternate method to the use of nuclear reactors for this purpose. Spallation technology has recently become increasingly focussed on pulsed spallation neutron sources. Pulsed spallation neutron sources avoid the limitations of high time-average heat removal by producing neutrons for only a small fraction of the time. Also, the amount of energy deposited per useful neutron produced from spallation is less than that from fission. During the pulse, the available neutron flux from a pulsed spallation source can be much more intense than that obtainable in a steady-state reactor. Furthermore, pulsed neutron sources have certain unique features, which open up qualitatively new areas of science, which are not accessible to steady-state reactors. We discuss here the spallation process and spallation neutron sources. We compare the qualitative differences between fission and spallation and provide absolute neutron intensities for cold neutron production from a liquid H2, moderator at the Manuel Lujan Jr. Neutron Scattering Center (Lujan Center) short-pulse pulsed spallation

  20. Random processes in nuclear reactors

    CERN Document Server

    Williams, M M R

    1974-01-01

    Random Processes in Nuclear Reactors describes the problems that a nuclear engineer may meet which involve random fluctuations and sets out in detail how they may be interpreted in terms of various models of the reactor system. Chapters set out to discuss topics on the origins of random processes and sources; the general technique to zero-power problems and bring out the basic effect of fission, and fluctuations in the lifetime of neutrons, on the measured response; the interpretation of power reactor noise; and associated problems connected with mechanical, hydraulic and thermal noise sources

  1. Perspectives on reactor safety

    Energy Technology Data Exchange (ETDEWEB)

    Haskin, F.E. [New Mexico Univ., Albuquerque, NM (United States). Dept. of Chemical and Nuclear Engineering; Camp, A.L. [Sandia National Labs., Albuquerque, NM (United States)

    1994-03-01

    The US Nuclear Regulatory Commission (NRC) maintains a technical training center at Chattanooga, Tennessee to provide appropriate training to both new and experienced NRC employees. This document describes a one-week course in reactor, safety concepts. The course consists of five modules: (1) historical perspective; (2) accident sequences; (3) accident progression in the reactor vessel; (4) containment characteristics and design bases; and (5) source terms and offsite consequences. The course text is accompanied by slides and videos during the actual presentation of the course.

  2. Advanced Neutron Source: Plant Design Requirements

    Energy Technology Data Exchange (ETDEWEB)

    1990-07-01

    The Advanced Neutron Source will be a new world-class facility for research using hot, thermal, cold, and ultra-cold neutrons. The heart of the facility will be a 330-MW (fission), heavy-water cooled and heavy-water moderated reactor. The reactor will be housed in a central reactor building, with supporting equipment located in an adjoining reactor support building. An array of cold neutron guides will fan out into a large guide hall, housing about 30 neutron research stations. Appropriate office, laboratory, and shop facilities will be included to provide a complete facility for users. The ANS is scheduled to begin operation at the Oak Ridge National Laboratory early in the next decade. This PDR document defines the plant-level requirements for the design, construction, and operation of ANS. It also defines and provides input to the individual System Design Description (SDD) documents. Together, this PDR document and the set of SDD documents will define and control the baseline configuration of ANS.

  3. Quantification of Noise Sources in EMI Surveys Technology Demonstration Report Army Research Laboratory Blossom Point Facility, Maryland July - September, 2006

    Science.gov (United States)

    2010-01-14

    disease – The U.S. Army declared the Army Research Laboratory Blossom Point Facility to be a high risk area for contracting Lyme disease in a 1993...assessment report. The assessment found the Lyme disease tick, the Lyme disease bacteria in the ticks and mammals at the facility, and cases of human Lyme ...2.5 Factors Affecting Cost and Performance.......................................................................... 15 2.6 Advantages and

  4. Experimental Breeder Reactor I Preservation Plan

    Energy Technology Data Exchange (ETDEWEB)

    Julie Braun

    2006-10-01

    Experimental Breeder Reactor I (EBR I) is a National Historic Landmark located at the Idaho National Laboratory, a Department of Energy laboratory in southeastern Idaho. The facility is significant for its association and contributions to the development of nuclear reactor testing and development. This Plan includes a structural assessment of the interior and exterior of the EBR I Reactor Building from a preservation, rather than an engineering stand point and recommendations for maintenance to ensure its continued protection.

  5. Armouring facility? Nuclear-weapon and reactor reseach at the Kaiser-Wilhelm Institute for Physics; Eine Waffenschmiede? Kernwaffen- und Reaktorforschung am Kaiser-Wilhelm-Institut fuer Physik

    Energy Technology Data Exchange (ETDEWEB)

    Hachtmann, R. (ed.); Walker, M.

    2005-07-01

    The Kaiser Wilhelm Institute for Physics is best known as the place where Werner Heisenberg worked on nuclear weapons for Hitler. Although this is essentially true, there is more to the story. At the start of World War II this institute was taken over by the German Army Ordnance to be the central, but not exclusive site for a research project into the economic and military applications of nuclear fission. The Army physicist Kurt Diebner was installed in the institute as its commissarial director. Heisenberg was affiliated with the institute as an advisor at first, and became the director in 1942. Heisenberg and his colleagues, including in particular Karl-Heinz Hoecker, Carl Friedrich von Weizsaecker, and Karl Wirtz, worked on nuclear reactors and isotope separation with the clear knowledge that these were two different paths to atomic bombs [Atombomben]. However, they were clearly ambivalent about what they were doing. New documents recently returned from Russian archives shed new light on this work and the scientists' motivations. (orig.)

  6. NUCLEAR REACTOR

    Science.gov (United States)

    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.

  7. Utilisation of British University Research Reactors.

    Science.gov (United States)

    Duncton, P. J.; And Others

    British experience relating to the employment of university research reactors and subcritical assemblies in the education of nuclear scientists and technologists, in the training of reactor operators and for fundamental pure and applied research in this field is reviewed. The facilities available in a number of British universities and the uses…

  8. Implementation and training methodology of subcritical reactors neutronic calculations triggered by external neutron source and applications; Implementacao e qualificacao de metodologia de calculos neutronicos em reatores subcriticos acionados por fonte externa de neutrons e aplicacoes

    Energy Technology Data Exchange (ETDEWEB)

    Carluccio, Thiago

    2011-07-01

    This works had as goal to investigate calculational methodologies on subcritical source driven reactor, such as Accelerator Driven Subcritical Reactor (ADSR) and Fusion Driven Subcritical Reactor (FDSR). Intense R and D has been done about these subcritical concepts, mainly due to Minor Actinides (MA) and Long Lived Fission Products (LLFP) transmutation possibilities. In this work, particular emphasis has been given to: (1) complement and improve calculation methodology with neutronic transmutation and decay capabilities and implement it computationally, (2) utilization of this methodology in the Coordinated Research Project (CRP) of the International Atomic Energy Agency Analytical and Experimental Benchmark Analysis of ADS and in the Collaborative Work on Use of Low Enriched Uranium in ADS, especially in the reproduction of the experimental results of the Yalina Booster subcritical assembly and study of a subcritical core of IPEN / MB-01 reactor, (3) to compare different nuclear data libraries calculation of integral parameters, such as k{sub eff} and k{sub src}, and differential distributions, such as spectrum and flux, and nuclides inventories and (4) apply the develop methodology in a study that may help future choices about dedicated transmutation system. The following tools have been used in this work: MCNP (Monte Carlo N particle transport code), MCB (enhanced version of MCNP that allows burnup calculation) and NJOY to process nuclear data from evaluated nuclear data files. (author)

  9. Assessment of the Technical Maturity of Generation IV Concepts for Test or Demonstration Reactor Applications, Revision 2

    Energy Technology Data Exchange (ETDEWEB)

    Gougar, Hans David [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-10-01

    The United States Department of Energy (DOE) commissioned a study the suitability of different advanced reactor concepts to support materials irradiations (i.e. a test reactor) or to demonstrate an advanced power plant/fuel cycle concept (demonstration reactor). As part of the study, an assessment of the technical maturity of the individual concepts was undertaken to see which, if any, can support near-term deployment. A Working Group composed of the authors of this document performed the maturity assessment using the Technical Readiness Levels as defined in DOE’s Technology Readiness Guide . One representative design was selected for assessment from of each of the six Generation-IV reactor types: gas-cooled fast reactor (GFR), lead-cooled fast reactor (LFR), molten salt reactor (MSR), supercritical water-cooled reactor (SCWR), sodium-cooled fast reactor (SFR), and very high temperature reactor (VHTR). Background information was obtained from previous detailed evaluations such as the Generation-IV Roadmap but other technical references were also used including consultations with concept proponents and subject matter experts. Outside of Generation IV activity in which the US is a party, non-U.S. experience or data sources were generally not factored into the evaluations as one cannot assume that this data is easily available or of sufficient quality to be used for licensing a US facility. The Working Group established the scope of the assessment (which systems and subsystems needed to be considered), adapted a specific technology readiness scale, and scored each system through discussions designed to achieve internal consistency across concepts. In general, the Working Group sought to determine which of the reactor options have sufficient maturity to serve either the test or demonstration reactor missions.

