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Sample records for activation analysis facility

  1. High-capacity neutron activation analysis facility

    A high-capacity neutron activation analysis facility, the Reactor Activation Facility, was designed and built and has been in operation for about a year at one of the Savannah River Plant's production reactors. The facility determines uranium and about 19 other elements in hydrogeochemical samples collected in the National Uranium Resource Evaluation program, which is sponsored and funded by the United States Department of Energy, Grand Junction Office. The facility has a demonstrated average analysis rate of over 10,000 samples per month, and a peak rate of over 16,000 samples per month. Uranium is determined by cyclic activation and delayed neutron counting of the U-235 fission products; other elements are determined from gamma-ray spectra recorded in subsequent irradiation, decay, and counting steps. The method relies on the absolute activation technique and is highly automated for round-the-clock unattended operation

  2. KFUPM fast neutron activation analysis facility

    A newly established Fast Neutron Activation Analysis facility at the Energy Research Laboratory is described. The facility mainly consists of a fast neutron irradiation station and a gamma ray counting station. Both stations are connected by a fast pneumatic sample transfer system which transports the sample from the irradiation station to the counting station in a short time of 3 s. The fast neutron activation analysis facility has been tested by measuring the 27A(n, α)24Na and 115In(n, n')115mIn cross sections at 14.8 and 2.5 MeV neutron energies, respectively. Within the experimental uncertainties, the measured cross sections for these elements agree with the published values. (orig.)

  3. Establishment of rabbit radiation facility for neutron activation analysis

    The transfer principle and the composition of a rabbit radiation facility for neutron activation analysis in a reactor were introduced. The functions and security designs of the pneumatic transfer system and automatic control system in the irradiation device were studied. By the testing,the transfer speed of the facility is 7.0 m/s. The facility has advantages of steady transmission, simple operation, easy maintenance, etc. The facility satisfies the demand of the neutron activation analysis for short half-life nuclides. (authors)

  4. In-beam activation analysis facility at MLZ, Garching

    Révay, Zs., E-mail: zsolt.revay@frm2.tum.de [Heinz Maier-Leibniz Zentrum (MLZ), Technische Universität München, 85748 Garching (Germany); Kudějová, P.; Kleszcz, K.; Söllradl, S. [Heinz Maier-Leibniz Zentrum (MLZ), Technische Universität München, 85748 Garching (Germany); Genreith, Christoph [Heinz Maier-Leibniz Zentrum (MLZ), Technische Universität München, 85748 Garching (Germany); Institute of Energy and Climate Research, IEK-6: Nuclear Waste and Reactor Safety Fuel Cycle, Forschungszentrum Jülich GmbH in der Helmholtz-Gemeinschaft, 52428 Jülich (Germany)

    2015-11-01

    The reconstruction of the prompt gamma activation analysis facility and the construction of the new low-background counting chamber at MLZ, Garching is presented. The improvement of the shielding and its effect on the radiation background is shown. The setting up and the fine-tuning of the electronics and their characterization are also discussed. The upgraded facility has been demonstrated to be applicable for both PGAA and neutron activation analysis using in-beam activation and decay counting in the low-background counting chamber. - Highlights: • Radiation background at the PGAA facility was efficiently reduced. • In-beam irradiation facility in the strongest neutron beam. • The best signal-to-background ratio at a PGAA facility was achieved.

  5. A description of the BNL active surface analysis facility

    Berkeley Nuclear Laboratories has a responsibility for the assessment of radioactive specimens arising both from post irradiation examination of power reactor components and structures and experimental programmes concerned with fission and activation product transport. Existing analytical facilities have been extended with the commissioning of an active surface analysis instrument (XSAM 800pci, Kratos Analytical). Surface analysis involves the characterisation of the outer few atomic layers of a solid surface/interface whose chemical composition and electronic structure will probably be different from the bulk. The new instrument consists three interconnected chambers positioned in series; comprising of a high vacuum sample introduction chamber, an ultra-high vacuum sample treatment/fracture chamber and an ultra-high vacuum sample analysis chamber. The sample analysis chamber contains the electron, X-ray and ion-guns and the electron and ion detectors necessary for performing X-ray photoelectron spectroscopy, scanning Auger microscopy and secondary-ion mass spectroscopy. The chamber also contains a high stability manipulator to enable sub-micron imaging of specimens to be achieved and provide sample heating and cooling between - 180 and 6000C. (author)

  6. Inventory of activation analysis facilities available in the European Community to Industrial users

    This inventory includes lists of activation equipment produced in the European Community, facilities available for industrial users and activation laboratories existing in the European companies. The aim of this inventory is to provide all information that may be useful, to companies interested in activation analysis, as well as to give an idea on existing routine applications and on the European market in facilities

  7. Software for a measuring facility for activation analysis

    A software package has been developed for an APPLE P.C. The programs are intended to control an automated measuring station for photon activation analysis at GELINA, the linear accelerator of C.B.N.M. at Geel (Belgium). They allow to set-up a measuring scheme, to execute it under computer control, to accumulate and store 2 K-spectra using a built-in ADC and to output the results as listings, plots or evaluated reports

  8. The role of neutron activation analysis technique Ex Industrial applications using the egyptian research reactor facilities

    This report covers several papers which deal with the industrial applications of the Neutron Activation Analysis Technique (NAAT) in Egypt. The applications include: exploration, mining, industrial environment and multielemental analysis of different materials, just for quality control, optimization, safety uses and help in improving the efficiency and economic evaluation. The technique principles, instrumentation, neutron irradiation facilities and experience of analysis are reviewed. Also, the current research activities using the ET-RR-1 facilities as well as a proposal for cold neutron applications in this field on the ET-RR-2 are given

  9. Utilization and facility of neutron activation analysis in HANARO research reactor

    The facilities of neutron activation analysis within a multi-purpose research reactor (HANARO) are described and the main applications of Neutron activation analysis (NAA) in Korea are reviewed. The sample irradiation tube, automatic and manual pneumatic transfer system, are installed at three irradiation holes. One irradiation hole is lined with a cadmium tube for epithermal-nal NAA. The performance of the NAA facility was examined to identify the characteristics of tube transfer system, irradiation sites and polyethylene irradiation capsule. The available thermal neutron flux with each irradiation site are in the range of 3.9x1013-1.6x1014 n/cm2·s and cadmium ratios are 15-250. Neutron activation analysis has been applied in the trace component analysis of nuclear, geological, biological, environmental and high purity materials and various polymers for research and development. Analytical services and the latest analytical results are summarized. (author)

  10. Self-sustainability of a research reactor facility with neutron activation analysis

    Long-term self-sustainability of a small reactor facility is possible because there is a large demand for non-destructive chemical analysis of bulk materials that can only be achieved with neutron activation analysis (NAA). The Ecole Polytechnique Montreal SLOWPOKE Reactor Facility has achieved self-sustainability for over twenty years, benefiting from the extreme reliability, ease of use and stable neutron flux of the SLOWPOKE reactor. The industrial clientele developed slowly over the years, mainly because of research users of the facility. A reliable NAA service with flexibility, high accuracy and fast turn-around time was achieved by developing an efficient NAA system, using a combination of the relative and k0 standardisation methods. The techniques were optimized to meet the specific needs of the client, such as low detection limit or high accuracy at high concentration. New marketing strategies are presented, which aim at a more rapid expansion. (author)

  11. Phase 1 sampling and analysis plan for the 304 Concretion Facility closure activities

    This document provides guidance for the initial (Phase 1) sampling and analysis activities associated with the proposed Resource Conservation and Recovery Act of 1976 (RCRA) clean closure of the 304 Concretion Facility. Over its service life, the 304 Concretion Facility housed the pilot plants associated with cladding uranium cores, was used to store engineering equipment and product chemicals, was used to treat low-level radioactive mixed waste, recyclable scrap uranium generated during nuclear fuel fabrication, and uranium-titanium alloy chips, and was used for the repackaging of spent halogenated solvents from the nuclear fuels manufacturing process. The strategy for clean closure of the 304 Concretion Facility is to decontaminate, sample (Phase 1 sampling), and evaluate results. If the evaluation indicates that a limited area requires additional decontamination for clean closure, the limited area will be decontaminated, resampled (Phase 2 sampling), and the result evaluated. If the evaluation indicates that the constituents of concern are below action levels, the facility will be clean closed. Or, if the evaluation indicates that the constituents of concern are present above action levels, the condition of the facility will be evaluated and appropriate action taken. There are a total of 37 sampling locations comprising 12 concrete core, 1 concrete chip, 9 soil, 11 wipe, and 4 asphalt core sampling locations. Analysis for inorganics and volatile organics will be performed on the concrete core and soil samples. Separate concrete core samples will be required for the inorganic and volatile organic analysis (VOA). Analysis for inorganics only will be performed on the concrete chip, wipe, and asphalt samples

  12. The Prompt Gamma Neutron Activation Analysis Facility at ICN—Pitesti

    Bǎrbos, D.; Pǎunoiu, C.; Mladin, M.; Cosma, C.

    2008-08-01

    PGNAA is a very widely applicable technique for determining the presence and amount of many elements simultaneously in samples ranging in size from micrograms to many grams. PGNAA is characterized by its capability for nondestructive multi-elemental analysis and its ability to analyse elements that cannot be determined by INAA. By means of this PGNAA method we are able to increase the performace of INAA method. A facility has been developed at Institute for Nuclear Research—Piteşti so that the unique features of prompt gamma-ray neutron activation analysis can be used to measure trace and major elements in samples. The facility is linked at the radial neutron beam tube at ACPR-TRIGA reactor. During the PGNAA—facility is in use the ACPR reactor will be operated in steady-state mode at 250 KW maximum power. The facility consists of a radial beam-port, external sample position with shielding, and induced prompt gamma-ray counting system. Thermal neutron flux with energy lower than cadmium cut-off at the sample position was measured using thin gold foil is: φscd = 1.106 n/cm2/s with a cadmium ratio of:80. The gamma-ray detection system consist of an HpGe detector of 16% efficiency (detector model GC1518) with 1.85 keV resolution capability. The HpGe is mounted with its axis at 90° with respect to the incident neutron beam at distance about 200mm from the sample position. To establish the performance capabilities of the facility, irradiation of pure element or sample compound standards were performed to identify the gama-ray energies from each element and their count rates.

  13. Establishment of prompt gamma neutron activation analysis facility at PARR-1

    Prompt gamma neutron activation analysis facility at through tube of upgraded PARR-1 reactor has been established. The salient features of the facility have been described. The in-pile as well as external collimators, beam shutter, target assembly and beam catcher were designed and fabricated indigenously. The flux at target position is 1.8x10/sup 10/ neutrons/cm/sup 2/sec and the cadmium ratio is about 10 Anti-Compton/pair spectrometer has been installed at 90 deg. to the incident beam. The measurements of the prompt gamma rays from thermal neutron capture in chlorine, nitrogen and chromium were carried out to calibrate HPGe detector up to 10 MeV. The set-up will be used in the determination of non-metals that form the elements of geological and biological materials or other trace elements with high thermal capture cross sections with improved peak/compton ratio of the spectrometer. (author)

  14. Prompt gamma neutron activation analysis facility at the RA-6 research reactor

    A prompt gamma neutron activation activation analysis facility was developed at the 500 kw thermal power RA-6 research reactor of the Bariloche Atomic Center, Argentina.This facility consist of a radial beam port with external positioning of the sample.The gamma radiation is reduced by a bismuth filter placed inside the extraction tube and the beam diameter is limited by a set of two collimators up to 5 cm.The neutron flux at the sample position is 7 106 n/cm2s with a Cadmium ratio of 20/1.The gamma detector is a 50 % efficiency type p HPGe rounded by a NaI(Tl) for Compton suppressioning.The gamma spectra is measured through 0 to 8.5 MeV.The background have counting rate of 350 cps without sample. In this work is shown the efficiency curve, the calculed sensibilities and the lower detection limits for B, Cd, Sm, Gd, H, Cl, Hg, Eu, Ti, Ag, Au, Mo. The RA-6's PGNAA facility is fully working, although the analytic capacity is under improvement

  15. Design of Stopper of Prompt Gamma Neutron Activation Analysis Facility at China Advanced Research Reactor

    2011-01-01

    The PGNAA facility consists of the filtered collimated neutron beam, the shielding of the whole facility, the control system, the detecting equipment and the data acquisition and analysis system. The neutron beam is filtered by a mono-crystalline bismuth filter,

  16. Development of a prompt gamma activation analysis facility using diffracted polychromatic neutron beam

    Byun, S H; Choi, H D

    2002-01-01

    A prompt gamma activation analysis facility has recently been developed at Hanaro, the 24 MW research reactor in the Korea Atomic Energy Research Institute. Polychromatic thermal neutrons are extracted by setting pyrolytic graphite crystals at a Bragg angle of 45 deg. . The detection system comprises a large single n-type HPGe detector, signal electronics and a fast ADC. Neutron beam characterization was performed both theoretically and experimentally. The neutron flux was measured to be 7.9x10 sup 7 n/cm sup 2 s in a 1x1 cm sup 2 beam area at the sample position with a uniformity of 12%. The corresponding Cd-ratio for gold was found to be 266. The beam quality was compared with other representative thermal neutron prompt gamma activation analysis. The detection efficiency was calibrated up to 11 MeV using a set of radionuclides and the (n,gamma) reactions of N and Cl. Finally, the sensitivities and the detection limits were obtained for several elements.

  17. Development of a prompt gamma activation analysis facility using diffracted polychromatic neutron beam

    A prompt gamma activation analysis facility has recently been developed at Hanaro, the 24 MW research reactor in the Korea Atomic Energy Research Institute. Polychromatic thermal neutrons are extracted by setting pyrolytic graphite crystals at a Bragg angle of 45 deg. . The detection system comprises a large single n-type HPGe detector, signal electronics and a fast ADC. Neutron beam characterization was performed both theoretically and experimentally. The neutron flux was measured to be 7.9x107 n/cm2 s in a 1x1 cm2 beam area at the sample position with a uniformity of 12%. The corresponding Cd-ratio for gold was found to be 266. The beam quality was compared with other representative thermal neutron prompt gamma activation analysis. The detection efficiency was calibrated up to 11 MeV using a set of radionuclides and the (n,γ) reactions of N and Cl. Finally, the sensitivities and the detection limits were obtained for several elements

  18. Neutron activation analysis at the Californium User Facility for Neutron Science

    The Californium User Facility (CUF) for Neutron Science has been established to provide 252Cf-based neutron irradiation services and research capabilities including neutron activation analysis (NAA). A major advantage of the CUF is its accessibility and controlled experimental conditions compared with those of a reactor environment The CUF maintains the world's largest inventory of compact 252Cf neutron sources. Neutron source intensities of ≤ 1011 neutrons/s are available for irradiations within a contamination-free hot cell, capable of providing thermal and fast neutron fluxes exceeding 108 cm-2 s-1 at the sample. Total flux of ≥109 cm-2 s-1 is feasible for large-volume irradiation rabbits within the 252Cf storage pool. Neutron and gamma transport calculations have been performed using the Monte Carlo transport code MCNP to estimate irradiation fluxes available for sample activation within the hot cell and storage pool and to design and optimize a prompt gamma NAA (PGNAA) configuration for large sample volumes. Confirmatory NAA irradiations have been performed within the pool. Gamma spectroscopy capabilities including PGNAA are being established within the CUF for sample analysis

  19. Development of a prompt gamma activation analysis facility using diffracted polychromatic neutron beam

    Byun, S. H.; Sun, G. M.; Choi, H. D.

    2002-07-01

    A prompt gamma activation analysis facility has recently been developed at Hanaro, the 24 MW research reactor in the Korea Atomic Energy Research Institute. Polychromatic thermal neutrons are extracted by setting pyrolytic graphite crystals at a Bragg angle of 45°. The detection system comprises a large single n-type HPGe detector, signal electronics and a fast ADC. Neutron beam characterization was performed both theoretically and experimentally. The neutron flux was measured to be 7.9×10 7 n/cm 2 s in a 1×1 cm 2 beam area at the sample position with a uniformity of 12%. The corresponding Cd-ratio for gold was found to be 266. The beam quality was compared with other representative thermal neutron prompt gamma activation analysis. The detection efficiency was calibrated up to 11 MeV using a set of radionuclides and the (n,γ) reactions of N and Cl. Finally, the sensitivities and the detection limits were obtained for several elements.

  20. Laser Guidance Analysis Facility

    Federal Laboratory Consortium — This facility, which provides for real time, closed loop evaluation of semi-active laser guidance hardware, has and continues to be instrumental in the development...

  1. Prompt gamma ray activation analysis using neutron beam from THOR facility

    A reactor-based facility for neutron-capture prompt gamma-ray spectrometry for activation analysis has been installed at the one megawatt Tsing Hua Open-pool Reactor. The system consists a neutron beam port with collimators, irradiation stand, external beam tube, neutron beam dump, and counting system. The counting system consists of a 25 % n-type high purity germanium main gamma-ray detector, a 9'' x 10'' NaI(T1) anti-Compton detector shield, and Compton-suppressed electronics coupled to the CANBERRA S-88 Multi-parameter analyzer. Although the neutron beam at the sample irradiation station has an intensity of only 1,300,000 n/cm2s with a cadmium ratio of 26 : 1, the background levels of the on-line measurement in the mixed neutron/gamma field are sufficiently low, resulting a satisfactory detection of many elemental composition in samples. The lower limits of detection of 35 elements in sample matrix of the present system and the current applications are discussed. (author)

  2. Characterization Of Normalization Factor In TRIGA 2000 Bandung Reactor Pneumatic Facility for Neutron Activation Analysis

    Neutron activation analysis using synthetic multielement comparators is prevalent method for multielement analysis. This comparison method has several limitations such as preparation of synthetic standard is time consuming and needs high cost. In order to overcome such difficulties, the use of normalization factor of sample geometry and irradiation position as well need to be done. The normalization factor is used to overcome flux inhomogeneity, so that the used of standard reference material can be minimized. In this research, characterization of normalization factor in pneumatic facility of TRIGA 2000 Bandung reactor, have been done. The determination was done for two sample positions (bottom and top) using polyethylene container. The average normalization factor at 60,30 and 15 second irradiation at 1500 k Watt for Cu sample gave values of 1.2848, 1.2908 and 1.3348 respectively. The effect of power reactor fluctuation on normalization factor was also studied. Fluctuation of power reactor under 2 % for sample position top and bottom gave deviation values of 3.1699% and 1.6238% respectively. The determination of normalization factor for Ti, I, V and AI reference standards have also been done. Normalization factor at 60 second irradiation at 1500 k Watt for Ti, I, V and AI reference standards gave mean values of 1.2554, 1.2066, 1.3625 and 1.2475 respectively. Normalization factor obtained of this research have a narrow range (<6.2%). The results obtained can be use in developing the NAA method, to minimize the spent of time, energy and cost

  3. Medical Image Analysis Facility

    1978-01-01

    To improve the quality of photos sent to Earth by unmanned spacecraft. NASA's Jet Propulsion Laboratory (JPL) developed a computerized image enhancement process that brings out detail not visible in the basic photo. JPL is now applying this technology to biomedical research in its Medical lrnage Analysis Facility, which employs computer enhancement techniques to analyze x-ray films of internal organs, such as the heart and lung. A major objective is study of the effects of I stress on persons with heart disease. In animal tests, computerized image processing is being used to study coronary artery lesions and the degree to which they reduce arterial blood flow when stress is applied. The photos illustrate the enhancement process. The upper picture is an x-ray photo in which the artery (dotted line) is barely discernible; in the post-enhancement photo at right, the whole artery and the lesions along its wall are clearly visible. The Medical lrnage Analysis Facility offers a faster means of studying the effects of complex coronary lesions in humans, and the research now being conducted on animals is expected to have important application to diagnosis and treatment of human coronary disease. Other uses of the facility's image processing capability include analysis of muscle biopsy and pap smear specimens, and study of the microscopic structure of fibroprotein in the human lung. Working with JPL on experiments are NASA's Ames Research Center, the University of Southern California School of Medicine, and Rancho Los Amigos Hospital, Downey, California.

  4. The Prompt Gamma Neutron Activation Analysis Facility at the RA-6 reactor of the Bariloche Atomic Centre, Argentina

    The RA-6 is a research reactor with 500 kW of thermal power, located at the Bariloche Atomic Centre. In one of its five extraction tube facilities a prompt gamma neutron activation analysis system is now under construction. The neutron thermal flux in the position sample is 7 106 n/cm2s using a 5 cm thick bismuth filter. This work presents two facility designs, a preliminary one and another one with some improvements. Shielding optimizing experiences which justify the incorporated improvements are described. The applications of them allow the measurement of a borated sample. Also presented is a new design of the beam catcher and it is compared with the old one by MCNP modelling. New applications are being considered in the frame of the contract with the IAEA under the Co-ordinated Research Project (CRP) on 'New Applications of PGNAA'. (author)

  5. Sample registration software for process automation in the Neutron Activation Analysis (NAA) Facility in Malaysia nuclear agency

    Neutron Activation Analysis (NAA) had been established in Nuclear Malaysia since 1980s. Most of the procedures established were done manually including sample registration. The samples were recorded manually in a logbook and given ID number. Then all samples, standards, SRM and blank were recorded on the irradiation vial and several forms prior to irradiation. These manual procedures carried out by the NAA laboratory personnel were time consuming and not efficient. Sample registration software is developed as part of IAEA/CRP project on ‘Development of Process Automation in the Neutron Activation Analysis (NAA) Facility in Malaysia Nuclear Agency (RC17399)’. The objective of the project is to create a pc-based data entry software during sample preparation stage. This is an effective method to replace redundant manual data entries that needs to be completed by laboratory personnel. The software developed will automatically generate sample code for each sample in one batch, create printable registration forms for administration purpose, and store selected parameters that will be passed to sample analysis program. The software is developed by using National Instruments Labview 8.6

  6. Sample registration software for process automation in the Neutron Activation Analysis (NAA) Facility in Malaysia nuclear agency

    Rahman, Nur Aira Abd, E-mail: nur-aira@nuclearmalaysia.gov.my; Yussup, Nolida; Ibrahim, Maslina Bt. Mohd; Mokhtar, Mukhlis B.; Soh Shaari, Syirrazie Bin Che; Azman, Azraf B. [Technical Support Division, Malaysian Nuclear Agency, 43000, Kajang, Selangor (Malaysia); Salim, Nazaratul Ashifa Bt. Abdullah [Division of Waste and Environmental Technology, Malaysian Nuclear Agency, 43000, Kajang, Selangor (Malaysia); Ismail, Nadiah Binti [Fakulti Kejuruteraan Elektrik, UiTM Pulau Pinang, 13500 Permatang Pauh, Pulau Pinang (Malaysia)

    2015-04-29

    Neutron Activation Analysis (NAA) had been established in Nuclear Malaysia since 1980s. Most of the procedures established were done manually including sample registration. The samples were recorded manually in a logbook and given ID number. Then all samples, standards, SRM and blank were recorded on the irradiation vial and several forms prior to irradiation. These manual procedures carried out by the NAA laboratory personnel were time consuming and not efficient. Sample registration software is developed as part of IAEA/CRP project on ‘Development of Process Automation in the Neutron Activation Analysis (NAA) Facility in Malaysia Nuclear Agency (RC17399)’. The objective of the project is to create a pc-based data entry software during sample preparation stage. This is an effective method to replace redundant manual data entries that needs to be completed by laboratory personnel. The software developed will automatically generate sample code for each sample in one batch, create printable registration forms for administration purpose, and store selected parameters that will be passed to sample analysis program. The software is developed by using National Instruments Labview 8.6.

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

    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.

  8. Application of the fast activation analysis facility of the TRIGA Mark II reactor

    Activation analyses for decision making performed with short lived nuclides would be the ideal method and could be applied more generally, if three requirements could be met: Broad applicability; High speed transportation systems and processing of very high information densities. This last point has turned out to be the bottle neck, preventing a broader application of this method. Concentrating on the third requirement, the author describes a new high rate gamma spectroscopy system with real time compensation of both dead time and pile up losses which works properly up to input rates of 320 kc, which has been developed and tested

  9. Data analysis facility at LAMPF

    This report documents the discussions and conclusions of a study held in July 1977 to develop the requirements for a data analysis facility to support the experimental program in medium-energy physics at the Clinton P. Anderson Meson Physics Facility (LAMPF). 2 tables

  10. The prompt gamma neutron activation analysis facility at the RA-6 reactor of the Bariloche Atomic Center, Argentina

    The RA-6 is a pool type research reactor with high enrichment uranium fuel and 500 kW of nominal power. A Prompt Gamma Neutron Activation Analysis (PGNAA) facility is setting up at the radial beam tube no. 2. The development of this facility was motivated by the request of 10B analysis in the application of Boron Neutron Capture Therapy. New applications are being considered in the frame of the contract signed with IAEA under the CRP on 'New Applications of PGNAA'. The beam characteristics have been fully investigated both theoretically and experimentally. MCNP simulations were used to test different materials and geometries for each component of the facility. Gamma dose rate and neutron flux measurements were performed in order to validate the calculations and complete the design. The design goals were: To get a thermal neutron flux in the order of 107 n cm-2 s-1 at sample position; To reduce as much as possible the neutron and gamma background on the detection system and the surrounding areas; To achieve a detection limit of 1 mg of 10B. A schematic layout of the facility is shown. A bismuth filter is used to improve the neutron/gamma ratio of the free beam. The beam collimation is achieved by using two collimators: the first one is positioned 2 m downstream from the reactor core inside the biological shielding. This collimator consists of alternating rings of borated polyethylene and lead in order to absorb the neutrons and gamma rays. The inner diameter of each ring is gradually decreased. The last layer has an inner diameter of 6.8 cm. The total length of this collimator is 52.5 cm. A steel door situated just behind it isolates the beam tube allowing its flooding to shut the beam. The second collimator configures the beam size at the sample position. It is composed of alternating rings of lead and a mixture of lithium carbonate (66%wt) and paraffin. The last ring has an inner diameter of 40 mm. The beam shielding is completed by a massive box, situated around

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

    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

    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. Chemical Analysis Facility

    Federal Laboratory Consortium — FUNCTION: Uses state-of-the-art instrumentation for qualitative and quantitative analysis of organic and inorganic compounds, and biomolecules from gas, liquid, and...

  14. Design and Analysis Facility

    Federal Laboratory Consortium — Provides engineering design of aircraft components, subsystems and installations using Pro/E, Anvil 1000, CADKEY 97, AutoCAD 13. Engineering analysis tools include...

  15. Quality Assurance Project Plan for Facility Effluent Monitoring Plan activities

    This Quality Assurance Project Plan addresses the quality assurance requirements for the activities associated with the Facility Effluent Monitoring Plans, which are part of the overall Hanford Site Environmental Protection Plan. This plan specifically applies to the sampling and analysis activities and continuous monitoring performed for all Facility Effluent Monitoring Plan activities conducted by Westinghouse Hanford Company. It is generic in approach and will be implemented in conjunction with the specific requirements of the individual Facility Effluent Monitoring Plans

  16. Addendum to the composite analysis for the E-Area Vaults and Saltstone Disposal Facilities

    This report documents the composite analysis performed on the two active SRS low-level radioactive waste disposal facilities. The facilities are the Z-Area Saltstone Disposal Facility and the E-Area Vaults Disposal Facility

  17. Addendum to the composite analysis for the E-Area Vaults and Saltstone Disposal Facilities

    Cook, J.R.

    2000-03-13

    This report documents the composite analysis performed on the two active SRS low-level radioactive waste disposal facilities. The facilities are the Z-Area Saltstone Disposal Facility and the E-Area Vaults Disposal Facility.

  18. Active use of urban park facilities

    Lindberg, Michael; Schipperijn, Jasper

    2015-01-01

    Abstract Urban green spaces (UGS), and more specific a higher number of facilities in UGS, have been positively associated with physical activity (PA). However, more detailed studies of which facilities generate high levels of PA, for which type of users, are relevant as existing knowledge is...... mentioned as a key factor when designing facilities. Our results provide important knowledge to architects, planners and policy makers when aiming at designing activity-promoting facilities in UGS. Future studies need to further investigate the use of facilities among specific target groups, particularly...

  19. Activation analysis

    The neutron activation analysis, which appears to be in limits for further advance, is the most suitable for providing information on the principal as well as the microcomponents in any sample of solid form. Then, instrumental activation analysis is capable of determination of far many elements in various samples. Principally on the neutron activation analysis, the following are described in literature survey from 1982 to middle 1984: bibliography, review, data collection, etc.; problems in spectral analysis and measurement; activation analysis with neutrons; charged particle and photo-nucleus reactions; chemical separation, isotopic dilution activation analysis; molecular activation analysis; standard materials; life and its relation samples; environmental, food, court trial and archaeological samples; space and earth sciences. (Mori, K.)

  20. Soil sampling and analysis plan for the 3718-F Alkali Metal Treatment and Storage Facility closure activities

    Amendment V.13.B.b to the approved closure plan (DOE-RL 1995a) requires that a soil sampling and analysis plan be prepared and submitted to the Washington State Department of Ecology (Ecology) for review and approval. Amendment V.13.B.c requires that a diagram of the 3718-F Alkali Metal Treatment and Storage Facility unit (the treatment, storage, and disposal [TSD] unit) boundary that is to be closed, including the maximum extent of operation, be prepared and submitted as part is of the soil sampling and analysis plan. This document describes the sampling and analysis that is to be performed in response to these requirements and amends the closure plan. Specifically, this document supersedes Section 6.2, lines 43--46, and Section 7.3.6 of the closure plan. Results from the analysis will be compared to cleanup levels identified in the closure plan. These cleanup levels will be established using residential exposure assumptions in accordance with the Model Toxics Control Act (MTCA) Cleanup Regulation (Washington Administrative Code [WAC] 173-340) as required in Amendment V.13.B.I. Results of all sampling, including the raw analytical data, a summary of analytical results, a data validation package, and a narrative summary with conclusions will be provided to Ecology as specified in Amendment V.13.B.e. The results and process used to collect and analyze the soil samples will be certified by a licensed professional engineer. These results and a certificate of closure for the balance of the TSD unit, as outlined in Chapter 7.0 of the approved closure plan (storage shed, concrete pad, burn building, scrubber, and reaction tanks), will provide the basis for a closure determination

  1. Development of a prompt-gamma neutron activation analysis facility for small animal in vivo body composition studies using Am-Be Source

    Full text: The design, calibration, radiation dosimetry and preliminary performance evaluation of a prompt-gamma neutron activation analysis facility for in vivo body composition studies in small animals (i.e. rats or rabbits) are described. The system design was guided by Monte Carlo neutron and photon transport calculations performed using the MCNP-4C code. The facility utilizes a 555 GBq (15 Ci) Am-Be radionuclide neutron source positioned within a graphite collimator and appropriate shielding assembly. Prompt gamma rays produced by thermal neutron capture reactions within the animal are detected by a combination of a NaI(Tl) and a HPGe detectors positioned on either side of the sample, perpendicularly to the neutron beam. Small animal body nitrogen and hydrogen are determined by the NaI(Tl) detector by analysis of the 10.83 MeV and 2.22 MeV peaks, respectively, while calcium and chlorine are determined by the HPGe detector by analysis of the 6.42 MeV and 6.11 MeV peaks, respectively. Moreover, body potassium is determined independently by means of 40K measurement at a modified whole body counter facility. Appropriate corrections for animal body size and shape are applied. Mixed neutron and gamma radiation dosimetry was performed using a tissue-equivalent proportional counter. The facility described is a simple tool enabling us to perform in vivo analysis of the major body compartments of protein, bone mass, extra-cellular and intra-cellular space. It will be used to perform serial nutritional and metabolic studies in sets of small experimental animals under controlled conditions for an ethically accepted radiation dose and without the need to kill the animal. (author)

  2. Composite analysis E-area vaults and saltstone disposal facilities

    Cook, J.R.

    1997-09-01

    This report documents the Composite Analysis (CA) performed on the two active Savannah River Site (SRS) low-level radioactive waste (LLW) disposal facilities. The facilities are the Z-Area Saltstone Disposal Facility and the E-Area Vaults (EAV) Disposal Facility. The analysis calculated potential releases to the environment from all sources of residual radioactive material expected to remain in the General Separations Area (GSA). The GSA is the central part of SRS and contains all of the waste disposal facilities, chemical separations facilities and associated high-level waste storage facilities as well as numerous other sources of radioactive material. The analysis considered 114 potential sources of radioactive material containing 115 radionuclides. The results of the CA clearly indicate that continued disposal of low-level waste in the saltstone and EAV facilities, consistent with their respective radiological performance assessments, will have no adverse impact on future members of the public.

  3. Composite analysis E-area vaults and saltstone disposal facilities

    This report documents the Composite Analysis (CA) performed on the two active Savannah River Site (SRS) low-level radioactive waste (LLW) disposal facilities. The facilities are the Z-Area Saltstone Disposal Facility and the E-Area Vaults (EAV) Disposal Facility. The analysis calculated potential releases to the environment from all sources of residual radioactive material expected to remain in the General Separations Area (GSA). The GSA is the central part of SRS and contains all of the waste disposal facilities, chemical separations facilities and associated high-level waste storage facilities as well as numerous other sources of radioactive material. The analysis considered 114 potential sources of radioactive material containing 115 radionuclides. The results of the CA clearly indicate that continued disposal of low-level waste in the saltstone and EAV facilities, consistent with their respective radiological performance assessments, will have no adverse impact on future members of the public

  4. The development and medical applications of a simple facility for partial body in vivo neutron activation analysis using californium-252 sources

    A simple and cheap facility for partial body neutron activation analysis has been designed, based on the use of two 100 μg 252Cf neutron sources. The results reported show that calcium can be measured in parts of the body such as the tibia with a precision as good as +- 1.6 % for a radiation dose of 2 rem. The uniformity of the thermal neutron flux density is better than +- 3 % over 10 cm. Some applications of this irradiation facility for studies of trace elements, in particular cadmium in liver and aluminium in liver or brain, have also been explored. However, the sensitivity attainable is not yet sufficient for the study of normal levels, but could be of interest in toxicological investigations

  5. Quality Assurance Project Plan for Facility Effluent Monitoring Plan activities

    This Quality Assurance Project Plan addresses the quality assurance requirements for the Facility Monitoring Plans of the overall site-wide environmental monitoring plan. This plan specifically applies to the sampling and analysis activities and continuous monitoring performed for all Facility Effluent Monitoring Plan activities conducted by Westinghouse Hanford Company. It is generic in approach and will be implemented in conjunction with the specific requirements of individual Facility Effluent Monitoring Plans. This document is intended to be a basic road map to the Facility Effluent Monitoring Plan documents (i.e., the guidance document for preparing Facility Effluent Monitoring Plans, Facility Effluent Monitoring Plan determinations, management plan, and Facility Effluent Monitoring Plans). The implementing procedures, plans, and instructions are appropriate for the control of effluent monitoring plans requiring compliance with US Department of Energy, US Environmental Protection Agency, state, and local requirements. This Quality Assurance Project Plan contains a matrix of organizational responsibilities, procedural resources from facility or site manuals used in the Facility Effluent Monitoring Plans, and a list of the analytes of interest and analytical methods for each facility preparing a Facility Effluent Monitoring Plan. 44 refs., 1 figs., 2 tabs

  6. The Pacific Northwest story. [imagery analysis facilities

    Johnson, K. A.; Schrumpf, B. J.; Krebs, L.

    1981-01-01

    The establishment of image analysis facilities for the operational utilization of LANDSAT data in Idaho, Oregon, and Washington is discussed. The hardware and software resources are described for each facility along with the range of services.

  7. Active shooter in educational facility.

    Downs, Scott

    2015-01-01

    The last decade has seen several of the most heinous acts imaginable committed against our educational facilities. In light of the recent shooting in Sandy Hook Elementary School in Monroe (Newtown), CT, which took the lives of 20 children and six employees, a new heightened sense of awareness for safety and security among our educational facilities was created.(1) The law enforcement and public-safety community is now looking to work together with many of the educational representatives across the nation to address this issue, which affects the educational environment now and in the future. The US public and private elementary and secondary school systems' population is approximately 55.2 million students with an additional 19.1 million students attending a 2- and 4-year college or university. These same public and private school and degree-granting institutions employ approximately 7.6 million staff members who can be an enormous threshold of potential targets.(2) A terrorist's act, whether domestic, international, or the actions of a Lone Wolf against one of our educational facilities, would create a major rippling effect throughout our nation. Terrorists will stop at nothing to advance their ideology and they must continue to advance their most powerful tool-fear-to further their agenda and mission of destroying our liberty and the advanced civilization of the Western hemisphere. To provide the safety and security for our children and those who are employed to educate them, educational institutions must address this issue as well as nullify the possible threat to our national security. This thesis used official government reports and data interview methodologies to address various concerns from within our nation's educational system. Educational personnel along with safety and security experts identified, describe, and pinpointed the recommended measures that our educational institutions should include to secure our nation from within. These modifications of

  8. 303-K Storage Facility: Report on FY98 closure activities

    This report summarizes and evaluates the decontamination activities, sampling activities, and sample analysis performed in support of the closure of the 303-K Storage Facility. The evaluation is based on the validated data included in the data validation package (98-EAP-346) for the 303-K Storage Facility. The results of this evaluation will be used for assessing contamination for the purpose of closing the 303-K Storage Facility as described in the 303-K Storage Facility Closure Plan, DOE/RL-90-04. The closure strategy for the 303-K Storage Facility is to decontaminate the interior of the north half of the 303-K Building to remove known or suspected dangerous waste contamination, to sample the interior concrete and exterior soils for the constituents of concern, and then to perform data analysis, with an evaluation to determine if the closure activities and data meet the closure criteria. The closure criteria for the 303-K Storage Facility is that the concentrations of constituents of concern are not present above the cleanup levels. Based on the evaluation of the decontamination activities, sampling activities, and sample data, determination has been made that the soils at the 303-K Storage Facility meet the cleanup performance standards (WMH 1997) and can be clean closed. The evaluation determined that the 303-K Building cannot be clean closed without additional closure activities. An additional evaluation will be needed to determine the specific activities required to clean close the 303-K Storage Facility. The radiological contamination at the 303-K Storage Facility is not addressed by the closure strategy

  9. 303-K Storage Facility report on FY98 closure activities

    Adler, J.G.

    1998-07-17

    This report summarizes and evaluates the decontamination activities, sampling activities, and sample analysis performed in support of the closure of the 303-K Storage Facility. The evaluation is based on the validated data included in the data validation package (98-EAP-346) for the 303-K Storage Facility. The results of this evaluation will be used for assessing contamination for the purpose of closing the 303-K Storage Facility as described in the 303-K Storage Facility Closure Plan, DOE/RL-90-04. The closure strategy for the 303-K Storage Facility is to decontaminate the interior of the north half of the 303-K Building to remove known or suspected dangerous waste contamination, to sample the interior concrete and exterior soils for the constituents of concern, and then to perform data analysis, with an evaluation to determine if the closure activities and data meet the closure criteria. The closure criteria for the 303-K Storage Facility is that the concentrations of constituents of concern are not present above the cleanup levels. Based on the evaluation of the decontamination activities, sampling activities, and sample data, determination has been made that the soils at the 303-K Storage Facility meet the cleanup performance standards (WMH 1997) and can be clean closed. The evaluation determined that the 303-K Building cannot be clean closed without additional closure activities. An additional evaluation will be needed to determine the specific activities required to clean close the 303-K Storage Facility. The radiological contamination at the 303-K Storage Facility is not addressed by the closure strategy.

  10. Cold Vacuum Drying Facility hazard analysis report

    This report describes the methodology used in conducting the Cold Vacuum Drying Facility (CVDF) hazard analysis to support the CVDF phase 2 safety analysis report (SAR), and documents the results. The hazard analysis was performed in accordance with DOE-STD-3009-94, Preparation Guide for US Department of Energy Nonreactor Nuclear Facility Safety Analysis Reports, and implements the requirements of US Department of Energy (DOE) Order 5480.23, Nuclear Safety Analysis Reports

  11. 340 Waste handling Facility Hazard Categorization and Safety Analysis

    The analysis presented in this document provides the basis for categorizing the facility as less than Hazard Category 3. The final hazard categorization for the deactivated 340 Waste Handling Facility (340 Facility) is presented in this document. This hazard categorization was prepared in accordance with DOE-STD-1 027-92, Change Notice 1, Hazard Categorization and Accident Analysis Techniques for Compliance with Doe Order 5480.23, Nuclear Safety Analysis Reports. The analysis presented in this document provides the basis for categorizing the facility as less than Hazard Category (HC) 3. Routine nuclear waste receiving, storage, handling, and shipping operations at the 340 Facility have been deactivated, however, the facility contains a small amount of radioactive liquid and/or dry saltcake in two underground vault tanks. A seismic event and hydrogen deflagration were selected as bounding accidents. The generation of hydrogen in the vault tanks without active ventilation was determined to achieve a steady state volume of 0.33%, which is significantly less than the lower flammability limit of 4%. Therefore, a hydrogen deflagration is not possible in these tanks. The unmitigated release from a seismic event was used to categorize the facility consistent with the process defined in Nuclear Safety Technical Position (NSTP) 2002-2. The final sum-of-fractions calculation concluded that the facility is less than HC 3. The analysis did not identify any required engineered controls or design features. The Administrative Controls that were derived from the analysis are: (1) radiological inventory control, (2) facility change control, and (3) Safety Management Programs (SMPs). The facility configuration and radiological inventory shall be controlled to ensure that the assumptions in the analysis remain valid. The facility commitment to SMPs protects the integrity of the facility and environment by ensuring training, emergency response, and radiation protection. The full scale

  12. Europlanet Research Infrastructure: Planetary Sample Analysis Facilities

    Cloquet, C.; Mason, N. J.; Davies, G. R.; Marty, B.

    2008-09-01

    EuroPlanet The Europlanet Research Infrastructure consortium funded under FP7 aims to provide the EU Planetary Science community greater access for to research infrastructure. A series of networking and outreach initiatives will be complimented by joint research activities and the formation of three Trans National Access distributed service laboratories (TNA's) to provide a unique and comprehensive set of analogue field sites, laboratory simulation facilities, and extraterrestrial sample analysis tools. Here we report on the infrastructure that comprises the third TNA: Planetary Sample Analysis Facilities. The modular infrastructure represents a major commitment of analytical instrumentation by three institutes and together forms a state-of-the-art analytical facility of unprecedented breadth. These centres perform research in the fields of geochemistry and cosmochemistry, studying fluids and rocks in order to better understand the keys cof the universe. Europlanet Research Infrastructure Facilities: Ion Probe facilities at CRPG and OU The Cameca 1270 Ion microprobe is a CNRS-INSU national facility. About a third of the useful analytical time of the ion probe (about 3 months each year) is allocated to the national community. French scientists have to submit their projects to a national committee for selection. The selected projects are allocated time in the following 6 months twice a year. About 15 to 20 projects are run each year. There are only two such instruments in Europe, with cosmochemistry only performed at CRPG. Different analyses can be performed on a routine basis, such as U-Pb dating on Zircon, Monazite or Pechblende, Li, B, C, O, Si isotopic ratios determination on different matrix, 26Al, 60Fe extinct radioactivity ages, light and trace elements contents . The NanoSIMS 50L - producing element or isotope maps with a spatial resolution down to ≈50nm. This is one of the cornerstone facilities of UKCAN, with 75% of available instrument time funded and

  13. Safety analysis report for the Waste Storage Facility. Revision 2

    Bengston, S.J.

    1994-05-01

    This safety analysis report outlines the safety concerns associated with the Waste Storage Facility located in the Radioactive Waste Management Complex at the Idaho National Engineering Laboratory. The three main objectives of the report are: define and document a safety basis for the Waste Storage Facility activities; demonstrate how the activities will be carried out to adequately protect the workers, public, and environment; and provide a basis for review and acceptance of the identified risk that the managers, operators, and owners will assume.

  14. Research activities by INS cyclotron facility

    Research activities made by the cyclotron facility and the related apparatuses at Institute for Nuclear Study (INS), University of Tokyo, have been reviewed in terms of the associated scientific publications. This publication list, which is to be read as a continuation of INS-Rep.-608 (October, 1986), includes experimental works on low-energy nuclear physics, accelerator technology, instrumental developments, radiation physics and other applications in interdisciplinary fields. The publications are classified into the following four categories. (A) : Internal reports published in INS. (B) : Publications in international scientific journals on experimental research works done by the cyclotron facility and the related apparatuses at INS. Those made by outside users are also included. (C) : Publications in international scientific journals on experimental low-energy nuclear physics, which have been done by the staff of INS Nuclear Physics Division using facilities outside INS. (D) : Contributions to international conferences. (author)

  15. Analysis of facility-monitoring data

    Howell, J.A.

    1996-09-01

    This paper discusses techniques for analysis of data collected from nuclear-safeguards facility-monitoring systems. These methods can process information gathered from sensors and make interpretations that are in the best interests of the facility or agency, thereby enhancing safeguards while shortening inspection time.

  16. Operating procedures: Fusion Experiments Analysis Facility

    The Fusion Experiments Analysis Facility (FEAF) is a computer facility based on a DEC VAX 11/780 computer. It became operational in late 1982. At that time two manuals were written to aid users and staff in their interactions with the facility. This manual is designed as a reference to assist the FEAF staff in carrying out their responsibilities. It is meant to supplement equipment and software manuals supplied by the vendors. Also this manual provides the FEAF staff with a set of consistent, written guidelines for the daily operation of the facility

  17. Research and development activities of a neutron generator facility

    The neutron generator facility at YNRC is used for elemental analysis, nuclear data measurement and education. In nuclear data measurement the focus is on re-evaluating the existing scattered nuclear activation cross-section to obtain systematic data for nuclear reactions such as (n,p), (n,α), and (n,2n). In elemental analysis it is used for analyzing the Nitrogen (N), Phosphor (P) and Potassium (K) contents in chemical and natural fertilizers (compost), protein in rice, soybean, and corn and pollution level in rivers. The neutron generator is also used for education and training of BATAN staff and university students. The facility can also produce neutron generator components. (author)

  18. SRMAFTE facility checkout model flow field analysis

    Dill, Richard A.; Whitesides, Harold R.

    1992-07-01

    The Solid Rocket Motor Air Flow Equipment (SRMAFTE) facility was constructed for the purpose of evaluating the internal propellant, insulation, and nozzle configurations of solid propellant rocket motor designs. This makes the characterization of the facility internal flow field very important in assuring that no facility induced flow field features exist which would corrupt the model related measurements. In order to verify the design and operation of the facility, a three-dimensional computational flow field analysis was performed on the facility checkout model setup. The checkout model measurement data, one-dimensional and three-dimensional estimates were compared, and the design and proper operation of the facility was verified. The proper operation of the metering nozzles, adapter chamber transition, model nozzle, and diffuser were verified. The one-dimensional and three-dimensional flow field estimates along with the available measurement data are compared.