  10. Use of Zircaloy 4 material for the pressure vessels of hot and cold neutron sources and beam tubes for research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Gutsmiedl, Erwin [Technical University Munich, FRM-II (Germany)

    2001-03-01

    The material Zircaloy 4 can be used for the pressure retaining walls for the cold and hot neutron sources and beam tubes. For the research reactor FRM-II of the Technical University Munich, Germany, the material Zircaloy 4 were chosen for the vessels of the cold and hot neutron source and for the beam tube No. 6. The sheets and forgings of Zircaloy 4 were examinated in the temperature range between -256degC and 250degC. The thickness of the sheets are 3, 4, 5 and 10 mm, the maximum diameter of the forgings was 560 mm. This great forging diameters are not be treated in the ASTM rule B 351 for nuclear material, so a special approval with independent experts was necessary. The requirements for the material examinations were specified in a material specification and material test sheets which based on the ASTM rules B 351 and B 352 with additional restriction and additional requirements of the basic safety concept for nuclear power plants in Germany, which was taken into consideration in the nuclear licensing procedure. Charpy-V samples were carried out in the temperature range between -256degC and 150degC to get more information on the ductile behaviour of the Zircaloy 4. The results of the sheet examination confirm the requirements of the specifications, the results of the forging examination in the tangential testing direction are lower than specified and expected for the tensile strength. The axial and transverse values confirm the specification requirements. For the strength calculation of the pressure retaining wall a reduced material value for the forgings has to taken into consideration. The material behaviour of Zircaloy 4 under irradiation up to a fluence of {approx}1{center_dot}10{sup 22} n/cm{sup 2} was investigated. The loss of ductility was determined. As an additional criteria the variation of the fracture toughness was studies. Fracture mechanic calculations of the material were carried out in the licensing procedure with the focus to fulfill the leak

  11. Stabilized Spheromak Fusion Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Fowler, T

    2007-04-03

    The U.S. fusion energy program is focused on research with the potential for studying plasmas at thermonuclear temperatures, currently epitomized by the tokamak-based International Thermonuclear Experimental Reactor (ITER) but also continuing exploratory work on other plasma confinement concepts. Among the latter is the spheromak pursued on the SSPX facility at LLNL. Experiments in SSPX using electrostatic current drive by coaxial guns have now demonstrated stable spheromaks with good heat confinement, if the plasma is maintained near a Taylor state, but the anticipated high current amplification by gun injection has not yet been achieved. In future experiments and reactors, creating and maintaining a stable spheromak configuration at high magnetic field strength may require auxiliary current drive using neutral beams or RF power. Here we show that neutral beam current drive soon to be explored on SSPX could yield a compact spheromak reactor with current drive efficiency comparable to that of steady state tokamaks. Thus, while more will be learned about electrostatic current drive in coming months, results already achieved in SSPX could point to a productive parallel development path pursuing auxiliary current drive, consistent with plans to install neutral beams on SSPX in the near future. Among possible outcomes, spheromak research could also yield pulsed fusion reactors at lower capital cost than any fusion concept yet proposed.

  12. 40 CFR 63.1406 - Reactor batch process vent provisions.

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 11 2010-07-01 2010-07-01 true Reactor batch process vent provisions... § 63.1406 Reactor batch process vent provisions. (a) Emission standards. Owners or operators of reactor... reactor batch process vent located at a new affected source shall control organic HAP emissions by...

  13. A new MTR fuel for a new MTR reactor: UMo for the Jules Horowitz reactor

    Energy Technology Data Exchange (ETDEWEB)

    Guigon, B. [CEA Cadarache, Dir. de l' Energie Nucleaire DEN, Reacteur Jules Horowitz, 13 - Saint-Paul-lez-Durance (France); Vacelet, H. [Compagnie pour l' Etude et la Realisation de Combustibles Atomiques, CERCA, Etablissement de Romans, 26 (France); Dornbusch, D. [Technicatome, Service d' Architecture Generale, 13 - Aix-en-Provence (France)

    2003-07-01

    Within some years, the Jules Horowitz Reactor will be the only working experimental reactor (material and fuel testing reactor) in France. It will have to provide facilities for a wide range of needs: from activation analysis to power reactor fuel qualification. In this paper will be presented the main characteristics of the Jules Horowitz Reactor: its total power, neutron flux, fuel element... Safety criteria will be explained. Finally merits and disadvantages of UMo compared to the standard U{sub 3}Si{sub 2} fuel will be discussed. (authors)

  14. Scientific opportunities with advanced facilities for neutron scattering

    Energy Technology Data Exchange (ETDEWEB)

    Lander, G.H.; Emery, V.J. (eds.)

    1984-01-01

    The present report documents deliberations of a large group of experts in neutron scattering and fundamental physics on the need for new neutron sources of greater intensity and more sophisticated instrumentation than those currently available. An additional aspect of the Workshop was a comparison between steady-state (reactor) and pulsed (spallation) sources. The main conclusions were: (1) the case for a new higher flux neutron source is extremely strong and such a facility will lead to qualitatively new advances in condensed matter science and fundamental physics; (2) to a large extent the future needs of the scientific community could be met with either a 5 x 10/sup 15/ n cm/sup -2/s/sup -1/ steady state source or a 10/sup 17/ n cm/sup -2/s/sup -1/ peak flux spallation source; and (3) the findings of this Workshop are consistent with the recommendations of the Major Materials Facilities Committee.

  15. Installation and operation of a high-temperature surface ion source for the online coupling of TRIGA-SPEC to the TRIGA Mainz research reactor and high-precision mass measurements of transuranium nuclides at TRIGA-TRAP

    Energy Technology Data Exchange (ETDEWEB)

    Renisch, Dennis

    2016-06-23

    The TRIGA-SPEC setup is dedicated for high-precision measurements of ground-state properties of exotic nuclides, like fission products or transuranium nuclides. For the online coupling of TRIGA-SPEC to the TRIGA Mainz research reactor, fission products are transported from a target chamber close to the reactor core by a gas-jet system to an ion source, which provides the connected experiments with a radioactive ion beam for the actual measurements. The design, installation and operation of the online ion source was a major part of the work described in this thesis. In addition, investigations on the optimal conditions of the gas-jet system for a reliable ion source operation were performed and the coupling of the ion source part to the subsequent elements of the beamline was conducted. The second part of this thesis deals with high-precision mass measurements on transuranium nuclides with the Penning-trap mass spectrometer TRIGA-TRAP, which is one branch of the TRIGA-SPEC setup. These measurements contribute to a mapping of the region of the chart of nuclides around the deformed shell closure at N=152. The masses of some of the investigated nuclides are directly measured for the first time. Due to the appearance of systematic inconsistencies during the evaluation of the mass measurements, the focus of the final section of this thesis lies in the identification and correction of the sources of the observed inconsistencies.

  16. 76 FR 5218 - Advisory Committee on Reactor Safeguards

    Science.gov (United States)

    2011-01-28

    ... COMMISSION Advisory Committee on Reactor Safeguards In accordance with the purposes of Sections 29 and 182b of the Atomic Energy Act (42 U.S.C. 2039, 2232b), the Advisory Committee on Reactor Safeguards (ACRS... Facilities and Probabilistic Risk Assessments (PRAs) for Reactors (Open)--The Committee will hear...

  17. 10 CFR 1.44 - Office of New Reactors.

    Science.gov (United States)

    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...

  18. Environmental concerns regarding a materials test reactor fuel fabrication facility at the Nuclear and Energy Research Institute - IPEN; Atomos para el desarrollo de Mexico

    Energy Technology Data Exchange (ETDEWEB)

    Santos, G. R. T.; Durazzo, M.; Carvalho, E. F. U. [IPEN, CNEN-SP, P.O. Box 11049, CEP 05422-970, Sao Paulo (Brazil); Riella, H. G. [Universidade Federal de Santa Catarina, Departamento de Engenharia Quimica, Campus Universitario, Florianopolis, CEP 88040-900 (Brazil)]. e-mail: grsantos@ipen.br

    2008-07-01

    The aim of the industrial activities success, front to a more and more informed and demanding society and to a more and more competitive market demands an environmental administration policy which doesn't limit itself to assist the legislation but anticipate and prevent, in a responsible way, possible damages to the environment. One of the maim programs of the Institute of Energetic and Nuclear Research of the national Commission of Nuclear Energy located in Brazil, through the Center of Nuclear Fuel - CCN - is to manufacture MTR-type fuel elements using low-enrichment uranium (20 wt% {sup 2}35U), to supply its IEA-RI research reactor. Integrated in this program, this work aims at well developing and assuring a methodology to implant an environment, health and safety policy, foreseeing its management with the use of detailed data reports and through the adoption of new tools for improving the management, in order to fulfil the applicable legislation and accomplish all the environmental, operational and works aspects. The applied methodology for the effluents management comprises different aspects, including the specific environmental legislation of a country, main available effluents treatment techniques, process flow analyses from raw materials and intakes to products, generated effluents, residuals and emissions. Data collections were accomplished for points gathering and tests characterization, classification and compatibility of the generated effluents and their eventual environmental impacts. This study aims to implant the Sustainable Concept in order to guarantee access to financial resources, allowing cost reduction, maximizing long-term profits, preventing and reducing environmental accident risks and stimulating both the attraction and the keeping of a motivated manpower. Work on this project has already started and, even though many technical actions have not still ended, the results have being extremely valuable. These results can already give to

  19. Advanced Burner Reactor Preliminary NEPA Data Study.

    Energy Technology Data Exchange (ETDEWEB)

    Briggs, L. L.; Cahalan, J. E.; Deitrich, L. W.; Fanning, T. H.; Grandy, C.; Kellogg, R.; Kim, T. K.; Yang, W. S.; Nuclear Engineering Division

    2007-10-15

    The Global Nuclear Energy Partnership (GNEP) is a new nuclear fuel cycle paradigm with the goals of expanding the use of nuclear power both domestically and internationally, addressing nuclear waste management concerns, and promoting nonproliferation. A key aspect of this program is fast reactor transmutation, in which transuranics recovered from light water reactor spent fuel are to be recycled to create fast reactor transmutation fuels. The benefits of these fuels are to be demonstrated in an Advanced Burner Reactor (ABR), which will provide a representative environment for recycle fuel testing, safety testing, and modern fast reactor design and safeguard features. Because the GNEP programs will require facilities which may have an impact upon the environment within the meaning of the National Environmental Policy Act of 1969 (NEPA), preparation of a Programmatic Environmental Impact Statement (PEIS) for GNEP is being undertaken by Tetra Tech, Inc. The PEIS will include a section on the ABR. In support of the PEIS, the Nuclear Engineering Division of Argonne National Laboratory has been asked to provide a description of the ABR alternative, including graphics, plus estimates of construction and operations data for an ABR plant. The compilation of this information is presented in the remainder of this report. Currently, DOE has started the process of engaging industry on the design of an Advanced Burner Reactor. Therefore, there is no specific, current, vendor-produced ABR design that could be used for this PEIS datacall package. In addition, candidate sites for the ABR vary widely as to available water, geography, etc. Therefore, ANL has based its estimates for construction and operations data largely on generalization of available information from existing plants and from the environmental report assembled for the Clinch River Breeder Reactor Plant (CRBRP) design [CRBRP, 1977]. The CRBRP environmental report was chosen as a resource because it thoroughly

  20. Reactor Dosimetry State of the Art 2008

    Science.gov (United States)

    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

  1. Immediate relation of ING to fast breeder reactor programs

    Energy Technology Data Exchange (ETDEWEB)

    Lewis, W.B

    1969-07-01

    The future large-scale use of nuclear energy is linked in the United States and other major countries to their fast breeder reactor development. Very serious basic problems have been discovered within the last two years, limiting the life in the high fast neutron flux at appropriate temperatures of materials, in particular of metals suitable for fuel cladding in sodium coolant. There is therefore a most urgent need for materials testing facilities under controlled conditions of temperature and neutron flux at sufficiently high ratings to match or surpass those required in commercially competitive fast breeder reactors. None of the test facilities yet planned for 1976 or sooner in the western world appears to match these conditions. The problem is mainly the difficulty of providing the high neutron flux effectively continuously. The spallation reaction in heavy elements was chosen as the basis of ING - the intense neutron generator, because it is the only known reaction that promises a fast neutron source density that is higher than can be controlled from the fission process. It is suggested that several countries will wish to consider urgently whether they should also explore the spallation reaction for the purpose of a fast neutron irradiation test facility. In view of the discontinuance of the ING project in Canada a favourable opportunity will exist over the next few months 10 obtain from Canada by direct personal contact details of the significant study that has been carried on for ING over the last five years. In the event that satisfactory materials are established within the lifetime of the spallation facilities they may continue to be used for the production of selected isotopes more profitably produced in high neutron fluxes. The facilities may be also used for the desirable preirradiation of thorium reactor fuel. The other research purposes planned for ING could also be served. (author)

  2. Role of Halden Reactor Project for world-wide nuclear energy development

    Energy Technology Data Exchange (ETDEWEB)

    McGrath, M.A.; Volkov, B.

    2011-07-01

    The great interest for utilization of nuclear materials to produce energy in the middle of last century needed special investigations using first class research facilities. Common problems in the area of nuclear fuel development motivated the establishment of joint research efforts. The OECD Halden Reactor Project (HRP) is a good example of such a cooperative research effort, which has been performing for more than 50 years. During that time, the Halden Reactor evolved from a prototype heavy water reactor envisaged as a power source for different applications to a research reactor that is able to simulate in-core conditions of modern commercial power reactors. The adaptability of the Halden Reactor enables the HRP to be an important international test facility for nuclear fuels and materials development. The long-term international cooperation is based on the flexible HRP organizational structure which also provides the continued success. [1,2] This paper gives a brief history of the Halden Reactor Project and its contribution to world-wide nuclear energy development. Recent expansion of the Project to the East and Asian countries may also assist and stimulate the development of a nuclear industry within these countries. The achievements of the HRP rely on the versatility of the research carried out in the reactor with reliable testing techniques and in-pile instrumentation. Diversification of scientific activity in the areas of development of alternative energy resources and man-machine technology also provide the HRP with a stable position as one of the leaders in the world scientific community. All of these aspects are described in this paper together with current experimental works, including the investigation of ULBA (Kazakhstan) production fuel in comparison with other world fuel suppliers, as well as other future and prospective plans of the Project.(Author)

  3. sources

    Directory of Open Access Journals (Sweden)

    Shu-Yin Chiang

    2002-01-01

    Full Text Available In this paper, we study the simplified models of the ATM (Asynchronous Transfer Mode multiplexer network with Bernoulli random traffic sources. Based on the model, the performance measures are analyzed by the different output service schemes.

  4. Sources of secondary radionuclide releases from Hanford Operations

    Energy Technology Data Exchange (ETDEWEB)

    Heeb, C.M.; Gydesen, S.P.

    1994-05-01

    This report considers Hanford facilities and operations with the potential to be secondary radionuclide release sources. Facilities that produced radionuclides or processed products of fission reactions and were not covered in previous source term reports are included in this report. The following facilities are described and any potentially significant releases from them are estimated: PUREX (1956--1972, 1983--1988) and REDOX (1952--1967)--campaigns with non-standard feed material (materials other than fuel from single-pass reactors); C PLANT (Hot Semi-Works)--pilot plant and strontium recovery; Z Plant--plutonium finishing; U and UO{sub 3} Plants--uranium recovery; 108 B Plant--tritium extraction; 300 Area Plutonium Recycle Test Reactor (PRTR); 300 Area Low Power Test Reactors; Criticality Accidents; and 400 Area Fast Flux Test Facility (FFTF). The method of analysis was to examine each facility, give a brief description of its purpose and operations, and describe the types of material the facility processed as an indication of the radionuclides it had the potential to release. Where possible, specific radionuclides are estimated and values from the original documents are reported.

  5. Airborne respirable silica near a sand and gravel facility in central California: XRD and elemental analysis to distinguish source and background quartz.

    Science.gov (United States)

    Shiraki, Ryoji; Holmén, Britt A

    2002-12-01

    Despite the potential toxicity of respirable quartz to humans, little is known about the transport of airborne quartz from sources to receptors and how to distinguish anthropogenically generated quartz from natural background in a receptor sample. Airborne quartz emissions near a sand and gravel facility were determined using PM10 and PM2.5 filter samples collected at four downwind sites (D1: 22 m, D2: 62 m, D3: 259 m, and D4: 745 m from the facility) as well as one upwind site (U1: 1495 m) during summer sampling. X-ray diffraction was used to determine quartz concentration and elemental composition was analyzed using PIXE, XRF, PESA, and HIPS techniques. Elemental composition of the PM samples was used to determine the X-ray mass absorption coefficients that are essential for accurate quartz quantification by XRD. Elemental composition was found to be a useful tool to distinguish source and background crystalline silica. Both PM10 and PM2.5 samples collected at the D1, D2, and D3 sites contained more Si, Al, and Fe and less H, Na, and S, compared to those at the U1 site, whereas site D4 sample compositions were similar to those at the U1 site. A composite variable, SOIL/(H+Na+S), where SOIL = 2.20Al + 2.49Si + 1.63Ca + 1.94Ti + 2.42Fe, was used to distinguish source materials from background. Average dry season quartz concentrations in replicate PM10 samples were 4.6 +/- 0.9) microg m(-3) at U1, 60.6 (+/- 5.4) microg m(3) at D1, 62.4 (+/- 3.6) at D2, 32.6 (+/- 2.1) microg m(-3) at D3, and 9.41 (+/- 0.9) microg m(-3) at D4. The mass fraction of quartz was the highest at the D1 site and decreased with increasing distance from the facility. The mass of PM2.5 samples was too low to determine quartz concentrations. These results identify the facility as the main source of quartz and other silicate minerals downwind of the plant and that the air quality of the most remote sampling site, located approximately 750 m downwind, was still impacted by the facility's activity.

  6. Impact of Dust from Ore Processing Facilities on Rain Water Collection Tanks in a Tropical Environment—The Obvious Source “Ain’t Necessarily So”

    Directory of Open Access Journals (Sweden)

    Brian Gulson

    2016-02-01

    Full Text Available Concerns have been expressed that dust from the minerals processing facilities at Karumba Queensland Australia have resulted in elevated lead (Pb concentrations in rain water tanks. The ores derived from the Century mine some 304 km from the port. High precision Pb isotopic measurements on environmental samples have been undertaken to evaluate the source of Pb in rainwaters and acid digests from roof wipes and gutter wipes. There does not appear to be any relationship between sample location and the processing facility but samples from the area subject to the prevailing winds show the highest contribution of Century Pb. All gutter wipes (82 to 1270 µg Pb/wipe have contributions of Century ore ranging from 87% to 96%. The contribution of Century ore to five roof wipes (22 to 88 µg Pb/wipe ranges from 89% to 97% and in the other two samples there is a mix of Century and Broken Hill Pb. Three of the seven rainwater have contributions of Century ore Pb ranging from 33% to 75%. Two of the other four rainwater samples have the highest water Pb concentrations of 88 and 100 µg/L and their isotopic data show Broken Hill Pb contributions ranging from 77% to 80%. The source of the Broken Hill Pb is probably from the galvanized roofing material and/or brass fittings in the rainwater tanks. The discrimination between various sources is only detectable using high precision 204Pb-based isotopic ratios and not the now common inductively coupled plasma mass spectrometry (ICP-MS data presentations of the higher abundance isotopes 208Pb, 207Pb and 206Pb. Isotopic results for the waters demonstrate that apportioning blame where there is an obvious point source may not always be the correct conclusion. Nevertheless the isotopic data for the gutter wipes indicates that there was widespread contamination from the processing facilities throughout the town.

  7. Licensed reactor nuclear safety criteria applicable to DOE reactors

    Energy Technology Data Exchange (ETDEWEB)

    1993-11-01

    This document is a compilation and source list of nuclear safety criteria that the Nuclear Regulatory Commission (NRC) applies to licensed reactors; it can be used by DOE and DOE contractors to identify NRC criteria to be evaluated for application to the DOE reactors under their cognizance. The criteria listed are those that are applied to the areas of nuclear safety addressed in the safety analysis report of a licensed reactor. They are derived from federal regulations, USNRC regulatory guides, Standard Review Plan (SRP) branch technical positions and appendices, and industry codes and standards.

  8. Reactor Neutrinos

    CERN Document Server

    Lasserre, T; Lasserre, Thierry; Sobel, Henry W.

    2005-01-01

    We review the status and the results of reactor neutrino experiments, that toe the cutting edge of neutrino research. Short baseline experiments have provided the measurement of the reactor neutrino spectrum, and are still searching for important phenomena such as the neutrino magnetic moment. They could open the door to the measurement of coherent neutrino scattering in a near future. Middle and long baseline oscillation experiments at Chooz and KamLAND have played a relevant role in neutrino oscillation physics in the last years. It is now widely accepted that a new middle baseline disappearance reactor neutrino experiment with multiple detectors could provide a clean measurement of the last undetermined neutrino mixing angle theta13. We conclude by opening on possible use of neutrinos for Society: NonProliferation of Nuclear materials and Geophysics.

  9. NEUTRONIC REACTORS

    Science.gov (United States)

    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.