  19. DEMO Active Maintenance Facility concept progress 2012

    The DEMO Active Maintenance Facility (AMF) would be used for the storage, handling and processing of In-Vessel Components (IVC) throughout their time on site, the only exception being the time that they are installed in the vessel. It is anticipated that all handling operations associated with used components will have to be carried out using remote handling techniques. During plasma operations the In-Vessel Components are exposed to high levels of neutron activation. This activation results in high radiation dose rates and decay heating. This presents a significant problem for Remote Handling Equipment (RHE) in the AMF. The high dose rates require the equipment to be sufficiently radiation tolerant to allow it to work reliably for long periods. The decay heating requires forced cooling of newly removed IVC's while they are in storage. The duration of the storage is dependent on the decay heating reducing to a level that has been nominally set at <50 °C without active cooling in room temperature air. This paper summarises the progress made in 2012 on the conceptual design of the AMF and its facilities. The layout and proposed function of the main areas will be described along with the principles applied. The design of the AMF has evolved from a simple representation of the required facilities in 2011 to a concept that can be developed to support maintenance of DEMO

  20. DEMO Active Maintenance Facility concept progress 2012

    Thomas, Justin, E-mail: Justin.Thomas@ccfe.ac.uk [Euratom/CCFE Fusion Association, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Loving, Antony; Crofts, Oliver; Morgan, Robert [Euratom/CCFE Fusion Association, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Harman, Jon [EFDA, PPP and T, Boltzmannstraße 2, 85748 Garching (Germany)

    2014-10-15

    The DEMO Active Maintenance Facility (AMF) would be used for the storage, handling and processing of In-Vessel Components (IVC) throughout their time on site, the only exception being the time that they are installed in the vessel. It is anticipated that all handling operations associated with used components will have to be carried out using remote handling techniques. During plasma operations the In-Vessel Components are exposed to high levels of neutron activation. This activation results in high radiation dose rates and decay heating. This presents a significant problem for Remote Handling Equipment (RHE) in the AMF. The high dose rates require the equipment to be sufficiently radiation tolerant to allow it to work reliably for long periods. The decay heating requires forced cooling of newly removed IVC's while they are in storage. The duration of the storage is dependent on the decay heating reducing to a level that has been nominally set at <50 °C without active cooling in room temperature air. This paper summarises the progress made in 2012 on the conceptual design of the AMF and its facilities. The layout and proposed function of the main areas will be described along with the principles applied. The design of the AMF has evolved from a simple representation of the required facilities in 2011 to a concept that can be developed to support maintenance of DEMO.

  1. Buildings, fields of activity, testing facilities

    Since 1969 the activities of the Materialpruefungsanstalt Stuttgart (MPA) have grown quickly as planned, especially in the field of reactor safety research, which made it necessary to increase the staff to approximately 165 members, to supplement the machines and equipment and to extend the fields of activities occasioning a further departmental reorganization. At present the MPA has the following departments: 1. Teaching (materials testing, materials science and strength of materials) 2. Materials and Welding Technology 3. Materials Science and General Materials Testing with Tribology 4. Design and Strength 5. Creep and Fatigue Testing 6. Central Facilities 7. Vessel and Component Testing. (orig./RW)

  2. Hazard analysis in uranium hexafluoride production facility

    The present work provides a method for preliminary hazard analysis of nuclear fuel cycle facilities. The proposed method identify both chemical and radiological hazards, as well as the consequences associated with accident scenarios. To illustrate the application of the method, a uranium hexafluoride production facility was selected. The main hazards are identified and the potential consequences are quantified. It was found that, although the facility handles radioactive material, the main hazards as associated with releases of toxic chemical substances such as hydrogen fluoride, anhydrous ammonia and nitric acid. It was shown that a contention bung can effectively reduce the consequences of atmospheric release of toxic materials. (author)

  3. Sampling and Analysis Plan for the 221-U Facility

    This sampling and analysis plan (SAP) presents the rationale and strategy for the sampling and analysis activities proposed to be conducted to support the evaluation of alternatives for the final disposition of the 221-U Facility. This SAP will describe general sample locations and the minimum number of samples required. It will also identify the specific contaminants of potential concern (COPCs) and the required analysis. This SAP does not define the exact sample locations and equipment to be used in the field due to the nature of unknowns associated with the 221-U Facility

  4. Hot Cell Facility (HCF) Safety Analysis Report

    MITCHELL,GERRY W.; LONGLEY,SUSAN W.; PHILBIN,JEFFREY S.; MAHN,JEFFREY A.; BERRY,DONALD T.; SCHWERS,NORMAN F.; VANDERBEEK,THOMAS E.; NAEGELI,ROBERT E.

    2000-11-01

    This Safety Analysis Report (SAR) is prepared in compliance with the requirements of DOE Order 5480.23, Nuclear Safety Analysis Reports, and has been written to the format and content guide of DOE-STD-3009-94 Preparation Guide for U. S. Department of Energy Nonreactor Nuclear Safety Analysis Reports. The Hot Cell Facility is a Hazard Category 2 nonreactor nuclear facility, and is operated by Sandia National Laboratories for the Department of Energy. This SAR provides a description of the HCF and its operations, an assessment of the hazards and potential accidents which may occur in the facility. The potential consequences and likelihood of these accidents are analyzed and described. Using the process and criteria described in DOE-STD-3009-94, safety-related structures, systems and components are identified, and the important safety functions of each SSC are described. Additionally, information which describes the safety management programs at SNL are described in ancillary chapters of the SAR.

  5. Hot Cell Facility (HCF) Safety Analysis Report

    This Safety Analysis Report (SAR) is prepared in compliance with the requirements of DOE Order 5480.23, Nuclear Safety Analysis Reports, and has been written to the format and content guide of DOE-STD-3009-94 Preparation Guide for U. S. Department of Energy Nonreactor Nuclear Safety Analysis Reports. The Hot Cell Facility is a Hazard Category 2 nonreactor nuclear facility, and is operated by Sandia National Laboratories for the Department of Energy. This SAR provides a description of the HCF and its operations, an assessment of the hazards and potential accidents which may occur in the facility. The potential consequences and likelihood of these accidents are analyzed and described. Using the process and criteria described in DOE-STD-3009-94, safety-related structures, systems and components are identified, and the important safety functions of each SSC are described. Additionally, information which describes the safety management programs at SNL are described in ancillary chapters of the SAR

  6. Synthetic multi-element standards: a good tool for calibration and quality control of irradiation facilities used for neutron activation analysis

    Neutron activation analysis (NAA) is a physical technique used for the absolute measurement of the concentration of substances in solids and liquids. The method uses neutron irradiation which is commonly realised using a nuclear reactor in order to activate (make radioactive) different isotopes of the elements present in the sample. The radionuclides produced in this way emit gamma-rays that are characteristic of the elements present in the sample. Using gamma-ray spectrometry these radionuclides can then be identified and quantified, and hence their concentration in the sample can be determined. Although NAA is a straightforward method it requires a sound control of the many physical parameters involved to obtain accurate results and to guarantee a set accuracy in routine analysis. The accuracy of NAA depends on the specific measurement method used. One can perform NAA in a relative way by co-irradiating a known standard and the unknown sample in the same conditions and by comparing the ratio of gamma-rays they emit. Relative NAA has limited applicability since it requires reference standards with a comparable composition as the unknown. A more generally applicable method is the k0-NAA method. In the k0-NAA method all measurements are relative to the element Au resulting in 198Au when irradiated. The k0-NAA method further relies on the fact that the neutron energy spectrum produced in a given position in the reactor can be parameterised with two parameters: the shape factor of the epithermal neutron flux, indicating the deviation of the epithermal neutron spectrum from the ideal 1/E shape approximated by a 1/E1+a distribution, with E the neutron energy; f: the thermal-to-epithermal neutron flux ratio. The parameters f and a are characteristic for the irradiation facility (reactor and irradiation channels) and may change or fluctuate in time according to the irradiation conditions. The way elements activate (become radioactive) when interacting with neutrons is

  7. Nuclear fuel cycle facility accident analysis handbook

    The Accident Analysis Handbook (AAH) covers four generic facilities: fuel manufacturing, fuel reprocessing, waste storage/solidification, and spent fuel storage; and six accident types: fire, explosion, tornado, criticality, spill, and equipment failure. These are the accident types considered to make major contributions to the radiological risk from accidents in nuclear fuel cycle facility operations. The AAH will enable the user to calculate source term releases from accident scenarios manually or by computer. A major feature of the AAH is development of accident sample problems to provide input to source term analysis methods and transport computer codes. Sample problems and illustrative examples for different accident types are included in the AAH

  8. 303-K Storage facility sampling and analysis plan

    Adler, J.G.

    1997-07-01

    This document describes the cleanup, sampling, and analysis activities associated with the closure of the 303-K Storage Facility under the Washington Administrative Code (WAC) 173-303-610, ``Dangerous Waste Regulations.`` this document is a supplement to the 303-K Storage Facility Closure Plan (DOE-RL 1995a) (Closure Plan). The objective of these activities is to support clean closure of the 303 K Storage Facility. This document defines the information and activities needed to meet this objective, including: constituents of concern, cleanup performance standards, cleanup activities, sampling locations and methods, field screening locations and methods, field quality control requirements, laboratory analytical methods, and data validation methodology. This document supersedes the Closure Plan if the two conflict

  9. Seismic risk analysis for General Electric Plutonium Facility, Pleasanton, California

    This report presents the results of a seismic risk analysis that focuses on all possible sources of seismic activity, with the exception of the postulated Verona Fault. The best estimate curve indicates that the Vallecitos facility will experience 30% g with a return period of roughly 130 years and 60% g with a return period of roughly 700 years

  10. Preliminary safety analysis report for the Waste Characterization Facility

    This safety analysis report outlines the safety concerns associated with the Waste Characterization Facility located in the Radioactive Waste Management Complex at the Idaho National Engineering Laboratory. The three main objectives of the report are to: define and document a safety basis for the Waste Characterization Facility activities; demonstrate how the activities will be carried out to adequately protect the workers, public, and environment; and provide a basis for review and acceptance of the identified risk that the managers, operators, and owners will assume. 142 refs., 38 figs., 39 tabs

  11. Production Facility System Reliability Analysis Report

    Dale, Crystal Buchanan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Klein, Steven Karl [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-10-06

    This document describes the reliability, maintainability, and availability (RMA) modeling of the Los Alamos National Laboratory (LANL) design for the Closed Loop Helium Cooling System (CLHCS) planned for the NorthStar accelerator-based 99Mo production facility. The current analysis incorporates a conceptual helium recovery system, beam diagnostics, and prototype control system into the reliability analysis. The results from the 1000 hr blower test are addressed.

  12. In vivo neutron activation facility at Brookhaven National Laboratory

    Ma, R.; Yasumura, Seiichi; Dilmanian, F.A.

    1997-11-01

    Seven important body elements, C, N, Ca, P, K, Na, and Cl, can be measured with great precision and accuracy in the in vivo neutron activation facilities at Brookhaven National Laboratory. The facilities include the delayed-gamma neutron activation, the prompt-gamma neutron activation, and the inelastic neutron scattering systems. In conjunction with measurements of total body water by the tritiated-water dilution method several body compartments can be defined from the contents of these elements, also with high precision. In particular, body fat mass is derived from total body carbon together with total body calcium and nitrogen; body protein mass is derived from total body nitrogen; extracellular fluid volume is derived from total body sodium and chlorine; lean body mass and body cell mass are derived from total body potassium; and, skeletal mass is derived from total body calcium. Thus, we suggest that neutron activation analysis may be valuable for calibrating some of the instruments routinely used in clinical studies of body composition. The instruments that would benefit from absolute calibration against neutron activation analysis are bioelectric impedance analysis, infrared interactance, transmission ultrasound, and dual energy x-ray/photon absorptiometry.

  13. Quantifying Detection Probabilities for Proliferation Activities in Undeclared Facilities

    International Safeguards is currently in an evolutionary process to increase effectiveness and efficiency of the verification system. This is an obvious consequence of the inability to detect the Iraq's clandestine nuclear weapons programme in the early 90s. By the adoption of the Programme 93+2, this has led to the development of Integrated Safeguards and the State-level concept. Moreover, the IAEA's focus was extended onto proliferation activities outside the State's declared facilities. The effectiveness of safeguards activities within declared facilities can and have been quantified with respect to costs and detection probabilities. In contrast, when verifying the absence of undeclared facilities this quantification has been avoided in the past because it has been considered to be impossible. However, when balancing the allocation of budget between the declared and the undeclared field, explicit reasoning is needed why safeguards effort is distributed in a given way. Such reasoning can be given by a holistic, information and risk-driven approach to Acquisition Path Analysis comprising declared and undeclared facilities. Regarding the input, this approach relies on the quantification of several factors, i.e., costs of attractiveness values for specific proliferation activities, potential safeguards measures and detection probabilities for these measures also for the undeclared field. In order to overcome the lack of quantification for detection probabilities in undeclared facilities, the authors of this paper propose a general verification error model. Based on this model, four different approaches are explained and assessed with respect to their advantages and disadvantages: the analogy approach, the Bayes approach, the frequentist approach and the process approach. The paper concludes with a summary and an outlook on potential future research activities. (author)

  14. EPA Facility Registry Service (FRS): RCRA_ACTIVE

    U.S. Environmental Protection Agency — This web feature service contains location and facility identification information from EPA's Facility Registry Service (FRS) for the subset of active hazardous...

  15. Computational analysis of PARAMETR facility experiments

    Full text of publication follows: Results of calculation of PARAMETR experiments are given in the paper. The PARAMETR facility is designed to research the phenomena relevant to typical LOCA scenarios (including severe accident) of VVER type reactors. The investigations at PARAMETR facility are directed to experimental research of fuel rods and core materials behavior, hydrogen generation processes, melting and interaction of core materials during severe accidents. The main facility components are rod bundle of 1250 mm heated length (up to 37 rods can be used), electrical power source, steam and water supply systems and instrumentation. The bundle is a mix of fresh fuel rods and electrically heated rods with uranium tablets and tungsten heater inside. The main objectives of calculations are analysis of computer code capability, in particular, RELAP/SCDAPSIM, to model severe accidents, identification of major parameter impact on calculation results and thus accident analysis improvements. RELAP/SCDAPSIM calculations were used to choose key parameters of experiments. Analysis of influence of thermal insulation properties, uncertainties of heater geometry, insulation thermal conductivity was done. Conditions and parameters needed to burn up intensive zirconium reaction were investigated. As a whole, calculation results showed good agreement with experiments. Some key points were observed such as essential impact of preheating phase, importance of thermal insulation material properties. Proper modeling of particular processes during preheating phase was very important since this phase defined bundle temperature level at the heating phase. There were some difficulties here. For instance, overestimation of temperatures had been observed until axial profiling of thermal conductivity was introduced. Some more proper models were used to reach the better agreement with experiments. The work done can be used in safety analysis of VVER type reactors and allow improving of

  16. Pulse radiolysis facilities and activities in Japan

    Pulse radiolysis studies in Japan have been reviewed in special reference to the facilities and the people who have engaged in the experiments. Main achievement is summarized with the list of selected publications. (author)

  17. Forensic activation analysis

    Basic principles of neutron activation analysis are outlined. Examples of its use in police science include analysis for gunshot residues, toxic element determinations and multielement comparisons. Advantages of neutron activation analysis over other techniques are described. (R.L.)

  18. Radiochemical analysis of military nuclear facilities

    Full text : Radiochemical Analysis is a branch of analytical chemistry comprising an aggregate of methods for qualitatively determining the composition and content of radioisotopes in the products of transformations. Safety and minimization of radiation impact on human and environment are important demand of operation of Military Nuclear Facilities (MNF). In accordance of recommendations of International Commission on Radiological Protection there are next objects of radiochemical analysis: 1) potential sources of radiochemical pollution; 2) environment (objects of environment, human environment including buildings, agricultural production, water, air et al.); 3) human himself (determination of dose from external and internal radiation, chemical poisoning). The chemical analysis can be carried out using, for example, the Gas Chromatography instrument whish separates chemical mixtures and identifies the components at a molecular level. It is one of the most accurate tools for analyzing environmental samples. The Gas Chromatography works on the principle that a mixture will separate into individual substances when heated. The heated gases are carried through a column with an inert gas (such as helium). As the separated substances emerge from the column opening, they flow into the Mass Spectrometry. Mass spectrometry identifies compounds by the mass of the analyte molecule. Newly developed portable Gas Chromatography and Mass Spectrometry are techniques that can be used to separate volatile organic compounds and pesticides. Other uses of Gas Chromatography, combined with other separation and analytical techniques, have been developed for radionuclides, explosive compounds such as royal demolition explosive and trinitrotoluene, and metals. So, based on the many years experience of operation of dangerous MNF, in concordance with norms of radiation and chemical safety it was considered that the tasks of the radiochemical analysis of Military Nuclear Facilities include

  19. Neutron Activation Analysis

    Corliss, William R.

    1968-01-01

    In activation analysis, a sample of an unknown material is first irradiated (activated) with nuclear particles. In practice these nuclear particles are almost always neutrons. The success of activation analysis depends upon nuclear reactions which are completely independent of an atom's chemical associations. The value of activation analysis as a research tool was recognized almost immediately upon the discovery of artificial radioactivity. This book discusses activation analysis experiments, applications and technical considerations.

  20. Analysis - The new legislation for nuclear facilities

    The ministerial order dated February 7, 2012 settles the legal rules concerning nuclear facilities such as laboratories using nuclear materials or operating research reactors and nuclear power plants. These new rules are more a clearer, legal frame of nuclear activities than an extension of the present legislation. Among the changes we can quote the implementation of sanctions or the concept of global safety that means that the potential impacts on the environment must be taken into account all along the operating life of the plant and also during the dismantling and the management of the resulting radioactive wastes. Undeniably positive this legal framework should not be too rigid for small facilities and it must be considered as an help for plant operators to assume their responsibility. This legal framework can be considered as an harmonization at the European scale in terms of safety requirements because it allows the implementation in the French law of the WENRA standards. This document gathers a series of short articles describing the different aspects of this new regulation: the benefits, its preparation, its progressive implementation and the results that are expected. (A.C.)

  1. Evaluation of energy system analysis techniques for identifying underground facilities

    VanKuiken, J.C.; Kavicky, J.A.; Portante, E.C. [and others

    1996-03-01

    This report describes the results of a study to determine the feasibility and potential usefulness of applying energy system analysis techniques to help detect and characterize underground facilities that could be used for clandestine activities. Four off-the-shelf energy system modeling tools were considered: (1) ENPEP (Energy and Power Evaluation Program) - a total energy system supply/demand model, (2) ICARUS (Investigation of Costs and Reliability in Utility Systems) - an electric utility system dispatching (or production cost and reliability) model, (3) SMN (Spot Market Network) - an aggregate electric power transmission network model, and (4) PECO/LF (Philadelphia Electric Company/Load Flow) - a detailed electricity load flow model. For the purposes of most of this work, underground facilities were assumed to consume about 500 kW to 3 MW of electricity. For some of the work, facilities as large as 10-20 MW were considered. The analysis of each model was conducted in three stages: data evaluation, base-case analysis, and comparative case analysis. For ENPEP and ICARUS, open source data from Pakistan were used for the evaluations. For SMN and PECO/LF, the country data were not readily available, so data for the state of Arizona were used to test the general concept.

  2. Neutronics analysis of the Laboratory Microfusion Facility

    The radiological safety hazards of the experimental area (EA) for the proposed Inertial Confinement Fusion (ICF) Laboratory Microfusion Facility (LMF) have been examined. The EA includes those structures required to establish the proper pre-shot environment, point the beams, contain the pellet yield, and measure many different facets of the experiments. The radiation dose rates from neutron activation of representative target chamber materials, the laser beam tubes and the argon gas they contain, the air surrounding the chamber, and the concrete walls of the experimental area are given. Combining these results with the allowable dose rates for workers, we show how radiological considerations affect access to the inside of the target chamber and to the diagnostic platform area located outside the chamber. Waste disposal and tritium containment issues are summarized. Other neutronics issues, such as radiation damage to the final optics and neutron heating of materials placed close to the target, are also addressed. 16 refs., 2 figs., 1 tab

  3. Neutronics analysis of the laboratory microfusion facility

    The radiological safety hazards of the experimental area (EA) for the proposed Inertial Confinement Fusion (ICF) Laboratory Microfusion Facility (LMF) have been examined. The EA includes those structures required to establish the proper pre-shot environment, point the beams, contain the pellet yield, and measure many different facets of the experiments. The radiation dose rates from neutron activation of representative target chamber materials, the laser beam tubes and the argon gas they contain, the air surrounding the chamber, and the concrete walls of the experimental area are given. Combining these results with the allowable dose rates for workers, the authors show how radiological considerations affect access to the inside of the target chamber and to the diagnostic platform area located outside the chamber. Waste disposal and tritium containment issues are summarized. Other neutronics issues, such as radiation damage to the final optics and neutron heating of materials placed close to the target, are also addressed

  4. Exploratory Studies Facility Subsurface Fire Hazards Analysis

    The primary objective of this Fire Hazard Analysis (FHA) is to confirm the requirements for a comprehensive fire and related hazards protection program for the Exploratory Studies Facility (ESF) are sufficient to minimize the potential for: (1) The occurrence of a fire or related event. (2) A fire that causes an unacceptable on-site or off-site release of hazardous or radiological material that will threaten the health and safety of employees, the public or the environment. (3) Vital US. Department of Energy (DOE) programs suffering unacceptable interruptions as a result of fire and related hazards. (4) Property losses from a fire and related events exceeding limits established by DOE. (5) Critical process controls and safety class systems being damaged as a result of a fire and related events

  5. The data analysis facilities that astronomers want

    This paper discusses the need and importance of data analysis facilities and what astronomers ideally want. A brief survey is presented of what is available now and some of the main deficiencies and problems with today's systems are discussed. The main sources of astronomical data are presented incuding: optical photographic, optical TV/CCD, VLA, optical spectros, imaging x-ray satellite, and satellite planetary camera. Landmark discoveries are listed in a table, some of which include: our galaxy as an island, distance to stars, H-R diagram (stellar structure), size of our galaxy, and missing mass in clusters. The main problems at present are discussed including lack of coordinated effort and central planning, differences in hardware, and measuring performance

  6. Exploratory Studies Facility Subsurface Fire Hazards Analysis

    The primary objective of this Fire Hazard Analysis (FHA) is to confirm the requirements for a comprehensive fire and related hazards protection program for the Exploratory Studies Facility (ESF) are sufficient to minimize the potential for: The occurrence of a fire or related event; A fire that causes an unacceptable on-site or off-site release of hazardous or radiological material that will threaten the health and safety of employees, the public or the environment; Vital U.S. Department of Energy (DOE) programs suffering unacceptable interruptions as a result of fire and related hazards; Property losses from a fire and related events exceeding limits established by DOE; and Critical process controls and safety class systems being damaged as a result of a fire and related events

  7. Mission analysis report - deactivation facilities at Hanford

    Lund, D.P.

    1996-09-27

    This document examines the portion of the Hanford Site Cleanup Mission that deals with facility deactivation. How facilities get identified for deactivation, how they enter EM-60 for deactivation, programmatic alternatives to perform facility deactivation, the deactivation process itself, key requirements and objectives associated with the deactivation process, and deactivation planning are discussed.

  8. Facilities of management magnetoresistive transformer of active power

    Val. S. Vuntesmeri

    2009-03-01

    Full Text Available Management facilities are considered, spectral composition is certain and the form of коммутируемого signal of magnetoresistive transformer of active power is rotined.

  9. Waste sampling and characterization facility complex safety analysis

    Meloy, R.T., Westinghouse Hanford

    1996-06-04

    The Waste Sampling and Characterization Facility is a `Non-Nuclear, Radiological Facility. This document demonstrates, by analysis, that WSCF can meet the chemical and radiological inventory limits for a radiological facility. It establishes control that ensures those inventories are maintained below threshold values to preserve the `Non- Nuclear, Radiological` classification.

  10. Criticality safety analysis for mockup facility

    Benchmark calculations for SCALE4.4 CSAS6 module have been performed for 31 UO2 fuel, 15MOX fuel and 10 metal material criticality experiments and then calculation biases of the SCALE 4.4 CSAS6 module have been revealed to be 0.00982, 0.00579 and 0.02347, respectively. When CSAS6 is applied to the criticality safety analysis for the mockup facility in which several kinds of nuclear material components are included, the calculation bias of CSAS6 is conservatively taken to be 0.02347. With the aid of this benchmarked code system, criticality safety analyses for the mockup facility at normal and hypothetical accidental conditions have been carried out. It appears that the maximum Keff is 0.28356 well below than the critical limit, Keff=0.95 at normal condition. In a hypothetical accidental condition, the maximum Keff is found to be 0.73527 much lower than the subcritical limit. For another hypothetical accidental condition the nuclear material leaks out of container and spread or lump in the floor, it was assumed that the nuclear material is shaped into a slab and water exists in the empty space of the nuclear material. Keff has been calculated as function of slab thickness and the volume ratio of water to nuclear material. The result shows that the Keff increases as the water volume ratio increases. It is also revealed that the Keff reaches to the maximum value when water if filled in the empty space of nuclear material. The maximum Keff value is 0.93960 lower than the subcritical limit

  11. 300 Area fuel supply facilities deactivation mission analysis report

    This report presents the results of the 300 Area fuel supply facilities (formerly call ''N reactor fuel fabrication facilities'') Deactivation Project mission analysis. Hanford systems engineering (SE) procedures call for a mission analysis. The mission analysis is an important first step in the SE process

  12. Cold vacuum drying facility final hazard analysis report

    POWERS, T.B.

    1999-06-07

    This report describes the methodology used in conducting the Cold Vacuum Dlying Facility (CVDF) Hazard Analysis to support the CVDF Final Safety Analysis Report and documents the results. The hazard analysis was performed in accordance with DOE-STD-3009-94, ''Preparation Guide for U.S. Department of Energy Nonreactor Nuclear Facility Safety Analysis Reports,'' and implements the requirements of DOE Order 5480.23, ''Nuclear Safety Analysis Reports.''

  13. Cold vacuum drying facility final hazard analysis report

    This report describes the methodology used in conducting the Cold Vacuum Dlying Facility (CVDF) Hazard Analysis to support the CVDF Final Safety Analysis Report and documents the results. The hazard analysis was performed in accordance with DOE-STD-3009-94, ''Preparation Guide for U.S. Department of Energy Nonreactor Nuclear Facility Safety Analysis Reports,'' and implements the requirements of DOE Order 5480.23, ''Nuclear Safety Analysis Reports.''

  14. Cold Vacuum Drying (CVD) Facility Final Hazard Analysis Report

    This report describes the methodology used in conducting the Cold Vacuum Drying Facility (CVDF) hazard Analysis to support the CVDF Final Safety Analysis Report and documents the results. The hazard analysis was performed in accordance with DOE-STD-3009-94, ''Preparation Guide for US. Department of Energy Nonreactor Nuclear Facility Safety Analysis Reports,'' and implements the requirements of DOE Order 5480.23, ''Nuclear Safety Analysis Reports.''

  15. Cold Vacuum Drying (CVD) Facility Hazards Analysis Report

    CROWE, R.D.

    2000-08-07

    This report describes the methodology used in conducting the Cold Vacuum Drying Facility (CVDF) Hazard Analysis to support the CVDF Final Safety Analysis Report and documents the results. The hazard analysis was performed in accordance with DOE-STD-3009-94, ''Preparation Guide for US. Department of Energy Nonreactor Nuclear Facility Safety Analysis Reports,'' and implements the requirements of DOE Order 5480.23, ''Nuclear Safety Analysis Reports.''

  16. Cold Vacuum Drying (CVD) Facility Hazards Analysis Report

    This report describes the methodology used in conducting the Cold Vacuum Drying Facility (CVDF) Hazard Analysis to support the CVDF Final Safety Analysis Report and documents the results. The hazard analysis was performed in accordance with DOE-STD-3009-94, ''Preparation Guide for US. Department of Energy Nonreactor Nuclear Facility Safety Analysis Reports,'' and implements the requirements of DOE Order 5480.23, ''Nuclear Safety Analysis Reports.''

  17. 105-DR Large Sodium Fire Facility closure activities evaluation report

    This report evaluates the closure activities at the 105-DR Large Sodium Fire Facility. The closure activities discussed include: the closure activities for the structures, equipment, soil, and gravel scrubber; decontamination methods; materials made available for recycling or reuse; and waste management. The evaluation compares these activities to the regulatory requirements and closure plan requirements. The report concludes that the areas identified in the closure plan can be clean closed

  18. Environmental analysis of biomass-ethanol facilities

    Corbus, D.; Putsche, V.

    1995-12-01

    This report analyzes the environmental regulatory requirements for several process configurations of a biomass-to-ethanol facility. It also evaluates the impact of two feedstocks (municipal solid waste [MSW] and agricultural residues) and three facility sizes (1000, 2000, and 3000 dry tons per day [dtpd]) on the environmental requirements. The basic biomass ethanol process has five major steps: (1) Milling, (2) Pretreatment, (3) Cofermentation, (4) Enzyme production, (5) Product recovery. Each step could have environmental impacts and thus be subject to regulation. Facilities that process 2000 dtpd of MSW or agricultural residues would produce 69 and 79 million gallons of ethanol, respectively.

  19. INTEGRATION OF FACILITY MODELING CAPABILITIES FOR NUCLEAR NONPROLIFERATION ANALYSIS

    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.

  20. Integration of facility modeling capabilities for nuclear nonproliferation analysis

    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.

  1. Integration Of Facility Modeling Capabilities For Nuclear Nonproliferation Analysis

    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.

  2. Integration of Facility Modeling Capabilities for Nuclear Nonproliferation Analysis

    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.

  3. Compilation of historical information of 300 Area facilities and activities

    Gerber, M.S.

    1992-12-01

    This document is a compilation of historical information of the 300 Area activities and facilities since the beginning. The 300 Area is shown as it looked in 1945, and also a more recent (1985) look at the 300 Area is provided.

  4. Physical Activity Breaks and Facilities in US Secondary Schools

    Hood, Nancy E.; Colabianchi, Natalie; Terry-McElrath, Yvonne M.; O'Malley, Patrick M.; Johnston, Lloyd D.

    2014-01-01

    Background: Research on physical activity breaks and facilities (indoor and outdoor) in secondary schools is relatively limited. Methods: School administrators and students in nationally representative samples of 8th (middle school) and 10th/12th grade (high school) students were surveyed annually from 2008-2009 to 2011-2012. School administrators…

  5. Compilation of historical information of 300 Area facilities and activities

    This document is a compilation of historical information of the 300 Area activities and facilities since the beginning. The 300 Area is shown as it looked in 1945, and also a more recent (1985) look at the 300 Area is provided

  6. Cold Vacuum Drying Facility Design Basis Accident Analysis Documentation

    This document provides the detailed accident analysis to support HNF-3553, Annex B, Spent Nuclear Fuel Project Final Safety Analysis Report, ''Cold Vacuum Drying Facility Final Safety Analysis Report (FSAR).'' All assumptions, parameters and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the FSAR

  7. Cold Vacuum Drying (CVD) Facility Design Basis Accident Analysis Documentation

    PIEPHO, M.G.

    1999-10-20

    This document provides the detailed accident analysis to support HNF-3553, Annex B, Spent Nuclear Fuel Project Final Safety Analysis Report, ''Cold Vacuum Drying Facility Final Safety Analysis Report (FSAR).'' All assumptions, parameters and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the FSAR.

  8. The Remote Handled Immobilization Low-Activity Waste Disposal Facility Environmental Permits and Approval Plan

    The purpose of this document is to revise Document HNF-SD-ENV-EE-003, ''Permitting Plan for the Immobilized Low-Activity Waste Project, which was submitted on September 4, 1997. That plan accounted for the interim storage and disposal of Immobilized-Low Activity Waste at the existing Grout Treatment Facility Vaults (Project W-465) and within a newly constructed facility (Project W-520). Project W-520 was to have contained a combination of concrete vaults and trenches. This document supersedes that plan because of two subsequent items: (1) A disposal authorization that was received on October 25, 1999, in a U. S. Department of Energy-Headquarters, memorandum, ''Disposal Authorization Statement for the Department of Energy Hanford site Low-Level Waste Disposal facilities'' and (2) ''Breakthrough Initiative Immobilized Low-Activity Waste (ILAW) Disposal Alternative,'' August 1999, from Lucas Incorporated, Richland, Washington. The direction within the U. S. Department of Energy-Headquarters memorandum was given as follows: ''The DOE Radioactive Waste Management Order requires that a Disposal authorization statement be obtained prior to construction of new low-level waste disposal facility. Field elements with the existing low-level waste disposal facilities shall obtain a disposal authorization statement in accordance with the schedule in the complex-wide Low-Level Waste Management Program Plan. The disposal authorization statement shall be issued based on a review of the facility's performance assessment and composite analysis or appropriate CERCLA documentation. The disposal authorization shall specify the limits and conditions on construction, design, operations, and closure of the low-level waste facility based on these reviews. A disposal authorization statement is a part of the required radioactive waste management basis for a disposal facility. Failure to obtain a disposal authorization statement or record of decision shall result in shutdown of an operational

  9. Recent Activities at the ORNL Multicharged Ion Research Facility (MIRF)

    Recent activities at the ORNL Multicharged Ion Research Facility (MIRF) are summarized. A brief summary of the MIRF high voltage (HV) platform and floating beam line upgrade is provided. An expansion of our research program to the use of molecular ion beams in heavy-particle and electron collisions, as well as in ion-surface interactions is described, and a brief description is provided of the most recently added Ion Cooling and Characterization End-station (ICCE) trap. With the expansion to include molecular ion beams, the acronym MIRF for the facility, however, remains unchanged: M can now refer to either Multicharged or Molecular.

  10. CANISTER HANDLING FACILITY - VENTILATION CONFINEMENT ZONING ANALYSIS

    The purpose of this calculation is to calculate the necessary airflow distribution used to size the HVAC equipment for the Canister Handling Facility. These results will be compared to the Heating and Cooling Load Calculation in detailed design. The calculations contained in this document were developed by DandE/Mechanical HVAC and are intended solely for the use of the DandE/Mechanical HVAC department in its work regarding the HVAC system for the Canister Handling Facility. Yucca Mountain Project personnel from the DandE/Mechanical HVAC department should be consulted before use of the calculations for purposes other than those stated herein or used by individuals other than authorized personnel in DandE/Mechanical HVAC department

  11. 13. seminar 'Activation analysis'

    Collection of the abstracts of contributions to the seminar covering broad ranges of application of activation analysis and improvements of systems and process steps. Most of them have been prepared separately for the energy data bases. (RB)

  12. Financial Analysis of a Health-Care Facility

    Bezděková, Pavla

    2009-01-01

    The aim of this work is carried out using selected methods the financial analysis of a Health-Care facility of the nature a hospital, an assessment of its financial health, its operation and financing.

  13. Activation of Air and Utilities in the National Ignition Facility

    Khater, H; Pohl, B; Brererton, S

    2010-04-08

    Detailed 3-D modeling of the NIF facility is developed to accurately simulate the radiation environment within the NIF. Neutrons streaming outside the NIF Target Chamber will activate the air present inside the Target Bay and the Ar gas inside the laser tubes. Smaller levels of activity are also generated in the Switchyard air and in the Ar portion of the SY laser beam path. The impact of neutron activation of utilities located inside the Target Bay is analyzed for variety of shot types. The impact of activating TB utilities on dose received by maintenance personnel post-shot is analyzed. The current NIF facility model includes all important features of the Target Chamber, shielding system, and building configuration. Flow of activated air from the Target Bay is controlled by the HVAC system. The amount of activated Target Bay air released through the stack is very small and does not pose significant hazard to personnel or the environment. Activation of Switchyard air is negligible. Activation of Target Bay utilities result in a manageable dose rate environment post high yield (20 MJ) shots. The levels of activation generated in air and utilities during D-D and THD shots are small and do not impact work planning post shots.

  14. A new PIXE/PIGME analysis facility

    A new PIXE/PIGME ion beam facility is being developed for use on the 3MeV Van de Graaff accelerator. It will be used to analyse geological samples prepared on microscope slides. The samples will be movable in the X and Y axis using remote or computer controlled motorised micrometers. Other features of the rig include remote selection of X-ray detector filters and beam defining apertures. Beam current monitoring is by backscattering, whilst target positions are determined optically using a variable gain borescope. The rig will be a useful tool when analysing very small targets or when target scanning is necessary

  15. Restoration activities in uranium mining and milling facilities in Spain

    From the end of the 80's up to now, several tasks have been carried out in Spain on restoration in the field of uranium mining and milling, significant among them being Andujar Uranium Mill (FUA) closure and La Haba closure. Also, a study has been carried out on restoration of inoperative and abandoned uranium mine sites. At present, detailed plans are being worked out for the project on the closure of the Elefante plant. All activities have been developed in the common framework of national standards and regulations which are generally in compliance with the standards, regulations and recommendations of international organizations. This paper describes briefly the standards and the criteria applied to the restoration tasks at various sites of the uranium mining and milling facilities in Spain. The restoration activities have different characteristics La Haba facility is an isolated and conventional facility to produce uranium concentrate; in the case of old and abandoned uranium mines the intervention criteria is more relevant than the activities to be carried out; the closure (the first phase of licensing) and restoration activities of Elefante plant have to be developed taking into account that it is sited within the area of Quercus plant which is currently in operation. (author)

  16. Risk Analysis for the Radioactive Waste Management Facility

    Method of PSA has been applied to nuclear reactor for power reactor or research reactor. As IAEA recommendation, PSA could be used on non-reactor nuclear facility. In this paper, PSA method has been applied for the radioactive waste management facility. Purpose of this method is to determine the risk that is combination of probability and consequence. In these cases, discharge of radioactive material and chemical substance and overexposure are as consequence. Analysis is carried out by two stages, firstly it determines initiating event and secondly, it makes accident sequence modeling. Analysis has been done for 5 group of initiating events. Initiating event frequency is adopted from facility condition and NUREG data. As component reliability data is used from data of IAEA-TECDOC-478 and NUREG. Result of analysis, probability of consequence is about 10-10 per year to 10-5 per year. The radioactive waste management facility is safe enough because probability of consequence is very small

  17. Nuclear fuel cycle facility accident analysis handbook

    The purpose of this Handbook is to provide guidance on how to calculate the characteristics of releases of radioactive materials and/or hazardous chemicals from nonreactor nuclear facilities. In addition, the Handbook provides guidance on how to calculate the consequences of those releases. There are four major chapters: Hazard Evaluation and Scenario Development; Source Term Determination; Transport Within Containment/Confinement; and Atmospheric Dispersion and Consequences Modeling. These chapters are supported by Appendices, including: a summary of chemical and nuclear information that contains descriptions of various fuel cycle facilities; details on how to calculate the characteristics of source terms for releases of hazardous chemicals; a comparison of NRC, EPA, and OSHA programs that address chemical safety; a summary of the performance of HEPA and other filters; and a discussion of uncertainties. Several sample problems are presented: a free-fall spill of powder, an explosion with radioactive release; a fire with radioactive release; filter failure; hydrogen fluoride release from a tankcar; a uranium hexafluoride cylinder rupture; a liquid spill in a vitrification plant; and a criticality incident. Finally, this Handbook includes a computer model, LPF No.1B, that is intended for use in calculating Leak Path Factors. A list of contributors to the Handbook is presented in Chapter 6. 39 figs., 35 tabs

  18. Nuclear fuel cycle facility accident analysis handbook

    NONE

    1998-03-01

    The purpose of this Handbook is to provide guidance on how to calculate the characteristics of releases of radioactive materials and/or hazardous chemicals from nonreactor nuclear facilities. In addition, the Handbook provides guidance on how to calculate the consequences of those releases. There are four major chapters: Hazard Evaluation and Scenario Development; Source Term Determination; Transport Within Containment/Confinement; and Atmospheric Dispersion and Consequences Modeling. These chapters are supported by Appendices, including: a summary of chemical and nuclear information that contains descriptions of various fuel cycle facilities; details on how to calculate the characteristics of source terms for releases of hazardous chemicals; a comparison of NRC, EPA, and OSHA programs that address chemical safety; a summary of the performance of HEPA and other filters; and a discussion of uncertainties. Several sample problems are presented: a free-fall spill of powder, an explosion with radioactive release; a fire with radioactive release; filter failure; hydrogen fluoride release from a tankcar; a uranium hexafluoride cylinder rupture; a liquid spill in a vitrification plant; and a criticality incident. Finally, this Handbook includes a computer model, LPF No.1B, that is intended for use in calculating Leak Path Factors. A list of contributors to the Handbook is presented in Chapter 6. 39 figs., 35 tabs.

  19. Sampling and Analysis Plan for the 233-S Plutonium Concentration Facility

    This Sampling and Analysis Plan (SAP) provides the information and instructions to be used for sampling and analysis activities in the 233-S Plutonium Concentration Facility. The information and instructions herein are separated into three parts and address the Data Quality Objective (DQO) Summary Report, Quality Assurance Project Plan (QAP), and SAP

  20. Safety assessment for facilities and activities. General safety requirements. Pt. 4

    Fundamental Safety Principles. Section 3 describes the graded approach to implementation of the requirements for safety assessment for different facilities and activities. Section 4 establishes the overall requirements for a safety assessment and specific requirements that relate to the assessment of features relevant to safety. Section 4 also establishes the requirements to address defence in depth and safety margins, to perform safety analysis, to document the safety assessment and to carry out an independent verification. Section 5 establishes the requirements for the management, use and maintenance of the safety assessment

  1. Safety Assessment for Facilities and Activities. General Safety Requirements. Pt. 4

    from the Fundamental Safety Principles. Section 3 describes the graded approach to implementation of the requirements for safety assessment for different facilities and activities. Section 4 establishes the overall requirements for a safety assessment and specific requirements that relate to the assessment of features relevant to safety. Section 4 also establishes the requirements to address defence in depth and safety margins, to perform safety analysis, to document the safety assessment and to carry out an independent verification. Section 5 establishes the requirements for the management, use and maintenance of the safety assessment

  2. CHANGE OF CONTRACTOR FOR THE FACILITIES MANAGEMENT ACTIVITIES AT CERN

    2003-01-01

    The Facilities Management contract at CERN, under the responsibility of ST Division, Group FM, is in charge of the maintenance and minor works on tertiary installations (i.e. all structures and installations that have no direct relation to the running of the accelerators) for the following trades: - Technical: heating, ventilation, air conditioning, plumbing, electricity, civil engineering (painting, roofing, glazing, blinds, fencing, masonry etc.), cleansing, passenger and goods lifts, automatic and powered doors, kitchen equipment, roads, signs, keys and locks, office furniture, - Services: waste collection, security, green areas, cleaning and sanitary supplies, disinfection, rodent control and insect control. Starting from the 1st June the present contractor will stop some activities that will be taken under its responsibility by the new one, INGEST Facility. The remaining activities (including cleaning) will be moved on the 1st July. Minor perturbation in the service might occur. The contact number will ...

  3. CHANGE OF CONTRACTOR FOR THE FACILITIES MANAGEMENT ACTIVITIES AT CERN

    2003-01-01

    The Facilities Management contract at CERN, under the responsibility of ST Division, Group FM, is in charge of the maintenance and minor works on tertiary installations (i.e. all structures and installations that have no direct relation to the running of the accelerators) for the following trades: - Technical: heating, ventilation, air conditioning, plumbing, electricity, civil engineering (painting, roofing, glazing, blinds, fencing, masonry etc.), cleansing, passenger and goods lifts, automatic and powered doors, kitchen equipment, roads, signs, keys and locks, office furniture, - Services: waste collection, security, green areas, cleaning and sanitary supplies, disinfection, rodent control and insect control. Starting from the 1st June the present contractor will stop some activities that will be taken under its responsibility by the new one, INGEST Facility. Others activities will be moved on the 1st July. Minor perturbation in the service might occur. The contact number will not change and will be opera...

  4. Safety Assessment for Facilities and Activities. General Safety Requirements

    This publication describes the generally applicable requirements to be fulfilled in safety assessments for facilities and activities, with special attention paid to defence in depth, quantitative analyses and the application of a graded approach to the ranges of facilities and of activities that are addressed. The requirements provide a consistent and coherent basis for safety assessments, facilitating the transfer of good practices between organizations. A review of Safety Requirements publications was commenced in 2011 following the accident in the Fukushima Daiichi nuclear power plant in Japan. The review revealed no significant areas of weakness and resulted in just a small set of amendments to strengthen the requirements and facilitate their implementation, which are contained in the present publication

  5. Activation analysis in Greece

    A review of research and development on NAA as well as examples of applications of this method are presented, taken from work carried out over the last 21 years at the Radioanalytical Laboratory of the Department of Chemistry in the Greek Nuclear Research Center ''Demokritos''. Improved and faster radiochemical NAA methods have been developed for the determination of Au, Ni, Cl, As, Cu, U, Cr, Eu, Hg and Mo in several materials, for the simultaneous determination of Br and I; Mg, Sr and Ni; As and Cu; As, Sb and Hg; Mn, Sr and Ba; Cd and Zn; Se and As; Mo and Cr in biological materials. Instrumental NAA methods have also been developed for the determination of Ag, Cl and Na in lake waters, Al, Ca, Mg and V in wines, 7 trace elements in biological materials, 17 trace elements in sediments and 20 minor and trace elements in ceramics. A comprehensive computer program for routine activation analysis using Ge(Li) detectors have been worked out. A rather extended charged-particle activation analysis program is carried out for the last 10 years, including particle induced X-ray emission (PIXE) analysis, particle induced prompt gamma-ray emission analysis (PIGE), other nuclear reactions and proton activation analysis. A special neutron activation method, the delayed fission neutron counting method is used for the analysis of fissionable elements, as U, Th, Pu, in samples of the whole nuclear fuel cycle including geological, enriched and nuclear safeguards samples

  6. Radioisotope Power System Facility shielding analysis

    A series of calculations for the Radioisotope Power System Facility have been performed. These analyses have determined the shielding required for storage, testing, and transport of 238Pu heat source modules using the Monte Carlo code MCNP3B. The source terms and the assumptions used have been verified by comparison of calculated dose rates with measured ones. This paper describes the methodology used for shielding designs and the utilization of available variance reduction techniques to improve the computational efficiency. The new version of MCNP (MCNP3B) with a repeated structure capability was used. It decreased the chance for computer model errors and greatly decreased the model setup time. 2 refs., 3 figs., 2 tabs

  7. Seismic analysis of the MFTF facility

    Seismic analyses were performed on the Mirror Fusion Test Facility (MFTF-B) located at the Lawrence Livermore National Laboratory, Livermore, CA. The three major structures studied were the vacuum vessel, the concrete shielding vault, and the steel frame enclosure building. The analyses performed on these structures ranged from fixed-base response spectrum analyses to soil-structure interaction analyses including the effects of structure-to-structure interaction and foundation flexibility. The results of these studies showed that the presence of the vault significantly affects the response of the vessel; that modeling the flexibility of the vault footing is important when studying forces near the base of the wall; and that the vault had very little effect on the building response. (orig.)

  8. Facility management and energy efficiency -- analysis and recommendations; Facility Management und Energieeffizienz: Analyse und Handlungsempfehlungen

    Staub, P.; Weibel, K.; Zaugg, T. [Pom and Consulting Ltd., Zuerich (Switzerland); Lang, R. [Gruenberg and Partner Ltd., Zuerich (Switzerland); Frei, Ch. [Herzog Kull Group, Aarau (Switzerland)

    2001-07-01

    This final report presents the results of a study made on how facility management (FM) is positioned in enterprises and on how energy management can be integrated into the facility management process. Also, recommendations are made on the actions that are considered necessary to improve the understanding of facility management and energy management. The findings of an analysis made of the results of a survey among 200 enterprises, 20 interviews and 5 case studies are presented. The authors state that, in spite of the relatively small sample taken - mostly larger enterprises - trends in facility management and energy management could be shown. The findings of the survey, such as the relative importance of the integration of energy topics in facility management and the need for standardised indicators and benchmarking, are discussed in detail. Also, it is noted that the success of FM is in part due to delegation of responsibility to smaller business units or even to individual employees. The market potential for FM services is examined, with yearly growth rates of up to 20%. The importance of anchoring FM strategies at the top level of management is stressed, as is the need for promotion of the idea of facility management and training concepts for those responsible for its implementation.