  10. The Study on Domestic and Foreign Cases for Decommissioning of DPRK Nuclear Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Baek, Ye Ji; Hhu, Joo Youn; Lee, Jung Hyun; Hwang, Yong Soo [Korea Institute of Nuclear Non-proliferation and Control, Daejeon (Korea, Republic of)

    2016-05-15

    This study was able to analyze domestic and foreign cases, and collect data on the approximate amount of waste and time required time; however, data on applied technology, input manpower, required cost, and waste disposal method was insufficient. DPRK activities such as nuclear weapon development or nuclear testing not only threaten our country's security but also have an adverse effect on nuclear nonproliferation and security in the international society. Therefore, denuclearization of the DPRK is prior task that is essential to peace on the Korean Peninsula. The fundamental purpose of denuclearization of the DPRK is to safely decommission facilities related to developing nuclear weapons and to depose related radioactive waste and nuclear materials. Understanding descriptive references and physical properties of the facility and its purpose important for decommissioning nuclear facilities. Although it was impossible to collect data on DPRK nuclear facilities to perform complete decommissioning, we were able to understand the process used at DPRK nuclear facilities with open source data. This study has been conducted to establish overall measures for decommissioning DPRK nuclear facilities. DPRK nuclear facilities in this study include a IRT- 2000 type nuclear research reactor, a 5 MWe graphite moderated reactor, nuclear fuel fabrication facility, and a nuclear fuel reprocessing facility, which are considered as facilities that produce or manufacture nuclear materials needed for nuclear weapons or related to such activities.

  11. Reactor monitoring and safeguards using antineutrino detectors

    CERN Document Server

    Bowden, N S

    2008-01-01

    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 very 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 reactors, as part of International Atomic Energy Agency (IAEA) and other reactor safeguards regimes. Several efforts to develop this monitoring technique are underway across the globe.

  12. Intercomparison of monochromatic source facilities for the determination of the relative spectral response of erythemal broadband filter radiometers.

    Science.gov (United States)

    Schreder, Josef; Gröbner, Julian; Los, Alexander; Blumthaler, Mario

    2004-07-01

    The relative spectral responses of erythemally weighted broadband radiometers determined at three different laboratories are compared, and the systems are described. The results of measurements of four different broadband radiometers are discussed. Although the common dynamic range of the measured relative spectral responses is approximately 10(4), the differences in the relative spectral response functions are lower than 20%. These differences are related mostly to measurement uncertainties and differences in the spectral response facilities.

  13. Sources

    OpenAIRE

    2015-01-01

    Sources Fondation Pablo Iglesias. Alcala de Henares. Sections : Archives privées de Manuel ArijaArchives extérieuresArchives FNJS de EspañaPrensa Archives Générales de l’Administration. Alcala de Henares. Sections : Opposition au franquismeSig. 653 Sig TOP 82/68.103-68.602.Índice de las cartas colectivas, Relaciones, Cartas al Ministro de Información de Marzo de 1965. c.662. Sources cinématographiques Filmothèque Nationale d’Espagne.NO.DO. N° 1157C. 08/03/1965.aguirre Javier, Blanco vertical....

  14. N Reactor Deactivation Program Plan. Revision 4

    Energy Technology Data Exchange (ETDEWEB)

    Walsh, J.L.

    1993-12-01

    This N Reactor Deactivation Program Plan is structured to provide the basic methodology required to place N Reactor and supporting facilities {center_dot} in a radiologically and environmentally safe condition such that they can be decommissioned at a later date. Deactivation will be in accordance with facility transfer criteria specified in Department of Energy (DOE) and Westinghouse Hanford Company (WHC) guidance. Transition activities primarily involve shutdown and isolation of operational systems and buildings, radiological/hazardous waste cleanup, N Fuel Basin stabilization and environmental stabilization of the facilities. The N Reactor Deactivation Program covers the period FY 1992 through FY 1997. The directive to cease N Reactor preservation and prepare for decommissioning was issued by DOE to WHC on September 20, 1991. The work year and budget data supporting the Work Breakdown Structure in this document are found in the Activity Data Sheets (ADS) and the Environmental Restoration Program Baseline, that are prepared annually.

  15. Analysis of accident sequences and source terms at treatment and storage facilities for waste generated by US Department of Energy waste management operations

    Energy Technology Data Exchange (ETDEWEB)

    Mueller, C.; Nabelssi, B.; Roglans-Ribas, J.; Folga, S.; Policastro, A.; Freeman, W.; Jackson, R.; Mishima, J.; Turner, S.

    1996-12-01

    This report documents the methodology, computational framework, and results of facility accident analyses performed for the US Department of Energy (DOE) Waste Management Programmatic Environmental Impact Statement (WM PEIS). The accident sequences potentially important to human health risk are specified, their frequencies assessed, and the resultant radiological and chemical source terms evaluated. A personal-computer-based computational framework and database have been developed that provide these results as input to the WM PEIS for the calculation of human health risk impacts. The WM PEIS addresses management of five waste streams in the DOE complex: low-level waste (LLW), hazardous waste (HW), high-level waste (HLW), low-level mixed waste (LLMW), and transuranic waste (TRUW). Currently projected waste generation rates, storage inventories, and treatment process throughputs have been calculated for each of the waste streams. This report summarizes the accident analyses and aggregates the key results for each of the waste streams. Source terms are estimated, and results are presented for each of the major DOE sites and facilities by WM PEIS alternative for each waste stream. Key assumptions in the development of the source terms are identified. The appendices identify the potential atmospheric release of each toxic chemical or radionuclide for each accident scenario studied. They also discuss specific accident analysis data and guidance used or consulted in this report.

  16. Industrial sector-based volatile organic compound (VOC) source profiles measured in manufacturing facilities in the Pearl River Delta, China.

    Science.gov (United States)

    Zheng, Junyu; Yu, Yufan; Mo, Ziwei; Zhang, Zhou; Wang, Xinming; Yin, Shasha; Peng, Kang; Yang, Yang; Feng, Xiaoqiong; Cai, Huihua

    2013-07-01

    Industrial sector-based VOC source profiles are reported for the Pearl River Delta (PRD) region, China, based source samples (stack emissions and fugitive emissions) analyzed from sources operating under normal conditions. The industrial sectors considered are printing (letterpress, offset and gravure printing processes), wood furniture coating, shoemaking, paint manufacturing and metal surface coating. More than 250 VOC species were detected following US EPA methods TO-14 and TO-15. The results indicated that benzene and toluene were the major species associated with letterpress printing, while ethyl acetate and isopropyl alcohol were the most abundant compounds of other two printing processes. Acetone and 2-butanone were the major species observed in the shoemaking sector. The source profile patterns were found to be similar for the paint manufacturing, wood furniture coating, and metal surface coating sectors, with aromatics being the most abundant group and oxygenated VOCs (OVOCs) as the second largest contributor in the profiles. While OVOCs were one of the most significant VOC groups detected in these five industrial sectors in the PRD region, they have not been reported in most other source profile studies. Such comparisons with other studies show that there are differences in source profiles for different regions or countries, indicating the importance of developing local source profiles. Crown Copyright © 2013. Published by Elsevier B.V. All rights reserved.

  17. Neutronic reactor

    Science.gov (United States)

    Wende, Charles W. J.; Babcock, Dale F.; Menegus, Robert L.

    1983-01-01

    A nuclear reactor includes an active portion with fissionable fuel and neutron moderating material surrounded by neutron reflecting material. A control element in the active portion includes a group of movable rods constructed of neutron-absorbing material. Each rod is movable with respect to the other rods to vary the absorption of neutrons and effect control over neutron flux.

  18. Neutronic reactor

    Energy Technology Data Exchange (ETDEWEB)

    Babcock, D.F.; Menegus, R.L.; Wende, C.W.

    1983-01-04

    A nuclear reactor includes an active portion with fissionable fuel and neutron moderating material surrounded by neutron reflecting material. A control element in the active portion includes a group of movable rods constructed of neutron-absorbing material. Each rod is movable with respect to the other rods to vary the absorption of neutrons and effect control over neutron flux.

  19. Second session: operating European facilities and their programs; Session 2: Les installations Europeennes actuelles et leurs programmes

    Energy Technology Data Exchange (ETDEWEB)

    Iracane, D. [CEA Saclay, Dir. de l' Energie Nucleaire (DEN), 91 - Gif sur Yvette (France); Porracchia, A.; Fougeras, P.; Morey, J.M. [CEA Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. d' Etudes des Reacteurs; Loubiere, S.; Durande-Ayme, P. [CEA Saclay, Nuclear Energy Div., Reactors and Nuclear Services Deptment, 91 - Gif-sur-Yvette (France); Guidez, J.; Goux, D. [CEA Valrho, Dir. de l' Energie Nucleaire, DEN, 30 - Marcoule (France); Dupraz, R.; Brand, B. [FRAMATOME, AREVA-FANP, 69 - Lyon (France); Blanc, J.Y. [CEA Saclay, 91 - Gif sur Yvette (France); Perthuis, S. de [FRAMATOME ANP, 92 - Paris-La-Defence (France); Le Rouzic, J.F. [Electricite de France (EDF), 92 - Clamart (France)

    2005-07-01

    Most European nuclear post-irradiation examination facilities are ageing and the optimization of the remaining infrastructures may lead to connect them through a network implying consistent staff competence between countries and efficient nuclear material transport means. The second article describes the 3 very-low power research reactors operating in France: Minerve, Masurca and Eole. The Osiris reactor is presented in the third article, the author focuses on the devices available in Osiris to perform irradiation in light water reactor conditions and in high temperature reactor conditions and on the associated programs. Phenix reactor located on the Marcoule site had been performing from 1974 to 1990 the necessary technological qualifications required by the fast reactor system. An important upgrading program, led from 1994 to 2003, has allowed the reactor to begin a second life. Its investigation program encompasses research work on the transmutation of actinides and fission products and on new nuclear fuels and materials required for the future fourth generation of reactors. In Europe about 20 hot laboratories offer services to perform examination and qualification required by their national civil nuclear programs. Most are state-owned and show a large range of activities: nuclear fuels, materials, reprocessing, radio-nuclides, and radio-active sources. The last article reviews the main test loops operating in France and in neighboring countries. About 30 installations are reported and classified according to their activity fields : critical heat flux, hydro-mechanics, device testing, accidental situations, helium and severe reactor accidents. (A.C.)

  20. Source Terms for HFIR Beam Tube Shielding Analyses, and a Complete Shielding Analysis of the HB-3 Tube

    Energy Technology Data Exchange (ETDEWEB)

    Bucholz, J.A.

    2000-07-01

    The High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory is in the midst of a massive upgrade program to enhance experimental facilities. The reactor presently has four horizontal experimental beam tubes, all of which will be replaced or redesigned. The HB-2 beam tube will be enlarged to support more guide tubes, while the HB-4 beam tube will soon include a cold neutron source.

  1. POWER SYSTEMS DEVELOPMENT FACILITY

    Energy Technology Data Exchange (ETDEWEB)

    Unknown

    2002-11-01

    This report discusses test campaign GCT4 of the Kellogg Brown & Root, Inc. (KBR) transport reactor train with a Siemens Westinghouse Power Corporation (Siemens Westinghouse) particle filter system at the Power Systems Development Facility (PSDF) located in Wilsonville, Alabama. The transport reactor is an advanced circulating fluidized-bed reactor designed to operate as either a combustor or a gasifier using one of two possible particulate control devices (PCDs). The transport reactor was operated as a pressurized gasifier during GCT4. GCT4 was planned as a 250-hour test run to continue characterization of the transport reactor using a blend of several Powder River Basin (PRB) coals and Bucyrus limestone from Ohio. The primary test objectives were: Operational Stability--Characterize reactor loop and PCD operations with short-term tests by varying coal-feed rate, air/coal ratio, riser velocity, solids-circulation rate, system pressure, and air distribution. Secondary objectives included the following: Reactor Operations--Study the devolatilization and tar cracking effects from transient conditions during transition from start-up burner to coal. Evaluate the effect of process operations on heat release, heat transfer, and accelerated fuel particle heat-up rates. Study the effect of changes in reactor conditions on transient temperature profiles, pressure balance, and product gas composition. Effects of Reactor Conditions on Synthesis Gas Composition--Evaluate the effect of air distribution, steam/coal ratio, solids-circulation rate, and reactor temperature on CO/CO{sub 2} ratio, synthesis gas Lower Heating Value (LHV), carbon conversion, and cold and hot gas efficiencies. Research Triangle Institute (RTI) Direct Sulfur Recovery Process (DSRP) Testing--Provide syngas in support of the DSRP commissioning. Loop Seal Operations--Optimize loop seal operations and investigate increases to previously achieved maximum solids-circulation rate.

  2. Boron ion beam production with the supernanogan ECR ion source for the CERN BIO-LEIR facility

    CERN Document Server

    Stafford-Haworth, J; Scrivens, R; Toivanen, V; Röhrich, J

    2014-01-01

    To deliver B3+ ions for medical research the compounds decaborane and m-carborane were tested using the metal ions from volatile compounds (MIVOC) method with the Supernanogan 14.5 GHz ECR ion source. Using decaborane the source delivered less than 10 A intensity of B3+ and after operation large deposits of material were found inside the source. Using m-carborane 50 A of B3+ were delivered without support gas. For m-carborane, helium and oxygen support gasses were also tested, and the effects of different source tuning parameters are discussed. The average consumption of m-carborane was 0:1 mg/Ah over all operation.

  3. Participation in the U.S. Department of Energy Reactor Sharing Program

    Energy Technology Data Exchange (ETDEWEB)

    Mulder, R. U.; Benneche, P. E.; Hosticka, B.

    1998-09-30

    The objective of the DOE supported Reactor Sharing Program is to increase the availability of university nuclear reactor facilities to non-reactor-owning educational institutions. The educational and research programs of these users institutions is enhanced by the use of the nuclear facilities.

  4. Organic carbon movement through two SWRO facilities from source water to pretreatment to product with relevance to membrane biofouling

    KAUST Repository

    Alshahri, Abdullah Hassan Mohammed

    2016-12-29

    The presence of algae, bacteria, various fractions of natural organic matter (NOM), and transparent exopolymer particles (TEP) in the raw water, after each pretreatment process and in the permeate and concentrate streams, were measured at two SWRO plants to assess biofouling potential. It was found that the most significant process controlling the concentration of algae, bacteria, and the biopolymer and humic substances was the intake type with the subsurface intake discharge showing significant reductions. The mixed media filtration process was marginally useful in removing some TOC and NOM, but had little effect on TEP removal. Some bacterial regrowth may be occurring in the cartridge filters, but the evidence is inconsistent. Significant quantities of the biopolymer and humic substance concentrations were found to be retained in the membranes, but the concentrations were significantly greater in the facility using an open-ocean intake. Bacteria and TEP were found in the permeate stream, which may document bacterial regrowth and TEP production downstream of the membrane process. Measurements of the organic carbon passage through SWRO facilities can be successfully used to evaluate pretreatment process effectiveness and to make SWRO plant operational improvements.

  5. Advanced Safeguards Approaches for New Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Durst, Philip C.; Therios, Ike; Bean, Robert; Dougan, A.; Boyer, Brian; Wallace, Rick L.; Ehinger, Michael H.; Kovacic, Don N.; Tolk, K.

    2007-12-15

    This third report in the series reviews possible safeguards approaches for new fast reactors in general, and the ABR in particular. Fast-neutron spectrum reactors have been used since the early 1960s on an experimental and developmental level, generally with fertile blanket fuels to “breed” nuclear fuel such as plutonium. Whether the reactor is designed to breed plutonium, or transmute and “burn” actinides depends mainly on the design of the reactor neutron reflector and the whether the blanket fuel is “fertile” or suitable for transmutation. However, the safeguards issues are very similar, since they pertain mainly to the receipt, shipment and storage of fresh and spent plutonium and actinide-bearing “TRU”-fuel. For these reasons, the design of existing fast reactors and details concerning how they have been safeguarded were studied in developing advanced safeguards approaches for the new fast reactors. In this regard, the design of the Experimental Breeder Reactor-II “EBR-II” at the Idaho National Laboratory (INL) was of interest, because it was designed as a collocated fast reactor with a pyrometallurgical reprocessing and fuel fabrication line – a design option being considered for the ABR. Similarly, the design of the Fast Flux Facility (FFTF) on the Hanford Site was studied, because it was a successful prototype fast reactor that ran for two decades to evaluate fuels and the design for commercial-scale fast reactors.

  6. Participation in the U.S. Department of Energy Reactor Sharing Program. Progress report

    Energy Technology Data Exchange (ETDEWEB)

    Mulder, R.U.; Benneche, P.E.; Hosticka, B.

    1997-03-01

    The objective of the DOE supported Reactor Sharing Program is to increase the availability of university nuclear reactor facilities to non-reactor-owning educational institutions. The educational and research programs of these user institutions is enhanced by the use of the nuclear facilities. Several methods have been used by the UVA Reactor Facility to achieve this objective. First, many college and secondary school groups toured the Reactor Facility and viewed the UVAR reactor and associated experimental facilities. Second, advanced undergraduate and graduate classes from area colleges and universities visited the facility to perform experiments in nuclear engineering and physics which would not be possible at the user institution. Third, irradiation and analysis services at the Facility have been made available for research by faculty and students from user institutions. Fourth, some institutions have received activated material from UVA for use at their institutions. These areas are discussed here.

  7. Development of a gas-jet coupled ISOL facility with a /sup 252/Cf spontaneous fission source

    CERN Document Server

    Greenwood, R C; Novick, V J

    1981-01-01

    A mass separator at the INEL has been successfully coupled on-line to a source of /sup 252/Cf fission products via a He-gas jet transport arrangement using solid aerosols of NaCl as activity carriers. Initial tests of the ISOL system on-line to an approximately 7 mu g /sup 252 /Cf source are conducted using gamma-ray spectroscopic measurements of the separated /sup 138,139/Cs, /sup 141,142/Ba and /sup 142/La activities. The measured transport efficiencies through the system of approximately 3% and approximately 0.3% for the Cs and Ba isotopes, respectively, are comparable with the results of earlier tests conducted at INEL with a hollow-cathode ion source alone coupled to the He-gas jet transport arrangement. Following these tests, a general survey of the mass-separated activities is conducted with the ISOL system on-line to an approximately 600 mu g source of /sup 252/Cf. Gross beta - gamma activity is measured for samples collected at 73 mass positions. Gamma-ray spectra are measured with a Ge(Li) detector ...

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

    Energy Technology Data Exchange (ETDEWEB)

    Honma, George [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-10-01

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

  9. Reactors are indispensable for radioisotope production.

    Science.gov (United States)

    Mushtaq, Ahmad

    2010-12-01

    Radioisotopes can be produced by reactors and accelerators. For certain isotopes there could be an advantage to a certain production method. However, nowadays many reports suggest, that useful isotopes needed in medicine, industry and research could be produced efficiently and dependence on reactors using enriched U-235 may be eliminated. In my view reactors and accelerators will continue to play their role side by side in the supply of suitable and economical sources of isotopes.

  10. Sources

    OpenAIRE

    2014-01-01

    Archives Archivo Histórico del Estado de Jalisco Fondo Gobernación, Asunto Pasaportes y Salvoconductos : G-8-877/9773-9775 G-8-878/9774, 9776, 9777 et 9781 G-8-879/9782-9788 G-8-880/9789-9798 G-8-881-882/9803 G-8-882/9804-9805 G-8-883/9806-9811 G-8-884/9813 G-8-885/9817-9820 G-8-886/9822-9825 G-8-887/9826-9830 G-8-888/9835 G-8-889-890/9837 G-8-889/9839 Sources imprimées Livres et chroniques O’Farrill Romulo, (2004) Reseña histórica estadística y comercial de México y sus estados, directorio g...

  11. NEUTRONIC REACTOR CORE INSTRUMENT

    Science.gov (United States)

    Mims, L.S.

    1961-08-22

    A multi-purpose instrument for measuring neutron flux, coolant flow rate, and coolant temperature in a nuclear reactor is described. The device consists essentially of a hollow thimble containing a heat conducting element protruding from the inner wall, the element containing on its innermost end an amount of fissionsble materinl to function as a heat source when subjected to neutron flux irradiation. Thermocouple type temperature sensing means are placed on the heat conducting element adjacent the fissionable material and at a point spaced therefrom, and at a point on the thimble which is in contact with the coolant fluid. The temperature differentials measured between the thermocouples are determinative of the neutron flux, coolant flow, and temperature being measured. The device may be utilized as a probe or may be incorporated in a reactor core. (AE C)

  12. Safeguards instrument to monitor spent reactor fuel

    Science.gov (United States)

    Nicholson, N.; Dowdy, E. J.; Holt, D. M.; Stump, C.