  9. Fault tree analysis for red oil explosion in reprocessing facility

    Almost all spent fuel reprocessing facilities have adopted Purex process. The red oil explosion is a great concern in safety study of spent fuel reprocessing facilities adopting Purex process. The event tree and fault tree analysis was performed for the red oil explosion of a medium level radioactive waste liquid evaporator for the collective decontamination and separation cycle segment in a representative reprocessing facility in this paper. The results show that the occurrence frequency of a red oil explosion is extremely low, and human errors and common cause failures are major causes to a red oil explosion. Therefore, some relevant measures should be taken to prevent such accidents. (authors)

  10. Accident analysis for aircraft crash into hazardous facilities: DOE standard

    This standard provides the user with sufficient information to evaluate and assess the significance of aircraft crash risk on facility safety without expending excessive effort where it is not required. It establishes an approach for performing a conservative analysis of the risk posed by a release of hazardous radioactive or chemical material resulting from an aircraft crash into a facility containing significant quantities of such material. This can establish whether a facility has a significant potential for an aircraft impact and whether this has the potential for producing significant offsite or onsite consequences. General implementation guidance, screening and evaluation guidelines, and methodologies for the evaluations are included

  11. Spatio-temporal Facility Utilization Analysis from Exhaustive WiFi Monitoring

    Prentow, Thor Siiger; Ruiz-Ruiz, Antonio; Blunck, Henrik;

    2015-01-01

    realistic data to inform facility planning. In this paper, we propose analysis methods to extract knowledge from large sets of network collected WiFi traces to better inform facility management and planning in large building complexes. The analysis methods, which build on a rich set of temporal and spatial...... to inform facility-planning activities. To evaluate the proposed methods and visualization tools, we present facility utilization analysis results for a large hospital complex covering more than 10 hectares. The evaluation is based on WiFi traces collected in the hospital’s WiFi infrastructure over two...... weeks observing around 18000 different devices recording more than a billion individual WiFi measurements. We highlight the tools’ ability to deduce people’s presences and movements and how they can provide respective insights into the test-bed hospital by investigating utilization patterns globally...

  12. ESF [Exploratory Shaft Facility] flexibility analysis

    This report directs that uncertainty allowances be included within the ESF facilities. The recommendations herein developed are intended as input to Title II Design criteria. Flexibility is measured first by lineal ft of drift, and then by hoisting rate and capacity of supporting utilities and services. A defined probability of need shows an extra 10,000 ft of drift for the first level of flexibility responding to testing and operations, and over 60,000 ft of drift for the second level of flexibility which recognizes possible need for perimeter drifting to investigate geologic stratigraphy. Observing there will be time constraints, a single shaft muck hoisting rate up to 170 to 250 tons per hour is recommended. The potential hoisting rate recommended for flexibility should be satisfied by a hoist approximately equivalent to, or conveniently upgraded from those being considered for sinking and construction, or 1000 horsepower. The cost of flexibility is limited to engineering planning and design (mostly conceptual) which makes later expansion achievable, and to selected items for initial construction where later upgrading would be impractical, impossible, or very costly. The cost is fixed to the level of flexibility and does not vary with excavated footage. The incremental margin is only a small fraction of the additional footage made available. Flexibility presents a strategy and not a position of design or technology. Examples used in this report are intended to be illustrative only, and not to lead design or cost estimates. 7 tabs

  13. SLOWPOKE: neutron activation analysis

    Neutron activation analysis permits the non-destructive determination of trace elements in crude oil and its derivatives at high sensitivity (up to 10-9 g/g) and good precision. This article consists of a quick survey of the method followed by an illustration based on the results of recent work at the SLOWPOKE reactor laboratory at the Ecole Polytechnique

  14. Detailed measurements and modelling of thermo active components using a room size test facility

    Weitzmann, Peter; Svendsen, Svend

    This paper describes an investigation of thermo active components based on prefabricated hollow core concrete decks. Recent years have given an increased awareness of the use of thermo active components as an alternative to mechanical cooling systems in office buildings. The investigation covers...... measurements in an office sized test facility with thermo active ceiling and floor as well as modelling of similar conditions in a computer program designed for analysis of building integrated heating and cooling systems. A method for characterizing the cooling capacity of thermo active components is described...

  15. Analysis of personnel error occurrence reports across Defense Program facilities

    Stock, D.A.; Shurberg, D.A.; O`Brien, J.N.

    1994-05-01

    More than 2,000 reports from the Occurrence Reporting and Processing System (ORPS) database were examined in order to identify weaknesses in the implementation of the guidance for the Conduct of Operations (DOE Order 5480.19) at Defense Program (DP) facilities. The analysis revealed recurrent problems involving procedures, training of employees, the occurrence of accidents, planning and scheduling of daily operations, and communications. Changes to DOE 5480.19 and modifications of the Occurrence Reporting and Processing System are recommended to reduce the frequency of these problems. The primary tool used in this analysis was a coding scheme based on the guidelines in 5480.19, which was used to classify the textual content of occurrence reports. The occurrence reports selected for analysis came from across all DP facilities, and listed personnel error as a cause of the event. A number of additional reports, specifically from the Plutonium Processing and Handling Facility (TA55), and the Chemistry and Metallurgy Research Facility (CMR), at Los Alamos National Laboratory, were analyzed separately as a case study. In total, 2070 occurrence reports were examined for this analysis. A number of core issues were consistently found in all analyses conducted, and all subsets of data examined. When individual DP sites were analyzed, including some sites which have since been transferred, only minor variations were found in the importance of these core issues. The same issues also appeared in different time periods, in different types of reports, and at the two Los Alamos facilities selected for the case study.

  16. Safety analysis of the existing 851 Firing Facility

    A safety analysis was performed to determine if normal operations and/or potential accidents at the 851 Firing Facility at Site 300 could present undue hazards to the general public, personnel at Site 300, or have an adverse effect on the environment. The normal operations and credible accidents that might have an effect on these facilities or have off-site consequences were considered. It was determined by this analysis that all but two of the hazards were either low or of the type or magnitude routinely encountered and/or accepted by the public. The exceptions were the linear accelerator and explosives, which were classified as moderate hazards per the requirements given in DOE Order 5481.1A. This safety analysis concluded that the operation at this facility will present no undue risk to the health and safety of LLNL employees or the public

  17. Safety analysis of the existing 850 Firing Facility

    A safety analysis was performed to determine if normal operations and/or potential accidents at the 850 Firing Facility at Site 300 could present undue hazards to the general public, personnel at Site 300, or have an adverse effect on the environment. The normal operations and credible accidents that might have an effect on these facilities or have off-site consequences were considered. It was determined by this analysis that all but one of the hazards were either low or of the type or magnitude routinely encountered and/or accepted by the public. The exception was explosives, which was classified as a moderate hazard per the requirements given in DOE Order 5481.1A. This safety analysis concluded that the operation at this facility will present no undue risk to the health and safety of LLNL employees or the public

  18. Analysis of the zone approach for plutonium facilities

    In order to examine the effect of different inspection strategies on inspection effort, an analysis was carried out of the zone approach for the international safeguards verifications of a model nuclear fuel cycle. The fuel cycle includes the fabrication of mixed-oxide fresh fuel for nine light-water reactors and one experimental breeder reactor and the subsequent reprocessing of the spent fuel. There are thus two zones to be considered, a plutonium zone and an irradiated fuel zone. The zone approach entails many fewer verifications of nuclear material flows between different material balance areas (facilities) than the facility-oriented approach, and it requires an annual simultaneous physical inventory verification (PIV) and monthly simultaneous interim inventory verifications for timeliness at all the facilities. Therefore, the zone approach yields snapshots of the disposition of the nuclear materials at the time of the simultaneous inventory verifications, but less verified information than a facility-oriented approach encompassing frequent flow verification

  19. Business administration of PET facilities. A cost analysis of three facilities utilizing delivery FDG

    PET (positron emission tomography) has been proved to be a powerful imaging tool in clinical oncology. The number of PET facilities in Japan has remarkably increased over the last decade. Furthermore, the approval of delivery fluorodeoxyglucose (FDG) in 2005 resulted in a tremendous expansion of the PET institutions without a cyclotron facility. The aim of this study was to conduct a cost analysis of PET institutions that utilized delivery FDG. Three PET facilities using delivery FDG were investigated about the costs for PET service. Fixed costs included depreciation costs for construction and medical equipments such as positron camera. Variable costs consisted of costs for medical materials including delivery FDG. The break-even point was analyzed in each of three institutions. In the three hospitals (A, B and C), the annual number of PET scan was 1,591, 1,637 and 914, while cost per scan was accounted as 110,262 yen, 111,091 yen, and 134,192 yen, respectively. The break-even point was calculated to be 2,583, 2,679 and 2,081, respectively. PET facilities utilizing delivery FDG seemed to have difficulty in business administration. Such a situation suggests the possibility that the current supply of PET facilities might exceed actual demand for the service. The efficiency of resource allocation should be taken into consideration in the future health service researches on PET. (author)

  20. Facile synthesis and antibacterial activity of naturally occurring 5-methoxyfuroflavone.

    Alam, Mohammad Sayed; Lee, Dong-Ung

    2010-12-01

    A convenient synthesis of 5-methoxyfuroflavone (6, pongaglabol methyl ether), a constituent of some Pongamia or Millettia genus, was achieved by starting from 2,4-dihydroxy-6-methoxyacetophenone via a chalcone precursor, followed by treatment with 2,3-dichloro-5,6-dicyano-1,4-benzoquinone (DDQ). This five-step reaction (total yield: 21.6%) is more facile with that of previously utilized procedures using each different starting material. Antibacterial activities of the above compound and its precursor chalcones, which also belongs to the class of furoflavonoids, were tested by the disc diffusion method against Shigella dysenteriae, Salmonella typhi, Streptococcus-β-haemolyticus, and Staphylococcus aureus. 5-Methoxyfuroflavone showed moderate bactericidal activity against all tested bacterial strains, whereas its corresponding chalcone compound revealed a selective activity. PMID:21139271

  1. Final safety analysis report for the irradiated fuels storage facility

    A fuel storage facility has been constructed at the Idaho Chemical Processing Plant to provide safe storage for spent fuel from two commercial HTGR's, Fort St. Vrain and Peach Bottom, and from the Rover nuclear rocket program. The new facility was built as an addition to the existing fuel storage basin building to make maximum use of existing facilities and equipment. The completed facility provides dry storage for one core of Peach Bottom fuel (804 elements), 11/2 cores of Fort St. Vrain fuel (2200 elements), and the irradiated fuel from the 20 reactors in the Rover program. The facility is designed to permit future expansion at a minimum cost should additional storage space for graphite-type fuels be required. A thorough study of the potential hazards associated with the Irradiated Fuels Storage Facility has been completed, indicating that the facility is capable of withstanding all credible combinations of internal accidents and pertinent natural forces, including design basis natural phenomena of a 10,000 year flood, a 175-mph tornado, or an earthquake having a bedrock acceleration of 0.33 g and an amplification factor of 1.3, without a loss of integrity or a significant release of radioactive materials. The design basis accident (DBA) postulated for the facility is a complete loss of cooling air, even though the occurrence of this situation is extremely remote, considering the availability of backup and spare fans and emergency power. The occurrence of the DBA presents neither a radiation nor an activity release hazard. A loss of coolant has no effect upon the fuel or the facility other than resulting in a gradual and constant temperature increase of the stored fuel. The temperature increase is gradual enough that ample time (28 hours minimum) is available for corrective action before an arbitrarily imposed maximum fuel centerline temperature of 11000F is reached

  2. Insider threat to secure facilities: data analysis

    Three data sets drawn from industries that have experienced internal security breaches are analyzed. The industries and the insider security breaches are considered analogous in one or more respects to insider threats potentially confronting managers in the nuclear industry. The three data sets are: bank fraud and embezzlement (BF and E), computer-related crime, and drug theft from drug manufacturers and distributors. A careful analysis by both descriptive and formal statistical techniques permits certain general conclusions on the internal threat to secure industries to be drawn. These conclusions are discussed and related to the potential insider threat in the nuclear industry. 49 tabs

  3. Forensic Activation Analysis

    The high sensitivity of high-flux (reactor) thermal-neutron activation analysis (NAA) for the detection and quantitative measurement of a large number of elements has led, in recent years, to a considerable degree of application of the method in the area of scientific crime investigation (criminalistics). Thus, in a Forensic Activation Analysis Bibliography recently compiled by the author, some 135 publications in this field are listed - and more are appearing quite rapidly. The nondestructive character of the purely-instrumental form of the method is an added advantage in forensic work, since evidence samples involved in actual criminal cases are not destroyed during analysis, but are preserved intact for possible presentation in court. Quite aside from, or in addition to, use in court, NAA results can be very helpful in the investigative stage of particular criminal cases. The ultra sensitivity of the method often enables one to analyze evidence specimens that are too tiny for meaningful analysis by more conventional elemental analysis methods. Also, this high sensitivity often enables one to characterize, or individualize, evidence specimens as to the possibility of common origin - via the principle of multi-element trace-constituent characterization

  4. The Management System for Facilities and Activities. Safety Requirements

    This publication establishes requirements for management systems that integrate safety, health, security, quality assurance and environmental objectives. A successful management system ensures that nuclear safety matters are not dealt with in isolation but are considered within the context of all these objectives. The aim of this publication is to assist Member States to establish and implement effective management systems that integrate all aspects of managing nuclear facilities and activities in a coherent manner. It details the planned and systematic actions necessary to provide adequate confidence that all these requirements are satisfied

  5. The Management System for Facilities and Activities. Safety Requirements

    This publication establishes requirements for management systems that integrate safety, health, security, quality assurance and environmental objectives. A successful management system ensures that nuclear safety matters are not dealt with in isolation but are considered within the context of all these objectives. The aim of this publication is to assist Member States in establishing and implementing effective management systems that integrate all aspects of managing nuclear facilities and activities in a coherent manner. It details the planned and systematic actions necessary to provide adequate confidence that all these requirements are satisfied. Contents: 1. Introduction; 2. Management system; 3. Management responsibility; 4. Resource management; 5. Process implementation; 6. Measurement, assessment and improvement.

  6. Insider threat to secure facilities: data analysis

    1979-12-07

    This report is the culmination of a project in which data from several industries confronting internal security threats were collected and analyzed. The industries and threats involved are deemed to be analogous in one or more respects to potential threats confronting decision makers in the nuclear industry. The analog internal threats consist of bank frauds and embezzlements over $10,000, computer crimes of various types and insider drug thefts from drug manufactures and distributors. These data have been subjected to careful analysis utilizing both descriptive and formal statistical techniques. A number of findings are quite suggestive as to the general nature of the internal threat and are discussed and interpreted in terms of thenuclear industry analogy.

  7. Documented Safety Analysis for the Waste Storage Facilities March 2010

    Laycak, D T

    2010-03-05

    This Documented Safety Analysis (DSA) for the Waste Storage Facilities was developed in accordance with 10 CFR 830, Subpart B, 'Safety Basis Requirements,' and utilizes the methodology outlined in DOE-STD-3009-94, Change Notice 3. The Waste Storage Facilities consist of Area 625 (A625) and the Decontamination and Waste Treatment Facility (DWTF) Storage Area portion of the DWTF complex. These two areas are combined into a single DSA, as their functions as storage for radioactive and hazardous waste are essentially identical. The B695 Segment of DWTF is addressed under a separate DSA. This DSA provides a description of the Waste Storage Facilities and the operations conducted therein; identification of hazards; analyses of the hazards, including inventories, bounding releases, consequences, and conclusions; and programmatic elements that describe the current capacity for safe operations. The mission of the Waste Storage Facilities is to safely handle, store, and treat hazardous waste, transuranic (TRU) waste, low-level waste (LLW), mixed waste, combined waste, nonhazardous industrial waste, and conditionally accepted waste generated at LLNL (as well as small amounts from other DOE facilities).

  8. Documented Safety Analysis for the Waste Storage Facilities

    Laycak, D

    2008-06-16

    This documented safety analysis (DSA) for the Waste Storage Facilities was developed in accordance with 10 CFR 830, Subpart B, 'Safety Basis Requirements', and utilizes the methodology outlined in DOE-STD-3009-94, Change Notice 3. The Waste Storage Facilities consist of Area 625 (A625) and the Decontamination and Waste Treatment Facility (DWTF) Storage Area portion of the DWTF complex. These two areas are combined into a single DSA, as their functions as storage for radioactive and hazardous waste are essentially identical. The B695 Segment of DWTF is addressed under a separate DSA. This DSA provides a description of the Waste Storage Facilities and the operations conducted therein; identification of hazards; analyses of the hazards, including inventories, bounding releases, consequences, and conclusions; and programmatic elements that describe the current capacity for safe operations. The mission of the Waste Storage Facilities is to safely handle, store, and treat hazardous waste, transuranic (TRU) waste, low-level waste (LLW), mixed waste, combined waste, nonhazardous industrial waste, and conditionally accepted waste generated at LLNL (as well as small amounts from other DOE facilities).

  9. Modeling and Analysis of Facility Systems for A Hybrid Materials Test Program

    Congiardo, Jared F.

    2007-01-01

    Analytic modeling and analysis processes employed at NASA-SSC in rocket propulsion systems testing are discussed in this paper with application to test facility propellant supply system design, activation and test of a hybrid rocket motor provided. This paper discusses the analytic model employed, its utilization across project phases and reviews performance results.

  10. A systems analysis approach to nuclear facility siting

    An attempt is made to demonstrate an application of the techniques of systems analysis, which have been successful in solving a variety of problems, to nuclear facility siting. Within the framework of an overall regional land-use plan, a methodology for establishing the acceptability of a combination of site and facility is discussed. The consequences (e.g. the energy produced, thermal and chemical discharges, radioactive releases, aeshetic values, etc.) of the site-facility combination are identified and compared with formalized criteria in order to ensure 'legal acceptability'. Failure of any consequences to satisfy standard requirements results in a feedback channel which works to effect design changes in the facility. When 'legal acceptability' has been assured, the project enters the public sector for consideration. The responses of individuals and of various interested groups to the external attributes of the nuclear facility gradually emerge. The criteria by which interest groups judge technological advances reflect both their rational assessment and unconscious motivations. This process operates on individual, group, societal and international levels and may result in two basic feedback loops: one which might act to change regulatory criteria; the other which might influence facility design or site selection. Such reactions and responses on these levels result in a continuing process of confrontation, collaborative interchange and possible resolution in the direction of an acceptable solution. Finally, a Paretian approach to optimizing the site-facility combination is presented for the case where there are several possible combinations of site and facility. A hypothetical example of the latter is given, based upon typical preference functions determined for four interest groups. The research effort of the IIASA Energy Systems Project and the Joint IAEA/IIASA Research Project in the area of nuclear siting is summarized. (author)

  11. Cold Vacuum Drying facility design basis accident analysis documentation

    CROWE, R.D.

    2000-08-08

    This document provides the detailed accident analysis to support HNF-3553, Annex B, Spent Nuclear Fuel Project Final Safety Analysis Report (FSAR), ''Cold Vacuum Drying Facility Final Safety Analysis Report.'' All assumptions, parameters, and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the FSAR. The calculations in this document address the design basis accidents (DBAs) selected for analysis in HNF-3553, ''Spent Nuclear Fuel Project Final Safety Analysis Report'', Annex B, ''Cold Vacuum Drying Facility Final Safety Analysis Report.'' The objective is to determine the quantity of radioactive particulate available for release at any point during processing at the Cold Vacuum Drying Facility (CVDF) and to use that quantity to determine the amount of radioactive material released during the DBAs. The radioactive material released is used to determine dose consequences to receptors at four locations, and the dose consequences are compared with the appropriate evaluation guidelines and release limits to ascertain the need for preventive and mitigative controls.

  12. Cold Vacuum Drying facility design basis accident analysis documentation

    This document provides the detailed accident analysis to support HNF-3553, Annex B, Spent Nuclear Fuel Project Final Safety Analysis Report (FSAR), ''Cold Vacuum Drying Facility Final Safety Analysis Report.'' All assumptions, parameters, and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the FSAR. The calculations in this document address the design basis accidents (DBAs) selected for analysis in HNF-3553, ''Spent Nuclear Fuel Project Final Safety Analysis Report'', Annex B, ''Cold Vacuum Drying Facility Final Safety Analysis Report.'' The objective is to determine the quantity of radioactive particulate available for release at any point during processing at the Cold Vacuum Drying Facility (CVDF) and to use that quantity to determine the amount of radioactive material released during the DBAs. The radioactive material released is used to determine dose consequences to receptors at four locations, and the dose consequences are compared with the appropriate evaluation guidelines and release limits to ascertain the need for preventive and mitigative controls

  13. The Remote Handled Immobilization Low Activity Waste Disposal Facility Environmental Permits & Approval Plan

    DEFFENBAUGH, M.L.

    2000-08-01

    The purpose of this document is to revise Document HNF-SD-ENV-EE-003, ''Permitting Plan for the Immobilized Low-Activity Waste Project, which was submitted on September 4, 1997. That plan accounted for the interim storage and disposal of Immobilized-Low Activity Waste at the existing Grout Treatment Facility Vaults (Project W-465) and within a newly constructed facility (Project W-520). Project W-520 was to have contained a combination of concrete vaults and trenches. This document supersedes that plan because of two subsequent items: (1) A disposal authorization that was received on October 25, 1999, in a U. S. Department of Energy-Headquarters, memorandum, ''Disposal Authorization Statement for the Department of Energy Hanford site Low-Level Waste Disposal facilities'' and (2) ''Breakthrough Initiative Immobilized Low-Activity Waste (ILAW) Disposal Alternative,'' August 1999, from Lucas Incorporated, Richland, Washington. The direction within the U. S. Department of Energy-Headquarters memorandum was given as follows: ''The DOE Radioactive Waste Management Order requires that a Disposal authorization statement be obtained prior to construction of new low-level waste disposal facility. Field elements with the existing low-level waste disposal facilities shall obtain a disposal authorization statement in accordance with the schedule in the complex-wide Low-Level Waste Management Program Plan. The disposal authorization statement shall be issued based on a review of the facility's performance assessment and composite analysis or appropriate CERCLA documentation. The disposal authorization shall specify the limits and conditions on construction, design, operations, and closure of the low-level waste facility based on these reviews. A disposal authorization statement is a part of the required radioactive waste management basis for a disposal facility. Failure to obtain a disposal authorization statement

  14. Fire Hazard Analysis for the Cold Vacuum Drying (CVD) Facility

    This Fire Hazard Analysis assesses the risk from fire within individual fire areas in the Cold Vacuum Drying Facility at the Hanford Site in relation to existing or proposed fire protection features to ascertain whether the objectives of DOE Order 5480.7A Fire Protection are met

  15. Fire Hazard Analysis for the Cold Vacuum Drying (CVD) Facility

    JOHNSON, B.H.

    1999-08-19

    This Fire Hazard Analysis assesses the risk from fire within individual fire areas in the Cold Vacuum Drying Facility at the Hanford Site in relation to existing or proposed fire protection features to ascertain whether the objectives of DOE Order 5480.7A Fire Protection are met.

  16. Neutron activations at the neutron facility of TU-Dresden

    Domula, Alexander; Zuber, Kai [TU Dresden, Institut fuer Kern- und Teilchenphysik, 01069 Dresden (Germany); Gehre, Daniel [TU Dresden, Institut fuer Kern- und Teilchenphysik, 01069 Dresden (Germany); FZD, Institut fuer Strahlenphysik, 01314 Dresden (Germany); Klix, Axel [KIT, Institut fuer Neutronenphysik und Reaktortechnik, 76344 Eggenstein-Leopoldshafen (Germany)

    2010-07-01

    The Technical University of Dresden (TUD) operates at the Forschungszentrum Dresden-Rossendorf (FZD) a 14 MeV Neutron Generator (NG) with fast, mono energetic neutrons from the T(d,{alpha})n reaction and 2.5 MeV neutrons from the D(d,x)n reaction. Since its commissioning in 2004 the NG is involved in the validation of European Activation File and mockup experiments for validation of neutron transport data in collaborations with FZK/KIT, PTB, ENEA, JAEA, Osaka University and University Vienna. Cross section measurements have been limited to long living isotopes. An automated sample changer is currently set up in order to extend the capabilities to radioisotopes with half-lives in the range from seconds to a few minutes. The general layout of the neutron facility is described. First example activations for GERDA and SNO+ have been made and are presented here.

  17. Wageningen UR Unmanned Aerial Remote Sensing Facility - Overview of activities

    Bartholomeus, Harm; Keesstra, Saskia; Kooistra, Lammert; Suomalainen, Juha; Mucher, Sander; Kramer, Henk; Franke, Jappe

    2016-04-01

    To support environmental management there is an increasing need for timely, accurate and detailed information on our land. Unmanned Aerial Systems (UAS) are increasingly used to monitor agricultural crop development, habitat quality or urban heat efficiency. An important reason is that UAS technology is maturing quickly while the flexible capabilities of UAS fill a gap between satellite based and ground based geo-sensing systems. In 2012, different groups within Wageningen University and Research Centre have established an Unmanned Airborne Remote Sensing Facility. The objective of this facility is threefold: a) To develop innovation in the field of remote sensing science by providing a platform for dedicated and high-quality experiments; b) To support high quality UAS services by providing calibration facilities and disseminating processing procedures to the UAS user community; and c) To promote and test the use of UAS in a broad range of application fields like habitat monitoring, precision agriculture and land degradation assessment. The facility is hosted by the Laboratory of Geo-Information Science and Remote Sensing (GRS) and the Department of Soil Physics and Land Management (SLM) of Wageningen University together with the team Earth Informatics (EI) of Alterra. The added value of the Unmanned Aerial Remote Sensing Facility is that compared to for example satellite based remote sensing more dedicated science experiments can be prepared. This includes for example higher frequent observations in time (e.g., diurnal observations), observations of an object under different observation angles for characterization of BRDF and flexibility in use of camera's and sensors types. In this way, laboratory type of set ups can be tested in a field situation and effects of up-scaling can be tested. In the last years we developed and implemented different camera systems (e.g. a hyperspectral pushbroom system, and multispectral frame cameras) which we operated in projects all

  18. Comprehensive safety analysis for pressure and cryogenic systems facilities

    There have been many instances where serious injuries and fatalities have resulted from over-pressurization, thermal stress, asphyxiation and other potential hazards associated with testing, handling and storage of compressed gases and cryogenic liquids at numerous production and research facilities. These hazards are major issues that should be addressed in system design and in materials selection appropriate for high pressure or cryogenic temperature applications. Potential hazards may be mitigated through system analysis and design process which are the major factors in preventing thermal/pressure hazards caused by possible leaks and fragmentation, in the case of rupture. This paper presents a conceptual model and framework for developing a comprehensive safety analysis which will reduce potential hazards, accidents and legal liabilities. The proposed in-depth system Safety Analysis Report (SAR) is a proven systematic approach to identify hazards and influence design to provide timely documentation of potential hazards and risks associated with systems, facilities, and equipment. As a result of this hazard analysis process, provisions and actions for hazard prevention, elimination, mitigation, and control have been put in place, and all identifiable potential hazards have been reduced to a low risk level. These methods are demonstrated in the example of comprehensive safety analysis of Cryogenic Subsystem of Accelerator String Test facilities (ASST) at Superconducting Super Collider Laboratory by developing Safety Analysis Report (SSC Laboratory, 1992)

  19. Fuel Storage Facility Final Safety Analysis Report. Revision 1

    Linderoth, C.E.

    1984-03-01

    The Fuel Storage Facility (FSF) is an integral part of the Fast Flux Test Facility. Its purpose is to provide long-term storage (20-year design life) for spent fuel core elements used to provide the fast flux environment in FFTF, and for test fuel pins, components and subassemblies that have been irradiated in the fast flux environment. This Final Safety Analysis Report (FSAR) and its supporting documentation provides a complete description and safety evaluation of the site, the plant design, operations, and potential accidents.

  20. Study on activation analysis

    High purity aluminum has been analyzed by neutron activation analysis. The determination of copper contents is aluminum has been used to evaluate its purity level. A new sensitive method has been developed by using graphite thermal column to reduce or eliminate the interference of 24Na which is generated from 27Al (n,α) 24Na reaction by fast neutron. Influence for activity of 24Na due to above reaction is found to be between 2.3 - 2.8 %. Copper contents in the high purity aluminum come out 0.542±0.084 ppm. In addition, contents of 23 other impurity elements (<0.1 - 0.01 ppm) are measured using general method after detection limit and optimum conditions are established. (author)

  1. Fault Tree Analysis of an Accident Probability for Pyroprocessing Facility

    The pyroprocessing technology is one of the spent fuel recycling technologies. Korea Atomic Energy Research Institute(KAERI) started the R and D about the pyroprocessing technology in 1997. The physical protection system requirements based on the VAI should be prepared for applying the pyroprocessing facility in Korea. In this study, we have arranged the accidents which can be happened in pyroprocessing facility. Then, we have obtained the accident path according to the hazards. We can expect that this study will be taken to the VAI as a basic data. The fault tree is not complete yet. The fault tree for an accident probability of pyroprocessing facility is being made according to the hot cell area and each process. Conclusions will be handled after finishing the fault tree analysis

  2. Fire Hazard Analysis for the Cold Vacuum Drying facility (CVD) Facility

    Singh, G

    2000-01-01

    The CVDF is a nonreactor nuclear facility that will process the Spent Nuclear Fuels (SNF) presently stored in the 105-KE and 105-KW SNF storage basins. Multi-canister overpacks (MCOs) will be loaded (filled) with K Basin fuel transported to the CVDF. The MCOs will be processed at the CVDF to remove free water from the fuel cells (packages). Following processing at the CVDF, the MCOs will be transported to the CSB for interim storage until a long-term storage solution can be implemented. This operation is expected to start in November 2000. A Fire Hazard Analysis (FHA) is required for all new facilities and all nonreactor nuclear facilities, in accordance with U.S. Department of Energy (DOE) Order 5480.7A, Fire Protection. This FHA has been prepared in accordance with DOE 5480.7A and HNF-PRO-350, Fire Hazard Analysis Requirements. Additionally, requirements or criteria contained in DOE, Richland Operations Office (RL) RL Implementing Directive (RLID) 5480.7, Fire Protection, or other DOE documentation are cite...

  3. Integration of facility modeling capabilities for nuclear nonproliferation analysis

    Developing automated methods for data collection and analysis that can facilitate nuclearnonproliferation assessment is an important research area with significant consequences for the effective global deployment of nuclear energy. Facilitymodeling 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 facilitymodeling 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 facilitymodelingcapabilities and illustrates how they could be integrated and utilized for nonproliferationanalysis. The inverse problem of inferring facility conditions based on collected observations is described, along with a proposed architecture and computer framework for utilizing facilitymodeling tools. After considering a representative sampling of key facilitymodelingcapabilities, the proposed integration framework is illustrated with several examples.

  4. PROMPT DOSE ANALYSIS FOR THE NATIONAL IGNITION FACILITY

    Khater, H; Dauffy, L; Sitaraman, S; Brereton, S

    2008-09-23

    Detailed 3-D modeling of the NIF facility is developed to accurately understand the prompt radiation environment within NIF. Prompt dose values are calculated for different phases of NIF operation. Results of the analysis were used to determine the final thicknesses of the Target Bay (TB) and secondary doors as well as the required shield thicknesses for all unused penetrations. Integrated dose values at different locations within the facility are needed to formulate the personnel access requirements within different parts of the facility. The conclusions of this presentation are: (1) The current NIF facility model includes all important features of the Target Chamber, shielding system, and building configuration; (2) All shielding requirements for Phase I operation are met; (3) Negligible dose values (a fraction of mrem) are expected in normally occupied areas during Phase I; (4) In preparation for the Ignition Campaign and Phase IV of operation, all primary and secondary shield doors will be installed; (5) Unused utility penetrations in the Target Bay and Switchyard walls ({approx}50%) will be shielded by 1 foot thick concrete to reduce prompt dose inside and outside the NIF facility; (6) During Phase IV, a 20 MJ shot will produce acceptable dose levels in the occupied areas as well as at the nearest site boundary; (7) A comprehensive radiation monitoring plan will be put in place to monitor dose values at large number of locations; and (8) Results of the dose monitoring will be used to modify personnel access requirements if needed.

  5. CFB gasification of biomass. An analysis of available and necessary research facilities

    The aim of the title analysis is to inventorize the required and available Dutch laboratory facilities for research on Circulating Fluidized Beds (CFB) gasification of biomass. A literature study has been carried to assess the international state-of-the-art of the technology and research. Based on the results the required research facilities could be determined. Next, interviews were held with researchers at relevant Dutch research institutes and information was collected to compile an overview of available Dutch facilities. It appears that the introduction of CFB gasification technologies can take place under good conditions, although coordination of future research activities is desired, while knowledge and facilities are spread over several research institutes. 16 figs., 43 refs., 1 appendix

  6. The economic impacts of noxious facilities on wages and property values: An exploratory analysis

    Nieves, L.A.; Hemphill, R.C.; Clark, D.E.

    1991-05-01

    Recent assessments of socioeconomic impacts resulting from the location of potentially hazardous facilities have concentrated on the issue of negative public perceptions and their resulting economic consequences. This report presents an analysis designed to answer the question: Can economic impacts resulting from negative perceptions of noxious facilities'' be identified and measured To identify the impacts of negative perceptions, data on noxious facilities sited throughout the United States were compiled, and secondary economic and demographic data sufficient to analyze the economic impacts on the surrounding study areas were assembled. This study uses wage rate and property value differentials to measure impacts on social welfare so that the extent to which noxious facilities and their associated activities have affected surrounding areas can be determined.

  7. The economic impacts of noxious facilities on wages and property values: An exploratory analysis

    Nieves, L.A.; Hemphill, R.C.; Clark, D.E.

    1991-05-01

    Recent assessments of socioeconomic impacts resulting from the location of potentially hazardous facilities have concentrated on the issue of negative public perceptions and their resulting economic consequences. This report presents an analysis designed to answer the question: Can economic impacts resulting from negative perceptions of ``noxious facilities`` be identified and measured? To identify the impacts of negative perceptions, data on noxious facilities sited throughout the United States were compiled, and secondary economic and demographic data sufficient to analyze the economic impacts on the surrounding study areas were assembled. This study uses wage rate and property value differentials to measure impacts on social welfare so that the extent to which noxious facilities and their associated activities have affected surrounding areas can be determined.

  8. Fire Hazard Analysis for the Cold Vacuum Drying facility (CVD) Facility

    The CVDF is a nonreactor nuclear facility that will process the Spent Nuclear Fuels (SNF) presently stored in the 105-KE and 105-KW SNF storage basins. Multi-canister overpacks (MCOs) will be loaded (filled) with K Basin fuel transported to the CVDF. The MCOs will be processed at the CVDF to remove free water from the fuel cells (packages). Following processing at the CVDF, the MCOs will be transported to the CSB for interim storage until a long-term storage solution can be implemented. This operation is expected to start in November 2000. A Fire Hazard Analysis (FHA) is required for all new facilities and all nonreactor nuclear facilities, in accordance with U.S. Department of Energy (DOE) Order 5480.7A, Fire Protection. This FHA has been prepared in accordance with DOE 5480.7A and HNF-PRO-350, Fire Hazard Analysis Requirements. Additionally, requirements or criteria contained in DOE, Richland Operations Office (RL) RL Implementing Directive (RLID) 5480.7, Fire Protection, or other DOE documentation are cited, as applicable. This FHA comprehensively assesses the risk of fire at the CVDF to ascertain whether the specific objectives of DOE 5480.7A are met. These specific fire protection objectives are: (1) Minimize the potential for the occurrence of a fire. (2) Ensure that fire does not cause an onsite or offsite release of radiological and other hazardous material that will threaten the public health and safety or the environment. (3) Establish requirements that will provide an acceptable degree of life safety to DOE and contractor personnel and ensure that there are no undue hazards to the public from fire and its effects in DOE facilities. (4) Ensure that vital DOE programs will not suffer unacceptable delays as a result of fire and related perils. (5) Ensure that property damage from fire and related perils does not exceed an acceptable level. (6) Ensure that process control and safety systems are not damaged by fire or related perils. This FHA is based on the

  9. Fire Hazard Analysis for the Cold Vacuum Drying facility (CVD) Facility

    SINGH, G.

    2000-09-06

    The CVDF is a nonreactor nuclear facility that will process the Spent Nuclear Fuels (SNF) presently stored in the 105-KE and 105-KW SNF storage basins. Multi-canister overpacks (MCOs) will be loaded (filled) with K Basin fuel transported to the CVDF. The MCOs will be processed at the CVDF to remove free water from the fuel cells (packages). Following processing at the CVDF, the MCOs will be transported to the CSB for interim storage until a long-term storage solution can be implemented. This operation is expected to start in November 2000. A Fire Hazard Analysis (FHA) is required for all new facilities and all nonreactor nuclear facilities, in accordance with U.S. Department of Energy (DOE) Order 5480.7A, Fire Protection. This FHA has been prepared in accordance with DOE 5480.7A and HNF-PRO-350, Fire Hazard Analysis Requirements. Additionally, requirements or criteria contained in DOE, Richland Operations Office (RL) RL Implementing Directive (RLID) 5480.7, Fire Protection, or other DOE documentation are cited, as applicable. This FHA comprehensively assesses the risk of fire at the CVDF to ascertain whether the specific objectives of DOE 5480.7A are met. These specific fire protection objectives are: (1) Minimize the potential for the occurrence of a fire. (2) Ensure that fire does not cause an onsite or offsite release of radiological and other hazardous material that will threaten the public health and safety or the environment. (3) Establish requirements that will provide an acceptable degree of life safety to DOE and contractor personnel and ensure that there are no undue hazards to the public from fire and its effects in DOE facilities. (4) Ensure that vital DOE programs will not suffer unacceptable delays as a result of fire and related perils. (5) Ensure that property damage from fire and related perils does not exceed an acceptable level. (6) Ensure that process control and safety systems are not damaged by fire or related perils. This FHA is based on the

  10. Large-coil-test-facility fault-tree analysis

    An operating-safety study is being conducted for the Large Coil Test Facility (LCTF). The purpose of this study is to provide the facility operators and users with added insight into potential problem areas that could affect the safety of personnel or the availability of equipment. This is a preliminary report, on Phase I of that study. A central feature of the study is the incorporation of engineering judgements (by LCTF personnel) into an outside, overall view of the facility. The LCTF was analyzed in terms of 32 subsystems, each of which are subject to failure from any of 15 generic failure initiators. The study identified approximately 40 primary areas of concern which were subjected to a computer analysis as an aid in understanding the complex subsystem interactions that can occur within the facility. The study did not analyze in detail the internal structure of the subsystems at the individual component level. A companion study using traditional fault tree techniques did analyze approximately 20% of the LCTF at the component level. A comparison between these two analysis techniques is included in Section 7

  11. MA-1 unit for instrumental fast neutron activation analysis

    A description of the MA-1 facility intended for performing an instrumental multielement activation analysis is giveperation of facility is based on the spectrometry of samples activated with fast neutrons. The facility comprises the M-6000 control computer complex, the 10N fast neutron generator, the ARS-28G pneumatic transporter, spectrometric devices and detection units. The facility can be used to determine the contents of more than 50 elements in different combinations in compact and powdery substances. Maximum sensitivity is achieved when determining Pr, Cu, Br, Ba, Pb, Sb, Si, P, Al, O and F. The measuring equipment of the facility is universal and after proper modification can be used in activation analysis with the use of nuclear reactors, electron accelerators, a cyclotron and neutron sources

  12. Nuclear Rocket Facility Decommissioning Project: Controlled Explosive Demolition of Neutron-Activated Shield Wall

    Located in Area 25 of the Nevada Test Site (NTS), the Test Cell A (TCA) Facility (Figure 1) was used in the early to mid-1960s for testing of nuclear rocket engines, as part of the Nuclear Rocket Development Program, to further space travel. Nuclear rocket testing resulted in the activation of materials around the reactors and the release of fission products and fuel particles. The TCA facility, known as Corrective Action Unit 115, was decontaminated and decommissioned (D and D) from December 2004 to July 2005 using the Streamlined Approach for Environmental Restoration (SAFER) process, under the Federal Facility Agreement and Consent Order. The SAFER process allows environmental remediation and facility closure activities (i.e., decommissioning) to occur simultaneously, provided technical decisions are made by an experienced decision maker within the site conceptual site model. Facility closure involved a seven-step decommissioning strategy. First, preliminary investigation activities were performed, including review of process knowledge documentation, targeted facility radiological and hazardous material surveys, concrete core drilling and analysis, shield wall radiological characterization, and discrete sampling, which proved to be very useful and cost-effective in subsequent decommissioning planning and execution and worker safety. Second, site setup and mobilization of equipment and personnel were completed. Third, early removal of hazardous materials, including asbestos, lead, cadmium, and oil, was performed ensuring worker safety during more invasive demolition activities. Process piping was to be verified void of contents. Electrical systems were de-energized and other systems were rendered free of residual energy. Fourth, areas of high radiological contamination were decontaminated using multiple methods. Contamination levels varied across the facility. Fixed beta/gamma contamination levels ranged up to 2 million disintegrations per minute (dpm)/100

  13. Nuclear Rocket Facility Decommissioning Project: Controlled Explosive Demolition of Neutron-Activated Shield Wall

    Michael R. Kruzic

    2008-06-01

    Located in Area 25 of the Nevada Test Site (NTS), the Test Cell A (TCA) Facility (Figure 1) was used in the early to mid-1960s for testing of nuclear rocket engines, as part of the Nuclear Rocket Development Program, to further space travel. Nuclear rocket testing resulted in the activation of materials around the reactors and the release of fission products and fuel particles. The TCA facility, known as Corrective Action Unit 115, was decontaminated and decommissioned (D&D) from December 2004 to July 2005 using the Streamlined Approach for Environmental Restoration (SAFER) process, under the Federal Facility Agreement and Consent Order. The SAFER process allows environmental remediation and facility closure activities (i.e., decommissioning) to occur simultaneously, provided technical decisions are made by an experienced decision maker within the site conceptual site model. Facility closure involved a seven-step decommissioning strategy. First, preliminary investigation activities were performed, including review of process knowledge documentation, targeted facility radiological and hazardous material surveys, concrete core drilling and analysis, shield wall radiological characterization, and discrete sampling, which proved to be very useful and cost-effective in subsequent decommissioning planning and execution and worker safety. Second, site setup and mobilization of equipment and personnel were completed. Third, early removal of hazardous materials, including asbestos, lead, cadmium, and oil, was performed ensuring worker safety during more invasive demolition activities. Process piping was to be verified void of contents. Electrical systems were de-energized and other systems were rendered free of residual energy. Fourth, areas of high radiological contamination were decontaminated using multiple methods. Contamination levels varied across the facility. Fixed beta/gamma contamination levels ranged up to 2 million disintegrations per minute (dpm)/100

  14. Safeguardability analysis for an engineering scale pyroprocess facility

    A qualitative safeguardability analysis was undertaken to investigate the safeguards system and draw recommendations for enhancing the performance of the safeguards system for an engineering-scale pyroprocess model facility. The analysis utilized INPRO proliferation resistance (PR) assessment methodologies including diversion pathway analysis. Uranium and transuranic metal (U/TRU) products emit high neutrons and gamma-rays, which are strong enough to be detected by the passive nondestructive assay (NDA) measurements and the hot-cells can role as inherently robust containments. Even though the product materials could be attractive, the abrupt diversion of U/TRU ingots through the selected pathway from the model facility will be reasonably difficult and detectable by applying the appropriate safeguards measures. For the design features to support safeguards implementation of the facility, more effective utilization of the inherent containment and enhancement of portal monitoring as well as focusing on the accounting material flow into and out of the system will make it possible to satisfy the safeguards goal. (author)

  15. The scanning microbeam PIXE analysis facility at NIRS

    In March 1999, a HVEE Tandetron was installed in the Electrostatic Accelerator Building of National Institute of Radiological Sciences (NIRS) for particle induced X-ray emission (PIXE) analysis. The specifications of the Tandetron accelerator system operating at NIRS are as follows: the accelerating voltage is 0.4-1.7 MV, and the maximum beam current is 500 nA at 3.4 MeV. The accelerator facility incorporates three beam lines for conventional, in-air and microbeam PIXE analysis. The scanning microbeam PIXE analysis line is based around an Oxford Microbeams OM2000 nuclear microscope end stage. This system provides the ability of multi-elemental mapping over sample areas up to 2 x 2 mm area with spatial resolutions routinely at 1 x 1 μm. The scheduled operation of this facility started in April 2000 and is controlled by the Division of Technical Service and Development. The result of beam resolution tests carried out in 2001 are as follows: for scanning transmission ion microscopy, the estimated beam size is 100 x 200 nm, measured using a 2.6 MeV proton beam scanned over a 12.7 μm repeat distance copper grid. For PIXE operation at 50 pA beam current the estimated best spot size is 0.4 x 0.6 μm. The microbeam facility is being used for research into the elemental distribution of small biological samples such as biological cells and tissue

  16. Safety analysis of the existing 804 and 845 firing facilities

    A safety analysis was performed to determine if normal operations and/or potential accidents at the 804 and 845 Firing Facilities at Site 300 could present undue hazards to the general public, peronnel at Site 300, or have an adverse effect on the environment. The normal operation and credible accident that might have an effect on these facilities or have off-site consequence were considered. It was determined by this analysis that all but one of the hazards were either low or of the type or magnitude routinely encountered and/or accepted by the public. The exception was explosives. Since this hazard has the potential for causing significant on-site and minimum off-site consequences, Bunkers 804 and 845 have been classified as moderate hazard facilties per DOE Order 5481.1A. This safety analysis concluded that the operation at these facilities will present no undue risk to the health and safety of LLNL employees or the public

  17. Neutron activation diagnostics at the National Ignition Facility (invited).

    Bleuel, D L; Yeamans, C B; Bernstein, L A; Bionta, R M; Caggiano, J A; Casey, D T; Cooper, G W; Drury, O B; Frenje, J A; Hagmann, C A; Hatarik, R; Knauer, J P; Johnson, M Gatu; Knittel, K M; Leeper, R J; McNaney, J M; Moran, M; Ruiz, C L; Schneider, D H G

    2012-10-01

    Neutron yields are measured at the National Ignition Facility (NIF) by an extensive suite of neutron activation diagnostics. Neutrons interact with materials whose reaction cross sections threshold just below the fusion neutron production energy, providing an accurate measure of primary unscattered neutrons without contribution from lower-energy scattered neutrons. Indium samples are mounted on diagnostic instrument manipulators in the NIF target chamber, 25-50 cm from the source, to measure 2.45 MeV deuterium-deuterium fusion neutrons through the (115)In(n,n')(115 m) In reaction. Outside the chamber, zirconium and copper are used to measure 14 MeV deuterium-tritium fusion neutrons via (90)Zr(n,2n), (63)Cu(n,2n), and (65)Cu(n,2n) reactions. An array of 16 zirconium samples are located on port covers around the chamber to measure relative yield anisotropies, providing a global map of fuel areal density variation. Neutron yields are routinely measured with activation to an accuracy of 7% and are in excellent agreement both with each other and with neutron time-of-flight and magnetic recoil spectrometer measurements. Relative areal density anisotropies can be measured to a precision of less than 3%. These measurements reveal apparent bulk fuel velocities as high as 200 km/s in addition to large areal density variations between the pole and equator of the compressed fuel. PMID:23126840

  18. Advanced materials analysis facility at CSIRO HIAF laboratory

    Kenny, M.J.; Wielunski, L.S.; Baxter, G.R. [CSIRO, Lindfield, NSW (Australia). Applied Physics Div.; Sie, S.H.; Suter, G.F. [CSIRO, North Ryde, NSW (Australia). Exploration and Mining Div.