    1981-10-01

    A hand held instrument for monitoring irradiated nuclear fuel inventories located in water filled storage ponds has been developed. This instrument provides sufficient precise qualitative and quantitative information to be useful as a confirmatory technique to International Atomic Energy Agency inspectors, and is believed to be of potential use to nuclear fuel managers and to operators of spent fuel storage facilities, both at reactor and away from reactor, and to operators of nuclear fuel reprocessing plants. Because the Cerenkov radiation glow can barely be seen by the unaided eye under darkened conditions, a night vision device is incorporated to aid the operator in locating the fuel assembly to be measured. Beam splitting optics placed in front of the image intensifier and a preset aperture select a predetermined portion of the observed scene for measurement of the light intensity using a photomultiplier (PM) tube and digital readout. The PM tube gain is adjusted by use of an internal optical reference source, providing long term repeatability and instrument to instrument consistency. Interchangeable lenses accommodate various viewing and measuring conditions.

  13. Advanced Neutron Source: Plant Design Requirements. Revision 4

    Energy Technology Data Exchange (ETDEWEB)

    1990-07-01

    The Advanced Neutron Source will be a new world-class facility for research using hot, thermal, cold, and ultra-cold neutrons. The heart of the facility will be a 330-MW (fission), heavy-water cooled and heavy-water moderated reactor. The reactor will be housed in a central reactor building, with supporting equipment located in an adjoining reactor support building. An array of cold neutron guides will fan out into a large guide hall, housing about 30 neutron research stations. Appropriate office, laboratory, and shop facilities will be included to provide a complete facility for users. The ANS is scheduled to begin operation at the Oak Ridge National Laboratory early in the next decade. This PDR document defines the plant-level requirements for the design, construction, and operation of ANS. It also defines and provides input to the individual System Design Description (SDD) documents. Together, this PDR document and the set of SDD documents will define and control the baseline configuration of ANS.

  14. Advanced Neutron Source (ANS) Project. Progress report FY 1993

    Energy Technology Data Exchange (ETDEWEB)

    Campbell, J.H. [ed.; Selby, D.L.; Harrington, R.M. [Oak Ridge National Lab., TN (United States); Thompson, P.B. [Martin Marietta Energy Systems, Inc., Oak Ridge, TN (United States). Engineering Div.

    1994-01-01

    This report covers the progress made in 1993 in the following sections: (1) project management; (2) research and development; (3) design and (4) safety. The section on research and development covers the following: (1) reactor core development; (2) fuel development; (3) corrosion loop tests and analysis; (4) thermal-hydraulic loop tests; (5) reactor control and shutdown concepts; (6) critical and subcritical experiments; (7) material data, structure tests, and analysis; (8) cold source development; (9) beam tube, guide, and instrument development; (10) neutron transport and shielding; (11) I and C research and development; and (12) facility concepts.

  15. Nuclear Reactors. Revised.

    Science.gov (United States)

    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…

  16. Simplifying Microbial Electrosynthesis Reactor Design

    Directory of Open Access Journals (Sweden)

    Cloelle G.S. Giddings

    2015-05-01

    Full Text Available Microbial electrosynthesis, an artificial form of photosynthesis, can efficiently convert carbon dioxide into organic commodities; however, this process has only previously been demonstrated in reactors that have features likely to be a barrier to scale-up. Therefore, the possibility of simplifying reactor design by both eliminating potentiostatic control of the cathode and removing the membrane separating the anode and cathode was investigated with biofilms of Sporomusa ovata, which reduces carbon dioxide to acetate. In traditional ‘H-cell’ reactors, where the anode and cathode chambers were separated with a proton-selective membrane, the rates and columbic efficiencies of microbial electrosynthesis remained high when electron delivery at the cathode was powered with a direct current power source rather than with a poteniostat-poised cathode utilized in previous studies. A membrane-less reactor with a direct-current power source with the cathode and anode positioned to avoid oxygen exposure at the cathode, retained high rates of acetate production as well as high columbic and energetic efficiencies. The finding that microbial electrosynthesis is feasible without a membrane separating the anode from the cathode, coupled with a direct current power source supplying the energy for electron delivery, is expected to greatly simplify future reactor design and lower construction costs.

  17. Proceedings of the sixth Asian symposium on research reactors

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1999-08-01

    The symposium consisted of 16 sessions with 58 submitted papers. Major fields were: (1) status and future plan of research and testing reactors, (2) operating experiences, (3) design and modification of the facility, and reactor fuels, (4) irradiation studies, (5) irradiation facilities, (6) reactor characteristics and instrumentation, and (7) neutron beam utilization. Panel discussion on the 'New Trends on Application of Research and Test Reactors' was also held at the last of the symposium. About 180 people participated from China, Korea, Indonesia, Thailand, Bangladesh, Vietnam, Chinese Taipei, Belgium, France, USA, Japan and IAEA. The 58 of the presented papers are indexed individually. (J.P.N.)

  18. Development of a nuclear reactor control system simulator using virtual instruments

    Energy Technology Data Exchange (ETDEWEB)

    Pinto, Antonio Juscelino; Mesquita, Amir Zacarias; Lameiras, Fernando Soares, E-mail: ajp@cdtn.b, E-mail: amir@cdtn.b, E-mail: fsl@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2011-07-01

    The International Atomic Energy Agency recommends the use of safety and friendly interfaces for monitoring and controlling the operational parameters of the nuclear reactors. This article describes a digital system being developed to simulate the behavior of the operating parameters using virtual instruments. The control objective is to bring the reactor power from its source level (mW) to a full power (kW). It is intended for education of basic reactor neutronic and thermohydraulic principles such as the multiplication factor, criticality, reactivity, period, delayed neutron, control by rods, fuel and coolant temperatures, power, etc. The 250 kW IPR-R1 TRIGA research reactor at Nuclear Technology Development Centre - CDTN was used as reference. TRIGA reactors, developed by General Atomics (GA), are the most widely used research reactor in the world. The simulator system is being developed using the LabVIEW (Laboratory Virtual Instruments Engineering Workbench) software, considering the modern concept of virtual instruments (VI's) using electronic processor and visual interface in video monitor. The main purpose of the system is to provide training tools for instructors and students, allowing navigating by user-friendly operator interface and monitoring tendencies of the operational variables. It will be an interactive tool for training and teaching and could be used to predict the reactor behavior. Some scenarios are presented to demonstrate that it is possible to know the behavior of some variables from knowledge of input parameters. The TRIGA simulator system will allow the study of parameters, which affect the reactor operation, without the necessity of using the facility. (author)

  19. Simulation of a high energy neutron irradiation facility at beamline 11 of the China Spallation Neutron Source

    Energy Technology Data Exchange (ETDEWEB)

    Tairan, Liang [School of Physics and Electronic Information Inner Mongolia University for the Nationalities, Tongliao 028043 (China); Zhiduo, Li [Dongguan Branch, Institute of High Energy Physics, CAS, Beijing 100049 (China); Wen, Yin, E-mail: wenyin@aphy.iphy.ac.cn [Dongguan Branch, Institute of High Energy Physics, CAS, Beijing 100049 (China); Institute of Physics, CAS, P.O. Box 603, Beijing 100190 (China); Fei, Shen [Dongguan Branch, Institute of High Energy Physics, CAS, Beijing 100049 (China); Quanzhi, Yu [Dongguan Branch, Institute of High Energy Physics, CAS, Beijing 100049 (China); Institute of Physics, CAS, P.O. Box 603, Beijing 100190 (China); Tianjiao, Liang [Dongguan Branch, Institute of High Energy Physics, CAS, Beijing 100049 (China)

    2017-07-11

    The China Spallation Neutron Source (CSNS) will accommodate 20 neutron beamlines at its first target station. These beamlines serve different purposes, and beamline 11 is designed to analyze the degraded models and damage mechanisms, such as Single Event Effects in electronic components and devices for aerospace electronic systems. This paper gives a preliminary discussion on the scheme of a high energy neutron irradiation experiment at the beamline 11 shutter based on the Monte Carlo simulation method. The neutron source term is generated by calculating the neutrons scattering into beamline 11 with a model that includes the target-moderator-reflector area. Then, the neutron spectrum at the sample position is obtained. The intensity of neutrons with energy of hundreds of MeV is approximately 1E8 neutron/cm{sup 2}/s, which is useful for experiments. The displacement production rate and gas productions are calculated for common materials such as tungsten, tantalum and SS316. The results indicate that the experiment can provide irradiation dose rate ranges from 1E-5 to 1E-4 dpa per operating year. The residual radioactivity is also calculated for regular maintenance work. These results give the basic reference for the experimental design.

  20. Simulation of a high energy neutron irradiation facility at beamline 11 of the China Spallation Neutron Source

    Science.gov (United States)

    Tairan, Liang; Zhiduo, Li; Wen, Yin; Fei, Shen; Quanzhi, Yu; Tianjiao, Liang

    2017-07-01

    The China Spallation Neutron Source (CSNS) will accommodate 20 neutron beamlines at its first target station. These beamlines serve different purposes, and beamline 11 is designed to analyze the degraded models and damage mechanisms, such as Single Event Effects in electronic components and devices for aerospace electronic systems. This paper gives a preliminary discussion on the scheme of a high energy neutron irradiation experiment at the beamline 11 shutter based on the Monte Carlo simulation method. The neutron source term is generated by calculating the neutrons scattering into beamline 11 with a model that includes the target-moderator-reflector area. Then, the neutron spectrum at the sample position is obtained. The intensity of neutrons with energy of hundreds of MeV is approximately 1E8 neutron/cm2/s, which is useful for experiments. The displacement production rate and gas productions are calculated for common materials such as tungsten, tantalum and SS316. The results indicate that the experiment can provide irradiation dose rate ranges from 1E-5 to 1E-4 dpa per operating year. The residual radioactivity is also calculated for regular maintenance work. These results give the basic reference for the experimental design.