    1993-12-31

    The HIAF facility at North Ryde, based on a 3 MV Tandetron accelerator has been operating for several years. Initially three ion sources were in operation:- conventional duoplasmatrons for proton and helium beams and a sputter ion source for heavy ions. An electrostatic focusing system was designed and built in-house for providing microbeams. The research emphasis has been largely on microbeam PIXE with particular reference to the mining industry. An AMS system was added in 1990 which prevented the inclusion of the charge exchange canal required for helium beams. The facility has been operated by CSIRO Division of Exploration and Mining. At the beginning of 1992, the lon Beam Technology Group of CSIRO Division of Applied Physics was relocated at Lindfield and became a major user of the HIAF facility. Because the research activities of this group involved Rutherford Backscattering and Channeling, it was necessary to add a helium ion source and a new high vacuum beam line incorporating a precision goniometer. These facilities became operational in the second quarter of 1992. Currently a PIXE system is being added to the chamber containing the goniometer, making the accelerator an extremely versatile one for a wide range of IBA techniques. 3 refs.

  19. Analysis of occupational doses in radioactive and nuclear facilities

    Occupational doses were analyzed in the most important nuclear and radioactive facilities in Argentina, on the period 1988-1994. The areas associated with uranium mining and milling, and medical uses of radiation facilities were excluded from this analysis. The ICRP publication 60 recommendations, adopted in 1990, and enforced in Argentine in 1994, keep the basic criteria of dose limitation system and recommend a substantial reduction in the dose limits. The reduction of the dose limits will affect the individual dose distributions, principally in those installations with occupational doses close to 50 mSv. It were analyzed Occupational doses, principally in the following facilities: Atucha-I and Embalse Nuclear Power Plants, radioisotope production plants, research reactors and radioactive waste management plants. The highest doses were identified in each facility, as well as the task associated with them. Trends in the individual dose distribution and collective and average doses were analyzed. It is concluded, that no relevant difficulties should appear in accomplishing with the basic standards for radiological safety, except for the Atucha-I Nuclear Power Plant. In this NPP a significant effort for the optimization of radiological safety procedures in order to diminish the occupational doses, and a change of the fuel channels by new ones free of cobalt are being carried out. (authors). 4 refs., 3 figs., 3 tabs

  20. BUSTED BUTTE TEST FACILITY GROUND SUPPORT CONFIRMATION ANALYSIS

    The main purpose and objective of this analysis is to confirm the validity of the ground support design for Busted Butte Test Facility (BBTF). The highwall stability and adequacy of highwall and tunnel ground support is addressed in this analysis. The design of the BBTF including the ground support system was performed in a separate document (Reference 5.3). Both in situ and seismic loads are considered in the evaluation of the highwall and the tunnel ground support system. In this analysis only the ground support designed in Reference 5.3 is addressed. The additional ground support installed (still work in progress) by the constructor is not addressed in this analysis. This additional ground support was evaluated by the A/E during a site visit and its findings and recommendations are addressed in this analysis

  1. A decision analysis of an exploratory studies facility

    An Exploratory Studies Facility (ESF) is planned to support the characterization of a potential site for a high-level nuclear waste repository at Yucca Mountain, NV. The selection of a design for the ESF is a critical decision, because the ESF design may affect the accuracy of characterization testing and subsequent repository design. The assist the design process, a comparative evaluation was conducted to rank 34 alternative relied on techniques from formal decision analysis, including decision trees and multiattribute utility analysis (MUA). The results helped to identify favorable design features and convinced the Department of Energy to adopt the top-ranked option as the preferred ESF design

  2. Activation analysis in Europe: present and future perspectives

    A survey is given of the present-day European contribution to activation analysis, covering neutron activation analysis (NAA), charged particle activation analysis (CPAA) and photon activation analysis (PAA). Attention is paid to the available irradiation facilities, in particular nuclear reactors, cyclotrons and Van de Graaff accelerators, and linear electron accelerators. Mention is made of progress in fundamental fields, but the attention is especially focussed on practical applications: environmental, geochemical/cosmochemical, biological/medical, and high-purity materials. Eventually, the role of activation analysis in research projects of the Commission of the European Communities (CEC) and in the Reference Materials program of the Community Bureau of Reference (BCR) is emphasized

  3. Risk management activities at the DOE Class A reactor facilities

    The probabilistic risk assessment (PRA) and risk management group of the Association for Excellence in Reactor Operation (AERO) develops risk management initiatives and standards to improve operation and increase safety of the DOE Class A reactor facilities. Principal risk management applications that have been implemented at each facility are reviewed. The status of a program to develop guidelines for risk management programs at reactor facilities is presented

  4. Multicriteria analysis of thermal and energy systems for tourist facilities

    The introductory part of the paper briefly presents the technological, economic and environmental optimisation procedure of thermal and energy systems for tourist facilities with the multicriteria ranging method when choosing an optimum solution. The procedure described includes a systematic analysis of the system's structure, energy-mass balance, balance of costs, environmental impact analysis and the choice of an optimum solution. Special attention was paid to criteria quantification for the choice of solution and the most appropriate ranging method.The procedure's application has been illustrated on an example of a potential tourist facility on the Island of Loinj, i.e. the locality with a potential highest category tourist development. This example includes (a) consumers (heating of rooms, preparation of hot water, heating of swimming pool water and cooling of rooms), and (b) producers (boiler room, cooling engine-rooms, a cogeneration plant and heat pumps). The data have been supplied from the project documentation for the reconstruction of the existing facilities mainly preliminary designs. The multicriteria ranging was conducted based on an appropriate computer programme for problem solution. (author)

  5. Activation analysis in forensic studies

    Application of neutron activation analysis in forensics are grouped into 3 categories: firearms-discharge applications, elemental analysis of other nonbiological evidence materials (paint, other), and elemental analysis of biological evidence materials (multielemental analysis of hair, analysis of hair for As and Hg). 18 refs

  6. Radiation analysis for a generic centralized interim storage facility

    This paper documents the radiation analysis performed for the storage area of a generic Centralized Interim Storage Facility (CISF) for commercial spent nuclear fuel (SNF). The purpose of the analysis is to establish the CISF Protected Area and Restricted Area boundaries by modeling a representative SNF storage array, calculating the radiation dose at selected locations outside the storage area, and comparing the results with regulatory radiation dose limits. The particular challenge for this analysis is to adequately model a large (6000 cask) storage array with a reasonable amount of analysis time and effort. Previous analyses of SNF storage systems for Independent Spent Fuel Storage Installations at nuclear plant sites (for example in References 5.1 and 5.2) had only considered small arrays of storage casks. For such analyses, the dose contribution from each storage cask can be modeled individually. Since the large number of casks in the CISF storage array make such an approach unrealistic, a simplified model is required

  7. Safety analysis and code development for nuclear fuel cycle facilities

    Development effort of computer codes applicable to nuclear fuel cycle facilities for assisting the task of NISA has been carried out. The work consists of 1) verification of criticality safety analysis codes : MVP and SCALE, 2) studies on burn-up credit applied methods, 3) preparation of non-uniformity effect calculation for criticality safety, 4) development of the new convenient library for shielding calculation based on JENDL-3.3 nuclear data, 5) development of a numerical simulation code DYMPL for analyzing abnormal transients of PUREX processes, 6) radiation dose evaluation code development for reprocessing facilities, 7) updating the dose evaluation data for the probabilistic environmental assessment code MACCS2-JF by emergency scenario. (author)

  8. Dry Transfer Facility No.1 - Ventilation Confinement Zoning Analysis

    The purpose of this analysis is to establish the preliminary Ventilation Confinement Zone (VCZ) for the Dry Transfer Facility (DTF). The results of this document is used to determine the air quantities for each VCZ that will eventually be reflected in the development of the Ventilation Flow Diagrams. The calculations contained in this document were developed by D and E/Mechanical-HVAC and are intended solely for the use of the D and E/Mechanical-HVAC department in its work regarding the HVAC system for the Dry Transfer Facility. Yucca Mountain Project personnel from the D and E/Mechanical-HVAC department should be consulted before use of the calculation for purposes other than those stated herein or used by individuals other than authorized personnel in D and E/Mechanical-HVAC department

  9. Service activities of chemical analysis division

    Progress of the Division during the year of 1988 was described on the service activities for various R and D projects carrying out in the Institute, for the fuel fabrication and conversion plant, and for the post-irradiation examination facility. Relevant analytical methodologies developed for the chemical analysis of an irradiated fuel, safeguards chemical analysis, and pool water monitoring were included such as chromatographic separation of lanthanides, polarographic determination of dissolved oxygen in water, and automation on potentiometric titration of uranium. Some of the laboratory manuals revised were also included in this progress report. (Author)

  10. Cold Vacuum Drying (CVD) Facility Sampling and Analysis Plan

    The Cold Vacuum Drying (CVD) Facility provides the required process systems, supporting equipment, and facilities needed for the conditioning of spent nuclear fuel (SNF) from the Hanford K-Basins prior to storage at the Canister Storage Building (CSB). The process water conditioning (PWC) system collects and treats the selected liquid effluent streams generated by the CVD process. The PWC system uses ion exchange modules (IXMs) and filtration to remove radioactive ions and particulate from CVD effluent streams. Water treated by the PWC is collected in a 5000-gallon storage tank prior to shipment to an on-site facility for additional treatment and disposal. The purpose of this sampling and analysis plan is to document the basis for achieving the following data quality objectives: (1) Measurement of the radionuclide content of the water transferred from the multi-canister overpack (MCO), vacuum purge system (VPS) condensate tank, MCO/Cask annulus and deionized water flushes to the PWC system receiver tanks. (2) Trending the radionuclide inventory of IXMs to assure that they do not exceed the limits prescribed in HNF-2760, Rev. 0-D, ''Safety Analysis Report for Packaging (Onsite) Ion Exchange Modules,'' and HNF-EP-0063 Rev. 5, ''Hanford Site Solid Waste Acceptance Criteria'' for Category 3, non-TRU, low level waste (LLW). (3) Determining the radionuclide content of the PWC system bulk water storage tank to assure that it meets the limits set forth in HNF-3 172, Rev. 0, ''Hanford Site Liquid Waste Acceptance Criteria,'' to permit transfer and disposal at the Effluent Treatment Facility (ETF) located at the 200 East Area

  11. Analysis on the application and actual condition of facilities preservation system in each industry

    Oh, Yon Woo; Kim, Seon Duk; Nam, Ji Hee

    2000-11-01

    In order to secure the maximum of a company's benefit through increasing the efficiency and the productivity of it. the facility preservation system has been developed and used so that can find it's maximum efficiency with a series of activities which make a plan for, install, maintain, and improve for it. Factories are managed to be classified by operation and maintenance with great interest in the facility preservation in South Korea. and the facilities has taken up much part in the management. But it has not been researched how the facilities affects the management of a company. According to that reasons, the facility preservation is underestimated compared with what it is and is regarded just as a cost. This report has an object to construct a fundamental electronic-database on the facility preservation in order to obtain excellent results in KAERI with researches into the introduction of the TPM technology in South Korea, and analysis the effect of the TPM on a company.

  12. Analysis on the application and actual condition of facilities preservation system in each industry

    In order to secure the maximum of a company's benefit through increasing the efficiency and the productivity of it. the facility preservation system has been developed and used so that can find it's maximum efficiency with a series of activities which make a plan for, install, maintain, and improve for it. Factories are managed to be classified by operation and maintenance with great interest in the facility preservation in South Korea. and the facilities has taken up much part in the management. But it has not been researched how the facilities affects the management of a company. According to that reasons, the facility preservation is underestimated compared with what it is and is regarded just as a cost. This report has an object to construct a fundamental electronic-database on the facility preservation in order to obtain excellent results in KAERI with researches into the introduction of the TPM technology in South Korea, and analysis the effect of the TPM on a company

  13. Noble Gas Measurement and Analysis Technique for Monitoring Reprocessing Facilities

    Charlton, William S

    1999-09-01

    An environmental monitoring technique using analysis of stable noble gas isotopic ratios on-stack at a reprocessing facility was developed. This technique integrates existing technologies to strengthen safeguards at reprocessing facilities. The isotopic ratios are measured using a mass spectrometry system and are compared to a database of calculated isotopic ratios using a Bayesian data analysis method to determine specific fuel parameters (e.g., burnup, fuel type, fuel age, etc.). These inferred parameters can be used by investigators to verify operator declarations. A user-friendly software application (named NOVA) was developed for the application of this technique. NOVA included a Visual Basic user interface coupling a Bayesian data analysis procedure to a reactor physics database (calculated using the Monteburns 3.01 code system). The integrated system (mass spectrometry, reactor modeling, and data analysis) was validated using on-stack measurements during the reprocessing of target fuel from a U.S. production reactor and gas samples from the processing of EBR-II fast breeder reactor driver fuel. These measurements led to an inferred burnup that matched the declared burnup with sufficient accuracy and consistency for most safeguards applications. The NOVA code was also tested using numerous light water reactor measurements from the literature. NOVA was capable of accurately determining spent fuel type, burnup, and fuel age for these experimental results. Work should continue to demonstrate the robustness of this system for production, power, and research reactor fuels.

  14. Seismic risk analysis for the Westinghouse Electric facility, Cheswick, Pennsylvania

    This report presents the results of a detailed seismic risk analysis of the Westinghouse Electric plutonium fuel development facility at Cheswick, Pennsylvania. This report focuses on earthquakes. The historical seismic record was established after a review of available literature, consultation with operators of local seismic arrays and examination of appropriate seismic data bases. Because of the aseismicity of the region around the site, an analysis different from the conventional closest approach in a tectonic province was adapted. Earthquakes as far from the site as 1,000 km were included, as were the possibility of earthquakes at the site. In addition, various uncertainties in the input were explicitly considered in the analysis. For example, allowance was made for both the uncertainty in predicting maximum possible earthquakes in the region and the effect of the dispersion of data about the best fit attenuation relation. The attenuation relationship is derived from two of the most recent, advanced studies relating earthquake intensity reports and acceleration. Results of the risk analysis, which include a Bayesian estimate of the uncertainties, are presented as return period accelerations. The best estimate curve indicates that the Westinghouse facility will experience 0.05 g every 220 years and 0.10 g every 1400 years. The accelerations are very insensitive to the details of the source region geometries or the historical earthquake statistics in each region and each of the source regions contributes almost equally to the cumulative risk at the site

  15. Noble Gas Measurement and Analysis Technique for Monitoring Reprocessing Facilities

    An environmental monitoring technique using analysis of stable noble gas isotopic ratios on-stack at a reprocessing facility was developed. This technique integrates existing technologies to strengthen safeguards at reprocessing facilities. The isotopic ratios are measured using a mass spectrometry system and are compared to a database of calculated isotopic ratios using a Bayesian data analysis method to determine specific fuel parameters (e.g., burnup, fuel type, fuel age, etc.). These inferred parameters can be used by investigators to verify operator declarations. A user-friendly software application (named NOVA) was developed for the application of this technique. NOVA included a Visual Basic user interface coupling a Bayesian data analysis procedure to a reactor physics database (calculated using the Monteburns 3.01 code system). The integrated system (mass spectrometry, reactor modeling, and data analysis) was validated using on-stack measurements during the reprocessing of target fuel from a U.S. production reactor and gas samples from the processing of EBR-II fast breeder reactor driver fuel. These measurements led to an inferred burnup that matched the declared burnup with sufficient accuracy and consistency for most safeguards applications. The NOVA code was also tested using numerous light water reactor measurements from the literature. NOVA was capable of accurately determining spent fuel type, burnup, and fuel age for these experimental results. Work should continue to demonstrate the robustness of this system for production, power, and research reactor fuels

  16. DOE standard: Integration of environment, safety, and health into facility disposition activities. Volume 2: Appendices

    This volume contains the appendices that provide additional environment, safety, and health (ES and H) information to complement Volume 1 of this Standard. Appendix A provides a set of candidate DOE ES and H directives and external regulations, organized by hazard types that may be used to identify potentially applicable directives to a specific facility disposition activity. Appendix B offers examples and lessons learned that illustrate implementation of ES and H approaches discussed in Section 3 of Volume 1. Appendix C contains ISMS performance expectations to guide a project team in developing and implementing an effective ISMS and in developing specific performance criteria for use in facility disposition. Appendix D provides guidance for identifying potential Applicable or Relevant and Appropriate Requirements (ARARs) when decommissioning facilities fall under the Comprehensive Environmental Response, Compensation, Liability Act (CERCLA) process. Appendix E discusses ES and H considerations for dispositioning facilities by privatization. Appendix F is an overview of the WSS process. Appendix G provides a copy of two DOE Office of Nuclear Safety Policy and Standards memoranda that form the bases for some of the guidance discussed within the Standard. Appendix H gives information on available hazard analysis techniques and references. Appendix I provides a supplemental discussion to Sections 3.3.4, Hazard Baseline Documentation, and 3.3.6, Environmental Permits. Appendix J presents a sample readiness evaluation checklist

  17. DOE standard: Integration of environment, safety, and health into facility disposition activities. Volume 2: Appendices

    NONE

    1998-05-01

    This volume contains the appendices that provide additional environment, safety, and health (ES and H) information to complement Volume 1 of this Standard. Appendix A provides a set of candidate DOE ES and H directives and external regulations, organized by hazard types that may be used to identify potentially applicable directives to a specific facility disposition activity. Appendix B offers examples and lessons learned that illustrate implementation of ES and H approaches discussed in Section 3 of Volume 1. Appendix C contains ISMS performance expectations to guide a project team in developing and implementing an effective ISMS and in developing specific performance criteria for use in facility disposition. Appendix D provides guidance for identifying potential Applicable or Relevant and Appropriate Requirements (ARARs) when decommissioning facilities fall under the Comprehensive Environmental Response, Compensation, Liability Act (CERCLA) process. Appendix E discusses ES and H considerations for dispositioning facilities by privatization. Appendix F is an overview of the WSS process. Appendix G provides a copy of two DOE Office of Nuclear Safety Policy and Standards memoranda that form the bases for some of the guidance discussed within the Standard. Appendix H gives information on available hazard analysis techniques and references. Appendix I provides a supplemental discussion to Sections 3.3.4, Hazard Baseline Documentation, and 3.3.6, Environmental Permits. Appendix J presents a sample readiness evaluation checklist.

  18. Californium-252 neutron activation facility at the Savannah River Laboratory

    A neutron irradiation facility has been established to develop new analytical methods and for the support of research programs. A major component of this facility is a 252Cf source which provides both fission spectrum and thermal neutrons. (U.S.)

  19. Life cycle cost estimation and systems analysis of Waste Management Facilities

    This paper presents general conclusions from application of a system cost analysis method developed by the United States Department of Energy (DOE), Waste Management Division (WM), Waste Management Facilities Costs Information (WMFCI) program. The WMFCI method has been used to assess the DOE complex-wide management of radioactive, hazardous, and mixed wastes. The Idaho Engineering Laboratory, along with its subcontractor Morrison Knudsen Corporation, has been responsible for developing and applying the WMFCI cost analysis method. The cost analyses are based on system planning level life-cycle costs. The costs for life-cycle waste management activities estimated by WMFCI range from bench-scale testing and developmental work needed to design and construct a facility, facility permitting and startup, operation and maintenance, to the final decontamination, decommissioning, and closure of the facility. For DOE complex-wide assessments, cost estimates have been developed at the treatment, storage, and disposal module level and rolled up for each DOE installation. Discussions include conclusions reached by studies covering complex-wide consolidation of treatment, storage, and disposal facilities, system cost modeling, system costs sensitivity, system cost optimization, and the integration of WM waste with the environmental restoration and decontamination and decommissioning secondary wastes

  20. Life cycle cost estimation and systems analysis of waste management facilities

    This paper presents general conclusions from application of a system cost analysis method developed by the United States Department of Energy (DOE), Waste Management Division (WM), Waste Management Facilities Costs Information (WMFCI) program. The WMFCI method has been used to assess the DOE complex-wide management of radioactive, hazardous, and mixed wastes. The Idaho Engineering Laboratory, along with its subcontractor Morrison Knudsen Corporation, has been responsible for developing and applying the WMFCI cost analysis method. The cost analyses are based on system planning level life-cycle costs. The costs for life-cycle waste management activities estimated by WMFCI range from bench-scale testing and developmental work needed to design and construct a facility, facility permitting and startup, operation and maintenance, to the final decontamination, decommissioning, and closure of the facility. For DOE complex-wide assessments, cost estimates have been developed at the treatment, storage, and disposal module level and rolled up for each DOE installation. Discussions include conclusions reached by studies covering complex-wide consolidation of treatment, storage, and disposal facilities, system cost modeling, system costs sensitivity, system cost optimization, and the integration of WM waste with the environmental restoration and decontamination and decommissioning secondary wastes

  1. Criticality analysis for uranium-scrap recycling facilities

    KNFC planned to build a uranium scrap recycling facility in order to make its fuel manufacturing process efficient. An engineering design has been done by Human and Technologies Corp. during 6 months of the last year. A criticality analysis has been performed with Kyunghee University and report was reviewed by KINS. This paper summarized a criticality analysis part of this work for licensing. A criticality analysis was done for all processes in scrap recycling system with data from design specifications based on reasonable assumptions. As the first step, parametric study was done for a normal operational condition in order to find crucial variables which would be sensitive to the criticality safety. Hypothetical accident was also simulated with double contingency principle and multi-parameter control principle. Calculation was performed with Monte Carlo code, MCNP-4C/2 with point data cross section data library

  2. Asymmetrical sabotage tactics, nuclear facilities/materials, and vulnerability analysis

    Full text: The emerging paradigm of a global community wherein post-modern political violence is a fact of life that must be dealt with by safety and security planners is discussed. This paradigm shift in the philosophy of terrorism is documented by analysis of the emerging pattern of asymmetrical tactics being employed by terrorists. Such philosophical developments in violent political movements suggest a shift in the risks that security and safety personnel must account for in their planning for physical protection of fixed site nuclear source facilities like power generation stations and the eventual storage and transportation of the by-products of these facilities like spent nuclear fuel and other high level wastes. This paper presents a framework for identifying these new political realities and related threat profiles, suggests ways in which security planners and administrators can design physical protection practices to meet these emerging threats, and argues for global adoption of standards for the protection of nuclear facilities that could be used as a source site from which terrorists could inflict a mass contamination event and for standards related to the protection of the waste materials that can be used in the production of radiological weapons of mass victimization. (author)

  3. Waste Encapsulation and Storage Facility mission analysis report

    This report defines the mission for the Waste Encapsulation and Storage Facility (WESF). It contains summary information regarding the mission analysis which was performed by holding workshops attended by relevant persons involved in the WESF operations. The scope of the WESF mission is to provide storage of Cesium (Cs) and Strontium (Sr) capsules, previously produced at WESF, until every capsule has been removed from the facility either to another storage location, for disposal or for beneficial use by public or private enterprises. Since the disposition of the capsules has not yet been determined, they may be stored at WESF for many years, even decades. The current condition of the WESF facility must be upgraded and maintained to provide for storage which is safe, cost effective, and fully compliant with DOE direction as well as federal, state, and local laws and regulations. The Cs capsules produced at WESF were originally released to private enterprises for uses such as the sterilization of medical equipment; but because of the leakage of one capsule, all are being returned. The systems, subsystems, and equipment not required for the storage mission will be available for use by other projects or private enterprises. Beyond the storage of the Cs and Sr capsules, no future mission for the WESF has been identified

  4. Waste Receiving and Processing (WRAP) Facility Final Safety Analysis Report (FSAR)

    The Waste Receiving and Processing Facility (WRAP), 2336W Building, on the Hanford Site is designed to receive, confirm, repackage, certify, treat, store, and ship contact-handled transuranic and low-level radioactive waste from past and present U.S. Department of Energy activities. The WRAP facility is comprised of three buildings: 2336W, the main processing facility (also referred to generically as WRAP); 2740W, an administrative support building; and 2620W, a maintenance support building. The support buildings are subject to the normal hazards associated with industrial buildings (no radiological materials are handled) and are not part of this analysis except as they are impacted by operations in the processing building, 2336W. WRAP is designed to provide safer, more efficient methods of handling the waste than currently exist on the Hanford Site and contributes to the achievement of as low as reasonably achievable goals for Hanford Site waste management

  5. Receiving Basin for Offsite Fuels and the Resin Regeneration Facility Safety Analysis Report, Executive Summary

    Shedrow, C.B.

    1999-11-29

    The Safety Analysis Report documents the safety authorization basis for the Receiving Basin for Offsite Fuels (RBOF) and the Resin Regeneration Facility (RRF) at the Savannah River Site (SRS). The present mission of the RBOF and RRF is to continue in providing a facility for the safe receipt, storage, handling, and shipping of spent nuclear fuel assemblies from power and research reactors in the United States, fuel from SRS and other Department of Energy (DOE) reactors, and foreign research reactors fuel, in support of the nonproliferation policy. The RBOF and RRF provide the capability to handle, separate, and transfer wastes generated from nuclear fuel element storage. The DOE and Westinghouse Savannah River Company, the prime operating contractor, are committed to managing these activities in such a manner that the health and safety of the offsite general public, the site worker, the facility worker, and the environment are protected.

  6. Receiving Basin for Offsite Fuels and the Resin Regeneration Facility Safety Analysis Report, Executive Summary

    The Safety Analysis Report documents the safety authorization basis for the Receiving Basin for Offsite Fuels (RBOF) and the Resin Regeneration Facility (RRF) at the Savannah River Site (SRS). The present mission of the RBOF and RRF is to continue in providing a facility for the safe receipt, storage, handling, and shipping of spent nuclear fuel assemblies from power and research reactors in the United States, fuel from SRS and other Department of Energy (DOE) reactors, and foreign research reactors fuel, in support of the nonproliferation policy. The RBOF and RRF provide the capability to handle, separate, and transfer wastes generated from nuclear fuel element storage. The DOE and Westinghouse Savannah River Company, the prime operating contractor, are committed to managing these activities in such a manner that the health and safety of the offsite general public, the site worker, the facility worker, and the environment are protected

  7. Waste Receiving and Processing (WRAP) Facility Final Safety Analysis Report (FSAR)

    TOMASZEWSKI, T.A.

    2000-04-25

    The Waste Receiving and Processing Facility (WRAP), 2336W Building, on the Hanford Site is designed to receive, confirm, repackage, certify, treat, store, and ship contact-handled transuranic and low-level radioactive waste from past and present U.S. Department of Energy activities. The WRAP facility is comprised of three buildings: 2336W, the main processing facility (also referred to generically as WRAP); 2740W, an administrative support building; and 2620W, a maintenance support building. The support buildings are subject to the normal hazards associated with industrial buildings (no radiological materials are handled) and are not part of this analysis except as they are impacted by operations in the processing building, 2336W. WRAP is designed to provide safer, more efficient methods of handling the waste than currently exist on the Hanford Site and contributes to the achievement of as low as reasonably achievable goals for Hanford Site waste management.

  8. Bioaerosol releases from compost facilities: Evaluating passive and active source terms at a green waste facility for improved risk assessments

    Taha, M. P. M.; Drew, G. H.; Longhurst, P. J.; Smith, R.; Pollard, S. J. T.

    The passive and active release of bioaerosols during green waste composting, measured at source is reported for a commercial composting facility in South East (SE) England as part of a research programme focused on improving risk assessments at composting facilities. Aspergillus fumigatus and actinomycetes concentrations of 9.8-36.8×10 6 and 18.9-36.0×10 6 cfu m -3, respectively, measured during the active turning of green waste compost, were typically 3-log higher than previously reported concentrations from static compost windrows. Source depletion curves constructed for A. fumigatus during compost turning and modelled using SCREEN3 suggest that bioaerosol concentrations could reduce to background concentrations of 10 3 cfu m -3 within 100 m of this site. Authentic source term data produced from this study will help to refine the risk assessment methodologies that support improved permitting of compost facilities.

  9. Inventory difference analysis at Los Alamos Plutonium Facility

    The authors have developed a prototype computer program that reads directly the inventory entries from a Microsoft Access data base. Based on historical data, the program then displays temporal trends and constructs a library of rules that encapsulates the system behavior. The following analysis of inventory data is illustrated by using a combination of realistic and simulated facility examples. Potential payoffs of this methodology include a reduction in time and resources needed to perform statistical tests and broad applicability to Department of Energy needs--for example, treaty verification

  10. Seismic analysis of the mirror fusion test facility shielding vault

    This report presents a seismic analysis of the vault in Building 431 at Lawrence Livermore National Laboratory which houses the mirror Fusion Test Facility. The shielding vault structure is approximately 120 ft long by 80 ft wide and is constructed of concrete blocks approximately 7 x 7 x 7 ft. The north and south walls are approximately 53 ft high and the east wall is approximately 29 ft high. These walls are supported on a monolithic concrete foundation that surrounds a 21-ft deep open pit. Since the 53-ft walls appeared to present the greatest seismic problem they were the first investigated

  11. A test facility of active alignment system at KEK

    A test facility with one control axis has been constructed at KEK to investigate a super-accurate alignment technique for the JLC (Japan Linear Collider) project. The facility consists of a stabilized laser system and a vibration control stage equipped with piezo transducers. Results of the first test show that the distance of about 28 cm is kept stable to 50 nm or better up to the frequency of 20 Hz, against the sine wave disturbance with a 500 nm amplitude

  12. Space Station Furnace Facility. Volume 2: Requirements definition and conceptual design study. Appendix 3: Environment analysis

    1992-01-01

    A Preliminary Safety Analysis (PSA) is being accomplished as part of the Space Station Furnace Facility (SSFF) contract. This analysis is intended to support SSFF activities by analyzing concepts and designs as they mature to develop essential safety requirements for inclusion in the appropriate specifications, and designs, as early as possible. In addition, the analysis identifies significant safety concerns that may warrant specific trade studies or design definition, etc. The analysis activity to date concentrated on hazard and hazard cause identification and requirements development with the goal of developing a baseline set of detailed requirements to support trade study, specifications development, and preliminary design activities. The analysis activity will continue as the design and concepts mature. Section 2 defines what was analyzed, but it is likely that the SSFF definitions will undergo further changes. The safety analysis activity will reflect these changes as they occur. The analysis provides the foundation for later safety activities. The hazards identified will in most cases have Preliminary Design Review (PDR) applicability. The requirements and recommendations developed for each hazard will be tracked to ensure proper and early resolution of safety concerns.

  13. Safety analysis of the 700-horsepower combustion test facility

    Berkey, B.D.

    1981-05-01

    The objective of the program reported herein was to provide a Safety Analysis of the 700 h.p. Combustion Test Facility located in Building 93 at the Pittsburgh Energy Technology Center. Extensive safety related measures have been incorporated into the design, construction, and operation of the Combustion Test Facility. These include: nitrogen addition to the coal storage bin, slurry hopper, roller mill and pulverizer baghouse, use of low oxygen content combustion gas for coal conveying, an oxygen analyzer for the combustion gas, insulation on hot surfaces, proper classification of electrical equipment, process monitoring instrumentation and a planned remote television monitoring system. Analysis of the system considering these factors has resulted in the determination of overall probabilities of occurrence of hazards as shown in Table I. Implementation of the recommendations in this report will reduce these probabilities as indicated. The identified hazards include coal dust ignition by hot ductwork and equipment, loss of inerting within the coal conveying system leading to a coal dust fire, and ignition of hydrocarbon vapors or spilled oil, or slurry. The possibility of self-heating of coal was investigated. Implementation of the recommendations in this report will reduce the ignition probability to no more than 1 x 10/sup -6/ per event. In addition to fire and explosion hazards, there are potential exposures to materials which have been identified as hazardous to personal health, such as carbon monoxide, coal dust, hydrocarbon vapors, and oxygen deficient atmosphere, but past monitoring experience has not revealed any problem areas. The major environmental hazard is an oil spill. The facility has a comprehensive spill control plan.

  14. 12 CFR 204.122 - Secondary market activities of international banking facilities.

    2010-01-01

    ... 12 Banks and Banking 2 2010-01-01 2010-01-01 false Secondary market activities of international banking facilities. 204.122 Section 204.122 Banks and Banking FEDERAL RESERVE SYSTEM BOARD OF GOVERNORS OF...) Interpretations § 204.122 Secondary market activities of international banking facilities. (a) Questions have...

  15. Systems Analysis of Safeguards Effectiveness in a Uranium Conversion Facility

    Elayat, H A; Lambert, H; O' Connell, W J

    2004-06-16

    The U.S. Department of Energy (DOE) is interested in developing tools and methods for potential U.S. use in designing and evaluating safeguards systems. For this goal several DOE National Laboratories are defining the characteristics of typical facilities of several size scales, and the safeguards measures and instrumentation that could be applied. Lawrence Livermore National Laboratory is providing systems modeling and analysis of facility and safeguards operations, diversion path generation, and safeguards system effectiveness. The constituent elements of diversion scenarios are structured using directed graphs (digraphs) and fault trees. Safeguards indicator probabilities are based on sampling statistics and/or measurement accuracies. Scenarios are ranked based on value and quantity of material removed and the estimated probability of non-detection. Significant scenarios, especially those involving timeliness or randomly varying order of events, are transferred to simulation analysis. Simulations show the range of conditions encountered by the safeguards measurements and inspections, e.g., the quantities of intermediate materials in temporary storage and the time sequencing of material flow. Given a diversion campaign, simulations show how much the range of the same parameters observed by the safeguards system can differ from the base-case range. The combination of digraphs, fault trees, statistics and simulation constitute a method for evaluation of the estimated benefit of alternate or additional safeguards equipment or features. A generic example illustrates the method.

  16. Site-specific meteorology identification for DOE facility accident analysis

    Currently, chemical dispersion calculations performed for safety analysis of DOE facilities assume a Pasquill D-Stability Class with a 4.5 m/s windspeed. These meteorological conditions are assumed to conservatively address the source term generation mechanism as well as the dispersion mechanism thereby resulting in a net conservative downwind consequence. While choosing this Stability Class / Windspeed combination may result in an overall conservative consequence, the level of conservative can not be quantified. The intent of this paper is to document a methodology which incorporates site-specific meteorology to determine a quantifiable consequence of a chemical release. A five-year meteorological database, appropriate for the facility location, is utilized for these chemical consequence calculations, and is consistent with the approach used for radiological releases. The hourly averages of meteorological conditions have been binned into 21 groups for the chemical consequence calculations. These 21 cases each have a probability of occurrence based on the number of times each case has occurred over the five year sampling period. A code has been developed which automates the running of all the cases with a commercially available air modeling code. The 21 cases are sorted by concentration. A concentration may be selected by the user for a quantified level of conservatism. The methodology presented is intended to improve the technical accuracy and defensability of Chemical Source Term / Dispersion Safety Analysis work. The result improves the quality of safety analyses products without significantly increasing the cost

  17. Optimization model for air quality analysis in energy facility siting

    Emanuel, W. R.; Murphy, B. D.; Huff, D. D.; Begovich, C. L.; Hurt, J. F.

    1977-09-01

    The siting of energy facilities on a regional scale is discussed with particular attention to environmental planning criteria. A multiple objective optimization model is proposed as a framework for the analysis of siting problems. Each planning criterion (e.g., air quality, water quality, or power demand) is treated as an objective function to be minimized or maximized subject to constraints in this optimization procedure. The formulation of the objective functions is illustrated by the development of a siting model for the minimization of human exposure to air pollutants. This air quality siting model takes the form of a linear programming problem. A graphical analysis of this type of problem, which provides insight into the nature of the siting model, is given. The air quality siting model is applied to an illustrative siting example for the Tennessee Valley area.

  18. Distribution of physical activity facilities in Scotland by small area measures of deprivation and urbanicity

    Ogilvie David

    2010-10-01

    Full Text Available Abstract Background The aim of this study was to examine the distribution of physical activity facilities by area-level deprivation in Scotland, adjusting for differences in urbanicity, and exploring differences between and within the four largest Scottish cities. Methods We obtained a list of all recreational physical activity facilities in Scotland. These were mapped and assigned to datazones. Poisson and negative binomial regression models were used to investigate associations between the number of physical activity facilities relative to population size and quintile of area-level deprivation. Results The results showed that prior to adjustment for urbanicity, the density of all facilities lessened with increasing deprivation from quintiles 2 to 5. After adjustment for urbanicity and local authority, the effect of deprivation remained significant but the pattern altered, with datazones in quintile 3 having the highest estimated mean density of facilities. Within-city associations were identified between the number of physical activity facilities and area-level deprivation in Aberdeen and Dundee, but not in Edinburgh or Glasgow. Conclusions In conclusion, area-level deprivation appears to have a significant association with the density of physical activity facilities and although overall no clear pattern was observed, affluent areas had fewer publicly owned facilities than more deprived areas but a greater number of privately owned facilities.

  19. Application of neutron activation analysis

    The physical basis and analytical possibilities of neutron activation analysis have been performed. The number of applications in material engineering, geology, cosmology, oncology, criminology, biology, agriculture, environment protection, archaeology, history of art and especially in chemical analysis have been presented. The place of the method among other methods of inorganic quantitative chemical analysis for trace elements determination has been discussed

  20. Analysis of Precision of Activation Analysis Method

    Heydorn, Kaj; Nørgaard, K.

    1973-01-01

    The precision of an activation-analysis method prescribes the estimation of the precision of a single analytical result. The adequacy of these estimates to account for the observed variation between duplicate results from the analysis of different samples and materials, is tested by the statistic T...

  1. Measurement of radon voluminal activity in underground facilities. Methodological guide

    The measurement of radon voluminal activity in a building is codified by the AFNOR NF M60-771 norm, relative to the methodology enforced to the case of underground buildings. It applies to any type of buildings whatever be the type of interface, the area and the ventilation mode. To bring out the presence of radon in a building, by measures comparable to the values of interest given by public authorities, must be realised with a detection mean. The objective of this detection is to determine if all or part of the building presents a yearly average value of the radon voluminal activity over to one or several values of interest. Only the methods of integrated measurement with a passive sampling and a delayed analysis are used in the case of radon detection. These methods and the plans of associated measures must be in accordance with the AFNOR NF M-60-766 norm. The implementation of this methodology requires knowledge relative to radon and to the building. It is thus the responsibility of relevant agencies. It is to notice that the estimation of people exposure to ambient gamma radiation can be got by the adding of gamma integrator dosemeters of thermoluminescent type detectors to the devices of radon measurement in the conditions described in this document. (N.C.)

  2. Assessment of activity-based pyroprocess costs for an engineering-scale facility in Korea

    Kim, Sung Ki; Ko, Won Il [Nuclear Fuel Cycle Analysis Department, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Bang, Sung Sig [Dept. of Business and Technology Management, Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2015-12-15

    This study set the pyroprocess facility at an engineering scale as a cost object, and presented the cost consumed during the unit processes of the pyroprocess. For the cost calculation, the activity based costing (ABC) method was used instead of the engineering cost estimation method, which calculates the cost based on the conceptual design of the pyroprocess facility. The calculation results demonstrate that the pyroprocess facility's unit process cost is $194/kgHM for pretreatment, $298/kgHM for electrochemical reduction, $226/kgHM for electrorefining, and $299/kgHM for electrowinning. An analysis demonstrated that the share of each unit process cost among the total pyroprocess cost is as follows: 19% for pretreatment, 29% for electrochemical reduction, 22% for electrorefining, and 30% for electrowinning. The total unit cost of the pyroprocess was calculated at $1,017/kgHM. In the end, electrochemical reduction and the electrowinning process took up most of the cost, and the individual costs for these two processes was found to be similar. This is because significant raw material cost is required for the electrochemical reduction process, which uses platinum as an anode electrode. In addition, significant raw material costs are required, such as for Li3PO4, which is used a lot during the salt purification process.

  3. Quantitative risk analysis of oil storage facilities in seismic areas.

    Fabbrocino, Giovanni; Iervolino, Iunio; Orlando, Francesca; Salzano, Ernesto

    2005-08-31

    Quantitative risk analysis (QRA) of industrial facilities has to take into account multiple hazards threatening critical equipment. Nevertheless, engineering procedures able to evaluate quantitatively the effect of seismic action are not well established. Indeed, relevant industrial accidents may be triggered by loss of containment following ground shaking or other relevant natural hazards, either directly or through cascade effects ('domino effects'). The issue of integrating structural seismic risk into quantitative probabilistic seismic risk analysis (QpsRA) is addressed in this paper by a representative study case regarding an oil storage plant with a number of atmospheric steel tanks containing flammable substances. Empirical seismic fragility curves and probit functions, properly defined both for building-like and non building-like industrial components, have been crossed with outcomes of probabilistic seismic hazard analysis (PSHA) for a test site located in south Italy. Once the seismic failure probabilities have been quantified, consequence analysis has been performed for those events which may be triggered by the loss of containment following seismic action. Results are combined by means of a specific developed code in terms of local risk contour plots, i.e. the contour line for the probability of fatal injures at any point (x, y) in the analysed area. Finally, a comparison with QRA obtained by considering only process-related top events is reported for reference. PMID:15908107

  4. Radiation Shielding Analysis of Electron Beam Accelerator Facility

    The objective of this technical report are to establish the radiation shielding technology of a high-energy electron accelerator to the facilities which utilize with electron beam. The technologies of electron beam irradiation(300 KeV -10 MeV) demand on the diverse areas of material processing, surface treatment, treatments on foods or food processing, improvement of metal properties, semiconductors, and ceramics, sterilization of medical goods and equipment, treatment and control of contamination and pollution, and so on. In order to acquire safety design for the protection of personnel from the radiations produced by electron beam accelerators, it is important to develop the radiation shielding analysis technology. The shielding analysis are carried out by which define source term, calculation modelling and computer calculations for 2 MeV and 10 MeV accelerators. And the shielding analysis for irradiation dump shield with 10 MeV accelerators are also performed by solving the complex 3-D geometry and long computer run time problem. The technology development of shielding analysis will be contributed to extend the further high energy accelerator development

  5. Surface Fire Hazards Analysis Technical Report-Constructor Facilities

    The purpose of this Fire Hazards Analysis Technical Report (hereinafter referred to as Technical Report) is to assess the risk from fire within individual fire areas to ascertain whether the U.S. Department of Energy (DOE) fire safety objectives are met. The objectives identified in DOE Order 420.1, Change 2, Facility Safety, Section 4.2, establish requirements for a comprehensive fire and related hazards protection program for facilities sufficient to minimize the potential for: The occurrence of a fire or related event; A fire that causes an unacceptable on-site or off-site release of hazardous or radiological material that will threaten the health and safety of employees, the public, or the environment; Vital DOE programs suffering unacceptable interruptions as a result of fire and related hazards; Property losses from a fire and related events exceeding defined limits established by DOE; and Critical process controls and safety class systems being damaged as a result of a fire and related events

  6. ATLAS Tier-3 within IFIC-Valencia analysis facility

    Villaplana, M; The ATLAS collaboration; Fernández, A; Salt, J; Lamas, A; Fassi, F; Kaci, M; Oliver, E; Sánchez, J; Sánchez-Martínez, V

    2012-01-01

    The ATLAS Tier-3 at IFIC-Valencia is attached to a Tier-2 that has 50% of the Spanish Federated Tier-2 resources. In its design, the Tier-3 includes a GRID-aware part that shares some of the features of IFIC Tier-2 such as using Lustre as a file system. ATLAS users, 70% of IFIC users, also have the possibility of analysing data with a PROOF farm and storing them locally. In this contribution we discuss the design of the analysis facility as well as the monitoring tools we use to control and improve its performance. We also comment on how the recent changes in the ATLAS computing GRID model affect IFIC. Finally, how this complex system can coexist with the other scientific applications running at IFIC (non-ATLAS users) is presented.

  7. Hedonic Pricing Evaluation on Agritourism Activity in Italy: Local Culture-based or Facility-based?

    Ohe, Yasuo; Ciani, Adriano

    2010-01-01

    This paper focused on how and what diversified activities influence the price level of agritourism. A hypothesis that contrasts two directions was examined: facility-based and local culture-based activities. First, from the conceptual consideration, we defined that agritourism based on local cultural resources can internalize positive externalities, which are accompanied by local cultural resources, into income, unlike facility-based activity that has no connection with local cultural resourc...

  8. Development of activation analysis on the IBR-2 reactor

    Different examples of activation analysis (AA) application and probabilities of its further development using IBR-2 reactor (Dubna) with two facilities: REGATA pneumotransport facility desigued for instrumental AA and biophysical channel designed for element analysis using capture prompt quanta and radiography-are considered. Characteristics of irradiation channels, values of flux densities for thermal, resonance and fast neutrons are given. Application advantages concerning instrumental AA of resonance neutrons are considered. New application trend of AA for composition optimization of concretes used in shielding structures of nuclear reactors, for reduction of long-lived directed activity is pointed out. 20 refs.; 6 figs.; 7 tabs

  9. National Ignition Facility Shot Data Analysis Module Guidelines

    Azevedo, S; Glenn, S; Lopez, A; Warrick, A; Beeler, R

    2007-10-03

    This document provides the guidelines for software development of modules to be included in Shot Data Analysis (SDA) for the National Ignition Facility (NIF). An Analysis Module is a software entity that groups a set of (typically cohesive) functions, procedures and data structures for performing an analysis task relevant to NIF shot operations. Each module must have its own unique identification (module name), clear interface specifications (data inputs and outputs), and internal documentation. It is vitally important to the NIF Program that all shot-related data be processed and analyzed in a consistent way that is reviewed by scientific and engineering experts. SDA is part of a NIF Integrated Product Team (IPT) whose goal is to provide timely and accurate reporting of shot results to NIF campaign experimentalists. Other elements of the IPT include the Campaign Management Tool (CMT) for configuring experiments, a data archive and provisioning system called CMS, a calibration and configuration database (CDMS), and a shot data visualization tool (SDV). We restrict our scope at this time to guidelines for modules written in Interactive Data Language, or IDL1. This document has sections describing example IDL modules and where to find them, how to set up a development environment, IDL programming guidelines, shared IDL procedures for general use, and revision control.