  1. Solid State Reactor Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Mays, G.T.

    2004-03-10

    The Solid State Reactor (SSR) is an advanced reactor concept designed to take advantage of Oak Ridge National Laboratory's (ORNL's) recently developed graphite foam that has enhanced heat transfer characteristics and excellent high-temperature mechanical properties, to provide an inherently safe, self-regulated, source of heat for power and other potential applications. This work was funded by the U.S. Department of Energy's Nuclear Energy Research Initiative (NERI) program (Project No. 99-064) from August 1999 through September 30, 2002. The initial concept of utilizing the graphite foam as a basis for developing an advanced reactor concept envisioned that a suite of reactor configurations and power levels could be developed for several different applications. The initial focus was looking at the reactor as a heat source that was scalable, independent of any heat removal/power conversion process. These applications might include conventional power generation, isotope production and destruction (actinides), and hydrogen production. Having conducted the initial research on the graphite foam and having performed the scoping parametric analyses from neutronics and thermal-hydraulic perspectives, it was necessary to focus on a particular application that would (1) demonstrate the viability of the overall concept and (2) require a reasonably structured design analysis process that would synthesize those important parameters that influence the concept the most as part of a feasible, working reactor system. Thus, the application targeted for this concept was supplying power for remote/harsh environments and a design that was easily deployable, simplistic from an operational standpoint, and utilized the new graphite foam. Specifically, a 500-kW(t) reactor concept was pursued that is naturally load following, inherently safe, optimized via neutronic studies to achieve near-zero reactivity change with burnup, and proliferation resistant. These four major areas

  2. Facile synthesis of CdS nanocrystals using thioglycolic acid as a sulfur source and stabilizer in aqueous solution

    Directory of Open Access Journals (Sweden)

    Liya Zhou

    2011-12-01

    Full Text Available A novel method has been developed for the synthesis of CdS nanocrystals (NCs using thioglycolic acid (TGA as a sulfur source and stabilizer with the presence of hydrogen peroxide in an aqueous medium. The products were characterized by X-ray powder diffraction (XRD, transmission electron microscopy (TEM and Fourier transform infrared (FTIR spectroscopy. The results indicate that the product was of zinc-blend crystal structure in a sphere-like shape. The room-temperature luminescence spectra revealed that the emission peak of CdS NCs prepared in relatively short refluxing times (10-120 min could be tuned from 518 nm to 610 nm, and the photoluminescence quantum yield of the as-prepared CdS NCs could reach as high as 12.6%. In addition, the mechanism of the CdS nanocrystals formation was preliminarily discussed.

  3. Calculation Package: Derivation of Facility-Specific Derived Air Concentration (DAC) Values in Support of Spallation Neutron Source Operations

    Energy Technology Data Exchange (ETDEWEB)

    McLaughlin, David A [ORNL

    2009-12-01

    Derived air concentration (DAC) values for 175 radionuclides* produced at the Oak Ridge National Laboratory (ORNL) Spallation Neutron Source (SNS), but not listed in Appendix A of 10 CFR 835 (01/01/2009 version), are presented. The proposed DAC values, ranging between 1 E-07 {micro}Ci/mL and 2 E-03 {micro}Ci/mL, were calculated in accordance with the recommendations of the International Commission on Radiological Protection (ICRP), and are intended to support an exemption request seeking regulatory relief from the 10 CFR 835, Appendix A, requirement to apply restrictive DACs of 2E-13 {micro}Ci/mL and 4E-11 {micro}Ci/mL and for non-listed alpha and non-alpha-emitting radionuclides, respectively.

  4. Advanced Reactors Transition Program Resource Loaded Schedule

    Energy Technology Data Exchange (ETDEWEB)

    GANTT, D.A.

    2000-01-12

    The Advanced Reactors Transition (ART) Resource Loaded Schedule (RLS) provides a cost and schedule baseline for managing the project elements within the ART Program. The Fast Flux Test Facility (FETF) activities are delineated through the end of FY 2000, assuming continued standby. The Nuclear Energy (NE) Legacies and Plutonium Recycle Test Reactor (PRTR) activities are delineated through the end of the deactivation process. This revision reflects the 19 Oct 1999 baseline.

  5. Advanced Reactors Transition Program Resource Loaded Schedule

    Energy Technology Data Exchange (ETDEWEB)

    BOWEN, W.W.

    1999-11-08

    The Advanced Reactors Transition (ART) Resource Loaded Schedule (RLS) provides a cost and schedule baseline for managing the project elements within the ART Program. The Fast Flux Test Facility (FFTF) activities are delineated through the end of FY 2000, assuming continued standby. The Nuclear Energy (NE) Legacies and Plutonium Recycle Test Reactor (PRTR) activities are delineated through the end of the deactivation process. This document reflects the 1 Oct 1999 baseline.

  6. Source Water Protection Contaminant Sources

    Data.gov (United States)

    Iowa State University GIS Support and Research Facility — Simplified aggregation of potential contaminant sources used for Source Water Assessment and Protection. The data is derived from IDNR, IDALS, and US EPA program...

  7. Measurement of ion species in high current ECR H⁺/D⁺ ion source for IFMIF (International Fusion Materials Irradiation Facility).

    Science.gov (United States)

    Shinto, K; Senée, F; Ayala, J-M; Bolzon, B; Chauvin, N; Gobin, R; Ichimiya, R; Ihara, A; Ikeda, Y; Kasugai, A; Kitano, T; Kondo, K; Marqueta, A; Okumura, Y; Takahashi, H; Valette, M

    2016-02-01

    Ion species ratio of high current positive hydrogen/deuterium ion beams extracted from an electron-cyclotron-resonance ion source for International Fusion Materials Irradiation Facility accelerator was measured by the Doppler shift Balmer-α line spectroscopy. The proton (H(+)) ratio at the middle of the low energy beam transport reached 80% at the hydrogen ion beam extraction of 100 keV/160 mA and the deuteron (D(+)) ratio reached 75% at the deuterium ion beam extraction of 100 keV/113 mA. It is found that the H(+) ratio measured by the spectroscopy gives lower than that derived from the phase-space diagram measured by an Allison scanner type emittance monitor. The H(+)/D(+) ratio estimated by the emittance monitor was more than 90% at those extraction currents.

  8. Measurement of ion species in high current ECR H{sup +}/D{sup +} ion source for IFMIF (International Fusion Materials Irradiation Facility)

    Energy Technology Data Exchange (ETDEWEB)

    Shinto, K., E-mail: shinto.katsuhiro@jaea.go.jp; Ichimiya, R.; Ihara, A.; Ikeda, Y.; Kasugai, A.; Kitano, T.; Kondo, K.; Takahashi, H. [Rokkasho Fusion Institute, Japan Atomic Energy Agency, Rokkasho, Aomori 039-3212 (Japan); Senée, F.; Bolzon, B.; Chauvin, N.; Gobin, R.; Valette, M. [Commissariat à l’Energie Atomique et aux Energies Alternatives, CEA/Saclay, DSM/IRFU, 91191 Gif/Yvette (France); Ayala, J.-M.; Marqueta, A.; Okumura, Y. [IFMIF/EVEDA Project Team, Rokkasho, Aomori 039-3212 (Japan)

    2016-02-15

    Ion species ratio of high current positive hydrogen/deuterium ion beams extracted from an electron-cyclotron-resonance ion source for International Fusion Materials Irradiation Facility accelerator was measured by the Doppler shift Balmer-α line spectroscopy. The proton (H{sup +}) ratio at the middle of the low energy beam transport reached 80% at the hydrogen ion beam extraction of 100 keV/160 mA and the deuteron (D{sup +}) ratio reached 75% at the deuterium ion beam extraction of 100 keV/113 mA. It is found that the H{sup +} ratio measured by the spectroscopy gives lower than that derived from the phase-space diagram measured by an Allison scanner type emittance monitor. The H{sup +}/D{sup +} ratio estimated by the emittance monitor was more than 90% at those extraction currents.

  9. Hybrid adsorptive membrane reactor

    Energy Technology Data Exchange (ETDEWEB)

    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.

  10. D and DR Reactors

    Data.gov (United States)

    Federal Laboratory Consortium — The world's second full-scale nuclear reactor was the D Reactor at Hanford which was built in the early 1940's and went operational in December of 1944.D Reactor ran...

  11. Reactor transient

    Energy Technology Data Exchange (ETDEWEB)

    Menegus, R.L.

    1956-05-31

    The authors are planning a calculation to be done on the Univac at the Louviers Building to estimate the effect of xenon transients, a high reactor power. This memorandum outlines the reasons why they prefer to do the work at Louviers rather than at another location, such as N.Y.U. They are to calculate the response of the reactor to a sudden change in position of the half rods. Qualitatively, the response will be a change in the rooftop ratio of the neutron flux. The rooftop ratio may oscillate with high damping, or, instead, it may oscillate for many cycles. It has not been possible for them to determine this response by hand calculation because of the complexity of the problem, and yet it is important for them to be certain that high power operation will not lead us to inherently unstable operation. Therefore they have resorted to machine computation. The system of differential equations that describes the response has seven dependent variables; therefore there are seven equations, each coupled with one or more of the others. The authors have discussed the problem with R.R. Haefner at the plant, and it is his opinion that the IBM 650 cannot adequately handle the system of seven equations because the characteristic time constants vary over a range of about 10{sup 8}. The Univac located at the Louviers Building is said to be satisfactory for this computation.

  12. Education and Training on ISIS Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Foulon, F.; Badeau, G.; Lescop, B.; Wohleber, X. [French Atomic Energy and Alternative Energies Commission, Paris (France)

    2013-07-01

    In the frame of academic and vocational programs the National Institute for Nuclear Science and Technology uses the ISIS research reactor as a major tool to ensure a practical and comprehensive understanding of the nuclear reactor physics, principles and operation. A large set of training courses have been developed on ISIS, optimising both the content of the courses and the pedagogical approach. Programs with duration ranging from 3 hours (introduction to reactor operation) to 24 hours (full program for the future operators of research reactors) are carried out on ISIS reactor. The reactor is operated about 350 hours/year for education and training, about 40 % of the courses being carried out in English. Thus, every year about 400 trainees attend training courses on ISIS reactor. We present here the ISIS research reactor and the practical courses that have been developed on ISIS reactor. Emphasis is given to the pedagogical method which is used to focus on the operational and safety aspects, both in normal and incidental operation. We will present the curricula of the academic and vocational courses in which the practical courses are integrated, the courses being targeted to a wide public, including operators of research reactors, engineers involved in the design and operation of nuclear reactors as well as staff of the regulatory body. We address the very positive impact of the courses on the development of the competences and skills of participants. Finally, we describe the Internet Reactor Laboratories (IRL) that are under development and will consist in broadcasting the training courses via internet to remote facilities or institutions.

  13. Optimizing a neutron-beam focusing device for the direct geometry time-of-flight spectrometer TOFTOF at the FRM II reactor source

    DEFF Research Database (Denmark)

    Rasmussen, N. G.; Simeoni, G. G.; Lefmann, K.