  10. TIBER activation analysis

    TIBER-II is an engineering test reactor designed to establish the technical feasibility for fusion, and is a U.S. option for the prospective International Thermonuclear Test Reactor (ITER). The TIBER-II baseline design has a 3 m major radius, 3.6 aspect ratio, and 1.1 MW/m2 average neutron wall loading. The inboard shield is about .5 m thick and structurally consists of tungsten alloy and PCA alloy. The outboard is 1.52 m thick and utilizes PCA as structure and beryllium as a neutron multiplier. An aqueous solution of 160 g LiNO3/liter is used throughout as a coolant and breeder. A one-dimensional cylindrical model for TIBER is used to calculate the neutron flux and the radioactivities. Activities are calculated during and after 2.5 full power years (FPY) of operation

  11. Life-cycle cost analysis 200-West Weather Enclosure: Multi-function Waste Tank Facility

    The Multi-Function Waste Tank Facility (MWTF)will provide environmentally safe and acceptable storage capacity for handling wastes resulting from the remediation of existing single-shell and double-shell tanks on the Hanford Site. The MWTF will construct two tank farm facilities at two separate locations. A four-tank complex will be constructed in the 200-East Area of the Hanford Site; a two-tank complex will be constructed in the 200-West Area. This report documents the results of a life-cycle cost analysis performed by ICF Kaiser Hanford Company (ICF KH) for the Weather Enclosure proposed to be constructed over the 200-West tanks. Currently, all tank farm operations on the Hanford Site are conducted in an open environment, with weather often affecting tank farm maintenance activities. The Weather Enclosure is being proposed to allow year-round tank farm operation and maintenance activities unconstrained by weather conditions. Elimination of weather-related delays at the MWTF and associated facilities will reduce operational costs. The life-cycle cost analysis contained in this report analyzes potential cost savings based on historical weather information, operational and maintenance costs, construction cost estimates, and other various assumptions

  12. Business administration of PET facilities. A nationwide survey for prices of PET screening and a cost analysis of three facilities

    The purpose of this study is to analyze the business administration of positron emission tomography (PET) facilities based on the survey of the price of PET cancer screening and cost analysis of PET examination. The questionnaire survey of the price of PET cancer screening was implemented for all PET facilities in Japan. Cost data of PET examination, including fixed costs and variable costs, were obtained from three different medical institutions. The marked price of the PET cancer screening was yen111,499 in average, and the most popular range of prices was between yen80,000 and yen90,000. Costs of PET per examination were accounted for yen110,675, yen79,158 and yen111,644 in facility A, B and C, respectively. The results suggested that facilities with two or more PET/CT per a cyclotron could only secure profits. In Japan, the boom in PET facility construction could not continue in accordance with increasing number of PET facilities. It would become more essential to analyze the appropriate distribution of PET facilities and the adequate amount of PET procedures from the perspective of efficient utilization of the PET equipments and supply of PET-related healthcare. (author)

  13. Thermal-hydraulic analysis of the International Fusion Materials Irradiation Facility

    Since 1994 the International Fusion Materials Irradiation Facility is under development. Up till now only design activities have been performed aimed at providing a reference design, evaluating remaining design uncertainty, reducing the costs and the key technology risk factors to reach the specified requirements with sufficient availability and reliability. From the beginning ENEA is engaged in the design of all the systems. In particular for the Lithium Target System, its activities are mainly focused on risk analysis, transient analysis, thermal-hydraulics and stability of lithium jet. This paper deals with the analysis of the behaviour of the Lithium Target System under normal and incident conditions, performed with a version of the RELAP5/Mod3.2 code modified to allow for specific features of the system itself (Lithium and organic oil as cooling fluids).(author)

  14. 25 CFR 170.152 - What transit facilities and activities are eligible for IRR Program funding?

    2010-04-01

    ... IRR Program funding? 170.152 Section 170.152 Indians BUREAU OF INDIAN AFFAIRS, DEPARTMENT OF THE... funding? Transit facilities and activities eligible for IRR Program funding include, but are not limited... facilities for use in mass transportation; (f) Third-party contracts for otherwise eligible...

  15. Preliminary Closure Plan for the Immobilized Low Activity Waste (ILAW) Disposal Facility

    This document describes the preliminary plans for closure of the Immobilized Low-Activity Waste (ILAW) disposal facility to be built by the Office of River Protection at the Hanford site in southeastern Washington. The facility will provide near-surface disposal of up to 204,000 cubic meters of ILAW in engineered trenches with modified RCRA Subtitle C closure barriers

  16. Preliminary Closure Plan for the Immobilized Low Activity Waste (ILAW) Disposal Facility

    BURBANK, D.A.

    2000-08-31

    This document describes the preliminary plans for closure of the Immobilized Low-Activity Waste (ILAW) disposal facility to be built by the Office of River Protection at the Hanford site in southeastern Washington. The facility will provide near-surface disposal of up to 204,000 cubic meters of ILAW in engineered trenches with modified RCRA Subtitle C closure barriers.

  17. Documented Safety Analysis Addendum for the Neutron Radiography Reactor Facility Core Conversion

    Boyd D. Christensen

    2009-05-01

    The Neutron Radiography Reactor Facility (NRAD) is a Training, Research, Isotope Production, General Atomics (TRIGA) reactor which was installed in the Idaho National Laboratory (INL) Hot Fuels Examination Facility (HFEF) at the Materials and Fuels Complex (MFC) in the mid 1970s. The facility provides researchers the capability to examine both irradiated and non-irradiated materials in support of reactor fuel and components programs through non-destructive neutron radiography examination. The facility has been used in the past as one facet of a suite of reactor fuels and component examination facilities available to researchers at the INL and throughout the DOE complex. The facility has also served various commercial research activities in addition to the DOE research and development support. The reactor was initially constructed using Fuel Lifetime Improvement Program (FLIP)- type highly enriched uranium (HEU) fuel obtained from the dismantled Puerto Rico Nuclear Center (PRNC) reactor. In accordance with international non-proliferation agreements, the NRAD core will be converted to a low enriched uranium (LEU) fuel and will continue to utilize the PRNC control rods, control rod drives, startup source, and instrument console as was previously used with the HEU core. The existing NRAD Safety Analysis Report (SAR) was created and maintained in the preferred format of the day, combining sections of both DOE-STD-3009 and Nuclear Regulatory Commission Regulatory Guide 1.70. An addendum was developed to cover the refueling and reactor operation with the LEU core. This addendum follows the existing SAR format combining required formats from both the DOE and NRC. This paper discusses the project to successfully write a compliant and approved addendum to the existing safety basis documents.

  18. Activation analysis using Cornell TRIGA

    A major use of the Cornell TRIGA is for activation analysis. Over the years many varieties of samples have been analyzed from a number of fields of interest ranging from geology, archaeology and textiles. More recently the analysis has been extended to high technology materials for applications in optical and semiconductor devices. Trace analysis in high purity materials like Si wafers has been the focus in many instances, while in others analysis of major/minor components were the goals. These analysis has been done using the delayed mode. Results from recent measurements in semiconductors and other materials will be presented. In addition the near future capability of using prompt gamma activation analysis using the Cornell cold neutron beam will be discussed. (author)

  19. Standardizing Activation Analysis: New Software for Photon Activation Analysis

    Photon Activation Analysis (PAA) of environmental, archaeological and industrial samples requires extensive data analysis that is susceptible to error. For the purpose of saving time, manpower and minimizing error, a computer program was designed, built and implemented using SQL, Access 2007 and asp.net technology to automate this process. Based on the peak information of the spectrum and assisted by its PAA library, the program automatically identifies elements in the samples and calculates their concentrations and respective uncertainties. The software also could be operated in browser/server mode, which gives the possibility to use it anywhere the internet is accessible. By switching the nuclide library and the related formula behind, the new software can be easily expanded to neutron activation analysis (NAA), charged particle activation analysis (CPAA) or proton-induced X-ray emission (PIXE). Implementation of this would standardize the analysis of nuclear activation data. Results from this software were compared to standard PAA analysis with excellent agreement. With minimum input from the user, the software has proven to be fast, user-friendly and reliable.

  20. Safety analysis report for the Cold Vacuum Drying Facility, phase 1, supporting civil/structural construction

    The US Department of Energy established the K Basins Spent Nuclear Fuel Project to address safety and environmental concerns associated with deteriorating spent nuclear fuel presently stored under water in the Hanford Site's K Basins, which are located near the Columbia River. Recommendations for a series of aggressive projects to construct and operate systems and facilities to manage the safe removal of K Basins fuel were made in WHC-EP-0830, Hanford Spent Nuclear Fuel Recommended Path Forward,' and its subsequent update, WHC-SD-SNF-SP-005, Hanford Spent Nuclear Fuel Project Integrated Process Strategy for K Basins Fuel. The integrated process strategy recommendations include the following process steps: fuel preparation activities at the K Basins, including removing the fuel elements from their K Basin canisters, separating fuel particulate from fuel elements and fuel fragments greater than 0.6 cm (0.25 in.) in any dimension, removing excess sludge from the fuel and fuel fragments by means of flushing, as necessary, and packaging the fuel into multicanister overpacks; removal of free water by draining and vacuum drying at the Cold Vacuum Drying Facility (CVDF), a new facility in the 100 K Area of the Hanford Site. This report is contains the safety analysis for the Cold Vacuum Drying Facility, Phase 1

  1. Development of an auditable safety analysis in support of a radiological facility classification

    In recent years, U.S. Department of Energy (DOE) facilities commonly have been classified as reactor, non-reactor nuclear, or nuclear facilities. Safety analysis documentation was prepared for these facilities, with few exceptions, using the requirements in either DOE Order 5481.1B, Safety Analysis and Review System; or DOE Order 5480.23, Nuclear Safety Analysis Reports. Traditionally, this has been accomplished by development of an extensive Safety Analysis Report (SAR), which identifies hazards, assesses risks of facility operation, describes and analyzes adequacy of measures taken to control hazards, and evaluates potential accidents and their associated risks. This process is complicated by analysis of secondary hazards and adequacy of backup (redundant) systems. The traditional SAR process is advantageous for DOE facilities with appreciable hazards or operational risks. SAR preparation for a low-risk facility or process can be cost-prohibitive and quite challenging because conventional safety analysis protocols may not readily be applied to a low-risk facility. The DOE Office of Environmental Restoration and Waste Management recognized this potential disadvantage and issued an EM limited technical standard, No. 5502-94, Hazard Baseline Documentation. This standard can be used for developing documentation for a facility classified as radiological, including preparation of an auditable (defensible) safety analysis. In support of the radiological facility classification process, the Uranium Mill Tailings Remedial Action (UMTRA) Project has developed an auditable safety analysis document based upon the postulation criteria and hazards analysis techniques defined in DOE Order 5480.23

  2. Analysis of facilities in OFF research in participating countries of CORE Organic

    Nykänen, Arja; Canali, Stefano

    2006-01-01

    Report lists the following research facilities: research farms, experimental fields, on-farm studies, networks, animal research facilities, leaching fields and long-term experiments. Other facilities like facilities for laboratory analyses, food processing, greenhouses, climate chambers and growth cabinets are left out from this analysis, because they are seldom exclusively used for OFF research and because their use for OFF research does not require particular characteristics. On the other h...

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

    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. PMID:22154389

  4. Experiment archive, analysis, and visualization at the National Ignition Facility

    Hutton, Matthew S., E-mail: hutton1@llnl.gov [Lawrence Livermore National Laboratory, Livermore, CA (United States); Azevedo, Stephen; Beeler, Richard; Bettenhausen, Rita; Bond, Essex; Casey, Allan; Liebman, Judith; Marsh, Amber; Pannell, Thomas; Warrick, Abbie [Lawrence Livermore National Laboratory, Livermore, CA (United States)

    2012-12-15

    Highlights: Black-Right-Pointing-Pointer We show the computing architecture to manage scientific data from NIF experiments. Black-Right-Pointing-Pointer NIF laser 'shots' generate GBs of data for sub-microsec events separated by hours. Black-Right-Pointing-Pointer Results are archived, analyzed and displayed with parallel and scalable code. Black-Right-Pointing-Pointer Data quality and pedigree, based on calibration of each part, are tracked. Black-Right-Pointing-Pointer Web-based visualization tools present data across shots and diagnostics. - Abstract: The National Ignition Facility (NIF) at the Lawrence Livermore National Laboratory is the world's most energetic laser, providing a scientific research center to study inertial confinement fusion and matter at extreme energy densities and pressures. A target shot involves over 30 specialized diagnostics measuring critical x-ray, optical and nuclear phenomena to quantify ignition results for comparison with computational models. The Shot Analysis and Visualization System (SAVI) acquires and analyzes target diagnostic data for display within a time-budget of 30 min. Laser and target diagnostic data are automatically loaded into the NIF archive database through clustered software data collection agents. The SAVI Analysis Engine distributes signal and image processing tasks to a Linux cluster where computation is performed. Intermediate results are archived at each step of the analysis pipeline. Data is archived with metadata and pedigree. Experiment results are visualized through a web-based user interface in interactive dashboards tailored to single or multiple shot perspectives. The SAVI system integrates open-source software, commercial workflow tools, relational database and messaging technologies into a service-oriented and distributed software architecture that is highly parallel, scalable, and flexible. The architecture and functionality of the SAVI system will be presented along with examples.

  5. Experiment archive, analysis, and visualization at the National Ignition Facility

    Highlights: ► We show the computing architecture to manage scientific data from NIF experiments. ► NIF laser “shots” generate GBs of data for sub-microsec events separated by hours. ► Results are archived, analyzed and displayed with parallel and scalable code. ► Data quality and pedigree, based on calibration of each part, are tracked. ► Web-based visualization tools present data across shots and diagnostics. - Abstract: The National Ignition Facility (NIF) at the Lawrence Livermore National Laboratory is the world's most energetic laser, providing a scientific research center to study inertial confinement fusion and matter at extreme energy densities and pressures. A target shot involves over 30 specialized diagnostics measuring critical x-ray, optical and nuclear phenomena to quantify ignition results for comparison with computational models. The Shot Analysis and Visualization System (SAVI) acquires and analyzes target diagnostic data for display within a time-budget of 30 min. Laser and target diagnostic data are automatically loaded into the NIF archive database through clustered software data collection agents. The SAVI Analysis Engine distributes signal and image processing tasks to a Linux cluster where computation is performed. Intermediate results are archived at each step of the analysis pipeline. Data is archived with metadata and pedigree. Experiment results are visualized through a web-based user interface in interactive dashboards tailored to single or multiple shot perspectives. The SAVI system integrates open-source software, commercial workflow tools, relational database and messaging technologies into a service-oriented and distributed software architecture that is highly parallel, scalable, and flexible. The architecture and functionality of the SAVI system will be presented along with examples.

  6. IFMIF - International Fusion Materials Irradiation Facility Conceptual Design Activity/Interim Report

    Environmental acceptability, safety, and economic viability win ultimately be the keys to the widespread introduction of fusion power. This will entail the development of radiation- resistant and low- activation materials. These low-activation materials must also survive exposure to damage from neutrons having an energy spectrum peaked near 14 MeV with annual radiation doses in the range of 20 displacements per atom (dpa). Testing of candidate materials, therefore, requires a high-flux source of high energy neutrons. The problem is that there is currently no high-flux source of neutrons in the energy range above a few MeV. The goal, is therefore, to provide an irradiation facility for use by fusion material scientists in the search for low-activation and damage-resistant materials. An accellerator-based neutron source has been established through a number of international studies and workshops' as an essential step for materials development and testing. The mission of the International Fusion Materials Irradiation Facility (IFMIF) is to provide an accelerator-based, deuterium-lithium (D-Li) neutron source to produce high energy neutrons at sufficient intensity and irradiation volume to test samples of candidate materials up to about a full lifetime of anticipated use in fusion energy reactors. would also provide calibration and validation of data from fission reactor and other accelerator-based irradiation tests. It would generate material- specific activation and radiological properties data, and support the analysis of materials for use in safety, maintenance, recycling, decommissioning, and waste disposal systems

  7. IFMIF - International Fusion Materials Irradiation Facility Conceptual Design Activity/Interim Report

    Rennich, M.J.

    1995-12-01

    Environmental acceptability, safety, and economic viability win ultimately be the keys to the widespread introduction of fusion power. This will entail the development of radiation- resistant and low- activation materials. These low-activation materials must also survive exposure to damage from neutrons having an energy spectrum peaked near 14 MeV with annual radiation doses in the range of 20 displacements per atom (dpa). Testing of candidate materials, therefore, requires a high-flux source of high energy neutrons. The problem is that there is currently no high-flux source of neutrons in the energy range above a few MeV. The goal, is therefore, to provide an irradiation facility for use by fusion material scientists in the search for low-activation and damage-resistant materials. An accellerator-based neutron source has been established through a number of international studies and workshops` as an essential step for materials development and testing. The mission of the International Fusion Materials Irradiation Facility (IFMIF) is to provide an accelerator-based, deuterium-lithium (D-Li) neutron source to produce high energy neutrons at sufficient intensity and irradiation volume to test samples of candidate materials up to about a full lifetime of anticipated use in fusion energy reactors. would also provide calibration and validation of data from fission reactor and other accelerator-based irradiation tests. It would generate material- specific activation and radiological properties data, and support the analysis of materials for use in safety, maintenance, recycling, decommissioning, and waste disposal systems.

  8. Instrument Systems Analysis and Verification Facility (ISAVF) users guide

    Davis, J. F.; Thomason, J. O.; Wolfgang, J. L.

    1985-01-01

    The ISAVF facility is primarily an interconnected system of computers, special purpose real time hardware, and associated generalized software systems, which will permit the Instrument System Analysts, Design Engineers and Instrument Scientists, to perform trade off studies, specification development, instrument modeling, and verification of the instrument, hardware performance. It is not the intent of the ISAVF to duplicate or replace existing special purpose facilities such as the Code 710 Optical Laboratories or the Code 750 Test and Evaluation facilities. The ISAVF will provide data acquisition and control services for these facilities, as needed, using remote computer stations attached to the main ISAVF computers via dedicated communication lines.

  9. Analysis on the Present Status of Conceptually Designed Pyroprocessing Facilities for Determining a Reference Pyroprocessing Facility

    In this report, pyro processing facility concepts suggested by US, Japan, and Republic of Korea have been summarized and analyzed, and the determination principles were established to determine a reference pyro processing facility concept. Three proposals for a reference pyro processing facility concept were suggested based on these principles. The 1st proposal is based on the GEN-IV PR/PP model except the metal fuel fabrication process. It may be possible to later add the metal fuel fabrication process, UO2 recovery process of Japan, and continuous electrorefining process invented in Republic of Korea to be the generic model including all pyroprocessing facility concepts in the world. The 2nd proposal is based on INL and ANL model which is simple for the most part and has basic essential processes. The 3rd proposal is determined to be the ESPF of KAERI, which is almost identical with that of the 2nd proposal except in regards to utilization of an input accountability tank and continuous electrorefining process and the 3rd proposal is planned to be realized in 7 years. After the review of the IAEA and discussions at 3rd Working Group Meeting held in IAEA headquarters, the 3rd proposal has been determined as the final version of a reference pyroprocessing facility concept

  10. Activation studies of the light ion beam target development facility

    Biological dose calculations have been performed for the target chamber of the Target Development Facility (TDF). Placement of an neutron moderator structure in the interior of the target chamber for the moderation of the high energy neutrons has been investigated as a viable option for lowering the biological dose rates of the chamber wall materials, Al6061-T6 and 2 1/4Cr-1Mo steel. Two moderator materials are considered, one made of H-451 graphite and the other of titanium hydride. In particular, a 40% porosity, 1 m thick graphite structure within the aluminum wall reduces the dose rate at the chamber wall outer surface to 13.1 mrem/h at 1 week after shutdown as compared to 1.29 rem/h without the moderator. A suitable maintenance schedule based on the 40% porosity graphite moderator design and on the allowable average dose of 1.25 rem per quarter is presented. (orig.)

  11. Reactor Accident Analysis Methodology for the Advanced Test Reactor Critical Facility Documented Safety Analysis Upgrade

    The regulatory requirement to develop an upgraded safety basis for a DOE Nuclear Facility was realized in January 2001 by issuance of a revision to Title 10 of the Code of Federal Regulations Section 830 (10 CFR 830). Subpart B of 10 CFR 830, ''Safety Basis Requirements,'' requires a contractor responsible for a DOE Hazard Category 1, 2, or 3 nuclear facility to either submit by April 9, 2001 the existing safety basis which already meets the requirements of Subpart B, or to submit by April 10, 2003 an upgraded facility safety basis that meets the revised requirements. 10 CFR 830 identifies Nuclear Regulatory Commission (NRC) Regulatory Guide 1.70, ''Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants'' as a safe harbor methodology for preparation of a DOE reactor documented safety analysis (DSA). The regulation also allows for use of a graded approach. This report presents the methodology that was developed for preparing the reactor accident analysis portion of the Advanced Test Reactor Critical Facility (ATRC) upgraded DSA. The methodology was approved by DOE for developing the ATRC safety basis as an appropriate application of a graded approach to the requirements of 10 CFR 830

  12. Glovebox dismantling activities and decommissioning plan for plutonium fuel fabrication facility

    The gloveboxes and process equipment used at plutonium fuel handling facilities have had to be replaced due to deterioration or the need to make changes. So far, their removal and replacement has taken place more than 30 times in Plutonium Fuel Center, Japan Nuclear Cycle Development Institute (JNC). In most recent dismantling activities, we removed four giant gloveboxes (total size, 110 cubic meters) which possessed equipment to recover plutonium from mixed oxide (MOX) fuel scraps. We have implemented a number of procedural improvements in dismantling activities and collected various kinds of data, including type and amount of primary and secondary waste from dismantling, relation between waste volume and work force, etc. Plutonium Fuel Fabricating Facility (PFFF) is one of the three plutonium fuel handling facilities in Plutonium Fuel Center, JNC. Its final mission to produce MOX fuels for the advanced thermal reactor 'Fugen' Nuclear Power Station was successfully finished in 2002. Then, we started preparatory activities to draw up a Deactivation and Decommissioning (DD) plan for this facility and to construct a database with the experimental data of glovebox dismantling activities acquired in the past thirty years. The DD schedule for this facility can be broken down into three phases. Phase 1 (to 2010): Stabilization of all the special nuclear materials in the facility and remove them from the facility. Establish new and effective decontamination and volume reduction technologies in order to improve existing methods. Phase 2 (2010-2015): Applying the above-mentioned technologies to some of the glovebox dismantling activities and confirm their adaptability for the project. Draw up a detailed DD plan which meets to various regulations. Phase 3 (2015-2020): Dismantling of all the remaining gloveboxes in the facility and promote research and development of DD technologies for future projects. Decontamination of inner surfaces of the building in order to reuse the

  13. Calibration Report for the WRAP Facility Gamma Energy Analysis System

    The Waste Receiving And Processing facility (WRAP) adheres to providing gamma-ray spectroscopy instrument calibrations traceable to the National Institute for Standards and Technology (NIST) standards. The detectors are used to produce quantitative results for the Waste Isolation Pilot Plant (WIPP) and must meet calibration programmatic calibration goals. Instruments must meet portions of ANSI N42.14, 1978 guide for Germanium detectors. The Non-Destructive Assay (NDA) Gamma Energy Analysis (GEA) utilizes NIST traceable line source standards for the detector system calibrations. The counting configuration is a series of drums containing the line sources and different density filler matrices. The drums are used to develop system efficiencies with respect to density. The efficiency and density correction factors are required for the processing of drummed waste materials of similar densities. The calibration verification is carried out after the calibration is deemed final, by counting a second drum of NIST traceable sources. Three in-depth calibrations have been completed on one of the two systems to date, the first being the system acceptance plan. This report has a secondary function; that being the development of the instrument calibration errors which are to be folded into the Total Instrument Uncertainty document, HNF-4050

  14. Calibration Report for the WRAP Facility Gamma Energy Analysis System

    The Waste Receiving And Processing facility (WRAP) adheres to providing gamma-ray spectroscopy instrument calibrations traceable to the National Institute for Standards and Technology (NIST) standard(4). The detectors are used to produce quantitative results for the Waste Isolation Pilot Plant (WIPP) and must meet calibration programmatic calibration goals. Instruments must meet portions of ANSI N42.14, 1978 guide for Germanium detectors. The Non-Destructive Assay (NDA) Gamma Energy Analysis (GEA) utilizes NIST traceable line source standards for the detector system calibrations. The counting configuration is a series of drums containing the line sources and different density filler matrices. The drums are used to develop system efficiencies with respect to density. The efficiency and density correction factors are required for the processing of drummed waste materials of similar densities. The calibration verification is carried out after the calibration is deemed final, by counting a second drum of NIST traceable sources. Three in-depth calibrations have been completed on one of the two systems to date, the first being the system acceptance plan. This report has a secondary function; that being the development of the instrument calibration errors which are to be folded into the Total Instrument Uncertainty document, HNF-4050

  15. Application of prompt gamma-ray activation analysis

    Chung, Yong Sam; Park, Kwang Won; Moon, Jong Hwa; Kim, Sun Ha; Baek, Sung Ryel [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2002-03-01

    This technical report is written for the promotion to utilization of prompt gamma-ray activation analysis facility to be installed in HANARO reactor. It is described for a practical aspects including experiment and equipments, methodology, current status of the research and development and its applications. 102 refs., 32 figs., 25 tabs. (Author)

  16. Instrumental Neutron Activation Analysis Technique using Subsecond Radionuclides

    Nielsen, H.K.; Schmidt, J.O.

    1987-01-01

    The fast irradiation facility Mach-1 installed at the Danish DR 3 reactor has been used in boron determinations by means of Instrumental Neutron Activation Analysis using12B with 20-ms half-life. The performance characteristics of the system are presented and boron determinations of NBS standard...

  17. Analysis of the National Ignition Facility ignition hohlraum energetics experiments

    A series of 40 experiments on the National Ignition Facility (NIF) [E. I. Moses et al., Phys. Plasmas 16, 041006 (2009)] to study energy balance and implosion symmetry in reduced- and full-scale ignition hohlraums was shot at energies up to 1.3 MJ. This paper reports the findings of the analysis of the ensemble of experimental data obtained that has produced an improved model for simulating ignition hohlraums. Last year the first observation in a NIF hohlraum of energy transfer between cones of beams as a function of wavelength shift between those cones was reported [P. Michel et al., Phys. Plasmas 17, 056305 (2010)]. Detailed analysis of hohlraum wall emission as measured through the laser entrance hole (LEH) has allowed the amount of energy transferred versus wavelength shift to be quantified. The change in outer beam brightness is found to be quantitatively consistent with LASNEX [G. B. Zimmerman and W. L. Kruer, Comments Plasma Phys. Controlled Fusion 2, 51 (1975)] simulations using the predicted energy transfer when possible saturation of the plasma wave mediating the transfer is included. The effect of the predicted energy transfer on implosion symmetry is also found to be in good agreement with gated x-ray framing camera images. Hohlraum energy balance, as measured by x-ray power escaping the LEH, is quantitatively consistent with revised estimates of backscatter and incident laser energy combined with a more rigorous non-local-thermodynamic-equilibrium atomic physics model with greater emissivity than the simpler average-atom model used in the original design of NIF targets.

  18. Setup for thin layer activation at BARC-TIFR Pelletron Accelerator Facility

    Layout of drift space above analysing magnet of BARC-TIFR Pelletron accelerator facility was modified in year 2003 to accommodate an irradiation setup in tower area known as 6M irradiation setup. Proton beam of a few MeV energy having current in range of hundreds of nA can be obtained at this port to carry out specific experiments. Irradiation setup was modified to mount metal samples of different shape and sizes to study wear and corrosion rates using thin layer activation analysis technique. Special jigs were fabricated to irradiate samples i.e. disc gears, balls and rectangular shape coupons. The samples were irradiated by a proton beam of 13 MeV energy having 200 nA beam current. The irradiation resulted in production of a gamma emitting radionuclide Cobalt-56 (half- life:77.3 d, Energy: 847 KeV, 1.24 MeV) by the nuclear reaction 56Fe (p, n) 56Co. The irradiated samples were subjected to wear/corrosion environment under certain experimental conditions and activity loss was monitored periodically using gamma spectrometer. The reduced activity was correlated with thickness loss by generating a calibration curve. Details of setup and activation results will be presented in paper. (author)

  19. Analysis Methods for Extracting Knowledge from Large-Scale WiFi Monitoring to Inform Building Facility Planning

    Ruiz-Ruiz, Antonio; Blunck, Henrik; Prentow, Thor Siiger;

    2014-01-01

    realistic data to inform facility planning. In this paper, we propose analysis methods to extract knowledge from large sets of network collected WiFi traces to better inform facility management and planning in large building complexes. The analysis methods, which build on a rich set of temporal and spatial....... Spatio-temporal visualization tools built on top of these methods enable planners to inspect and explore extracted information to inform facility-planning activities. To evaluate the methods, we present results for a large hospital complex covering more than 10 hectares. The evaluation is based on WiFi...... traces collected in the hospital’s WiFi infrastructure over two weeks observing around 18000 different devices recording more than a billion individual WiFi measurements. For the presented analysis methods we present quantitative performance results, e.g., demonstrating over 95% accuracy for correct...

  20. Abbreviated sampling and analysis plan for planning decontamination and decommissioning at Test Reactor Area (TRA) facilities

    The objective is to sample and analyze for the presence of gamma emitting isotopes and hazardous constituents within certain areas of the Test Reactor Area (TRA), prior to D and D activities. The TRA is composed of three major reactor facilities and three smaller reactors built in support of programs studying the performance of reactor materials and components under high neutron flux conditions. The Materials Testing Reactor (MTR) and Engineering Test Reactor (ETR) facilities are currently pending D/D. Work consists of pre-D and D sampling of designated TRA (primarily ETR) process areas. This report addresses only a limited subset of the samples which will eventually be required to characterize MTR and ETR and plan their D and D. Sampling which is addressed in this document is intended to support planned D and D work which is funded at the present time. Biased samples, based on process knowledge and plant configuration, are to be performed. The multiple process areas which may be potentially sampled will be initially characterized by obtaining data for upstream source areas which, based on facility configuration, would affect downstream and as yet unsampled, process areas. Sampling and analysis will be conducted to determine the level of gamma emitting isotopes and hazardous constituents present in designated areas within buildings TRA-612, 642, 643, 644, 645, 647, 648, 663; and in the soils surrounding Facility TRA-611. These data will be used to plan the D and D and help determine disposition of material by D and D personnel. Both MTR and ETR facilities will eventually be decommissioned by total dismantlement so that the area can be restored to its original condition

  1. Abbreviated sampling and analysis plan for planning decontamination and decommissioning at Test Reactor Area (TRA) facilities

    NONE

    1994-10-01

    The objective is to sample and analyze for the presence of gamma emitting isotopes and hazardous constituents within certain areas of the Test Reactor Area (TRA), prior to D and D activities. The TRA is composed of three major reactor facilities and three smaller reactors built in support of programs studying the performance of reactor materials and components under high neutron flux conditions. The Materials Testing Reactor (MTR) and Engineering Test Reactor (ETR) facilities are currently pending D/D. Work consists of pre-D and D sampling of designated TRA (primarily ETR) process areas. This report addresses only a limited subset of the samples which will eventually be required to characterize MTR and ETR and plan their D and D. Sampling which is addressed in this document is intended to support planned D and D work which is funded at the present time. Biased samples, based on process knowledge and plant configuration, are to be performed. The multiple process areas which may be potentially sampled will be initially characterized by obtaining data for upstream source areas which, based on facility configuration, would affect downstream and as yet unsampled, process areas. Sampling and analysis will be conducted to determine the level of gamma emitting isotopes and hazardous constituents present in designated areas within buildings TRA-612, 642, 643, 644, 645, 647, 648, 663; and in the soils surrounding Facility TRA-611. These data will be used to plan the D and D and help determine disposition of material by D and D personnel. Both MTR and ETR facilities will eventually be decommissioned by total dismantlement so that the area can be restored to its original condition.

  2. Computer modeling for neutron activation analysis methods

    Full text: The INP AS RU develops databases for the neutron-activation analysis - ND INAA [1] and ELEMENT [2]. Based on these databases, the automated complex is under construction aimed at modeling of methods for natural and technogenic materials analysis. It is well known, that there is a variety of analysis objects with wide spectra, different composition and concentration of elements, which makes it impossible to develop universal methods applicable for every analytical research. The modelling is based on algorithm, that counts the period of time in which the sample was irradiated in nuclear reactor, providing the sample's total absorption and activity analytical peaks areas with given errors. The analytical complex was tested for low-elemental analysis (determination of Fe and Zn in vegetation samples, and Cu, Ag and Au - in technological objects). At present, the complex is applied for multielemental analysis of sediment samples. In this work, modern achievements in the analytical chemistry (measurement facilities, high-resolution detectors, IAEA and IUPAC databases) and information technology applications (Java software, database management systems (DBMS), internet technologies) are applied. Reference: 1. Tillaev T., Umaraliev A., Gurvich L.G., Yuldasheva K., Kadirova J. Specialized database for instrumental neutron activation analysis - ND INAA 1.0, The 3-rd Eurasian Conference Nuclear Science and its applications, 2004, pp.270-271.; 2. Gurvich L.G., Tillaev T., Umaraliev A. The Information-analytical database on the element contents of natural objects. The 4-th International Conference Modern problems of Nuclear Physics, Samarkand, 2003, p.337. (authors)

  3. Overview of the Facility Safeguardability Analysis (FSA) Process

    Bari, Robert A.; Hockert, John; Wonder, Edward F.; Johnson, Shirley J.; Wigeland, Roald; Zentner, Michael D.

    2011-10-10

    The safeguards system of the International Atomic Energy Agency (IAEA) provides the international community with credible assurance that a State is fulfilling its nonproliferation obligations. The IAEA draws such conclusions from the evaluation of all available information. Effective and cost-efficient IAEA safeguards at the facility level are, and will remain, an important element of this “State-level” approach. Efficiently used, the Safeguards by Design (SBD) methodologies , , , now being developed can contribute to effective and cost-efficient facility-level safeguards. The Facility Safeguardability Assessment (FSA) introduced here supports SBD in three areas. 1. It describes necessary interactions between the IAEA, the State regulator, and the owner / designer of a new or modified facility to determine where SBD efforts can be productively applied, 2. It presents a screening approach intended to identify potential safeguard issues for; a) design changes to existing facilities; b) new facilities similar to existing facilities with approved safeguards approaches, and c) new designs, 3. It identifies resources (the FSA toolkit), such as good practice guides, design guidance, and safeguardability evaluation methods that can be used by the owner/designer to develop solutions for potential safeguards issues during the interactions with the State regulator and IAEA. FSA presents a structured framework for the application of the SBD tools developed in other efforts. The more a design evolves, the greater the probability that new safeguards issues could be introduced. Likewise, for first-of-a-kind facilities or research facilities that involve previously unused processes or technologies, it is reasonable to expect that a number of possible safeguards issues might exist. Accordingly, FSA is intended to help the designer and its safeguards experts identify early in the design process: • Areas where elements of previous accepted safeguards approach(es) may be applied

  4. Waste management facility accident analysis (WASTE ACC) system: software for analysis of waste management alternatives

    This paper describes the Waste Management Facility Accident Analysis (WASTEunderscoreACC) software, which was developed at Argonne National Laboratory (ANL) to support the US Department of Energy's (DOE's) Waste Management (WM) Programmatic Environmental Impact Statement (PEIS). WASTEunderscoreACC is a decision support and database system that is compatible with Microsoft reg-sign Windows trademark. It assesses potential atmospheric releases from accidents at waste management facilities. The software provides the user with an easy-to-use tool to determine the risk-dominant accident sequences for the many possible combinations of process technologies, waste and facility types, and alternative cases described in the WM PEIS. In addition, its structure will allow additional alternative cases and assumptions to be tested as part of the future DOE programmatic decision-making process. The WASTEunderscoreACC system demonstrates one approach to performing a generic, systemwide evaluation of accident risks at waste management facilities. The advantages of WASTEunderscoreACC are threefold. First, the software gets waste volume and radiological profile data that were used to perform other WM PEIS-related analyses directly from the WASTEunderscoreMGMT system. Second, the system allows for a consistent analysis across all sites and waste streams, which enables decision makers to understand more fully the trade-offs among various policy options and scenarios. Third, the system is easy to operate; even complex scenario runs are completed within minutes

  5. Analysis of a nuclear accident: fission and activation product releases from the Fukushima Daiichi nuclear facility as remote indicators of source identification, extent of release, and state of damaged spent nuclear fuel.

    Schwantes, Jon M; Orton, Christopher R; Clark, Richard A

    2012-08-21

    Researchers evaluated radionuclide measurements of environmental samples taken from the Fukushima Daiichi nuclear facility and reported on the Tokyo Electric Power Co. Website following the 2011 tsunami-initiated catastrophe. This effort identified Units 1 and 3 as the major source of radioactive contamination to the surface soil near the facility. Radionuclide trends identified in the soils suggested that: (1) chemical volatility driven by temperature and reduction potential within the vented reactors' primary containment vessels dictated the extent of release of radiation; (2) all coolant had likely evaporated by the time of venting; and (3) physical migration through the fuel matrix and across the cladding wall were minimally effective at containing volatile species, suggesting damage to fuel bundles was extensive. Plutonium isotopic ratios and their distance from the source indicated that the damaged reactors were the major contributor of plutonium to surface soil at the source, decreasing rapidly with distance from the facility. Two independent evaluations estimated the fraction of the total plutonium inventory released to the environment relative to cesium from venting Units 1 and 3 to be ∼0.002-0.004%. This study suggests significant volatile radionuclides within the spent fuel at the time of venting, but not as yet observed and reported within environmental samples, as potential analytes of concern for future environmental surveys around the site. The majority of the reactor inventories of isotopes of less volatile elements like Pu, Nb, and Sr were likely contained within the damaged reactors during venting. PMID:22680069

  6. An analysis of heating, ventilation and air conditioning system for nuclear facilities

    An analysis of HVAC system was made on various nuclear facilities such as the existing nuclear power plants in Korea, Post Irradiation Examination Facility at KAERI and Midwest Fuel Recovery Plant in USA, to get basic data and information for the design of the spent fuel interim storage facility to be implemented as one of the radwaste management projects. With the results of this study, the HVAC system to be applied to the spent fuel interim storage facility was selected and the major design considerations of the facility were suggested. (Author)

  7. Survey on Neutron Activation Analysis Activities at the Dalat Nuclear Research Centre

    The Dalat Nuclear Research Centre (D.N.R.C.) during the past few years has been involved in conducting an activation analysis service. Work has been carried out in collaboration with other organizations. However, no rigid research programme of our own has been established and thus the Radiochemistry Division of the D.N.R.C. has no personnel and special facilities permanently engaged in this field. The equipment and facilities used are shared with other activities within the Division and the whole Centre. The activities in neutron activation analysis at the D.N.R.C. are sporadically revived by requests for analysis. Up to now, such analyses have been performed free of charge. Most of the work was carried out on biological materials such as vegetables, raw natural rubber (RES smoked sheets of different qualities, crepes and dried rubber films obtained from concentrated natural latex) from hevea tree leaves from various clones

  8. Charged-particle activation analysis

    The paper discusses the methodology and application of nuclear activation with ion beams (19 via 16O(3He,p)18F, 12C(3He,α)11C and 14N(p,α)11C respectively. Recently, triton activation has been shown to be inherently still superior to 3He activation for the determination of oxygen [16O(3H,n)18F]. Lithium, boron, carbon and sulphur can be detected rapidly, nondestructively and with high sensitivity (approximately 0.25ppm for Li and B) via ''quasi-prompt'' activation based on the detection of short-lived, high-energy beta emitters (10ms1H(7Li,n)7Be for example. Nondestructive multielement analysis: Proton activation has the inherent potential for meeting requirements of broad elemental coverage, sensitivity (ppm and sub-ppm range) and selectivity. Up to 30 elements have been determined in Al, Co, Ag, Nb, Rh, Ta and biological samples, using 12-MeV proton activation followed by gamma-ray spectrometry. These capabilities are further enhanced with the counting of X-ray emitters, 28 elements (269) and accuracy using proton activation. 204Pb/206Pb ratios can also be determined with a relative precision of a few per cent. Although charged-particle activation analysis is a well-established trace analysis technique, broad potential capabilities remain to be explored, e.g. those arising from ultrashort-lived nuclides, heavy ion interactions and the combination of delayed and prompt methods. (author)

  9. Automated activation-analysis system

    An automated delayed neutron counting and instrumental neutron activation analysis system has been developed at Los Alamos National Laboratory's Omega West Reactor (OWR) to analyze samples for uranium and 31 additional elements with a maximum throughput of 400 samples per day. The system and its mode of operation for a large reconnaissance survey are described

  10. Neutron activation analysis of coins

    Activation analysis was applied to the study of coins using 14MeV neutrons produced by an accelerator for the determination of oxygen and neutrons emitted from a 252Cf source for the determination of the other elements (Au, Ag, Cu, As etc...). The advantages of this technique are presented

  11. Economic Analysis for the Establishment of a Dry Dates Irradiation Facility at El Wadi - El Gedid Governorate

    The present study discus the economic analysis of the establishing dry dates irradiation facility at El Wadi El Gadid governorate. This study was divided into three sections the first section includes the arrangement of the equation of simple regression foretelling the future production for dry dates and radiation source activity , the second section was studied the financial analysis for the project. The third section includes the suitable commodities mix to full off the capacity

  12. GAS MIXING ANALYSIS IN A LARGE-SCALED SALTSTONE FACILITY

    Lee, S

    2008-05-28

    Computational fluid dynamics (CFD) methods have been used to estimate the flow patterns mainly driven by temperature gradients inside vapor space in a large-scaled Saltstone vault facility at Savannah River site (SRS). The purpose of this work is to examine the gas motions inside the vapor space under the current vault configurations by taking a three-dimensional transient momentum-energy coupled approach for the vapor space domain of the vault. The modeling calculations were based on prototypic vault geometry and expected normal operating conditions as defined by Waste Solidification Engineering. The modeling analysis was focused on the air flow patterns near the ventilated corner zones of the vapor space inside the Saltstone vault. The turbulence behavior and natural convection mechanism used in the present model were benchmarked against the literature information and theoretical results. The verified model was applied to the Saltstone vault geometry for the transient assessment of the air flow patterns inside the vapor space of the vault region using the potential operating conditions. The baseline model considered two cases for the estimations of the flow patterns within the vapor space. One is the reference nominal case. The other is for the negative temperature gradient between the roof inner and top grout surface temperatures intended for the potential bounding condition. The flow patterns of the vapor space calculated by the CFD model demonstrate that the ambient air comes into the vapor space of the vault through the lower-end ventilation hole, and it gets heated up by the Benard-cell type circulation before leaving the vault via the higher-end ventilation hole. The calculated results are consistent with the literature information. Detailed results and the cases considered in the calculations will be discussed here.

  13. Sensitivity analysis for activation problems

    A study has been made about how to develop further the techniques for sensitivity analysis used by FISPACT-II. FISPACT-II is a software suite for the analysis of nuclear activation and transmutation problems, developed for all nuclear applications. The software already permits sensitivity analysis to be performed by Monte Carlo sampling, and a faster uncertainty analysis is made possible by a powerful graph-based approach which generates a reduced set of nuclides on pathways leading to significant contributions to radiological quantities. The peculiar aspects of the sensitivity analysis problem for activation are the large number, typically thousands, of rate equation parameters (decay rates and reaction cross-sections) which all have some degree of associated error, and the fact that activity as a function of time varies as a sum of exponentials, so appears discontinuous as rate parameters are varied unless the sampling frequency is impracticably fast. Nevertheless, Monte Carlo sampling is a generic approach and it is therefore conceivable that techniques more targeted to the activation problem might be beneficial. Moreover, recent theoretical developments have highlighted the importance of a two-stage approach to mathematically similar problems, where in the first stage, information is collected about the global behaviour of the problem, such as the identification of the rate parameters which cause the greatest variation in dose or nuclear activity, before a second stage examines a problem with its scope restricted by the information from the first. In the second stage, for example, Quasi-Monte Carlo sampling may be used in a restricted parameter space. The current work concentrates on the first stage and consists of a review of possible techniques with a detailed examination of the most promising pathways reduction approach, examined directly using FISPACT-II. All the evidence obtained demonstrates the strong potential of this approach. (authors)

  14. Safety analysis of the Los Alamos critical experiments facility

    The safety of Pajarito Site critical assembly operations depends upon protection built into the facility, upon knowledgeable personnel, and upon good practice as defined by operating procedures and experimental plans. Distance, supplemented by shielding in some cases, would protect personnel against an extreme accident generating 1019 fissions. During the facility's 28-year history, the direct cost of criticality accidents has translated to a risk of less than $200 per year

  15. Stanford Synchrotron Radiation Laboratory 1991 activity report. Facility developments January 1991--March 1992

    SSRL is a national facility supported primarily by the Department of Energy for the utilization of synchrotron radiation for basic and applied research in the natural sciences and engineering. It is a user-oriented facility which welcomes proposals for experiments from all researchers. The synchrotron radiation is produced by the 3.5 GeV storage ring, SPEAR, located at the Stanford Linear Accelerator Center (SLAC). SPEAR is a fully dedicated synchrotron radiation facility which operates for user experiments 7 to 9 months per year. SSRL currently has 24 experimental stations on the SPEAR storage ring. There are 145 active proposals for experimental work from 81 institutions involving approximately 500 scientists. There is normally no charge for use of beam time by experimenters. This report summarizes the activity at SSRL for the period January 1, 1991 to December 31, 1991 for research. Facility development through March 1992 is included

  16. The interim storage facility with dry storage casks and its safeguards activity

    Recyclable-Fuel Storage Company (RFS) is constructing an interim storage facility of spent fuel at Recyclable-Fuel Storage Center (RFSC) in Aomori Prefecture. Metallic dry casks are employed to contain the spent fuel from nuclear power plants and to serve for about 50 years in RFSC. Metallic dry casks have already been used for dry cask storage facility at Tokai No.2 power station of Japan Atomic Power Company. But, RFSC is not exactly the same as the dry cask storage facility at Tokai No.2 power station, for example, cask transportation between facilities and no hot cells. Therefore, additional safeguards activities are necessary. The outline of the design and handling of metallic dry casks at RFSC and the currently developing status of safeguards activity such as containment and surveillance for the cask receipt and storage at RFSC, etc are described. (author)

  17. Stanford Synchrotron Radiation Laboratory 1991 activity report. Facility developments January 1991--March 1992

    Cantwell, K.; St. Pierre, M. [eds.

    1992-12-31

    SSRL is a national facility supported primarily by the Department of Energy for the utilization of synchrotron radiation for basic and applied research in the natural sciences and engineering. It is a user-oriented facility which welcomes proposals for experiments from all researchers. The synchrotron radiation is produced by the 3.5 GeV storage ring, SPEAR, located at the Stanford Linear Accelerator Center (SLAC). SPEAR is a fully dedicated synchrotron radiation facility which operates for user experiments 7 to 9 months per year. SSRL currently has 24 experimental stations on the SPEAR storage ring. There are 145 active proposals for experimental work from 81 institutions involving approximately 500 scientists. There is normally no charge for use of beam time by experimenters. This report summarizes the activity at SSRL for the period January 1, 1991 to December 31, 1991 for research. Facility development through March 1992 is included.

  18. An in situ neutron irradiation facility for non-destructive elemental analysis

    Design and construction of an irradiation facility for nondestructive neutron activation analysis have been investigated. Two facilities have been constructed, using the available 5 Ci, Pu-Be, neutron sources and paraffin as modderator. The first unit has been designed mainly for neutron flux distribution studies. This unit consists of paraffin bricks of various sizes to facilitate placing the Dy, In and gold foils for flux maping. The second unit consists of a single paraffin block 50 x 50 x 50 cm3, having the same dimensions as the first unit. Two horizontal beam ports and one vertical were constructed in the paraffin moderator for sample irradiation. The maximum total absolute flux measured in the median plane was found to be about 1.32 x 104 n/cm2. Sec. at a distance of about 3.5 cm from the source surface. For test on geological sample of commercial interest gold sample were irradiated and measured. More than 10 elements have been clearly identified. Gold which is known to exist in this sample at about 10 parts per million was also clearly observed. This work is a part of joint programme with the international Atomic Energy Agency Research Contract No. 1697/RB. The title of this project is Elemental Activation Analysis with Decay and Prompt Gamma Ray Techniques, Using Isotopic Neutron Sources and Nuclear Research Reactors

  19. Immobilized low-activity waste interim storage facility, Project W-465 conceptual design report

    Pickett, W.W.