    2016-01-01

    A dedicated beam-focusing device has been designed for the direct geometry thermal-cold neutron time-of-flight spectrometer TOFTOF at the neutron facility FRM II (Garching, Germany). The prototype, based on the compressed Archimedes' mirror concept, benefits from the adaptive-optics technology (a...

  14. Reference worldwide model for antineutrinos from reactors

    Science.gov (United States)

    Baldoncini, Marica; Callegari, Ivan; Fiorentini, Giovanni; Mantovani, Fabio; Ricci, Barbara; Strati, Virginia; Xhixha, Gerti

    2015-03-01

    Antineutrinos produced at nuclear reactors constitute a severe source of background for the detection of geoneutrinos, which bring to the Earth's surface information about natural radioactivity in the whole planet. In this framework, we provide a reference worldwide model for antineutrinos from reactors, in view of reactors operational records yearly published by the International Atomic Energy Agency. We evaluate the expected signal from commercial reactors for ongoing (KamLAND and Borexino), planned (SNO +), and proposed (Juno, RENO-50, LENA, and Hanohano) experimental sites. Uncertainties related to reactor antineutrino production, propagation, and detection processes are estimated using a Monte Carlo-based approach, which provides an overall site-dependent uncertainty on the signal in the geoneutrino energy window on the order of 3%. We also implement the off-equilibrium correction to the reference reactor spectra associated with the long-lived isotopes, and we estimate a 2.4% increase of the unoscillated event rate in the geoneutrino energy window due to the storage of spent nuclear fuels in the cooling pools. We predict that the research reactors contribute to less than 0.2% to the commercial reactor signal in the investigated 14 sites. We perform a multitemporal analysis of the expected reactor signal over a time lapse of ten years using reactor operational records collected in a comprehensive database published at www.fe.infn.it/antineutrino.

  15. Tracking the sources of polybrominated diphenyl ethers in birds: Foraging in waste management facilities results in higher DecaBDE exposure in males

    Energy Technology Data Exchange (ETDEWEB)

    Gentes, Marie-Line, E-mail: gentes.marie_line@courrier.uqam.ca [Centre de recherche en toxicologie de l’environnement (TOXEN), Département des sciences biologiques, Université du Québec à Montréal, P.O. Box 8888, Station Centre-ville, Montreal, QC, Canada H3C 3P8 (Canada); Mazerolle, Marc J., E-mail: Marc.Mazerolle@uqat.ca [Centre for Forest Research, Université du Québec en Abitibi-Témiscamingue, 445 boulevard de l’Université, Rouyn-Noranda, QC, Canada J9X 5E9 (Canada); Giroux, Jean-François, E-mail: giroux.jean-francois@uqam.ca [Groupe de recherche en écologie comportementale et animale, Département des sciences biologiques, Université du Québec à Montréal, P.O. Box 8888, Station Centre-ville, Montreal, QC, Canada H3C 3P8 (Canada); Patenaude-Monette, Martin [Groupe de recherche en écologie comportementale et animale, Département des sciences biologiques, Université du Québec à Montréal, P.O. Box 8888, Station Centre-ville, Montreal, QC, Canada H3C 3P8 (Canada); and others

    2015-04-15

    Differences in feeding ecology are now recognized as major determinants of inter-individual variations in contaminant profiles of free-ranging animals, but exceedingly little attention has been devoted to the role of habitat use. Marked inter-individual variations and high levels of polybrominated diphenyl ethers (PBDEs) (e.g., DecaBDE) have previously been documented in ring-billed gulls (Larus delawarensis) breeding in a colony near Montreal (QC, Canada). However, the environmental sources of these compounds, and thus the reasons causing these large inter-individual variations remain unidentified. In the present study, we used GPS-based telemetry (±5 to 10 m precision) to track ring-billed gulls from this colony to reconstruct their movements at the landscape level. We related habitat use of individual gulls (n=76) to plasma concentrations (ng/g ww) and relative contributions (percentages) to Σ{sub 38}PBDEs of major congeners in the internationally restricted PentaBDE and current-use DecaBDE mixtures. Male gulls that visited waste management facilities (WMFs; i.e., landfills, wastewater treatment plants and related facilities; 25% of all GPS-tracked males) exhibited greater DecaBDE (concentrations and percentages) and lower PentaBDE (percentages) relative to those that did not. In contrast, no such relationships were found in females. Moreover, in males, DecaBDE (concentrations and percentages) increased with percentages of time spent in WMFs (i.e., ~5% of total foraging time), while PentaBDE (percentages) decreased. No relationships between percentages of time spent in other habitats (i.e., urban areas, agriculture fields, and St. Lawrence River) were found in either sex. These findings suggest that animals breeding in the vicinity of WMFs as well as mobile species that only use these sites for short stopovers to forage, could be at risk of enhanced DecaBDE exposure. - Highlights: • The study was conducted on breeding gulls with high levels of flame

  16. Assessment of Startup Fuel Options for a Test or Demonstration Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Carmack, Jon [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hayes, Steven [Idaho National Lab. (INL), Idaho Falls, ID (United States); Walters, L. C. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-09-01

    This document explores startup fuel options for a proposed test/demonstration fast reactor. The fuel options considered are the metallic fuels U-Zr and U-Pu-Zr and the ceramic fuels UO2 and UO2-PuO2 (MOX). Attributes of the candidate fuel choices considered were feedstock availability, fabrication feasibility, rough order of magnitude cost and schedule, and the existing irradiation performance database. The reactor-grade plutonium bearing fuels (U-Pu-Zr and MOX) were eliminated from consideration as the initial startup fuels because the availability and isotopics of domestic plutonium feedstock is uncertain. There are international sources of reactor grade plutonium feedstock but isotopics and availability are also uncertain. Weapons grade plutonium is the only possible source of Pu feedstock in sufficient quantities needed to fuel a startup core. Currently, the available U.S. source of (excess) weapons-grade plutonium is designated for irradiation in commercial light water reactors (LWR) to a level that would preclude diversion. Weapons-grade plutonium also contains a significant concentration of gallium. Gallium presents a potential issue for both the fabrication of MOX fuel as well as possible performance issues for metallic fuel. Also, the construction of a fuel fabrication line for plutonium fuels, with or without a line to remove gallium, is expected to be considerably more expensive than for uranium fuels. In the case of U-Pu-Zr, a relatively small number of fuel pins have been irradiated to high burnup, and in no case has a full assembly been irradiated to high burnup without disassembly and re-constitution. For MOX fuel, the irradiation database from the Fast Flux Test Facility (FFTF) is extensive. If a significant source of either weapons-grade or reactor-grade Pu became available (i.e., from an international source), a startup core based on Pu could be reconsidered.

  17. Site Suitability and Hazard Assessment Guide for Small Modular Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Wayne Moe

    2013-10-01

    Commercial nuclear reactor projects in the U.S. have traditionally employed large light water reactors (LWR) to generate regional supplies of electricity. Although large LWRs have consistently dominated commercial nuclear markets both domestically and abroad, the concept of small modular reactors (SMRs) capable of producing between 30 MW(t) and 900 MW(t) to generating steam for electricity is not new. Nor is the idea of locating small nuclear reactors in close proximity to and in physical connection with industrial processes to provide a long-term source of thermal energy. Growing problems associated continued use of fossil fuels and enhancements in efficiency and safety because of recent advancements in reactor technology suggest that the likelihood of near-term SMR technology(s) deployment at multiple locations within the United States is growing. Many different types of SMR technology are viable for siting in the domestic commercial energy market. However, the potential application of a particular proprietary SMR design will vary according to the target heat end-use application and the site upon which it is proposed to be located. Reactor heat applications most commonly referenced in connection with the SMR market include electric power production, district heating, desalinization, and the supply of thermal energy to various processes that require high temperature over long time periods, or a combination thereof. Indeed, the modular construction, reliability and long operational life purported to be associated with some SMR concepts now being discussed may offer flexibility and benefits no other technology can offer. Effective siting is one of the many early challenges that face a proposed SMR installation project. Site-specific factors dealing with support to facility construction and operation, risks to the plant and the surrounding area, and the consequences subsequent to those risks must be fully identified, analyzed, and possibly mitigated before a license

  18. 10 CFR 170.21 - Schedule of fees for production and utilization facilities, review of standard referenced design...

    Science.gov (United States)

    2010-01-01

    .... Amendment, Renewal, Other Approvals Full cost. C. Test Facility/Research Reactor/Critical Facility..., approvals of facility standard reference designs, re-qualification and replacement examinations for reactor... of fees Fees 1,2 A. Nuclear Power Reactors Application for Construction Permit Full cost. Early Site...

  19. Challenges regarding the start-up of an anaerobic biological sulphate reactor using H2 and CO2 as electron donor and carbon sources

    CSIR Research Space (South Africa)

    Roux, SP

    2009-05-01

    Full Text Available and energy source. Synthesis gas, comprise of H2 the preferred electron donor of sulphate reducing bacteria (SRB), CO2 a carbon source and CO and is available from coal-burning processes. Biological sulphate reduction technology using synthesis gas should...

  20. Decommissioning of commercial reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yui, Kohei [Japan Atomic Power Co., Tokyo (Japan)

    1997-02-01

    In the case of nuclear reactors, the diversion is often difficult as they are highly purposive, the disassembling is not easy as they are robust, and attention is required to handle the equipment containing radioactive substances. Decommissioning is defined as all the measures taken from the state that facilities become unused to the state of becoming green field. In Japan, already 40 years have elapsed since the effort for nuclear power was begun, and in this paper, the present state and future subjects of the decommissioning of nuclear power stations are summarized at the opportunity that the stop of commercial operation of Tokai Nuclear Power Station was decided recently. In the Tokai Nuclear Power Station, 166 MWe graphite-moderated, carbon dioxide-cooled reactor called improved Calder Hall type is installed, which started the operation in 1966. The circumstances of the decision to stop its operation are explained. The basic policy of the decommissioning of commercial nuclear power stations has been already published by the Advisory Committee for Energy. The state of the decommissioning in various foreign countries is reported. In Japan, the state of green field was realized in 1996 in the decommissioning of the JPDR in Japan Atomic Energy Research institute, and the decommissioning of the atomic powered ship ``Mutsu`` was completed. (K.I.)