    1997-12-30

    This report outlines the design and Total Estimated Cost to modify the four unused grout vaults for the remote handling and interim storage of immobilized low-activity waste (ILAW). The grout vault facilities in the 200 East Area of the Hanford Site were constructed in the 1980s to support Tank Waste disposal activities. The facilities were to serve project B-714 which was intended to store grouted low-activity waste. The existing 4 unused grout vaults, with modifications for remote handling capability, will provide sufficient capacity for approximately three years of immobilized low activity waste (ILAW) production from the Tank Waste Remediation System-Privatization Vendors (TWRS-PV). These retrofit modifications to the grout vaults will result in an ILAW interim storage facility (Project W465) that will comply with applicable DOE directives, and state and federal regulations.

  20. Immobilized low-activity waste interim storage facility, Project W-465 conceptual design report

    This report outlines the design and Total Estimated Cost to modify the four unused grout vaults for the remote handling and interim storage of immobilized low-activity waste (ILAW). The grout vault facilities in the 200 East Area of the Hanford Site were constructed in the 1980s to support Tank Waste disposal activities. The facilities were to serve project B-714 which was intended to store grouted low-activity waste. The existing 4 unused grout vaults, with modifications for remote handling capability, will provide sufficient capacity for approximately three years of immobilized low activity waste (ILAW) production from the Tank Waste Remediation System-Privatization Vendors (TWRS-PV). These retrofit modifications to the grout vaults will result in an ILAW interim storage facility (Project W465) that will comply with applicable DOE directives, and state and federal regulations

  1. Facile synthesis and photocatalytic activity of zinc oxide hierarchical microcrystals

    Xu, Xinjiang

    2013-04-04

    ZnO microcrystals with hierarchical structure have been synthesized by a simple solvothermal approach. The microcrystals were studied by means of X-ray diffraction, transmission electron microscopy, and scanning electron microscopy. Research on the formation mechanism of the hierarchical microstructure shows that the coordination solvent and precursor concentration have considerable influence on the size and morphology of the microstructures. A possible formation mechanism of the hierarchical structure was suggested. Furthermore, the catalytic activity of the ZnO microcrystals was studied by treating low concentration Rhodamine B (RhB) solution under UV light, and research results show the hierarchical microstructures of ZnO display high catalytic activity in photocatalysis, the catalysis process follows first-order reaction kinetics, and the apparent rate constant k = 0.03195 min-1.

  2. Transfer station and storage facility for medium active solid waste

    In the Muehlheim Kaerlich nuclear power plant a temporary store has been erected for production-related medium active solid waste (MAW). This paper reports that the following were the main criteria behind this step: creating a temporary store capacity for specific MAW for the duration of the entire operating time, the temporary storage of MAW over several years to achieve an appropriate dose reduction (decay period), and market independence

  3. Facile synthesis of novel benzotriazole derivatives and their antibacterial activities

    Jun Wan; Peng-Cheng Lv; Na-Na Tian; Hai-Liang Zhu

    2010-07-01

    A series of benzotriazole derivatives (compounds 1-27) were synthesized, and 24 (compounds 1-5, 9-27) of which were first reported. Their chemical structures were confirmed by means of 1H NMR, IR and elemental analyses, coupled with one selected single crystal structure (compound 1). All the compounds were assayed for antibacterial activities against three Gram positive bacterial strains (Bacillus subtilis, Staphylococcus aureus and Streptococcus faecalis) and three Gram negative bacterial strains (Escherichia coli, Pseudomonas aeruginosa and Enterobacter cloacae) by MTT method. Among the compounds tested, most of them exhibited potent antibacterial activity against the six bacterial strains. Most importantly, compound 3-benzotriazol-1-yl-1-(4-bromo-phenyl)-2-[1,2,4]triazol-1-ylpropan-1-one (19) showed the most favourable antibacterial activity against B. subtilis, S. aureus, S. faecalis, P. aeruginosa, E. coli and E. cloacae with MIC of 1.56 g/mL, 1.56 g/mL, 1.56 g/mL, 3.12 g/mL, 6.25 g/mL and 6.25 g/mL, respectively.

  4. Los Alamos National Laboratory corregated metal pipe saw facility preliminary safety analysis report. Volume I

    NONE

    1990-09-19

    This Preliminary Safety Analysis Report addresses site assessment, facility design and construction, and design operation of the processing systems in the Corrugated Metal Pipe Saw Facility with respect to normal and abnormal conditions. Potential hazards are identified, credible accidents relative to the operation of the facility and the process systems are analyzed, and the consequences of postulated accidents are presented. The risk associated with normal operations, abnormal operations, and natural phenomena are analyzed. The accident analysis presented shows that the impact of the facility will be acceptable for all foreseeable normal and abnormal conditions of operation. Specifically, under normal conditions the facility will have impacts within the limits posted by applicable DOE guidelines, and in accident conditions the facility will similarly meet or exceed the requirements of all applicable standards. 16 figs., 6 tabs.

  5. Activation analysis using γ photons

    This report summarizes all the data required for using photonuclear reactions in the field of analysis. After a brief review of the elementary properties of nuclear reactions induced by photon irradiation, the main characteristics are given of high energy (E > 20 MeV) Bremsstrahlung sources. The principle of activation analysis based on the use of photons is given. Actual examples of the analytic possibilities are described in detail, in particular in the case of the determination of very small quantities (-6) of C, N, O and F. The influence of interfering nuclear reactions is discussed. (author)

  6. State-of-the-art dry active waste processing facility

    Palo Verde Nuclear Generating Station (PVNGS) is operated by Arizona Public Service for a consortium of seven owners. The site consists of three identical single unit power plants. Each unit is a Combustion Engineering Series 80 pressurized water reactor (PWR) rated at 1270 Megawatts electric. The site is located 100 kilometers west of Phoenix, Arizona in the arid southwest desert region of the United States of America. Since the start up of Unit One in 1985, Palo Verde has aggressively pursued waste volume reduction. This includes a dry active waste (DAW) segregation program that locates and separates nonradioactive and reusable materials that have been mixed with the radioactive DAW. The DAW program is described in further detail in the paper

  7. FACILE SYNTHESIS, DOCKING STUDIES AND ANTIOXIDANT ACTIVITY OF FGVR

    Himaja M

    2011-08-01

    Full Text Available A rational designing of linear Tetrapeptide FGVR was done and was synthesized by solution phase peptide synthesis. The docking studies of designed linear tetrapeptide FGVR was carried out by using Schrodinger Software Solutions, USA. Qikprop results show the ligand FGVR mostly act as antihypertensive and anti coagulant properties. The solution phase synthesis of FGVR is carried out by using 1-Ethyl-3-(3-dimethylaminopropyl carbodiimide (EDC as coupling agents and N-Methyl morpholine (NMM as base. Structure of synthesized FGVR was confirmed by FTIR, 1H NMR and Mass spectral data, and evaluated for antioxidant property by using 1,1-diphenyl-2-picryl-hydrazil (DPPH. method and the synthesized peptides FGVR possess moderate antioxidant activity.

  8. Waste Sampling & Characterization Facility (WSCF) Complex Safety Analysis

    MELOY, R.T.

    2002-04-01

    This document was prepared to analyze the Waste Sampling and Characterization Facility for safety consequences by: Determining radionuclide and highly hazardous chemical inventories; Comparing these inventories to the appropriate regulatory limits; Documenting the compliance status with respect to these limits; and Identifying the administrative controls necessary to maintain this status. The primary purpose of the Waste Sampling and Characterization Facility (WSCF) is to perform low-level radiological and chemical analyses on various types of samples taken from the Hanford Site. These analyses will support the fulfillment of federal, Washington State, and Department of Energy requirements.

  9. Modeling and analysis of inertial-confinement-fusion facilities

    Approximate analytic models are used to explore relations among technical and economic characteristics of Inertial Confinement Fusion (ICF) facilities. Presented are attainable pulse rates for different reactor cavities and dependencies of the unit production cost of electricity on ICF driver pulse energy and repetition rate and on the facility size and the performance of the driver-pellet combination. The results indicate that economic electricity production with ICF reactors may require repetition rates of approx. 15 Hz or 20 Hz but that it may be achieved with values of the driver efficiency-pellet gain product as low as approx. 3 or 4

  10. Seismic design and analysis of nuclear fuel cycle facilities in France

    Methodology for seismic design of nuclear fuel facilities and power plants in France is described. After the description of regulatory and normative texts for seismic design, different elements are examined: definition of ground motion, analysis methods, new trends, reevaluation and specificity of Fuel Cycle Facilities. R/D developments are explicated in each part. Their final objective are to better quantify the margins of each step which, in relation with safety analysis,lead to balanced design, analysis and retrofit rules. (author)

  11. Post test analysis of counterpart tests in LOBI, SPES, BETHSY, LSTF facilities performed with the CATHARE2 code

    The present paper deals with the evaluation of results from the application of the thermalhydraulic system code CATHARE2V1.3U to the post test analysis of six co-called counterpart tests performed in four PWR facilities. These are LOBI/Mod2, SPES, BETHSY and LSTF facilities operated respectively at the European Research Centre of Ispra (Italy), at Piacenza (Italy), at Grenoble (France) and at Tokaj Mura (Japan). The activity is a part of a wide research having as main objective the evaluation of scaling capabilities of the CATHARE2 code. (author)

  12. Qualitative and quantitative analysis of CATHARE2 code results of counterpart test calculations in LOBI, SPES, BETHSY, LSTF facilities

    The present paper deals with the qualitative and quantitative accuracy evaluation of results from the application of the thermalhydraulic system code CATHARE2V1.3U to the post test analysis of six so-called counterpart tests performed in four PWR facilities. These are LOBI/Mod2, SPES, BETHSY and LSTF facilities operated respectively at the European Research Centre of Ispra (Italy), at Piacenza (Italy), at Grenoble (France) and at Tokai Mura (Japan). The activity is a part of a wide research having as main objective the evaluation of scaling capabilities of the CATHARE2 code. (author)

  13. Facile route to silver submicron-sized particles and their catalytic activity towards 4-nitrophenol reduction

    Research highlights: → Submicron-sized Ag particles can be prepared by using EDTA as a reducing agent. → By varying the amount of EDTA, the size of Ag particles can be controlled. → By varying the hydrothermal reaction time, the size of Ag particles can be controlled. → In comparison with Ag nanoparticles, the submicron-sized Ag particles have a comparable catalytic activity. - Abstract: A facile, efficient, and environmentally friendly synthetic route was developed to fabricate silver submicron-sized particles by reducing silver nitrate with EDTA in aqueous solution. X-ray diffraction (XRD), field emission scanning electron microscopy (FESEM), and transmission electron microscopy (TEM) analysis revealed the formation of silver particles, with sizes ranging from 100 to 800 nm. By varying the amount of EDTA utilized in the reaction medium and/or hydrothermal reaction time, the size of prepared silver particles can be readily controlled. Compared with silver nanoparticles, the as-synthesized submicron-sized silver particles were found to show a comparable catalytic activity towards the reduction of 4-nitrophenol to 4-aminophenol in the presence of an excess amount of NaBH4.

  14. Facile route to silver submicron-sized particles and their catalytic activity towards 4-nitrophenol reduction

    Jiang Deli [School of Chemistry and Chemical Engineering, Jiangsu University, Zhenjiang 212013 (China); Xie Jimin, E-mail: Xiejm391@sohu.com [School of Chemistry and Chemical Engineering, Jiangsu University, Zhenjiang 212013 (China); Chen Min; Li Di; Zhu Jianjun; Qin Huiru [School of Chemistry and Chemical Engineering, Jiangsu University, Zhenjiang 212013 (China)

    2011-02-03

    Research highlights: > Submicron-sized Ag particles can be prepared by using EDTA as a reducing agent. > By varying the amount of EDTA, the size of Ag particles can be controlled. > By varying the hydrothermal reaction time, the size of Ag particles can be controlled. > In comparison with Ag nanoparticles, the submicron-sized Ag particles have a comparable catalytic activity. - Abstract: A facile, efficient, and environmentally friendly synthetic route was developed to fabricate silver submicron-sized particles by reducing silver nitrate with EDTA in aqueous solution. X-ray diffraction (XRD), field emission scanning electron microscopy (FESEM), and transmission electron microscopy (TEM) analysis revealed the formation of silver particles, with sizes ranging from 100 to 800 nm. By varying the amount of EDTA utilized in the reaction medium and/or hydrothermal reaction time, the size of prepared silver particles can be readily controlled. Compared with silver nanoparticles, the as-synthesized submicron-sized silver particles were found to show a comparable catalytic activity towards the reduction of 4-nitrophenol to 4-aminophenol in the presence of an excess amount of NaBH{sub 4}.

  15. Facile Synthesis and Antimicrobial Evaluation of Some New Heterocyclic Compounds Incorporating a Biologically Active Sulfamoyl Moiety

    Elham S. Darwish

    2014-01-01

    Full Text Available A facile and convenient synthesis of new heterocyclic compounds containing a sulfamoyl moiety suitable for use as antimicrobial agents was reported. The precursor 3-oxo-3-phenyl-N-(4-sulfamoylphenylpropionamide was coupled smoothly with arenediazonium salt producing hydrazones which reacted with malononitrile or triethylorthoformate affording pyridazine and triazine derivatives, respectively. Also, the reactivity of the same precursor with DMF-DMA was followed by aminotriazole; aromatic aldehydes was followed by hydrazine hydrate, triethylorthoformate, or thiourea affording triazolo[1,5-a]pyrimidine, pyrazole, acrylamide, and dihydropyrimidine derivatives, respectively. On the other hand, treatment of the precursor propionamide with phenyl isothiocyanate and KOH in DMF afforded the intermediate salt which was treated with dilute HCl followed by 2-bromo-1-phenylethanone affording carboxamide derivative. While the same intermediate salt reacted in situ with chloroacetone, ethyl 2-chloroacetate, 3-(2-bromoacetyl-2H-chromen-2-one, methyl iodide, or 2-oxo-N-phenylpropane hydrazonoyl chloride afforded the thiophene, ketene N,S-acetal, and thiadiazole derivatives, respectively. The structure of the new products was established based on elemental and spectral analysis. Antimicrobial evaluation of some selected examples from the synthesized products was carried out whereby four compounds were found to have moderate activities and one compound showed the highest activity.

  16. Higher Education Facilities Systems Building Analysis. Summary Report.

    Texas A and M Univ., College Station. Coll. of Architecture and Environmental Design.

    This document is an extract of EA 004 095 which reports a yearlong study of possible benefits in cost, time, and facility utilization of a systems building approach for Texas college and university construction. (Photographs on pages 7,8,9, and 13 may reproduce poorly.) (Author)

  17. Higher Education Facilities: Systems Building Analysis. Documentary Work Report.

    Trost, F. J.

    This document reports a yearlong study of possible benefits in cost, time, and facility utilization of a systems building approach for Texas college and university construction. The first part of the report deals with trends and needs in higher education and the related architectural implications. A subsequent discussion of alternative building…

  18. A study on safety analysis methodology in spent fuel dry storage facility

    Collection and review of the domestic and foreign technology related to spent fuel dry storage facility. Analysis of a reference system. Establishment of a framework for criticality safety analysis. Review of accident analysis methodology. Establishment of accident scenarios. Establishment of scenario analysis methodology

  19. A study on safety analysis methodology in spent fuel dry storage facility

    Che, M. S.; Ryu, J. H.; Kang, K. M.; Cho, N. C.; Kim, M. S. [Hanyang Univ., Seoul (Korea, Republic of)

    2004-02-15

    Collection and review of the domestic and foreign technology related to spent fuel dry storage facility. Analysis of a reference system. Establishment of a framework for criticality safety analysis. Review of accident analysis methodology. Establishment of accident scenarios. Establishment of scenario analysis methodology.

  20. Introduction and preparation of the nuclear fuel cycle facility risk analysis code: STAR

    STAR code is a computer program, by which one can perform the probabilistic safety assessment (PSA) for the nuclear fuel cycle facility in both the normal and the accidental event of environmental radioactive material release. This code was originally developed by NUKEM GmbH in West Germany as a fruit of the PSE (Projekt Sicherheitsstudien Entsorgung) aiming at R and D of safety analysis methods for use in nuclear fuel cycle facilities such as reprocessing plants. In JAERI, efforts have been made to research and develop safety assessment methods applicable to the accidental situations assumed to happen in the reprocessing plants. In this line of objectives, the STAR code was introduced from NUKEM GmbH in 1986 and, since then, has been improved and prepared to add an ability to analyze public radiation exposure by released activities from the plants. At the first stage of this code preparation, the program conversion was made to adapt the STAR code, originally operative on IBM-compatible PC's and Hewlett Packard 7550A plotters, to NEC PC 9801RX and NEC PR 602R page printers installed in the Fuel Cycle Safety Assessment Laboratory of JAERI. This report describes calculational performances of the STAR code, results of the improvement and preparation works together with input/output data format in illustration of a sample HALW (High Activity Liquid Waste) tank PSA problem, thus making a users' manual for the STAR code. (author)

  1. Optimization of instrumental activation analysis

    Activation analysis is one of the most well-understood methods available to the analyst. It should, therefore, be possible to infer, from prior information about the sample, what procedure should be followed in its analysis. The accuracy of this process is naturally limited by the extent and accuracy of the prior information available. Better results should be obtained in this way, however, than by ignoring prior information. It is the task of optimization to discover the analytical procedure that best suits the sample being analyzed. Optimization can be conveniently conceptualized if each experimental parameter is considered as a dimension of a geometric space. In activation analysis, if only irradiation and decay times are to be adjusted, the parameter space will be two dimensional. Each point in the parameter space corresponds to a possible procedure for carrying out a determination and each such procedure will perform more or less satisfactorily than others. Optimization, then, consists of a search for a point or a region in parameter space where performance meets the analyst's requirements. Practicality is an important consideration in designing a procedure for activation analysis. There are limits to the amount of radioactive material that can be handled safely and to the count rate that be accurately measured. Circumstances often impose further limits. It is, therefore, necessary to constrain the search of parameter space to those regions that correspond to practical procedures. In attempting an optimization, one must consider a number of aspects. A set of experimental parameters must be chosen for adjustment and others set at fixed values, often due to practical constraints. The way in which quality of analytical performance (the response function) is to be evaluated must be decided. A means of locating the optimum must be chosen and, finally, this optimization scheme must be implemented in a practical, convenient manner. These aspects are discussed

  2. NASA Johnson Space Center Usability Testing and Analysis Facility (WAF) Overview

    Whitmore, M.

    2004-01-01

    The Usability Testing and Analysis Facility (UTAF) is part of the Space Human Factors Laboratory at the NASA Johnson Space Center in Houston, Texas. The facility provides support to the Office of Biological and Physical Research, the Space Shuttle Program, the International Space Station Program, and other NASA organizations. In addition, there are ongoing collaborative research efforts with external businesses and universities. The UTAF provides human factors analysis, evaluation, and usability testing of crew interfaces for space applications. This includes computer displays and controls, workstation systems, and work environments. The UTAF has a unique mix of capabilities, with a staff experienced in both cognitive human factors and ergonomics. The current areas of focus are: human factors applications in emergency medical care and informatics; control and display technologies for electronic procedures and instructions; voice recognition in noisy environments; crew restraint design for unique microgravity workstations; and refinement of human factors processes. This presentation will provide an overview of ongoing activities, and will address how the projects will evolve to meet new space initiatives.

  3. A new microcomputer-controlled neutron activation and analysis system

    A microcomputer-controlled irradiation and measurement system and a microprocessor-controlled sample changer have been installed at the SLOWPOKE-2 Facility at the Royal Military College of Canada (RMC). These systems can provide the gamut of instrumental neutron activation analysis (INAA) techniques for the analyst. Custom software has been created for system control, data acquisition, and off-line spectral analysis using programs that incorporate Gaussian peak-fitting methods of analysis. The design and use of the equipment is discussed, and the performance is illustrated with results obtained from the analysis of marine sediment and biological reference materials

  4. Analysis of the formation, expression, and economic impacts of risk perceptions associated with nuclear facilities

    Allison, T.; Hunter, S.; Calzonetti, F.J.

    1992-10-01

    This report investigates how communities hosting nuclear facilities form and express perceptions of risk and how these risk perceptions affect local economic development. Information was collected from site visits and interviews with plant personnel, officials of local and state agencies, and community activists in the hosting communities. Six commercial nuclear fuel production facilities and five nuclear facilities operated for the US Department of Energy by private contractors were chosen for analysis. The results presented in the report indicate that the nature of risk perceptions depends on a number of factors. These factors are (1) level of communication by plant officials within the local community, (2) track record of the facility. operator, (3) process through which community and state officials receive information and form opinions, (4) level of economic links each plant has with the local community, and (15) physical characteristics of the facility itself. This report finds that in the communities studied, adverse ask perceptions have not affected business location decisions, employment levels in the local community, tourism, or agricultural development. On the basis of case-study findings, this report recommends that nuclear facility siting programs take the following observations into account when addressing perceptions of risk. First, the quality of a facility`s participation with community activists, interest groups, and state agencies helps to determine the level of perceived risk within a community. Second, the development of strong economic links between nuclear facilities and their host communities will produce a higher level of acceptance of the nuclear facilities.

  5. Preclosure radiological safety analysis for the exploratory shaft facilities

    This study assesses which structures, systems, and components of the exploratory shaft facility (ESF) are important to safety when the ESF is converted to become part of the operating waste repository. The assessment follows the methodology required by DOE Procedure AP-6.10Q. Failures of the converted ESF during the preclosure period have been evaluated, along with other underground accidents, to determine the potential offsite radiation doses and associated probabilities. The assessment indicates that failures of the ESF will not result in radiation doses greater than 0.5 rem at the nearest unrestricted area boundary. Furthermore, credible accidents in other underground facilities will not result in radiation doses larger than 0.5 rem, even if any structure, system, or component of the converted ESF fails at the same time. Therefore, no structure, system, or component of the converted ESF is important to safety

  6. Overview of the Facility Safeguardability Analysis (FSA) Process

    Bari, Robert A.; Hockert, John; Wonder, Edward F.; Johnson, Scott J.; Wigeland, Roald; Zentner, Michael D.

    2012-08-01

    Executive Summary The safeguards system of the International Atomic Energy Agency (IAEA) is intended to provide the international community with credible assurance that a State is fulfilling its safeguards obligations. Effective and cost-efficient IAEA safeguards at the facility level are, and will remain, an important element of IAEA safeguards as those safeguards evolve towards a “State-Level approach.” The Safeguards by Design (SBD) concept can facilitate the implementation of these effective and cost-efficient facility-level safeguards (Bjornard, et al. 2009a, 2009b; IAEA, 1998; Wonder & Hockert, 2011). This report, sponsored by the National Nuclear Security Administration’s Office of Nuclear Safeguards and Security, introduces a methodology intended to ensure that the diverse approaches to Safeguards by Design can be effectively integrated and consistently used to cost effectively enhance the application of international safeguards.

  7. Beam positioning stability analysis on large laser facilities

    Fang; Liu; Zhigang; Liu; Liunian; Zheng; Hongbiao; Huang; Jianqiang; Zhu

    2013-01-01

    Beam positioning stability in a laser-driven inertial confinement fusion(ICF) facility is a vital problem that needs to be fixed. Each laser beam in the facility is transmitted in lots of optics for hundreds of meters, and then targeted in a micro-sized pellet to realize controllable fusion. Any turbulence in the environment in such long-distance propagation would affect the displacement of optics and further result in beam focusing and positioning errors. This study concluded that the errors on each of the optics contributed to the target, and it presents an efficient method of enhancing the beam stability by eliminating errors on error-sensitive optics. Optimizations of the optical system and mechanical supporting structures are also presented.

  8. Analysis of fuel management in the KIPT neutron source facility

    Research highlights: → Fuel management of KIPT ADS was analyzed. → Core arrangement was shuffled in stage wise. → New fuel assemblies was added into core periodically. → Beryllium reflector could also be utilized to increase the fuel life. - Abstract: Argonne National Laboratory (ANL) of USA and Kharkov Institute of Physics and Technology (KIPT) of Ukraine have been collaborating on the conceptual design development of an experimental neutron source facility consisting of an electron accelerator driven sub-critical assembly. The neutron source driving the sub-critical assembly is generated from the interaction of 100 KW electron beam with a natural uranium target. The sub-critical assembly surrounding the target is fueled with low enriched WWR-M2 type hexagonal fuel assemblies. The U-235 enrichment of the fuel material is <20%. The facility will be utilized for basic and applied research, producing medical isotopes, and training young specialists. With the 100 KW electron beam power, the total thermal power of the facility is ∼360 kW including the fission power of ∼260 kW. The burnup of the fissile materials and the buildup of fission products continuously reduce the system reactivity during the operation, decrease the neutron flux level, and consequently impact the facility performance. To preserve the neutron flux level during the operation, the fuel assemblies should be added and shuffled for compensating the lost reactivity caused by burnup. Beryllium reflector could also be utilized to increase the fuel life time in the sub-critical core. This paper studies the fuel cycles and shuffling schemes of the fuel assemblies of the sub-critical assembly to preserve the system reactivity and the neutron flux level during the operation.

  9. Analysis for cover design of a near surface disposal facility

    Model Simulation based on NETEC concept design of disposal facility was performed by using HELP Code and the infiltration rate through cover system was estimated with method of water budget. The final leakage through bottom of bentonite mixed layer estimated lower than 1.0mm/year. But long term integrity of geomembrane and asphalt as engineered barrier is not guaranteed yet. So assuming that these two barrier lost their function, the final leakage will be increased to 35 mm/year

  10. Lunar Mission Analysis for a Wallops Flight Facility Launch

    Dolan, John Martin

    2008-01-01

    Recently there is an increase in interest in the Moon as a destination for space missions. This increased interest is in the composition and geography of the Moon as well as using the Moon to travel beyond the Earth to other planets in the solar system. This thesis explores the mechanics behind a lunar mission and the costs and benefits of different approaches. To constrain this problem, the launch criteria are those of Wallops Flight Facility (WFF), which has expressed interest in launching...

  11. BUDAPEST, BRATISLAVA AND VIENNA CONFERENCE FACILITIES, COMPARATIVE ANALYSIS

    Endre György Bártfai

    2011-01-01

    The aim of this study is to give an overview of conference facilities in three capital cities in the Central European area, along the Danube, analyse and compare their possibilities and venues. The utilized data within the study was collected from different sources, like websites of the Hungarian, Slovakian and Austrian Convention Bureaus, books dealing with convention and event management and statistics, ICCA publications. Budapest is highly ranked between cities transacting conferences for ...

  12. Experimental facility for analysis of biomass combustion characteristics

    Miljković Biljana M.

    2015-01-01

    Full Text Available The objective of the present article is to present an experimental facility which was designed and built at the Faculty of Technical Sciences in order to study the combustion of different sorts of biomass and municipal solid waste. Despite its apparent simplicity, direct combustion is a complex process from a technological point of view. Conventional combustion equipment is not designed for burning agricultural residues. Devices for agricultural waste combustion are still in the development phase, which means that adequate design solution is presently not available at the world market. In order to construct a boiler and achieve optimal combustion conditions, it is necessary to develop a mathematical model for biomass combustion. Experimental facility can be used for the collection of data necessary for detailed modelling of real grate combustor of solid biomass fuels. Due to the complexity of the grate combustion process, its mathematical models and simulation software tools must be developed and verified using experimental data. This work highlights the properties required for the laboratory facility designed for the examination of biomass combustion and discusses design and operational issues.

  13. Safety Analysis Report: X17B2 beamline Synchrotron Medical Research Facility

    This report contains a safety analysis for the X17B2 beamline synchrotron medical research facility. Health hazards, risk assessment and building systems are discussed. Reference is made to transvenous coronary angiography

  14. Mixed Waste Management Facility Preliminary Safety Analysis Report. Chapters 1 to 20

    This document provides information on waste management practices, occupational safety, and a site characterization of the Lawrence Livermore National Laboratory. A facility description, safety engineering analysis, mixed waste processing techniques, and auxiliary support systems are included

  15. Mixed Waste Management Facility Preliminary Safety Analysis Report. Chapters 1 to 20

    1994-09-01

    This document provides information on waste management practices, occupational safety, and a site characterization of the Lawrence Livermore National Laboratory. A facility description, safety engineering analysis, mixed waste processing techniques, and auxiliary support systems are included.

  16. Safety Analysis Report: X17B2 beamline Synchrotron Medical Research Facility

    Gmuer, N.F.; Thomlinson, W.

    1990-02-01

    This report contains a safety analysis for the X17B2 beamline synchrotron medical research facility. Health hazards, risk assessment and building systems are discussed. Reference is made to transvenous coronary angiography. (LSP)

  17. Instrumentation in neutron activation analysis

    The rise of neutron activation analysis (NAA) as a tool in geochemical research has parallelled advances in detector, multi-channel analyzer, and computer technology. Micro-computers are now being integrated into NAA systems, and gamma-ray spectrometer instrumentation is evolving towards direct-reading systems. The investigator is faced with a wide range of possibilities and choices when equipping or re-equipping a laboratory. The geoscientist is provided with an overview of the available instrumentation and what soon may be feasible. (L.L.)

  18. RCRA and CERCLA requirements affecting cleanup activities at a federal facility superfund site

    The Fernald Environmental Management Project (FEMP) achieved success on an integrated groundwater monitoring program which addressed both RCRA and CERCLA requirements. The integrated plan resulted in a cost savings of approximately $2.6 million. At present, the FEMP is also working on an integrated closure process to address Hazardous Waste Management Units (HWMUs) at the site. To date, Ohio EPA seems willing to discuss an integrated program with some stipulations. If an integrated program is implemented, a cost savings of several million dollars will be realized since the CERCLA documents can be used in place of a RCRA closure plan. The success of an integrated program at the FEMP is impossible without the support of DOE and the regulators. Since DOE is an owner/operator of the facility and Ohio EPA regulates hazardous waste management activities at the FEMP, both parties must be satisfied with the proposed integration activities. Similarly, US EPA retains CERCLA authority over the site along with a signed consent agreement with DOE, which dictates the schedule of the CERCLA activities. Another federal facility used RCRA closure plans to satisfy CERCLA activities. This federal facility was in a different US EPA Region than the FEMP. While this approach was successful for this site, an integrated approach was required at the FEMP because of the signed Consent Agreement and Consent Decree. For federal facilities which have a large number of HWMUs along with OUs, an integrated approach may result in a timely and cost-effective cleanup

  19. DETERMINANTS OF FARMER-TO-CONSUMER DIRECT MARKET VISITS BY TYPE OF FACILITY: A LOGIT ANALYSIS

    Govindasamy, Ramu; Nayga, Rodolfo M.

    1997-01-01

    This study identifies several socioeconomic and demographic characteristics of individuals who visited farmer-to-consumer direct markets in New Jersey. The analysis was performed for each type of direct marketing facility: pick-your-own farms, roadside stands, farmers' markets, and direct farm markets. Logit analysis results indicate that various factors affect visitation to each type of facility. Factors examined include consumers' consumption and variety of fruits and vegetables, price expe...

  20. An Extended Industry Analysis of the Water Facilities Design Industry in British Columbia

    Ibrahim, Imtiaz

    2012-01-01

    This paper presents an extended industry analysis of British Columbia’s water facilities design industry. The paper analyzes the industry using two models to determine the competitive position, profitability, linkages and competitive scopes of typical engineering consulting firms with water facilities design practices in British Columbia. Strategies to be pursued are then recommended based on these analyses. The extended industry analysis is first conducted using the Five Forces model. This m...

  1. Preliminary siting activities for new waste handling facilities at the Idaho National Engineering Laboratory

    Taylor, D.D.; Hoskinson, R.L.; Kingsford, C.O.; Ball, L.W.

    1994-09-01

    The Idaho Waste Processing Facility, the Mixed and Low-Level Waste Treatment Facility, and the Mixed and Low-Level Waste Disposal Facility are new waste treatment, storage, and disposal facilities that have been proposed at the Idaho National Engineering Laboratory (INEL). A prime consideration in planning for such facilities is the selection of a site. Since spring of 1992, waste management personnel at the INEL have been involved in activities directed to this end. These activities have resulted in the (a) identification of generic siting criteria, considered applicable to either treatment or disposal facilities for the purpose of preliminary site evaluations and comparisons, (b) selection of six candidate locations for siting,and (c) site-specific characterization of candidate sites relative to selected siting criteria. This report describes the information gathered in the above three categories for the six candidate sites. However, a single, preferred site has not yet been identified. Such a determination requires an overall, composite ranking of the candidate sites, which accounts for the fact that the sites under consideration have different advantages and disadvantages, that no single site is superior to all the others in all the siting criteria, and that the criteria should be assigned different weighing factors depending on whether a site is to host a treatment or a disposal facility. Stakeholder input should now be solicited to help guide the final selection. This input will include (a) siting issues not already identified in the siting, work to date, and (b) relative importances of the individual siting criteria. Final site selection will not be completed until stakeholder input (from the State of Idaho, regulatory agencies, the public, etc.) in the above areas has been obtained and a strategy has been developed to make a composite ranking of all candidate sites that accounts for all the siting criteria.

  2. Preliminary siting activities for new waste handling facilities at the Idaho National Engineering Laboratory

    The Idaho Waste Processing Facility, the Mixed and Low-Level Waste Treatment Facility, and the Mixed and Low-Level Waste Disposal Facility are new waste treatment, storage, and disposal facilities that have been proposed at the Idaho National Engineering Laboratory (INEL). A prime consideration in planning for such facilities is the selection of a site. Since spring of 1992, waste management personnel at the INEL have been involved in activities directed to this end. These activities have resulted in the (a) identification of generic siting criteria, considered applicable to either treatment or disposal facilities for the purpose of preliminary site evaluations and comparisons, (b) selection of six candidate locations for siting,and (c) site-specific characterization of candidate sites relative to selected siting criteria. This report describes the information gathered in the above three categories for the six candidate sites. However, a single, preferred site has not yet been identified. Such a determination requires an overall, composite ranking of the candidate sites, which accounts for the fact that the sites under consideration have different advantages and disadvantages, that no single site is superior to all the others in all the siting criteria, and that the criteria should be assigned different weighing factors depending on whether a site is to host a treatment or a disposal facility. Stakeholder input should now be solicited to help guide the final selection. This input will include (a) siting issues not already identified in the siting, work to date, and (b) relative importances of the individual siting criteria. Final site selection will not be completed until stakeholder input (from the State of Idaho, regulatory agencies, the public, etc.) in the above areas has been obtained and a strategy has been developed to make a composite ranking of all candidate sites that accounts for all the siting criteria

  3. Recent National Solar Thermal Test Facility activities, in partnership with industry

    Ghanbari, Cherly; Cameron, Christopher P.; Ralph, Mark E.; Pacheco, James E.; Rawlinson, K. Scott; Evans, Lindsey R.

    The National Solar Thermal Test Facility (NSTTF) at Sandia National Laboratories in Albuquerque, New Mexico, USA conducts testing of solar thermal components and systems, funded primarily by the US Department of Energy. Activities are conducted in support of Central Receiver Technology, Distributed Receiver Technology and Design Assistance projects. All activities are performed in support of various cost-shared government/industry joint ventures and, on a design assistance basis, in support of a number of other industry partners.

  4. The activity record against the 2011 off the pacific coast of Tohoku Earthquake in the Department of Hot Laboratories and Facilities. Part 2. Damage situations and restoration activities for facilities

    The Department of Hot Laboratories and Facilities in the Nuclear Science Research Institute was supporting research activities on the safety and basic researches for nuclear fuel materials and reactor structural materials and so on. Our department was in charge of operation and management of 11 research facilities including 4 hot laboratories, in which nuclear fuel materials such as uranium and plutonium, and various radioactive isotopes were handled. This document describes the activity record of the detailed damage situations and their restorations of 11 facilities against the 2011 off the Pacific coast of Tohoku earthquake. About the restoration situation of facilities, restoration activities of 8 facilities out of 11 facilities were completed by March 31st, 2013. This document is a continuation of the Part 1 Emergency Activities at Onset (JAEA-Review 2011-048) which was described just after the earthquake. (author)

  5. Upgrades to the Radiochemistry Analysis of Gas Samples (RAGS) diagnostic at the National Ignition Facility

    Jedlovec, Donald; Christensen, Kim; Velsko, Carol; Cassata, Bill; Stoeffl, Wolfgang; Shaughnessy, Dawn; Lugten, John; Golod, Tony; Massey, Warren

    2015-08-01

    The Radiochemical Analysis of Gaseous Samples (RAGS) diagnostic apparatus operates at the National Ignition Facility (NIF). At the NIF, xenon is injected into the target chamber as a tracer, used as an analyte in the NIF targets, and generated as a fission product from 14 MeV neutron fission of depleted uranium contained in the NIF hohlraum. Following a NIF shot, the RAGS apparatus used to collect the gas from the NIF target chamber and then to cryogenically fractionate xenon gas. Radio-xenon and other activation products are collected and counted via gamma spectrometry, with the results used to determine critical physics parameters including: capsule areal density, fuel-ablator mix, and nuclear cross sections.

  6. Wound center facility billing: A retrospective analysis of time, wound size, and acuity scoring for determining facility level of service.

    Fife, Caroline E; Walker, David; Farrow, Wade; Otto, Gordon

    2007-01-01

    Outpatient wound center facility reimbursement for Medicare beneficiaries can be a challenge to determine and obtain. To compare methods of calculating facility service levels for outpatient wound centers and to demonstrate the advantages of an acuity-based billing system (one that incorporates components of facility work that is non-reimbursable by procedure codes and that represents an activity-based costing approach to medical billing), a retrospective study of 5,098 patient encounters contained in a wound care-specific electronic medical record database was conducted. Approximately 500 patient visits to the outpatient wound center of a Texas regional hospital between April 2003 and November 2004 were categorized by service level in documentation and facility management software. Visits previously billed using a time-based system were compared to the Centers for Medicare and Medicaid Services' proposed three-tiered wound size-based system. The time-based system also was compared to an acuity-based scoring system. The Pearson correlation coefficient between billed level of service by time and estimated level of service by acuity was 0.442 and the majority of follow-up visits were billed as Level 3 and above (on a time level of 1 to 5) , confirming that time is not a surrogate for actual work performed. Wound size also was found to be unrelated to service level (Pearson correlation = 0.017) and 97% of wound areas were based scoring system produced a near-normal distribution of results, producing more mid-range billings than extremes; no other method produced this distribution. Hospital-based outpatient wound centers should develop, review, and refine acuity score-based models on which to determine billed level of service. PMID:17264354

  7. Applied research and development of neutron activation analysis

    Chung, Yong Sam; Moon, Jong Hwa; Kim, Sun Ha; Baek, Sung Ryel; Kim, Young Gi; Jung, Hwan Sung; Park, Kwang Won; Kang, Sang Hun; Lim, Jong Myoung

    2003-05-01

    The aims of this project are to establish the quality control system of Neutron Activation Analysis(NAA) due to increase of industrial needs for standard analytical method and to prepare and identify the standard operation procedure of NAA through practical testing for different analytical items. R and D implementations of analytical quality system using neutron irradiation facility and gamma-ray measurement system and automation of NAA facility in HANARO research reactor are as following ; 1) Establishment of NAA quality control system for the maintenance of best measurement capability and the promotion of utilization of HANARO research reactor 2) Improvement of analytical sensitivity for industrial applied technologies and establishment of certified standard procedures 3) Standardization and development of Prompt Gamma-ray Activation Analysis (PGAA) technology.

  8. Applied research and development of neutron activation analysis

    The aims of this project are to establish the quality control system of Neutron Activation Analysis(NAA) due to increase of industrial needs for standard analytical method and to prepare and identify the standard operation procedure of NAA through practical testing for different analytical items. R and D implementations of analytical quality system using neutron irradiation facility and gamma-ray measurement system and automation of NAA facility in HANARO research reactor are as following ; 1) Establishment of NAA quality control system for the maintenance of best measurement capability and the promotion of utilization of HANARO research reactor 2) Improvement of analytical sensitivity for industrial applied technologies and establishment of certified standard procedures 3) Standardization and development of Prompt Gamma-ray Activation Analysis (PGAA) technology

  9. Practical aspects of operating a neutron activation analysis laboratory

    This book is intended to advise in everyday practical problems related to operating a neutron activation analysis (NAA) laboratory. It gives answers to questions like ''what to use NAA for'', ''how to find relevant research problems'', ''how to find users for the technique'', ''how to estimate the cost of the analysis and how to finance the work'', ''how to organize the work in a rational way'' and ''how to perform the quality control''. It gives advice in choosing staff, equipment, and consumables and how to design facilities and procedures according to need and available resources. Potential applications of economic or environmental importance, reactor facilities, counting and measuring equipment of the lab, cooperation with other analytical groups and competitiveness of NAA are discussed by experienced analysts. The compiled 8 tables of data useful for neutron activation analysts are a valuable asset for research labs as well as industrial quality control units. Refs, figs and tabs

  10. Development of boron concentration analysis system and techniques for testing performance of BNCT facility

    Choi, Hee Dong; Kim, Chang Shuk; Byun, Soo Hyun; Lee, Jae Yun; Sun, Gwang Min; Kim, Suk Kwon [Seoul National University, (Korea)

    2000-04-01

    I. Objectives and Necessity of the Project. Development of a boron concentration analysis system used for BNCT. Development of test techniques for BNCT facility. II. Contents and Scopes of the Project. (1) Design of a boron concentration analysis system at HANARO. (2) Component machining and instruments purchase, performance test. (3) Calculation and measurement of diffracted polychromatic beam quality. (4) Test procedures for boron concentration analysis system and BNCT facility. III. Result of the Project (1) Diffracted neutron beam quality for boron concentration analysis. (neutron flux: 1.2 * 10{sup 8} n/cm{sup 2}s, Cd-ratio : 1,600) (2) Components and instruments of the boron concentration analysis system. (3) Diffracted neutron spectrum and flux. (4) Test procedures for boron concentration analysis system and BNCT facility. 69 refs., 44 figs., 14 tabs. (Author)

  11. Thermal performance analysis of an MHD simulation test facility

    To evaluate the performance of the downstream components of a coal-fired, baseline, open cycle MHD power plant, Mississippi State University has set up a simulation test facility. Reduced thermal data from this test stand for steady-state operating conditions are presented in the paper. A thermal model to predict the variation of important thermal parameters in the test stand is shown. Results from the reduced experimental data and the predictive thermal model are compared. In addition, results for calibration runs and from recent secondary combustion tests are discussed. 7 refs

  12. Seismic fragility analysis of structural components for HFBR facilities

    The paper presents a summary of recently completed seismic fragility analyses of the HFBR facilities. Based on a detailed review of past PRA studies, various refinements were made regarding the strength and ductility evaluation of structural components. Available laboratory test data were analysed to evaluate the formulations used to predict the ultimate strength and deformation capacities of steel, reinforced concrete and masonry structures. The biasness and uncertainties were evaluated within the framework of the fragility evaluation methods widely accepted in the nuclear industry. A few examples of fragility calculations are also included to illustrate the use of the presented formulations

  13. Analysis of factors related to man-induced hazard for nuclear facilities

    Lee, Young Soon; Jung, Jea Hee; Lee, Keun O; Son, Ki Sang; Wang, Sang Chul; Lee, Chang Jin; Ku, Min Ho; Park, Nam Young [Seoul National Univ. of Technology, Seoul (Korea, Republic of)

    2003-03-15

    This study is to show a guide for installing hazardous facilities adjoined atomic power plant after finding out how much these facilities could impact to the atomic plant. Nuclear power plant is an important facility which is closely connected with public life, industrial activity, and the conduct of public business, so it should not be damaged. Therefore, if there are hazardous and harmful facilities near the plant, then they must be evaluated by the size, the type, and the shape. First of all, any factors that could cause man induced accident must be investigated. And they must be exactly evaluated from how much it will damage the plant facilities. The purpose of this study is to set a technical standard for the installation of these facilities by evaluating the man induced accident. Also, it is to make out the evaluation methods by investigating the hazardous facilities which are placed near the plant. Our country is now using CFR standard : reg. guide and IAEA safety series. However, not only the standard of technology which is related to man induced accident but also the evaluation methods for facilities are not yet layed down. As It was mentioned above, we should evaluate these facilities adequately, and these methods must be made out.

  14. THACO, a Test Facility for Characterizing the Noise Performance of Active Antenna Arrays

    Woestenburg, E E M; Ruiter, M; Ivashina, M V; Witvers, R H

    2011-01-01

    This paper discusses an outdoor test facility for the noise characterization of active antenna arrays, using measurement results of array noise temperatures in the order of 50 K for a number of small aperture arrays. The measurement results are obtained by a Y-factor method with hot and a cold noise sources, with an absorber at room temperature as the hot load and the cold sky as the cold load. The effect of shielding the arrays by the test facility, with respect to noise and RFI from the environment, will also be discussed.

  15. Research and development activities for transmutation physics experimental facility in J-PARC

    The Japan Atomic Energy Agency (JAEA) has the plan to construct Transmutation Physics Experimental Facility (TEF-P) under a framework of J-PARC (Japan Proton Accelerator Research Complex) project. TEF-P is a critical assembly which can load Minor Actinide (MA) fuels to perform reactor physics experiments for transmutation systems such as Accelerator-Driven System (ADS) or Fast Reactor (FR). The facility can also use proton beam from the J-PARC accelerator to investigate the controllability of ADS. Current status and activities for TEF-P are described. (author)

  16. A new shunt DC active filter of power supply in a steady high magnetic field facility

    A DC active power filter is an indispensable part in a high power and high stability power supply system, especially in the power supply system of the Steady High Magnetic Field Facility, which requires that the current ripple should be limited to 50 parts per million. In view of the disadvantages of the series DC active power filter and shunt Pulse Width Modulation DC active filter, this paper puts forward a novel DC active filter by combining the advantages of the transistor regulator and the shunt type. The structure and principle of the new shunt linear active filter are introduced. Meanwhile, the design of several key components that construct the new shunt linear active filter is also analyzed. The simulation model and an experimental prototype of the shunt linear active filter are developed, and the results verify that the parameter design is reasonable and the shunt active filter has a good filter effect. (authors)

  17. Analysis of the formation, expression, and economic impacts of risk perceptions associated with nuclear facilities

    This report investigates how communities hosting nuclear facilities form and express perceptions of risk and how these risk perceptions affect local economic development. Information was collected from site visits and interviews with plant personnel, officials of local and state agencies, and community activists in the hosting communities. Six commercial nuclear fuel production facilities and five nuclear facilities operated for the US Department of Energy by private contractors were chosen for analysis. The results presented in the report indicate that the nature of risk perceptions depends on a number of factors. These factors are (1) level of communication by plant officials within the local community, (2) track record of the facility. operator, (3) process through which community and state officials receive information and form opinions, (4) level of economic links each plant has with the local community, and (15) physical characteristics of the facility itself. This report finds that in the communities studied, adverse ask perceptions have not affected business location decisions, employment levels in the local community, tourism, or agricultural development. On the basis of case-study findings, this report recommends that nuclear facility siting programs take the following observations into account when addressing perceptions of risk. First, the quality of a facility's participation with community activists, interest groups, and state agencies helps to determine the level of perceived risk within a community. Second, the development of strong economic links between nuclear facilities and their host communities will produce a higher level of acceptance of the nuclear facilities

  18. Derivation of activity limits for the disposal of radioactive waste in near surface disposal facilities

    criteria for disposal of radioactive wastes to near surface facilities. These criteria are qualitative in nature and, for example, they do not address limitations on radionuclide content of waste, waste packages or the facility as a whole. This publication is to present an approach for establishing radiological waste acceptance criteria using a safety assessment methodology and to illustrate its application in establishing limits on the total activity and the activity concentrations of radioactive waste to be disposed in near surface disposal facilities. The approach makes use of accepted methods and computational schemes currently used in assessing the safety of near surface disposal facilities both during the operational and post-closure periods. The scope of this publication covers the use of safety assessment methodology to calculate total and specific activities limits for radioactive waste in near surface disposal facilities. It is used to evaluate the potential operational and post-closure radiological impact of solid and solidified radioactive waste in near surface facilities. The radioactive waste types used to illustrate the approach range from waste containing radionuclides used for medical, industrial and research purposes to waste arising from nuclear fuel cycle activities. They also include waste arising from the decommissioning of nuclear facilities. The focus of the publication is on using of safety assessment methodology in derivation of quantitative radioactivity limits. This report deals with the role of activity limits in disposal system safety (Section 2), the relevant radiation protection criteria (Section 3), the approach to derive activity limits (Section 4), illustrations of the application of this approach (Section 5), and guidance on the use of the approach (Section 6)

  19. Decision support and analysis tool for planning in a semiconductor manufacturing facility

    Fargher, Hugh E.; Smith, Richard A.

    1994-03-01

    As part of the recently completed Microelectronics Manufacturing Science and Technology (MMST) project, a decision support and analysis tool for planning in a semiconductor manufacturing facility has been developed. Design of the planning system uses an object- oriented approach, and implementation is performed in the Smalltalk programming environment. The system continually maintains a plan for wafer release into a facility, and predicts processing completion dates. The system has been built to run in a distributed environment, allowing simultaneous users in different parts of the facility. The system also provides several types of what-if analysis, both on the existing production plan and on production data. Production plan analysis is used to assist in making operational decisions related to the facility in its current state, such as determining the least disruptive time to take a piece of equipment down for maintenance. Production data analysis, which can be performed independent of the production plan, determines information such as equipment throughput rates to achieve given product cycle-times. All planning is performed using artificial intelligence search techniques, and is based on a time-phased capacity model of the facility. Uncertainty inherent in production data, such as cycle-times, is modeled using fuzzy arithmetic. This tool was used during the final 1000 wafer demonstration for MMST, and is currently being installed in other semiconductor manufacturing facilities. This paper describes the main goals of the planning system, the overall approach to planning and analysis, and a brief description of the current status.

  20. Shielding analysis of high level waste water storage facilities using MCNP code

    The neutron and gamma-ray transport analysis for the facility as a reprocessing facility with large buildings having thick shielding was made. Radiation shielding analysis consists of a deep transmission calculation for the concrete wall and a skyshine calculation for the space out of the buildings. An efficient analysis with a short running time and high accuracy needs a variance reduction technique suitable for all the calculation regions and structures. In this report, the shielding analysis using MCNP and a discrete ordinate transport code is explained and the idea and procedure of decision of variance reduction parameter is completed. (J.P.N.)

  1. Nuclear Rocket Test Facility Decommissioning Including Controlled Explosive Demolition of a Neutron-Activated Shield Wall

    Michael Kruzic

    2007-09-01

    Located in Area 25 of the Nevada Test Site, the Test Cell A Facility was used in the 1960s for the testing of nuclear rocket engines, as part of the Nuclear Rocket Development Program. The facility was decontaminated and decommissioned (D&D) in 2005 using the Streamlined Approach For Environmental Restoration (SAFER) process, under the Federal Facilities Agreement and Consent Order (FFACO). Utilities and process piping were verified void of contents, hazardous materials were removed, concrete with removable contamination decontaminated, large sections mechanically demolished, and the remaining five-foot, five-inch thick radiologically-activated reinforced concrete shield wall demolished using open-air controlled explosive demolition (CED). CED of the shield wall was closely monitored and resulted in no radiological exposure or atmospheric release.

  2. Nuclear Rocket Test Facility Decommissioning Including Controlled Explosive Demolition of a Neutron-Activated Shield Wall

    Located in Area 25 of the Nevada Test Site, the Test Cell A Facility was used in the 1960s for the testing of nuclear rocket engines, as part of the Nuclear Rocket Development Program. The facility was decontaminated and decommissioned (D and D) in 2005 using the Streamlined Approach For Environmental Restoration (SAFER) process, under the Federal Facilities Agreement and Consent Order (FFACO). Utilities and process piping were verified void of contents, hazardous materials were removed, concrete with removable contamination decontaminated, large sections mechanically demolished, and the remaining five-foot, five-inch thick radiologically-activated reinforced concrete shield wall demolished using open-air controlled explosive demolition (CED). CED of the shield wall was closely monitored and resulted in no radiological exposure or atmospheric release

  3. Safety research activities for Japanese regulations of spent fuel interim storage facilities

    Japan Nuclear Energy Safety Organization (JNES) carries out (a) preparation of technical documents, (b) technical evaluations of standards (prepared by academic societies), etc. and (c) other R and D activities, to support Nuclear and Industrial Safety Agency (NISA: which prepares necessary regulations for Spent Fuel Interim Storage Facilities). In 2010 fiscal year, JNES completed technical evaluation of the standard (prepared by Atomic Energy Society of Japan) used for the storage facility (dual purpose cask system) being constructed in Mutsu-City and R and D for UT test of welded canister lids which is required for concrete cask storage facilities. And also, JNES is preparing dynamic test of spent fuel to examine the integrity of spent fuel at cask drop accidents and PWR spent fuel storage test to prove long term integrity of spent fuel and cask itself. The results of these tests will be reported in 2011 and 2012 fiscal year. (author)

  4. The new prompt gamma-ray activation facility at the Paul Scherrer Institute, Switzerland

    Crittin, M; Schenker, J L

    2000-01-01

    Since October 1997, a new Prompt Gamma-ray Activation (PGA) facility at the neutron spallation source SINQ of the Paul Scherrer Institute (PSI) in Villigen, Switzerland, is operational. The detection system includes a Compton-suppression spectrometer and a pair spectrometer. An interesting feature of this PGA facility is the capillary-based neutron focusing optics which permits scanning of samples and nuclear spectroscopy of isotopes having small capture cross sections. During the beam periods 1997 and 1998, measurements were undertaken to characterize the PGA facility (gamma-ray background, efficiencies of the two spectrometers, analytical sensitivities and detection limits for several elements, performances of the neutron lens). Elemental analyses of standards were also performed.

  5. Radioactive waste package assay facility. Volume 2. Investigation of active neutron and active gamma interrogation

    Volume 2 of this report describes the theoretical and experimental work carried out at Harwell on active neutron and active gamma interrogation of 500 litre cemented intermediate level waste drums. The design of a suitable neutron generating target in conjunction with a LINAC was established. Following theoretical predictions of likely neutron responses, an experimental assay assembly was built. Responses were measured for simulated drums of ILW, based on CAGR, Magnox and PCM wastes. Good correlations were established between quantities of 235-U, nat-U and D2O contained in the drums, and the neutron signals. Expected sensitivities are -1g of fissile actinide and -100g of total actinide. A measure of spatial distribution is obtainable. The neutron time spectra obtained during neutron interrogation were more complex than expected, and more analysis is needed. Another area of discrepancy is the difference between predicted and measured thermal neutron flux in the drum. Clusters of small 3He proportional counters were found to be much superior for fast neutron detection than larger diameter counters. It is necessary to ensure constancy of electron beam position relative to target(s) and drum, and prudent to measure the target neutron or gamma output as appropriate. 59 refs., 77 figs., 11 tabs

  6. A facile reflux procedure to increase active surface sites form highly active and durable supported palladium@platinum bimetallic nanodendrites

    Wang, Qin; Li, Yingjun; Liu, Baocang; Xu, Guangran; Zhang, Geng; Zhao, Qi; Zhang, Jun

    2015-11-01

    A series of well-dispersed bimetallic Pd@Pt nanodendrites uniformly supported on XC-72 carbon black are fabricated by using different capping agents. These capping agents are essential for the branched morphology control. However, the surfactant adsorbed on the nanodendrites surface blocks the access of reactant molecules to the active surface sites, and the catalytic activities of these bimetallic nanodendrites are significantly restricted. Herein, a facile reflux procedure to effectively remove the capping agent molecules without significantly affecting their sizes is reported for activating supported nanocatalysts. More significantly, the structure and morphology of the nanodendrites can also be retained, enhancing the numbers of active surface sites, catalytic activity and stability toward methanol and ethanol electro-oxidation reactions. The as-obtained hot water reflux-treated Pd@Pt/C catalyst manifests superior catalytic activity and stability both in terms of surface and mass specific activities, as compared to the untreated catalysts and the commercial Pt/C and Pd/C catalysts. We anticipate that this effective and facile removal method has more general applicability to highly active nanocatalysts prepared with various surfactants, and should lead to improvements in environmental protection and energy production.

  7. Bubble-condenser experimental qualification at EREC test facility. Analysis with ATHLET code

    In the framework of PHARE/TACIS project No. PH 2.13/95 the experimental test facility for bubble condenser experimental qualification is under construction at Electrogorsk Research and Engineering Centre. The test facility will contain high pressure system, compartments upstream of the bubble condenser and a section of the bubble condenser system. The scaling of the test facility is 1:100. The high pressure system consists of five vessels to appropriately model the leak functions (mass flow rate and enthalpy) during the loss of coolant accidents postulated in the design of VVER-440/V213. Pre-test analysis has been carried out with ATHLET code in order to properly design the high pressure system of the test facility. ATHLET code was applied to the NPP and to the test facility configuration and for both geometries the selected accidents were calculated. (author)

  8. Economic Feasibility Analysis of the Application of Geothermal Energy Facilities to Public Building Structures

    Sangyong Kim

    2014-03-01

    Full Text Available This study aims to present an efficient plan for the application of a geothermal energy facility at the building structure planning phase. Energy consumption, energy cost and the primary energy consumption of buildings were calculated to enable a comparison of buildings prior to the application of a geothermal energy facility. The capacity for energy savings and the costs related to the installation of such a facility were estimated. To obtain more reliable criteria for economic feasibility, the lifecycle cost (LCC analysis incorporated maintenance costs (reflecting repair and replacement cycles based on construction work specifications of a new renewable energy facility and initial construction costs (calculated based on design drawings for its practical installation. It is expected that the findings of this study will help in the selection of an economically viable geothermal energy facility at the building construction planning phase.

  9. Guidelines for job and task analysis for DOE nuclear facilities

    The guidelines are intended to be responsive to the need for information on methodology, procedures, content, and use of job and task analysis since the establishment of a requirement for position task analysis for Category A reactors in DOE 5480.1A, Chapter VI. The guide describes the general approach and methods currently being utilized in the nuclear industry and by several DOE contractors for the conduct of job and task analysis and applications to the development of training programs or evaluation of existing programs. In addition other applications for job and task analysis are described including: operating procedures development, personnel management, system design, communications, and human performance predictions

  10. Battery Test Facility- Electrochemical Analysis and Diagnostics Laboratory

    Federal Laboratory Consortium — The Electrochemical Analysis and Diagnostics Laboratory (EADL) provides battery developers with reliable, independent, and unbiased performance evaluations of their...

  11. Analysis of RA-8 critical facility core in some configurations

    The RA-8 critical facility was designated and built to be used in the experimental plan of the 'CAREM' Project but is, in itself, very versatile and adequate to perform many types of other experiments. The present paper includes calculated estimates of some critical configurations and comparisons with experimental results obtained during its start up. Results for Core 1 with homogeneous arrangement of rods containing 1.8 % enriched uranium, showed very good agreement. In fact, an experimentally critical configuration was reached with 1.300 rods and calculated values were: 1.310 using the WIMS code and 1.148 from the CONDOR code. Moreover, it was verified that the estimated number of 3.4% enriched uranium rods to be fabricated is enough to build a heterogeneous core or even a homogeneous core with this enrichment. The replacement of 3.4 % enriched uranium by 3.6 % will not present problems related with the original plan. (author)

  12. Improved Methodology Application for 12-Rad Analysis in a Shielded Facility at SRS

    The DOE Order 420.1 requires establishing 12-rad evacuation zone boundaries and installing Criticality Accident Alarm System (CAAS) per ANS-8.3 standard for facilities having a probability of criticality greater than 10-6 per year. The H-Canyon at the Savannah River Site (SRS) is one of the reprocessing facilities where SRS reactor fuels, research reactor fuels, and other fissile materials are processed and purified using a modified Purex process called H-Modified or HM Process. This paper discusses an improved methodology for 12-rad zone analysis and its implementation within this large shielded facility that has a large variety of criticality sources and scenarios

  13. Safety analysis of IFR fuel processing in the Argonne National Laboratory Fuel Cycle Facility

    The Integral Fast Reactor (IFR) concept developed by Argonne National Laboratory (ANL) includes on-site processing and recycling of discharged core and blanket fuel materials. The process is being demonstrated in the Fuel Cycle Facility (FCF) at ANL's Idaho site. This paper describes the safety analyses that were performed in support of the FCF program; the resulting safety analysis report was the vehicle used to secure authorization to operate the facility and carry out the program, which is now under way. This work also provided some insights into safety-related issues of a commercial IFR fuel processing facility. These are also discussed

  14. The verification of neutron activation analysis support system (cooperative research)

    Neutron activation analysis support system is the system in which even the user who has not much experience in the neutron activation analysis can conveniently and accurately carry out the multi-element analysis of the sample. In this verification test, subjects such functions, usability, precision and accuracy of the analysis and etc. of the neutron activation analysis support system were confirmed. As a method of the verification test, it was carried out using irradiation device, measuring device, automatic sample changer and analyzer equipped in the JRR-3M PN-3 facility, and analysis software KAYZERO/SOLCOI based on the k0 method. With these equipments, calibration of the germanium detector, measurement of the parameter of the irradiation field and analysis of three kinds of environmental standard sample were carried out. The k0 method adopted in this system is primarily utilized in Europe recently, and it is the analysis method, which can conveniently and accurately carried out the multi-element analysis of the sample without requiring individual comparison standard sample. By this system, total 28 elements were determined quantitatively, and 16 elements with the value guaranteed as analytical data of the NIST (National Institute of Standards and Technology) environment standard sample were analyzed in the accuracy within 15%. This report describes content and verification result of neutron activation support system. (author)

  15. Final safety analysis report (FSAR) for waste receiving and processing (WRAP) facility

    This safety analysis report provides a summary description of the WRAP Facility, focusing on significant safety-related characteristics of the location and facility design. This report demonstrates that adherence to the safety basis wi11 ensure necessary operational safety considerations have been addressed sufficiently and justifies the adequacy of the safety basis in protecting the health and safety of the public, workers, and the environment

  16. Fuel Assemblies Thermal Analysis in the New Spent Fuel Storage Facility at Inshass Site

    New Wet Storage Facility (NSF) is constructed at Inshass site to solve the problem of spent fuel storage capacity of ETRR-1 reactor . The Engineering Safety Heat Transfer Features t hat characterize the new facility are presented. Thermal analysis including different scenarios of pool heat load and safety limits are discussed . Cladding temperature limit during handling and storage process are specified for safe transfer of fuel

  17. A Strategic Analysis of a Facility Supply Distributor in British Columbia

    Brooks, Derek

    2014-01-01

    This paper presents a business level strategic analysis of RST Corporation’s facility supplydivision, in British Columbia Canada.The facility supply industry is in decline. Firms find it increasingly difficult to achieveprofitability. Despite its decline, this industry will provide opportunities for a smaller number offirms with the correct strategy to take advantage of them. This paper explores the industryenvironment, RST’s resources, current strategy, and performance. It then goes on to de...

  18. An Analysis of Marketing Strategies of an Integrated Facility Services Company : The Case of ISS, Sweden

    Amanze, Collins Nwakanma; B.K.A, Sondengam

    2008-01-01

    TOPIC: An Analysis of Marketing Strategies of an Integrated Facility Services Company: The Case of ISS, Sweden AUTHORS: Collins Nwakanma Amanze, Sondengam B.K.A SUPERVISOR: Anders Hederstierna COURSE: Master Thesis in Business Administration DEPARTMENT: School of Management, Blekinge Institute of Technology, Sweden PROGRAMME: Masters in Business Administration (MBA) PURPOSE: The purpose of our research is to understand how Integrated Facility service companies (using ISS, Sweden as our focus)...

  19. Status Report on the Neutron Activation Analysis Activities in the Philippines

    The Philippines has a one megawatt open-pool type nuclear research reactor which is presently utilized in the conduct of nuclear research and development activities. The reactor is operated by the Philippine Atomic Research Center, the research arm of the Philippine Atomic Energy Commission. The reactor is presently utilized in the production of some radioisotopes, nuclear physics experiments and neutron activation analysis. For activation analysis the facilities available include the two 2 inch pneumatic tubes and a 2-inch central core dry-pipe. Although the reactor has been operative since 1963 it was only in the latter part of 1966 that a neutron activation analysis group was organized and almost immediately the training of personnel and setting up of a radiochemical laboratory and nucleonic counting assembly were initiated. Today, the counting system include a 100 channel analyzer with a 3 x 3 inch Nal(Tl) crystal

  20. Analysis of marketing communications in the selected accommodation facilities

    UHROVÁ, Pavlína

    2010-01-01

    The aim of this thesis is an analysis of marketing communication in selected accommodation and suggest possible changes. For this purpose I chose the hotel U Sladka. Estimation of marketing communication was made on the basis of information supplied by directors, employees and the hotel's own observations. An analysis of the present situation results that the hotel uses these elements of promotion: advertising, sales promotion, public relations, sponsorship, trade fairs and personal selling.D...

  1. Analysis of Operational and Management Cybersecurity Controls for Nuclear Facilities

    U.S. NRC developed this RG 5.71 by tailoring the baseline security controls described in NIST Special Publication 800-53 'Recommended Security Controls for Federal Information Systems and Organizations' to provide an acceptable method to comply with the 10 CFR 73.54. The purpose of this publication is to provide guidelines for selecting and specifying security controls for information systems. In this paper, we are going to analyze and compare the NRC RG 5.71 and the NIST SP800-53, in particular, for operational security controls and management security controls. If RG 5.71 omits the specific security control that is included in SP800-53, we would review that omitting is adequate or not. If RG 5.71 includes the specific security control that is not included in SP800-53, we would also review the rationale. And we are going to consider some security controls to strengthen cybersecurity of nuclear facilities

  2. 77 FR 48107 - Workshop on Performance Assessments of Near-Surface Disposal Facilities: FEPs Analysis, Scenario...

    2012-08-13

    ...-Surface Disposal Facilities: FEPs Analysis, Scenario and Conceptual Model Development, and Code Selection... Radioactive Waste.'' These regulations were published in the Federal Register on December 27, 1982 (47 FR... on three aspects of a performance assessment: (1) Features, Events, and Processes (FEPs) analysis,...

  3. A mobile prompt-gamma in-vivo neutron activation facility

    A significant environmental-medical problem is the internal deposition of Cd in industrially exposed populations. With the development of 238Pu, Be neutron sources, it has become possible to design facilities that are significantly reduced in size and hence portable. The paper describes the development and fabrication of a self-contained 'neutron-capture' facility in a mobile trailer which is readily transportable. Although the initial project was designed for the measurement of Cd deposited in the liver and kidney of industrial workers, the sensitivity of the system will allow for measurements in normal populations. The limits of detection of Cd are 2.0 mg for kidney and 1.5 μg/g (wet weight) for liver (associated organ radiation dose is 660 mrem). The portable facility makes it possible to study Cd-exposed populations in any geographic location. With minor modifications of the collimator, detectors and counting geometry, it is also possible to make in-vivo measurements of total body nitrogen. The neutron source used in the mobile facility is 85 Ci 238Pu, Be shielded by epoxy resin doped with LiCO3 and 6LiF. Additional shielding is provided by Pb bricks and polyethylene bricks with 1% boron and 80% Pb. For Cd measurement, the detection system consists of two Ge(Li) detectors (25% efficiency each) shielded by Bi and a 1.5-cm layer of paraffin heavily doped with 6LiF. For body nitrogen, two shielded 15.2cmx15.2cm NaI(Tl) detectors are positioned above the scanning bed. This mobile activation facility could be usefully employed in countries which need to monitor industrial populations but do not have ready access to large neutron research facilities and permanently installed whole-body counters

  4. Spatial accessibility to physical activity facilities and to food outlets and overweight in French youth

    Casey, R.; Chaix, B.; Weber, C.; Schweitzer, B; Charreire, H.; Salze, P; Badariotti, D; Banos, A.; Oppert, J-M; Simon, C

    2012-01-01

    Objective: Some characteristics of the built environment have been associated with obesity in youth. Our aim was to determine whether individual and environmental socio-economic characteristics modulate the relation between youth overweight and spatial accessibility to physical activity (PA) facilities and to food outlets. Design: Cross-sectional study. Subjects: 3293 students, aged 12±0.6 years, randomly selected from eastern France middle schools. Measurements and methods: Using geographica...

  5. Research on semi-active seismic isolation and vibration damping technology using new system for nuclear facilities

    Seismic isolation structure has been widely used in civil engineering and construction industry and its validity was observed at the Kobe Earthquake in 1995. A layered rubber isolator, which has been mostly deployed, is not good at uncertain loading conditions of external forces because of its seismic isolation structure of a resonance period and also becomes unstable at limiting deformation. Recently a ball isolator with guide or roller-guide has been introduced instead of it. As for a new seismic isolation system for nuclear facilities, semi-active seismic isolation and vibration damping system has been proposed with a ball isolator and a controllable friction damper using smart materials (magneto-rheological fluid). Effects of adhesion of a ball isolator due to aging and semi-active control on smart materials damper have been tested using small test modules. Analysis model of a ball isolator dependent on contact stress has been also developed. Larger mockup tests and their detailed analysis will be needed for their deployment for nuclear facilities. (T. Tanaka)

  6. CSER 94-013: Classification and access to PFP 232-Z Incinerator Facility and limits on characterization and disassembly activities in 232-Z burning hood

    This CSER justifies the Limited Control Facility designation for the closed Burning Hood in the PFP 232-Z Incinerator Facility. If the Burning Hood is opened to characterize the plutonium distribution and geometric integrity of the internals or for disassembly of the internals, then the more rigorous Fissionable Material Facility classification is required. Two sets of requirements apply for personnel access, criticality firefighting category for water use, and fissile material movement for the two states of the Burning Hood. The parameters used in the criticality analysis are listed to establish the limits under which this CSER is valid. Determination that the Burning Hood fissile material, moderation, or internal arrangements are outside these limits requires reevaluation of these parameter values and activities at the 232-Z Incinerator Facility. When the Burning Hood is open, water entry is to be prevented by two physical barriers for each water source

  7. The CERN analysis facility-a PROOF cluster for day-one physics analysis

    ALICE (A Large Ion Collider Experiment) at the LHC plans to use a PROOF cluster at CERN (CAF - CERN Analysis Facility) for analysis. The system is especially aimed at the prototyping phase of analyses that need a high number of development iterations and thus require a short response time. Typical examples are the tuning of cuts during the development of an analysis as well as calibration and alignment. Furthermore, the use of an interactive system with very fast response will allow ALICE to extract physics observables out of first data quickly. An additional use case is fast event simulation and reconstruction. A test setup consisting of 40 machines is used for evaluation since May 2006. The PROOF system enables the parallel processing and xrootd the access to files distributed on the test cluster. An automatic staging system for files either catalogued in the ALICE file catalog or stored in the CASTOR mass storage system has been developed. The current setup and ongoing development towards disk quotas and CPU fairshare are described. Furthermore, the integration of PROOF into ALICE's software framework (AliRoot) is discussed

  8. Analysis of Sunspot Activity Cycles

    Greenkorn, Robert A.

    2009-04-01

    A nonlinear analysis of the daily sunspot number for each of cycles 10 to 23 is used to indicate whether the convective turbulence is stochastic or chaotic. There is a short review of recent papers considering sunspot statistics and solar activity cycles. The differences in the three possible regimes - deterministic laminar flow, chaotic flow, and stochastic flow - are discussed. The length of data sets necessary to analyze the regimes is investigated. Chaos is described and a chronology of recent results that utilize chaos and fractals to analyze sunspot numbers follows. The parameters necessary to describe chaos - time lag, phase space, embedding dimension, local dimension, correlation dimension, and the Lyapunov exponents - are determined for the attractor for each cycle. Assuming the laminar regime is unlikely if chaos is not indicated in a cycle by the calculations, the regime must be stochastic. The sunspot numbers in each of cycles 10 to 19 indicate stochastic behavior. There is a transition from stochastic to chaotic behavior of the sunspot numbers in cycles 20, 21, 22, and 23. These changes in cycles 20 - 23 may indicate a change in the scale of turbulence in the convection zone that could result in a change in the convective heat transfer and a change in the size of the convection region for these four cycles.

  9. Neutron activation analysis in Bulgaria

    The development of instrumental neutron activation analysis (INAA) as a routine method started in 1960 with bringing into use of the experimental nuclear reactor 2 MW -IRT-2000. For the purposes of INAA the vertical channels were used. The neutron flux vary from 1 to 6x1012n/cm2s, with Cd ratio for gold of about 4,4. In one of the channels the neutron flux is additionally thermalised with grafite, in others - a pneumatic double-tube rabbit system is installed. One of the irradiation positions is equiped with 1 mm Cd shield constantly. With the pressure of the working gas (air) of 2 bar the transport time in one direction is 2,5 sec. Because of lack of special system for uniform irradiation an accuracy of 3% can be reached by use of iron monitors for long irradiations and copper monitors for use in the rabbit system. Two neutron generators are also working but the application of 14 MeV neutrons for INAA is still quite limited. The most developed are the applications of INAA in the fields of geology and paedology, medicine and biology, environment and pollution, archaeology, metallurgy, metrology and hydrology, criminology

  10. Special Analysis for Disposal of High-Concentration I-129 Waste in the Intermediate-Level Vaults at the E-Area Low-Level Waste Facility

    This revision was prepared to address comments from DOE-SR that arose following publication of revision 0. This Special Analysis (SA) addresses disposal of wastes with high concentrations of I-129 in the Intermediate-Level (IL) Vaults at the operating, low-level radioactive waste disposal facility (the E-Area Low-Level Waste Facility or LLWF) on the Savannah River Site (SRS). This SA provides limits for disposal in the IL Vaults of high-concentration I-129 wastes, including activated carbon beds from the Effluent Treatment Facility (ETF), based on their measured, waste-specific Kds

  11. Soil-structure interaction for seismic analysis of a nuclear facility

    Chen, J.Z.; Rosidi, D. [CH2M Hill Engineering Ltd., Toronto, ON (Canada); Lee, L. [ARES Corp., Toronto, ON (Canada)

    2010-07-01

    If the foundation material of a nuclear facility is not rock or rock-like soil to support the structure, then the effects of soil-structure interaction (SSI) must be considered in the seismic design of the facility. The direct or impedance method is generally used for SSI seismic analysis. For the direct method, the complete soil-foundation-structure system is modeled and analyzed in a single step based on the input of ground motion at the boundaries while the impedance method uses multiple steps to combine two primary causes of SSI, such as the inability of the foundation to match the free-field deformation and the effect of the dynamic response of the structure-foundation system on the movement of the supporting soil. This paper presented a case study for dynamic analysis of a nuclear facility located in a high seismic hazard zone. The direct method and lumped spring models were utilized in dynamic analysis of SSI. The paper discussed the seismic response of the nuclear facility. Specific topics that were discussed related to the analysis model included the structural model and soil spring model. The discussion regarding the soil spring model included the foundation stiffness, shear stiffness, ultimate foundation capacity, and equivalent damping coefficients. Seismic analysis was also outlined, with particular reference to dynamic analysis procedure; input ground motion; and results of seismic analysis. The results demonstrated a slight variation in peak structural response due to effects of foundation stiffness. 10 refs., 1 tab., 4 figs.

  12. Fast flux test facility final safety analysis report amendment 79

    This document is provided to replace, remove, or add applicable pages to the chapters on: Heat Transport System; Containment and Structures; Auxiliary Systems; Reactor Refueling System; Conduct of Operations; Safety Analysis; Quality Assurance; FFTF Criticality Specifications; and Appendix H's TRIGA Fuel Storage System

  13. Synthesis of sulfonated porous carbon nanospheres solid acid by a facile chemical activation route

    Chang, Binbin, E-mail: changbinbin806@163.com; Guo, Yanzhen; Yin, Hang; Zhang, Shouren; Yang, Baocheng, E-mail: baochengyang@yahoo.com

    2015-01-15

    Generally, porous carbon nanospheres materials are usually prepared via a template method, which is a multi-steps and high-cost strategy. Here, we reported a porous carbon nanosphere solid acid with high surface area and superior porosity, as well as uniform nanospheical morphology, which prepared by a facile chemical activation with ZnCl{sub 2} using resorcinol-formaldehyde (RF) resins spheres as precursor. The activation of RF resins spheres by ZnCl{sub 2} at 400 °C brought high surface area and large volume, and simultaneously retained numerous oxygen-containing and hydrogen-containing groups due to the relatively low processing temperature. The presence of these functional groups is favorable for the modification of –SO{sub 3}H groups by a followed sulfonation treating with sulphuric acid and organic sulfonic acid. The results of N{sub 2} adsorption–desorption and electron microscopy clearly showed the preservation of porous structure and nanospherical morphology. Infrared spectra certified the variation of surface functional groups after activation and the successful modification of –SO{sub 3}H groups after sulfonation. The acidities of catalysts were estimated by an indirect titration method and the modified amount of –SO{sub 3}H groups were examined by energy dispersive spectra. The results suggested sulfonated porous carbon nanospheres catalysts possessed high acidities and –SO{sub 3}H densities, which endowed their significantly catalytic activities for biodiesel production. Furthermore, their excellent stability and recycling property were also demonstrated by five consecutive cycles. - Graphical abstract: Sulfonated porous carbon nanospheres with high surface area and superior catalytic performance were prepared by a facile chemical activation route. - Highlights: • Porous carbon spheres solid acid prepared by a facile chemical activation. • It owns high surface area, superior porosity and uniform spherical morphology. • It possesses

  14. Synthesis of sulfonated porous carbon nanospheres solid acid by a facile chemical activation route

    Generally, porous carbon nanospheres materials are usually prepared via a template method, which is a multi-steps and high-cost strategy. Here, we reported a porous carbon nanosphere solid acid with high surface area and superior porosity, as well as uniform nanospheical morphology, which prepared by a facile chemical activation with ZnCl2 using resorcinol-formaldehyde (RF) resins spheres as precursor. The activation of RF resins spheres by ZnCl2 at 400 °C brought high surface area and large volume, and simultaneously retained numerous oxygen-containing and hydrogen-containing groups due to the relatively low processing temperature. The presence of these functional groups is favorable for the modification of –SO3H groups by a followed sulfonation treating with sulphuric acid and organic sulfonic acid. The results of N2 adsorption–desorption and electron microscopy clearly showed the preservation of porous structure and nanospherical morphology. Infrared spectra certified the variation of surface functional groups after activation and the successful modification of –SO3H groups after sulfonation. The acidities of catalysts were estimated by an indirect titration method and the modified amount of –SO3H groups were examined by energy dispersive spectra. The results suggested sulfonated porous carbon nanospheres catalysts possessed high acidities and –SO3H densities, which endowed their significantly catalytic activities for biodiesel production. Furthermore, their excellent stability and recycling property were also demonstrated by five consecutive cycles. - Graphical abstract: Sulfonated porous carbon nanospheres with high surface area and superior catalytic performance were prepared by a facile chemical activation route. - Highlights: • Porous carbon spheres solid acid prepared by a facile chemical activation. • It owns high surface area, superior porosity and uniform spherical morphology. • It possesses high acidity and high –SO3H density. • It

  15. Linde FUSRAP Site Remediation: Engineering Challenges and Solutions of Remedial Activities on an Active Industrial Facility - 13506

    The Linde FUSRAP Site (Linde) is located in Tonawanda, New York at a major research and development facility for Praxair, Inc. (Praxair). Successful remediation activities at Linde combines meeting cleanup objectives of radiological contamination while minimizing impacts to Praxair business operations. The unique use of Praxair's property coupled with an array of active and abandoned utilities poses many engineering and operational challenges; each of which has been overcome during the remedial action at Linde. The U.S. Army Corps of Engineers - Buffalo District (USACE) and CABRERA SERVICES, INC. (CABRERA) have successfully faced engineering challenges such as relocation of an aboveground structure, structural protection of an active water line, and installation of active mechanical, electrical, and communication utilities to perform remediation. As remediation nears completion, continued success of engineering challenges is critical as remaining activities exist in the vicinity of infrastructure essential to business operations; an electrical substation and duct bank providing power throughout the Praxair facility. Emphasis on engineering and operations through final remediation and into site restoration will allow for the safe and successful completion of the project. (authors)

  16. IFMIF : International Fusion Materials Irradiation Facility Conceptual Design Activity: Executive summary

    This report is a summary of the results of the Conceptual Design Activity (CDA) on the International Fusion Materials Irradiation Facility (IFMIF), conducted during 1995 and 1996. The activity is under the auspices of the International Energy Agency (IEA) Implementing Agreement for a Programme of Research and Development on Fusion Materials. An IEA Fusion Materials Executive Subcommittee was charged with overseeing the IFMIF-CDA work. Participants in the CDA are the European Union, Japan, and the United States, with the Russian Federation as an associate member

  17. IFMIF : International Fusion Materials Irradiation Facility Conceptual Design Activity: Final report

    Martone, M. [ENEA, Centro Ricerche Frascati, Rome (Italy)

    1997-01-01

    This report documents the results of the Conceptual Design Activity (CDA) on the International Fusion Materials Irradiation Facility (IFMIF), conducted during 1995 and 1996. The activity is under the auspices of the International Energy Agency (IEA) Implementing Agreement for a Programme of Research and Development on Fusion Materials. An IEA Fusion Materials Executive Subcommittee was charged with overseeing the IFMIF-CDA work. Participants in the CDA are the European Union, Japan, and the United States, with the Russian Federation as an associate member.

  18. IFMIF : International Fusion Materials Irradiation Facility Conceptual Design Activity: Final report

    This report documents the results of the Conceptual Design Activity (CDA) on the International Fusion Materials Irradiation Facility (IFMIF), conducted during 1995 and 1996. The activity is under the auspices of the International Energy Agency (IEA) Implementing Agreement for a Programme of Research and Development on Fusion Materials. An IEA Fusion Materials Executive Subcommittee was charged with overseeing the IFMIF-CDA work. Participants in the CDA are the European Union, Japan, and the United States, with the Russian Federation as an associate member

  19. Conference on instrumental activation analysis IAA 92

    The publication contains 26 abstracts primarily concerned with neutron activation analysis, although other analytical techniques based on X-ray fluorescence analysis, PIXE, PIGE, RBS are also included. Some contributions deal with aspects of quality practice and assurance in radioanalytical laboratories, with marketing of instrumental neutron activation analysis services, with hard- and software aspects of radiation detection, etc. (Z.S.)

  20. Acquisition Path Analysis for a SFR Fuel Manufacturing Facility

    Chang, H. L.; Kwon, E. H.; Ahn, S. K.; Ko, W. I.; Kim, H. D. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    The coarse acquisition path analysis does not claim to be complete, but it identifies plausible acquisition paths detailed enough to show that the acquisition path analysis can provide reasonable insights regarding the safeguardability assessment, and demonstrates the availability of safeguards tools and measures, although not complete, required for the implementation of effective and efficient safeguards, including the coverage of the nuclear energy system (NES) by multiple intrinsic features and extrinsic measures. It also identifies strengths, weaknesses and gaps of a system in the area of proliferation resistance in a generally understandable form. The acquisition path analysis demonstrates that all acceptance limits for the safeguardability, in principle, are met although the acceptance limit for the efficiency of the IAEA safeguards can be answered only at the end of the Safeguards-by-Design process, including interaction with IAEA operations. However, procedures for destructive assay (DA) for the verification by the IAEA are not defined. Target values for non-destructive assay (NDA) for this type of nuclear material are also not defined. Therefore, there is a need to finish demonstrations of NDA measurements on novel material types and material flows. The acquisition path analysis also shows some concerns that need to be assured in the system design process: e. g., the ID number of all storage containers in all storage positions can be read or checked without moving the storage container, transfer of TRU fuel and heel/scrap (product stream) should be strictly separated from transfer routes for waste, to make the transfer of TRU fuel and heel/scrap into waste container impossible, etc.

  1. Transuranic-contaminated solid waste Treatment Development Facility. Final safety analysis report

    The Final Safety Analysis Report (FSAR) for the Transuranic-Contaminated Solid-Waste Treatment Facility has been prepared in compliance with the Department of Energy (DOE) Manual Chapter 0531, Safety of Nonreactor Nuclear Facilities. The Treatment Development Facility (TDF) at the Los Alamos Scientific Laboratory is a research and development facility dedicated to the study of radioactive-waste-management processes. This analysis addresses site assessment, facility design and construction, and the design and operating characteristics of the first study process, controlled air incineration and aqueous scrub off-gas treatment with respect to both normal and accident conditions. The credible accidents having potentially serious consequences relative to the operation of the facility and the first process have been analyzed and the consequences of each postulated credible accident are presented. Descriptions of the control systems, engineered safeguards, and administrative and operational features designed to prevent or mitigate the consequences of such accidents are presented. The essential features of the operating and emergency procedures, environmental protection and monitoring programs, as well as the health and safety, quality assurance, and employee training programs are described

  2. Transuranic-contaminated solid waste Treatment Development Facility. Final safety analysis report

    Warner, C.L. (comp.)

    1979-07-01

    The Final Safety Analysis Report (FSAR) for the Transuranic-Contaminated Solid-Waste Treatment Facility has been prepared in compliance with the Department of Energy (DOE) Manual Chapter 0531, Safety of Nonreactor Nuclear Facilities. The Treatment Development Facility (TDF) at the Los Alamos Scientific Laboratory is a research and development facility dedicated to the study of radioactive-waste-management processes. This analysis addresses site assessment, facility design and construction, and the design and operating characteristics of the first study process, controlled air incineration and aqueous scrub off-gas treatment with respect to both normal and accident conditions. The credible accidents having potentially serious consequences relative to the operation of the facility and the first process have been analyzed and the consequences of each postulated credible accident are presented. Descriptions of the control systems, engineered safeguards, and administrative and operational features designed to prevent or mitigate the consequences of such accidents are presented. The essential features of the operating and emergency procedures, environmental protection and monitoring programs, as well as the health and safety, quality assurance, and employee training programs are described.

  3. Thermal neutron activation experiments on Ag, In, Cs, Eu, V, Mo, Zn, Sn and Zr in the MINERVE facility

    Leconte Pierre

    2016-01-01

    Full Text Available The MAESTRO experimental program has been designed to improve nuclear data uncertainty on a large range of materials used for detection, absorption, moderation and structures in LWRs. It consists of pile-oscillation and neutron activation experiments, carried out in the MINERVE low power facility. For this program, the core configuration has been designed to be representative of HZP (Hot Zero Power conditions of a typical PWR. Samples of high purity elements have been manufactured with severe technological constraints to reach a target accuracy of ±2% (1σ on the measurement. This paper presents a preliminary analysis of activation experiments, based on TRIPOLI4 Monte-Carlo calculations and various nuclear data libraries.

  4. Thermal neutron activation experiments on Ag, In, Cs, Eu, V, Mo, Zn, Sn and Zr in the MINERVE facility

    Leconte, Pierre; Geslot, Benoit; Gruel, Adrien; Pépino, Alexandra

    2016-03-01

    The MAESTRO experimental program has been designed to improve nuclear data uncertainty on a large range of materials used for detection, absorption, moderation and structures in LWRs. It consists of pile-oscillation and neutron activation experiments, carried out in the MINERVE low power facility. For this program, the core configuration has been designed to be representative of HZP (Hot Zero Power) conditions of a typical PWR. Samples of high purity elements have been manufactured with severe technological constraints to reach a target accuracy of ±2% (1σ) on the measurement. This paper presents a preliminary analysis of activation experiments, based on TRIPOLI4 Monte-Carlo calculations and various nuclear data libraries.

  5. Clustering Logistics Facilities in a Metropolitan Area via a Hot - Spot Analysis

    İsmail Önden

    2014-12-01

    Full Text Available Hot spots can be described as the high attraction points. Defining the hot spots and clustering approaches in a metropolitan area helps to provide solutions for balancing the freight flows between the sub-areas in the city center. It also helps to provide solutions for secondary problems, such as traffic congestion and air pollution. The technique assists decision-makers in making inferences about the city’s future and taking precautions for sustainability. In this paper, a geographic information systems (GIS analysis tool, spatial statistics based on the Getis Ord* statistics, is used to illustrate which part of Istanbul has hot spots. The hot spot identification is based on logistics activities at the locations of the logistics facilities. The outputs of the analysis are discussed within the context of logistics costs and environmental effects.Anahtar Kelimeler: Mağaza yeri seçimi, delphi tekniği, makro analiz, mağaza yeri seçim kriterleri .

  6. Safety analysis report for the 238PuO2 fuel form facility

    The Plutonium Fuel Form (PuFF) Facility has been constructed at the Savannah River Plant to manufacture 30 to 60 kg/yr of 238Pu fuel forms for space power applications. This facility is located in the existing Building 235-F near the geographical center of the Savannah River Plant (SRP) site. Pilot production is scheduled to begin in July 1977, with full-scale production in April 1978. The process line of the facility consists of nine separate, interconnected shielded cells; five shielded wing cabinets or glove boxes; three hoods; and contained auxiliary equipment. These process enclosures will be, for the most part, under an atmosphere of recirculating inert gas. The products of the facility will be dense fuel forms manufactured from PuO2 powder with a nominal isotopic composition of 80% 238Pu-20% 239Pu. This powder, made from calcined oxalate, has been produced safely at a rate of about 20 kg 238Pu/yr in the H-Area B-Line in Building 221-H for approximately ten years. This report describes design objectives, nature of operations, potential hazards and limiting factors, facility response to postulated accidents and failures, and environmental effects. The results of the analyses described in this report indicate that the facility has the capacity to prevent or sufficiently reduce accidents that represent potential risks to health and safety. The safety analysis in conjunction with process requirements provides the bases for Technical Standards for operation. The analysis also documents the degree of conformance of the facility design with the General Design Criteria - Plutonium Facilities and the Environmental Statement

  7. Training and accreditation activities at the Department of Energy Category A Reactors and Nuclear Facilities

    A new era dawned within the Department of Energy (DOE) in 1989 when DOE Order 5480.18A 'Accreditation of Performance-Based Training for Category A Reactors and Nuclear Facilities' was issued. This new era emphasized the importance of personnel training and qualification in maintaining the continued safe and efficient operation of the diverse nuclear facilities within the DOE complex. This approach to the design, development, and implementation of training is very similar to the approach that has proven to be very successful within the commercial nuclear industry. During the 1980s in the aftermath of Three Mile Island (TMI), DOE made a significant effort to conduct its mission in an environmentally safe manner and to increase the existing level of protection of the workers and the public. The DOE, like the commercial nuclear industry, realized that a nuclear accident anywhere in the U.S.A. would negatively impact the public confidence in the entire industry. This effort has not been easy because of changes within DOE and changes in regulatory requirements. Difficulties include aging facilities, outdated equipment, ingrained operating habits, inflexible culture, outdated or nonexistent procedures, stagnated management structure, lack of technical specifications and safety analysis reports, no configuration control, informal and undocumented training and qualification processes, and a wide diversity of operating facilities. The commercial industry had problems adjusting to the new regulations after TMI, but none as challenging as those facing DOE. This paper centers on the importance and status of accreditation within the DOE community and the efforts to develop and implement a performance-based approach to the training of the personnel at these facilities

  8. The calibration of DD neutron indium activation diagnostic for Shenguang-III facility

    Song, Zi-Feng; Liu, Zhong-Jie; Zhan, Xia-Yu; Tang, Qi

    2014-01-01

    The indium activation diagnostic was calibrated on an accelerator neutron source in order to diagnose deuterium-deuterium (DD) neutron yields of implosion experiments on Shenguang-III facility. The scattered neutron background of the accelerator room was measured by placing a polypropylene shield in front of indium sample, in order to correct the calibrated factor of this activation diagnostic. The proper size of this shield was given by Monte Carlo simulation software. The affect from some other activated nuclei on the calibration was verified by judging whether the measured curve obeys exponential decay and contrasting the half life of the activated sample. The calibration results showed that the linear range reached up to 100 cps net count rate in the full energy peak of interest, the scattered neutron background of accelerator room was about 9% of the total neutrons and the possible interferences mixed scarcely in the sample. Subtracting the portion induced by neutron background, the calibrated factor of ...

  9. Safety research activities for Japanese regulations of spent fuel interim storage facilities

    Japan Nuclear Energy Safety Organization (JNES) carries out (a) preparation of technical documents, (b) technical evaluations of standards (prepared by academic societies), etc. and (c) other R and D activities, to support Nuclear and Industrial Safety Agency (NISA: which prepares necessary regulations for Spent Fuel Interim Storage Facilities). In 2011 fiscal year, JNES carried out R and D for UT test of welded canister lids which is required for concrete cask storage facilities. And also, JNES carried out dynamic test of spent fuel to examine the integrity of spent fuel at cask drop accidents and PWR spent fuel storage test to prove long term integrity of spent fuel and cask itself. Some of these tests will be carried out in 2012 fiscal year and after. (author)

  10. The physical protection of nuclear material and nuclear facilities including activities to combat nuclear terrorism

    The paper describes present of physical protection of nuclear facilities and materials in the Czech Republic; the basic concept and regulation in physical protection and the effort made to strengthen the national regulatory programmes; the role of the police as a response force and the role of the new private security companies; the upgrading of the physical protection systems at the different types of the nuclear installations to fulfill the more strict requirements of the new Atomic Law No. 18/1997 Coll. and Regulation No. 144/1997 Coll., on physical protection of nuclear materials and nuclear facilities; activities carried out in connection with governmental decision No. 479 dated 19 May 2004 on National action plan to combat terrorism. (author)

  11. Appendix 1. The initial and final state of the nuclear facility and the planned follow-on and time-binding activities to achieve the ultimate state of the nuclear facility in that phase, including their impact on employees of nuclear facilities and surroundings nuclear facility

    In this chapter the initial and final state of the NPP A-1 and the planned follow-on and time-binding activities to achieve the ultimate state of the nuclear facility in that phase, including their impact on employees of nuclear facilities and surroundings nuclear facility are reviewed.

  12. Instrumental neutron activation analysis - a routine method

    This thesis describes the way in which at IRI instrumental neutron activation analysis (INAA) has been developed into an automated system for routine analysis. The basis of this work are 20 publications describing the development of INAA since 1968. (Auth.)

  13. Data analysis for remote monitoring of safeguarded facilities

    The International Remote Monitoring Project (IRMP) sponsored by the US DOE allows DOE and its international partners to gain experience with the remote collection, transmission, and interpretation of safeguards-relevant data. This paper focuses on the interpretation of the data from these remote monitoring systems. Users of these systems need to be able to ascertain that the remote monitoring system is functioning as expected and that the events generated by the sensors are consistent with declared activity. The initial set of analytical tools being provided for IRMP installations this year include a suite of automatically generated views of user-selected data. The baseline set of tools, with illustrative examples, will be discussed. Plans for near-term enhancements will also be discussed. Finally, the applicability of more advanced analytical techniques such as expert systems will be discussed

  14. Criticality safety analysis of WWER spent fuel facilities

    Criticality safety analysis of the WWER transport casks as well as of WWER wet spent fuel storage at the Kozloduy NPP site is performed under conservative assumptions. For criticality calculations the methodology based on the modular code system SCALE4.4 and MCNP code has been applied. The criticality parameters for WWER-440 and WWER-1000 transport casks, and for WWER wet spent fuel storage have been calculated by SCALE. The verification of the results has been carried out based on the comparison with the MCNP results. This comparison shows good coincidence of the results, calculated by SCALE and MCNP, in the uncertainty margins. The results presented lead to the conclusion that the criticality safety criteria Keff<0.95 for WWER transport casks and WWER wet spent fuel storage are satisfied quite well under conservative assumptions

  15. Development of HANARO Activation Analysis System and Utilization Technology

    1. Establishment of evaluation system using a data for a neutron activation analysis : Improvement of NAA measurement system and its identification, Development of combined data evaluation code of NAA/PGAA, International technical cooperation project 2. Development of technique for a industrial application of high precision gamma nuclide spectroscopic analysis : Analytical quality control, Development of industrial application techniques and its identification 3. Industrial application research for a prompt gamma-ray activation analysis : Improvement of Compton suppression counting system (PGAA), Development of applied technology using a PGAA system 4. Establishment of NAA user supporting system and KOLAS management : Development and validation of KOLAS/ISO accreditation testing and identification method, Cooperation researches for a industrial application, Establishment of integrated user analytical supporting system, Accomplishment of sample irradiation facility

  16. Development of HANARO Activation Analysis System and Utilization Technology

    Chung, Y. S.; Moon, J. H.; Cho, H. J. (and others)

    2007-06-15

    1. Establishment of evaluation system using a data for a neutron activation analysis : Improvement of NAA measurement system and its identification, Development of combined data evaluation code of NAA/PGAA, International technical cooperation project 2. Development of technique for a industrial application of high precision gamma nuclide spectroscopic analysis : Analytical quality control, Development of industrial application techniques and its identification 3. Industrial application research for a prompt gamma-ray activation analysis : Improvement of Compton suppression counting system (PGAA), Development of applied technology using a PGAA system 4. Establishment of NAA user supporting system and KOLAS management : Development and validation of KOLAS/ISO accreditation testing and identification method, Cooperation researches for a industrial application, Establishment of integrated user analytical supporting system, Accomplishment of sample irradiation facility.

  17. Computational Tools and Facilities for the Next-Generation Analysis and Design Environment

    Noor, Ahmed K. (Compiler); Malone, John B. (Compiler)

    1997-01-01

    This document contains presentations from the joint UVA/NASA Workshop on Computational Tools and Facilities for the Next-Generation Analysis and Design Environment held at the Virginia Consortium of Engineering and Science Universities in Hampton, Virginia on September 17-18, 1996. The presentations focused on the computational tools and facilities for analysis and design of engineering systems, including, real-time simulations, immersive systems, collaborative engineering environment, Web-based tools and interactive media for technical training. Workshop attendees represented NASA, commercial software developers, the aerospace industry, government labs, and academia. The workshop objectives were to assess the level of maturity of a number of computational tools and facilities and their potential for application to the next-generation integrated design environment.

  18. Current status of neutron activation analysis in HANARO Research Reactor

    The facilities for neutron activation analysis in the HANARO (Hi-flux Advanced Neutron Application Research Reactor) are described and the main applications of NAA (Neutron Activation Analysis) are reviewed. The sample irradiation tube, automatic and manual pneumatic transfer system were installed at three irradiation holes of HANARO at the end of 1995. The performance of the NAA facility was examined to identify the characteristics of the tube transfer system, irradiation sites and custom-made polyethylene irradiation capsule. The available thermal neutron fluxes at irradiation sites are in the range of 3 x 1013 - 1 x 1014 n/cm2·s and cadmium ratios are in 15 - 250. For an automatic sample changer for gamma-ray counting, a domestic product was designed and manufactured. An integrated computer program (Labview) to analyse the content was developed. In 2001, PGNAA (Prompt Gamma Neutron Activation Analysis) facility has been installed using a diffracted neutron beam of ST1. NAA has been applied in the trace component analysis of nuclear, geological, biological, environmental and high purity materials, and various polymers for research and development. The improvement of analytical procedures and establishment of an analytical quality control and assurance system were studied. Applied research and development for the environment, industry and human health by NAA and its standardization were carried out. For the application of the KOLAS (Korea Laboratory Accreditation Scheme), evaluation of measurement uncertainty and proficiency testing of reference materials were performed. Also to verify the reliability and to validate analytical results, intercomparison studies between laboratories were carried out. (author)

  19. Current status of neutron activation analysis in HANARO Research Reactor

    Chung, Yong Sam; Moon, Jong Hwa; Sohn, Jae Min [Korea Atomic Energy Research Institute, Daejeon (Korea)

    2003-03-01

    The facilities for neutron activation analysis in the HANARO (Hi-flux Advanced Neutron Application Research Reactor) are described and the main applications of NAA (Neutron Activation Analysis) are reviewed. The sample irradiation tube, automatic and manual pneumatic transfer system were installed at three irradiation holes of HANARO at the end of 1995. The performance of the NAA facility was examined to identify the characteristics of the tube transfer system, irradiation sites and custom-made polyethylene irradiation capsule. The available thermal neutron fluxes at irradiation sites are in the range of 3 x 10{sup 13} - 1 x 10{sup 14} n/cm{sup 2}{center_dot}s and cadmium ratios are in 15 - 250. For an automatic sample changer for gamma-ray counting, a domestic product was designed and manufactured. An integrated computer program (Labview) to analyse the content was developed. In 2001, PGNAA (Prompt Gamma Neutron Activation Analysis) facility has been installed using a diffracted neutron beam of ST1. NAA has been applied in the trace component analysis of nuclear, geological, biological, environmental and high purity materials, and various polymers for research and development. The improvement of analytical procedures and establishment of an analytical quality control and assurance system were studied. Applied research and development for the environment, industry and human health by NAA and its standardization were carried out. For the application of the KOLAS (Korea Laboratory Accreditation Scheme), evaluation of measurement uncertainty and proficiency testing of reference materials were performed. Also to verify the reliability and to validate analytical results, intercomparison studies between laboratories were carried out. (author)

  20. Who Enters Campus Recreation Facilities: A Demographic Analysis

    Paul Rohe Milton

    2011-06-01

    Full Text Available The purpose of this study was to examine student entry into a campus recreation center based on seven demographics (gender, ethnicity, age, class standing, intercollegiate athlete vs. non-athlete, students with self-reported disability vs. non-disability, and campus residence in order to determine who would be most likely to enter the recreation center. Subjects were from a mid-western, four year state-assisted institution with combined enrollment of 23,932 undergraduate and graduate students. Of the 23,932 enrolled, 14,032 students were examined in this study. Information on student entry to the recreation center was collected through the university’s student information system. Data was analyzed and interpreted using chi-square analysis. Results of the study show statistically significant differences in the demographics except the student disability demographic. More males than females, more African Americans than other ethnicities, more traditionally aged (18-25 students than non-traditional students, more underclassmen than seniors, more athletes and non-athletes, more residents than commuters were likely to enter the campus recreation center. The findings in this study could be used by collegiate recreational sport directors and administrators, in the United States and internationally, for future ideas about programming in similar recreation settings.

  1. Analysis facility infrastructure (Tier-3) for ATLAS experiment

    González de la Hoza, S; Ros, E; Sánchez, J; Amorós, G; Fassi, F; Fernández, A; Kaci, M; Lamas, A; Salt, J

    2008-01-01

    In the ATLAS computing model the tiered hierarchy ranged from the Tier-0 (CERN) down to desktops or workstations (Tier-3). The focus on defining the roles of each tiered component has evolved with the initial emphasis on the Tier-0 and Tier-1 definition and roles. The various LHC (Large Hadron Collider) projects, including ATLAS, then evolved the tiered hierarchy to include Tier-2’s (Regional centers) as part of their projects. Tier-3 centres, on the other hand, have been defined as whatever an institution could construct to support their Physics goals using institutional and otherwise leveraged resources and therefore have not been considered to be part of the official ATLAS computing resources. However, Tier-3 centres are going to exist and will have implications on how the computing model should support ATLAS physicists. Tier-3 users will want to access LHC data and simulations and will want to enable their resources to support their analysis and simulation work. This document will define how IFIC (Insti...

  2. Opening address [International conference on lessons learned from the decommissioning of nuclear facilities and the safe termination of nuclear activities

    With the end of life approaching for many facilities, the development and implementation of a holistic approach to decommissioning and the termination of nuclear activities is essential, not only for large nuclear facilities, but also - and in particular - for small facilities, for which resources and safety and security measures are limited. The holistic approach refers not only to the time dimension of the life cycle of a specific facility, but also to the long-term sustainability of the whole system in the country and the region, including the possible recycling of material and multinational or regional cooperation. It should also comprehensively cover the technical, financial, social and political aspects of decommissioning

  3. Availability of exercise facilities and physical activity in 2,037 adults: cross-sectional results from the Swedish neighborhood and physical activity (SNAP) study

    Eriksson Ulf; Arvidsson Daniel; Sundquist Kristina

    2012-01-01

    Abstract Background Exercise facilities may have the potential to promote physical activity among residents, and to support an active lifestyle throughout the year. We investigated the association between objectively assessed availability of exercise facilities and objectively assessed physical activity outcomes, and whether time of year had a modifying effect on these associations. Methods A total of 2,037 adults (55% females) wore an accelerometer for seven days. Time spent in moderate to v...

  4. Safety analysis report for the cold vacuum drying facility, phase 1, supporting civil/structural construction

    The Cold Vacuum Drying Facility is a subproject of the overall Spent Nuclear Fuel Project. This Phase 2 Safety Analysis Report incorporates the CVD systems design and will update the SAR per DOE Order 5480.23 for manual and other Hanford infrastructure changes

  5. Hazards Analysis for the Spent Nuclear Fuel L-Experimental Facility

    The purpose of this Hazard Analysis (HA) is to identify and assess potential hazards associated with the operations of the Spent Nuclear Fuels (SNF) Treatment and Storage Facility LEF. Additionally, this HA will be used for identifying and assessing potential hazards and specifying functional attributes of SSCs for the LEF project

  6. Sampling and Analysis Instruction for Asbestos-Containing Materials from Surveillance Maintenance and Transition Facilities

    The purpose of this sampling and analysis instruction is to define the waste characterization requirements for disposition of asbestos-containing material in the form of thermal system insulation and transite cement asbestos board found in or near the Hanford Site facilities

  7. Evaluation of syngas production unit cost of bio-gasification facility using regression analysis techniques

    Deng, Yangyang; Parajuli, Prem B.

    2011-08-10

    Evaluation of economic feasibility of a bio-gasification facility needs understanding of its unit cost under different production capacities. The objective of this study was to evaluate the unit cost of syngas production at capacities from 60 through 1800Nm 3/h using an economic model with three regression analysis techniques (simple regression, reciprocal regression, and log-log regression). The preliminary result of this study showed that reciprocal regression analysis technique had the best fit curve between per unit cost and production capacity, with sum of error squares (SES) lower than 0.001 and coefficient of determination of (R 2) 0.996. The regression analysis techniques determined the minimum unit cost of syngas production for micro-scale bio-gasification facilities of $0.052/Nm 3, under the capacity of 2,880 Nm 3/h. The results of this study suggest that to reduce cost, facilities should run at a high production capacity. In addition, the contribution of this technique could be the new categorical criterion to evaluate micro-scale bio-gasification facility from the perspective of economic analysis.

  8. Safety analysis of the Los Alamos critical experiments facility: burst operation of Skua

    A detailed consideration of the Skua burst assembly is presented, thereby supplementing the facility safety analysis report covering the operation of other critical assemblies at Los Alamos. As with these assemblies the small fission-product inventory, ambient pressure, and moderate temperatures in Skua are amenable to straightforward measures to ensure the protection of the public

  9. Safety analysis of the Los Alamos critical experiments facility: burst operation of Skua

    Detailed consideration of the Skua burst assembly is provided, thereby supplementing the facility Safety Analysis Report covering the operation of other critical assemblies at the Los Alamos Scientific Laboratory. As with these assemblies the small fission-product inventory, ambient pressure, and moderate temperatures in Skua are amenable to straightforward measures to ensure the protection of the public

  10. Application of the Management System for Facilities and Activities. Safety Guide

    This publication provides guidance for following the requirements for management systems that integrate safety, health, security, quality assurance and environmental objectives. A successful management system ensures that nuclear safety matters are not dealt with in isolation but are considered within the context of all these objectives. The aim of this publication is to assist Member States to establish and implement effective management systems that coherently integrate all aspects of managing nuclear facilities and activities. Contents: 1. Introduction; 2. Management system; 3. Management responsibility; 4. Resource management; 5. Process implementation; 6. Measurement, assessment and improvement; Appendix I: Transition to an integrated management system; Appendix II: Activities in the document control process; Appendix III: Activities in the procurement process; Appendix IV: Performance of independent assessments; Annex I: Electronic document management system; Annex II: Media for record storage; Annex III: Record retention and storage; Glossary.

  11. Denitrification: a Clean-Up Mechanism for High Nitrate Ground Water Near an Active Swine Facility?

    Townsend, M. A.

    2001-05-01

    An active swine facility in south central Kansas appears to be cleaning up nitrate in regional ground water in an area with shallow ground water (water chemistry to determine the impact of a bentonite lined hog lagoon on shallow ground-water chemistry. Regional ground water surrounding the facility had nitrate-nitrogen values routinely measured above 10 ppm. Chloride concentrations in the area ranged from 3 to 25 ppm and bicarbonate values ranged from 45 to 200 ppm. Two periods of sampling in the area showed nitrogen isotope values in the fertilizer range (waste range (+13 to +20) which is similar to the value measured for the waste lagoon (+18). Chloride and bicarbonate values at all of the monitoring wells, except the well downgradient from the lagoon, were similar to the regional ground water. The lagoon water had >500 ppm chloride and >1400 ppm ammonium-N. The downgradient monitoring well had chloride values > 100 ppm and bicarbonate values above 400 ppm for the two sampling periods. Use of chloride ratios showed that approximately 30% of the water contributing to the downgradient well sample was from lagoon leakage. Preliminary calculations of the amount of bicarbonate resulting from denitrification processes, chloride ratios, and nitrogen isotope values suggest that the sampled water is a mixture of denitrified regional ground water plus lagoon water. Although the nitrate values near the swine facility appear to be decreasing, the long-term impact of increased salt load on the regional ground water is unknown at this time.

  12. Facile synthesis of AIE-active amphiphilic polymers: Self-assembly and biological imaging applications.

    Long, Zi; Liu, Meiying; Wang, Ke; Deng, Fengjie; Xu, Dazhuang; Liu, Liangji; Wan, Yiqun; Zhang, Xiaoyong; Wei, Yen

    2016-09-01

    In this work, we reported a rather facile method for fabrication of ultrabright, well dispersible and biocompatible fluorescent organic nanoparticles (FONs) with aggregation-induced emission (AIE) properties through combination of esterification and ring-opening reaction. The hydroxyl groups of Pluronic F127 was first reacted with the chloride of trimellitic anhydride chloride (TMAC), and its anhydride groups were further reacted with the amino groups of amino-terminated AIE dye (PhNH2) through ring-opening reaction. The optical properties, biocompatibility as well as cell uptake behavior of these obtained AIE-active nanoparticles (F127-TMAC-PhNH2 FONs) were examined by a series of characterization techniques and assays. We demonstrated that uniform organic nanoparticles with high water dispersibility, strong luminescence and desirable biocompatibility can be facilely obtained, which are promising for biological imaging applications. More importantly, a number of carboxyl groups were introduced into these AIE-active nanoparticles, which can be further utilized for further conjugation reaction and carrying anticancer drugs such as cisplatin. Therefore, the strategy of described in this work should be a simple and useful route for fabrication of multifunctional AIE-active luminescent nanotheranostic systems. PMID:27207057

  13. Access to water source, latrine facilities and other risk factors of active trachoma in Ankober, Ethiopia.

    Ilya Golovaty

    Full Text Available OBJECTIVE: This study aims to determine the prevalence and correlates of active trachoma in Ankober, Ethiopia. METHODS: 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. RESULTS: 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. CONCLUSION: 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.

  14. Determination of Importance Evaluation for Exploratory Studies Facility (ESF) Subsurface Testing Activities

    This Determination of Importance Evaluation (DIE) applies to the Subsurface Exploratory Studies Facility (ESF), encompassing the Topopah Spring (TS) Loop from Station 0+00 meters (m) at the North Portal to breakthrough at the South Portal (approximately 78+77 m), and ancillary test and operation support areas including the Enhanced Characterization of the Repository Block (ECRB) Cross Drift. This evaluation applies specifically to site characterization testing activities ongoing and planned in the Subsurface ESF. ESF site characterization activities are being performed to obtain the information necessary to determine whether the Yucca Mountain Site is suitable as a geologic repository for spent nuclear fuel and high-level radioactive waste. A more detailed description of these testing activities is provided in Section 6 of this DIE. Generally, the construction and operation of excavations associated with these testing activities are evaluated in the DIE for the Subsurface ESF (CRWMS M and O 1999a) and the DIE for the ESF ECRB Cross Drift (CRWMS M and O 2000a). The scope of this DIE also entails the proposed Unsaturated Zone (UZ) Transport Test at Busted Butte. Although, not a part of the TS Loop or ECRB Cross Drift, the associated testing activities are Subsurface testing activities. Busted Butte is located to the south south-east of the TS Loop and is outside the Conceptual Controlled Area Boundary (CCAB). These activities provide access to the Calico Hills (CH) geologic structure. In the case of Busted Butte, construction and operation of excavations are evaluated herein (since this activity was not previously evaluated in CRWMS M and O 1999a). The objectives of this DIE are to determine whether Subsurface ESF testing, and associated activities, could potentially impact site characterization testing and/or the waste isolation capabilities of the site. Controls needed to limit any potential impacts are identified in Section 13. The validity and veracity of the

  15. Determination of Importance Evaluation for Exploratory Studies Facility (ESF) Subsurface Testing Activities

    C.J. Byrne

    2001-02-20

    This Determination of Importance Evaluation (DIE) applies to the Subsurface Exploratory Studies Facility (ESF), encompassing the Topopah Spring (TS) Loop from Station 0+00 meters (m) at the North Portal to breakthrough at the South Portal (approximately 78+77 m), and ancillary test and operation support areas including the Enhanced Characterization of the Repository Block (ECRB) Cross Drift. This evaluation applies specifically to site characterization testing activities ongoing and planned in the Subsurface ESF. ESF site characterization activities are being performed to obtain the information necessary to determine whether the Yucca Mountain Site is suitable as a geologic repository for spent nuclear fuel and high-level radioactive waste. A more detailed description of these testing activities is provided in Section 6 of this DIE. Generally, the construction and operation of excavations associated with these testing activities are evaluated in the DIE for the Subsurface ESF (CRWMS M&O 1999a) and the DIE for the ESF ECRB Cross Drift (CRWMS M&O 2000a). The scope of this DIE also entails the proposed Unsaturated Zone (UZ) Transport Test at Busted Butte. Although, not a part of the TS Loop or ECRB Cross Drift, the associated testing activities are Subsurface testing activities. Busted Butte is located to the south south-east of the TS Loop and is outside the Conceptual Controlled Area Boundary (CCAB). These activities provide access to the Calico Hills (CH) geologic structure. In the case of Busted Butte, construction and operation of excavations are evaluated herein (since this activity was not previously evaluated in CRWMS M&O 1999a). The objectives of this DIE are to determine whether Subsurface ESF testing, and associated activities, could potentially impact site characterization testing and/or the waste isolation capabilities of the site. Controls needed to limit any potential impacts are identified in Section 13. The validity and veracity of the individual

  16. Annual Report To Congress. Department of Energy Activities Relating to the Defense Nuclear Facilities Safety Board, Calendar Year 2003

    None, None

    2004-02-28

    The Department of Energy (Department) submits an Annual Report to Congress each year detailing the Department’s activities relating to the Defense Nuclear Facilities Safety Board (Board), which provides advice and recommendations to the Secretary of Energy (Secretary) regarding public health and safety issues at the Department’s defense nuclear facilities. In 2003, the Department continued ongoing activities to resolve issues identified by the Board in formal recommendations and correspondence, staff issue reports pertaining to Department facilities, and public meetings and briefings. Additionally, the Department is implementing several key safety initiatives to address and prevent safety issues: safety culture and review of the Columbia accident investigation; risk reduction through stabilization of excess nuclear materials; the Facility Representative Program; independent oversight and performance assurance; the Federal Technical Capability Program (FTCP); executive safety initiatives; and quality assurance activities. The following summarizes the key activities addressed in this Annual Report.

  17. Safety research activities for Japanese regulations of spent fuel interim storage facilities

    Japan Nuclear Energy Safety Organization (JNES) carries out (a) preparation of technical documents, (b) technical evaluations of standards (prepared by academic societies), etc. and (c) other R and D activities, to support Nuclear Regulation Authority (NRA: which controls the regulations for Spent Fuel Interim Storage Facilities). In 2012 fiscal year, JNES carried out dynamic test of spent fuel to examine the integrity of spent fuel under cask drop accidents, and preparation for PWR spent fuel storage test to prove long term integrity of spent fuel and cask itself. Some of these tests will be also carried out in 2013 fiscal year and after. (author)

  18. Transportation activity analysis using smartphones

    Xiao, Yu; Low, David; Bandara, Thusitha; Pathak, Parth; Lim, Hock Beng; Goyal, Devendra; Santos, Jorge Oliveira; Cottrill, Caitlin; Pereira, Francisco C.; Zegras, P. Christopher; Ben-Akiva, Moshe E.

    2012-01-01

    Transportation activity surveys investigate when, where and how people travel in urban areas to provide information necessary for urban transportation planning. In Singapore, the Land Transport Authority (LTA) carries out such a survey amongst households every four years. The survey is conducted through conventional questionnaires and travel diaries. However, the conventional surveys are problematic and error-prone. We are developing a smartphone-based transportation activity survey system to...

  19. Activation analysis with reactor neutrons

    The potentialities of neutron as an analytical probe are indicated, pointing out the need for development of other approaches, besides the conventional activation method. Development of instrumental approach to activation and applications, carried out at Analytical Chemistry Division are outlined. The role of, and the need for, the development and application of mathematical methods in enhancing the information content, and in turn the interpretation of the analytical results, is demonstrated. (author)

  20. Sociospatial distribution of access to facilities for moderate and vigorous intensity physical activity in Scotland by different modes of transport

    Lamb Karen E

    2012-07-01

    Full Text Available Abstract Background People living in neighbourhoods of lower socioeconomic status have been shown to have higher rates of obesity and a lower likelihood of meeting physical activity recommendations than their more affluent counterparts. This study examines the sociospatial distribution of access to facilities for moderate or vigorous intensity physical activity in Scotland and whether such access differs by the mode of transport available and by Urban Rural Classification. Methods A database of all fixed physical activity facilities was obtained from the national agency for sport in Scotland. Facilities were categorised into light, moderate and vigorous intensity activity groupings before being mapped. Transport networks were created to assess the number of each type of facility accessible from the population weighted centroid of each small area in Scotland on foot, by bicycle, by car and by bus. Multilevel modelling was used to investigate the distribution of the number of accessible facilities by small area deprivation within urban, small town and rural areas separately, adjusting for population size and local authority. Results Prior to adjustment for Urban Rural Classification and local authority, the median number of accessible facilities for moderate or vigorous intensity activity increased with increasing deprivation from the most affluent or second most affluent quintile to the most deprived for all modes of transport. However, after adjustment, the modelling results suggest that those in more affluent areas have significantly higher access to moderate and vigorous intensity facilities by car than those living in more deprived areas. Conclusions The sociospatial distributions of access to facilities for both moderate intensity and vigorous intensity physical activity were similar. However, the results suggest that those living in the most affluent neighbourhoods have poorer access to facilities of either type that can be reached on foot

  1. The effect of health facility delivery on neonatal mortality: systematic review and meta-analysis

    Tura Gurmesa

    2013-01-01

    Full Text Available Abstract Background Though promising progress has been made towards achieving the Millennium Development Goal four through substantial reduction in under-five mortality, the decline in neonatal mortality remains stagnant, mainly in the middle and low-income countries. As an option, health facility delivery is assumed to reduce this problem significantly. However, the existing evidences show contradicting conclusions about this fact, particularly in areas where enabling environments are constraint. Thus, this review was conducted with the aim of determining the pooled effect of health facility delivery on neonatal mortality. Methods The reviewed studies were accessed through electronic web-based search strategy from PUBMED, Cochrane Library and Advanced Google Scholar by using combination key terms. The analysis was done by using STATA-11. I2 test statistic was used to assess heterogeneity. Funnel plot, Begg’s test and Egger’s test were used to check for publication bias. Pooled effect size was determined in the form of relative risk in the random-effects model using DerSimonian and Laird's estimator. Results A total of 2,216 studies conducted on the review topic were identified. During screening, 37 studies found to be relevant for data abstraction. From these, only 19 studies fulfilled the preset criteria and included in the analysis. In 10 of the 19 studies included in the analysis, facility delivery had significant association with neonatal mortality; while in 9 studies the association was not significant. Based on the random effects model, the final pooled effect size in the form of relative risk was 0.71 (95% CI: 0.54, 0.87 for health facility delivery as compared to home delivery. Conclusion Health facility delivery is found to reduce the risk of neonatal mortality by 29% in low and middle income countries. Expansion of health facilities, fulfilling the enabling environments and promoting their utilization during childbirth are

  2. Determination of Importance Evaluation for Exploratory Studies Facility (ESF) Subsurface Testing Activities

    This Determination of Importance Evaluation (DIE) applies to the Subsurface Exploratory Studies Facility (ESF), encompassing the Topopah Spring (TS) Loop from Station 0+00 meters (m) at the North Portal to breakthrough at the South Portal (approximately 78+77 m), and ancillary test and operation support areas including the Enhanced Characterization of the Repository Block (ECRB) Cross Drift. This evaluation applies specifically to site characterization testing activities ongoing and planned in the Subsurface ESF. ESF site characterization activities are being performed to obtain the information necessary to determine whether the Yucca Mountain Site is suitable as a geologic repository for spent nuclear fuel and high-level radioactive waste. A more detailed description of these testing activities is provided in Section 6 of this DIE. Generally, the construction and operation of excavations associated with these testing activities are evaluated in the DIE for the Subsurface ESF (CRWMS M andO 1999a) and the DIE for the ESF ECRB Cross Drift (CRWMS M andO 2000a). The scope of this DIE also entails the proposed Unsaturated Zone (UZ) Transport Test at Busted Butte. Although, not a part of the TS Loop or ECRB Cross Drift, the associated testing activities are Subsurface testing activities. Busted Butte is located to the south south-east of the TS Loop and is outside the Conceptual Controlled Area Boundary (CCAB). These activities provide access to the Calico Hills (CH) geologic structure. In the case of Busted Butte, construction and operation of excavations are evaluated herein (since this activity was not previously evaluated in CRWMS M andO 1999a). The objectives of this DIE are to determine whether Subsurface ESF testing, and associated activities, could potentially impact site characterization testing and/or the waste isolation capabilities of the site. Controls needed to limit any potential impacts are identified in Section 13. The validity and veracity of the

  3. Determination of Importance Evaluation for Exploratory Studies Facility (ESF) Subsurface Testing Activities

    S. Goodin

    2002-07-22

    This Determination of Importance Evaluation (DIE) applies to the Subsurface Exploratory Studies Facility (ESF), encompassing the Topopah Spring (TS) Loop from Station 0+00 meters (m) at the North Portal to breakthrough at the South Portal (approximately 78+77 m), and ancillary test and operation support areas including the Enhanced Characterization of the Repository Block (ECRB) Cross Drift. This evaluation applies specifically to site characterization testing activities ongoing and planned in the Subsurface ESF. ESF site characterization activities are being performed to obtain the information necessary to determine whether the Yucca Mountain Site is suitable as a geologic repository for spent nuclear fuel and high-level radioactive waste. A more detailed description of these testing activities is provided in Section 6 of this DIE. Generally, the construction and operation of excavations associated with these testing activities are evaluated in the DIE for the Subsurface ESF (CRWMS M&O 1999a) and the DIE for the ESF ECRB Cross Drift (CRWMS M&O 2000a). The scope of this DIE also entails the proposed Unsaturated Zone (UZ) Transport Test at Busted Butte. Although, not a part of the TS Loop or ECRB Cross Drift, the associated testing activities are Subsurface testing activities. Busted Butte is located to the south south-east of the TS Loop and is outside the Conceptual Controlled Area Boundary (CCAB). These activities provide access to the Calico Hills (CH) geologic structure. In the case of Busted Butte, construction and operation of excavations are evaluated herein (since this activity was not previously evaluated in CRWMS M&O 1999a). The objectives of this DIE are to determine whether Subsurface ESF testing, and associated activities, could potentially impact site characterization testing and/or the waste isolation capabilities of the site. Controls needed to limit any potential impacts are identified in Section 13. The validity and veracity of the individual

  4. Preparation of Phased and Merged Safety Analysis Reports for New DOE Nuclear Facilities

    The Spent Nuclear Fuels Project (SNFP) is charged with moving to storage 2,100 metric tons of spent nuclear fuel elements left over from plutonium production at DOE'S Hanford site in Washington state. Two new facilities, the Cold Vacuum Drying Facility (CVDF) and the Canister Storage Building (CSB) are in final construction. In order to meet aggressive schedule commitments, the SNFP chose to prepare the safety analysis reports (SAR's) in phases that covered only specific portions of each facility's design as it was built. Each SAR also merged the preliminary and final safety analysis reports into a single SAR, thereby covering all aspects of design, construction, and operation for that portion (phase) of the facility. A policy of ''NRC equivalency'' was also implemented in parallel with this effort, with the goal of achieving a rigor of safety analysis equivalent to that of NRC-licensed fuel processing facilities. DOE Order 5480.23. ''Nuclear Safety Analysis Reports'' allows preparation of both a phased and a merged SAR to accelerate construction schedules. However, project managers must be aware that such acceleration is not guaranteed. Managers considering this approach for their project should be cognizant of numerous obstacles that will be encountered. Merging and phasing SAR's will create new, unique, and unanticipated difficulties which may actually slow construction unless expeditiously and correctly managed. Pitfalls to be avoided and good practices to be implemented in preparing phased and merged SAR's are presented. The value of applying NRC requirements to the DOE safety analysis process is also discussed. As of December, 1999, the SNFP has completed and approved a SAR for the CVDF. Approval of the SAR for the CSB is pending

  5. Preparation of Phased and Merged Safety Analysis Reports for New DOE Nuclear Facilities

    BISHOP, G.E.

    2000-04-04

    The Spent Nuclear Fuels Project (SNFP) is charged with moving to storage 2,100 metric tons of spent nuclear fuel elements left over from plutonium production at DOE'S Hanford site in Washington state. Two new facilities, the Cold Vacuum Drying Facility (CVDF) and the Canister Storage Building (CSB) are in final construction. In order to meet aggressive schedule commitments, the SNFP chose to prepare the safety analysis reports (SAR's) in phases that covered only specific portions of each facility's design as it was built. Each SAR also merged the preliminary and final safety analysis reports into a single SAR, thereby covering all aspects of design, construction, and operation for that portion (phase) of the facility. A policy of ''NRC equivalency'' was also implemented in parallel with this effort, with the goal of achieving a rigor of safety analysis equivalent to that of NRC-licensed fuel processing facilities. DOE Order 5480.23. ''Nuclear Safety Analysis Reports'' allows preparation of both a phased and a merged SAR to accelerate construction schedules. However, project managers must be aware that such acceleration is not guaranteed. Managers considering this approach for their project should be cognizant of numerous obstacles that will be encountered. Merging and phasing SAR's will create new, unique, and unanticipated difficulties which may actually slow construction unless expeditiously and correctly managed. Pitfalls to be avoided and good practices to be implemented in preparing phased and merged SAR's are presented. The value of applying NRC requirements to the DOE safety analysis process is also discussed. As of December, 1999, the SNFP has completed and approved a SAR for the CVDF. Approval of the SAR for the CSB is pending.

  6. Neutron activation analysis of biological substances

    A Bowen cabbage sample was used as a reference material for the neutron activation studies, and the method was checked by the analysis of other biological substances (blood or serum etc.). For nondestructive measurements also some non-trace elements were determined in order to decide whether the activation analysis is a useful means for such measurements. The new activation analysis procedure was used for biomedical studies as, e.g., for trace element determination in body fluids, and for the analysis of inorganic components in air samples. (R.P.)

  7. Analysis of kinetics experiments in LEU-HTR configurations of the PROTEUS facility

    On the recommendation of the International Atomic Energy Agency's (IAEA's) working group on gas-cooled reactors, the IAEA has established a coordinated research program (CRP) on the validation of safety-related physics calculations for low-enriched uranium (LEU)-fueled high-temperature gas-cooled reactors (HTGRs). The objective of the CRP is to provide safety-related physics data for LEU-fueled HTGRs for use in validating reactor physics codes and methods used by participating countries for analysis of their designs. At present, the main activities within the CRP are being carried out by the international project now under way at the PROTEUS critical facility at the Paul Scherrer Institute in Villigen, Switzerland. Within this project, critical experiments will be conducted for HTGR-LEU systems to determine core reactivity; flux and power profiles; reaction rate ratios; worth of control rods, including reflector control rods; worth of burnable poisons; and the effects of water ingress on these parameters. Of particular interest are quality-assured (benchmark type) measurements of the worth of control rods located in the graphite reflector. Two independent techniques have been selected for this purpose; the pulsed neutron source (PNS) and the inverse kinetics methods

  8. Guidance for preparation of safety analysis reports for nonreactor facilities and operations

    Department of Energy (DOE) Orders 5480.23, ''Nuclear Safety Analysis Reports,'' and 5481.1B, ''Safety Analysis and Review System'' require the preparation of appropriate safety analyses for each DOE operation and subsequent significant modifications including decommissioning, and independent review of each safety analysis. The purpose of this guide is to assist in the preparation and review of safety documentation for Oak Ridge Field Office (OR) nonreactor facilities and operation. Appendix A lists DOE Orders, NRC Regulatory Guides and other documents applicable to the preparation of safety analysis reports

  9. Pre-test analysis for the KNGR DVI performance test facility using FLUENT

    Pre-test analysis using a FLUENT code has been performed for the KGNR(Korean Next Generation Reactor) DVI(Direct Vessel Injection) performance test facility which is a full height and 1/24.3 volume scaled separate effect test facility. The ideal gas discharge condition is considered to simulation a steam discharge condition. The scale effects on the flow pattern, pressure distribution, and similarity for scaled model are numerically tested. From the various results for the scale effects, it was found that the similarity of hydraulics is founded

  10. Analysis of HR activities in selected company

    Jandová, Šárka

    2012-01-01

    The aim of this thesis is based on theoretical knowledge and analysis of the basic HR activities in the selected company to identify the strengths and weaknesses of the policies implemented personnel work. Then propose an effective ways of improving HR activities and employee satisfaction. Basic personal activities analyzed in this thesis are the adaptation, acquisition and selection of employees, performance management and staff appraisal, remuneration and training of staff. Analysis of pers...

  11. Analysis of Debris Trajectories at the Scaled Wind Farm Technology (SWiFT) Facility

    White, Jonathan R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Burnett, Damon J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2016-01-01

    Sandia National Laboratories operates the Scaled Wind Farm Technology Facility (SWiFT) on behalf of the Department of Energy Wind and Water Power Technologies Office. An analysis was performed to evaluate the hazards associated with debris thrown from one of SWiFT’s operating wind turbines, assuming a catastrophic failure. A Monte Carlo analysis was conducted to assess the complex variable space associated with debris throw hazards that included wind speed, wind direction, azimuth and pitch angles of the blade, and percentage of the blade that was separated. In addition, a set of high fidelity explicit dynamic finite element simulations were performed to determine the threshold impact energy envelope for the turbine control building located on-site. Assuming that all of the layered, independent, passive and active engineered safety systems and administrative procedures failed (a 100% failure rate of the safety systems), the likelihood of the control building being struck was calculated to be less than 5/10,000 and ballistic simulations showed that the control building would not provide passive protection for the majority of impact scenarios. Although options exist to improve the ballistic resistance of the control building, the recommendation is not to pursue them because there is a low probability of strike and there is an equal likelihood personnel could be located at similar distances in other areas of the SWiFT facility which are not passively protected, while the turbines are operating. A fenced exclusion area has been created around the turbines which restricts access to the boundary of the 1/100 strike probability. The overall recommendation is to neither relocate nor improve passive protection of the control building as the turbine safety systems have been improved to have no less than two independent, redundant, high quality engineered safety systems. Considering this, in combination with a control building strike probability of less than 5/10,000, the

  12. Xenon diffusion studies with prompt gamma activation analysis

    Developing a better understanding of xenon transport through porous systems is critical to predicting how this gas will enter the atmosphere after a below ground nuclear weapons test. Radioxenon monitoring is a vital part of the Comprehensive Nuclear-Test-Ban Treaty (CTBT) International Monitoring System. This work details the development of prompt gamma activation analysis for measuring the diffusion rates of xenon and argon gases through a porous medium. The University of Texas at Austin maintains a prompt gamma activation analysis facility with a peak neutron flux of ∼ 1.5 x 107 cm-2 s-1 and a beam diameter of 1 cm. Due to the relatively large prompt gamma cross sections of many stable xenon isotopes at thermal and sub-thermal neutron energies, prompt gamma activation analysis is a suitable technique for in situ non-destructive analysis of natural xenon. A test chamber has been designed and constructed to utilize prompt gamma activation analysis to measure xenon and argon diffusion through geological materials (e.g., sand, soil, etc.). Initial experiments have been conducted to determine the detection limits for stable gas measurements. The results from these experiments will be utilized to benchmark parts of a xenon transport model that is being used to determine diffusion coefficients for xenon and argon. (author)

  13. Conference on instrumental activation analysis - IAA 89

    The proceedings contain 40 abstracts of papers all of which have been incorporated in INIS. The papers were centred on the applications of radioanalytical methods, especialy on neutron activation analysis, X-ray fluorescence analysis, PIXE analysis and tracer techniques in biology, medicine and metallurgy, measuring instruments including microcomputers, and data processing methods. (J.P.)

  14. Conference on Instrumental Activation Analysis: IAA 89

    Vobecky, M.; Obrusnik, I.

    1989-05-01

    The proceedings contain 40 abstracts of papers all of which have been incorporated in INIS. The papers were centred on the applications of radioanalytical methods, especially on neutron activation analysis, x ray fluorescence analysis, PIXE analysis and tracer techniques in biology, medicine and metallurgy, measuring instruments including microcomputers, and data processing methods.

  15. The European contribution to the development and validation activities for the design of IFMIF lithium facility

    Miccichè, Gioacchino, E-mail: gioacchino.micciche@enea.it [EURATOM-ENEA, CR Brasimone I-40035 Camugnano, BO (Italy); Aiello, Antonio; Bernardi, Davide; Favuzza, Paolo; Agostini, Pietro [EURATOM-ENEA, CR Brasimone I-40035 Camugnano, BO (Italy); Frisoni, Manuela [EURATOM-ENEA, CR Bologna I-40129, BO Italy (Italy); Pinna, Tonio; Porfiri, MariaTeresa [EURATOM-ENEA, CR Frascati I-0044 Frascati, Roma (Italy); Tincani, Amelia [EURATOM-ENEA, CR Brasimone I-40035 Camugnano, BO (Italy); Di Maio, PieroAlessandro [University of Palermo, I-90128 Palermo (Italy); Knaepen, Bernard [Université libre de Bruxelles, I-1050 Bruxelles (Belgium)

    2013-10-15

    Highlights: • Engineering design of the target assembly. • Erosion, corrosion phenomena promoted by the lithium are studied. • Purification system implemented in the LiFus6 loop. • Study of the remote handling maintenance for the IFMIF TA. -- Abstract: The International Fusion Materials Irradiation Facility (IFMIF) is an accelerator-driven intense neutron source where candidate materials for fusion reactors will be tested and validated. The high energy neutron flux is produced by means of two deuteron beams (total current of 250 mA, energy of 40 MeV) that strikes a liquid lithium target circulating in a lithium loop of IFMIF plant. The European (EU) contribution to the development of the lithium facility comprises five procurement packages, as follow: (1) participation to the experimental activities of the EVEDA lithium test loop in Oarai (Japan); (2) study aimed at evaluating the corrosion and erosion phenomena, promoted by lithium, for structural fusion reference materials like AISI 316L and Eurofer; (3) design and validation of the lithium purification method with the aim to provide input data for the design of the purification system of IFIMF lithium loop; (4) design and validation of the remote handling (RH) procedures for the refurbishment/replacement of the EU concept of IFMIF target assembly including the design of the remote handling tools; (5) the engineering design of the European target assembly for IFMIF and the safety and RAMI analyses for the entire IFMIF lithium facility. The paper gives an overview of the status of the activities and of the main outcomes achieved so far.

  16. Control of pneumatic transfer system for neutron activation analysis

    Pneumatic transfer system(PTS) is one of the facilities to be used in irradiation of target materials for neutron activation analysis(NAA) in the research reactor. There are two systems the manual and the automatic system in PTS of HANARO research reactor. The pneumatic transfer system consists of many devices, sends and loads the capsules from NAA laboratory into three holes in the reflector tank of reactor and retrieves irradiated capsules after irradiation. This report describes the part's design, control system and the operation procedures. All the algorithm described in the text will be used for maintenance and upgrading

  17. Control of pneumatic transfer system for neutron activation analysis

    Jung, H. S.; Chung, Y. S.; Wu, J. S.; Kim, H. K.; Choi, Y. S.; Kim, S. H.; Moon, J. H.; Baek, S. Y

    2000-06-01

    Pneumatic transfer system(PTS) is one of the facilities to be used in irradiation of target materials for neutron activation analysis(NAA) in the research reactor. There are two systems the manual and the automatic system in PTS of HANARO research reactor. The pneumatic transfer system consists of many devices, sends and loads the capsules from NAA laboratory into three holes in the reflector tank of reactor and retrieves irradiated capsules after irradiation. This report describes the part's design, control system and the operation procedures. All the algorithm described in the text will be used for maintenance and upgrading.

  18. Assessment of soil and ground-water activation in the underground facility of the linear accelerator at RAON

    Lee, Sangjin; Nam, Shinwoo; Chung, Yonsei; Kim, Suna; Lee, Cheol Woo

    2015-10-01

    RAON is a heavy-ion accelerator complex that is being constructed in Daejeon, Korea. The superconducting linear accelerator of RAON will provide various heavy-ion beams with a maximum power of 400 kW. In order to determine the design requirements of the underground facility for the accelerator, we considered the radiation's influence on the soil and the ground-water under the condition of long-term operation of the accelerator. A source term for prompt neutrons generated by heavy-ion beams losses along the beam lines at a rate of 1 W/m was applied to obtain the activation level of nearby material outside the tunnel by using the simulation codes MCNPX and SP-FISPACT. This report presents the analysis and the result for the tunnel shielding condition obtained from the assessment.

  19. Radiohygienic aspects of the safety analysis of the Puespoekszilagy radioactive waste disposal and treatment facility, Hungary

    A temporary disposal was established for low level radioactive waste (LLW) at Solymar close to Budapest in 1960. Approx. 900 m3 LLW was disposed in concrete ring bells on the site until 1975. A new disposal (Radwaste Treatment and Disposal Facility, RWTDF) for low and intermediate radioactive waste (L/ILW) was put into operation at Puespoekszilagy, about 40 km to Budapest in 1976. The site was operated by the Metropolitan Institute of National Public Health and Medical Officer Service until 1997, when according to the new Hungarian Act on Atomic Energy the Public Agency for Radioactive Waste Management was established to perform the tasks connected to radwaste management and decommissioning of nuclear installations. The Solymar facility was dismantled and the radioactive waste transported to Puespoekszilagy. The RWTDF is situated on the ridge of a hill in a clay formation with conductivity from 10-8 to 10-6 cm.s-1; the groundwater depth is 17-20 m from the bottom of the disposal units. The waste is deposited in near surface disposal units (trenches, cells, and wells) with engineered barriers. Up to now about 4900 m3 of solid and solidified waste has been emplaced and 2 trenches of about 3000 m3 has been temporary sealed. More than 80% of the disposed waste is of low level. Approx. 700 TBq is the total activity of the radwaste including long-lived and alpha emitting radionuclides with the activity of the order of magnitude of 10 TBq. As the safety analysis was performed in a simple way in 1970's during the commissioning of the facility a comprehensive safety analysis was prescribed to get the license for the operation of the storage units extended at the end of 1980's. ETV-EROETERV Ltd. has won the tender for the safety analysis and the NRIRR was involved in the biosphere characterisation of the region and in the dose estimations for different accidental scenarios as well. The biosphere characterisation included the following categories: meteorology, geography, land

  20. Groundwater flow analysis using mixed hybrid finite element method for radioactive waste disposal facilities

    In safety assessments of radioactive waste disposal facilities, ground water flow analysis are used for calculating the radionuclide transport pathway and the infiltration flow rate of groundwater into the disposal facilities. For this type of calculations, the mixed hybrid finite element method has been used and discussed about the accuracy of ones in Europe. This paper puts great emphasis on the infiltration flow rate of groundwater into the disposal facilities, and describes the accuracy of results obtained from mixed hybrid finite element method by comparing of local water mass conservation and the reliability of the element breakdown numbers among the mixed hybrid finite element method, finite volume method and nondegenerated finite element method. (author)

  1. Modelling activities of experimental facilities related to advanced reactors. Considerations on 1D/3D issues

    The state of art of modelling activities related to integral experimental facilities of advanced passive reactors show to date important open items. The main advantage of using 1D plant codes is the capability of simulating the full interaction between components traditionally correctly modelled (condensers, heat exchangers, pipes and vessels) and other components for which codes are not 100% suitable (pools and containments). Polytechnical University of Catalonia (UPC) and Polytechnical University of Valencia (UPV) cooperated with other European research organizations in the 'Technology Enhancement for Passive Safety Systems' (TEPSS) project, within the European Fourth Framework Programme. It was a task of both Universities to supply analytical support of PANDA tests. The paper deals with the 1D/3D discussion in the framework of modelling activities related to integral passive facilities like PANDA. It starts choosing reference tests among those corresponding to our participation in TEPSS project. The discrepancies observed in a 1D simulation of the selected tests will be shown and analyzed. An evaluation of how the 3D version can lead to a better agreement with data will be included. Disadvantages of 3D codes will be shown too. Combining the use of different codes, and considering analyst criteria, will make possible to establish suitable recommendations from both engineering and scientific point of view. (author)

  2. Applied research and service activities at the University of Missouri Research Reactor Facility (MURR)

    The University Of Missouri operates MURR to provide an intense source of neutron and gamma radiation for research and applications by experimenters from its four campuses and by experimenters from other universities, government and industry. The 10 MW reactor, which has been operating an average of 155 hours per week for the past eight years, produces thermal neutron fluxes up to 6-7x1014 n/cm2-s in the central flux trap and beamport source fluxes of up to 1.2x1014 n/cm2-s. The mission of the reactor facility, to promote research, education and service, is the same as the overall mission of the university and therefore, applied research and service supported by industrial firms have been welcomed. The university recognized after a few years of reactor operation that in order to build utilization, it would be necessary to develop in-house research programs including people, equipment and activity so that potential users could more easily and quickly obtain the results needed. Nine research areas have been developed to create a broadly based program to support the level of activity needed to justify the cost of operating the facility. Applied research and service generate financial support for about one-half of the annual budget. The applied and service programs provide strong motivation for university/industry association in addition to the income generated. (author)

  3. DOE standard: Integration of environment, safety, and health into facility disposition activities. Volume 1: Technical standard

    NONE

    1998-05-01

    This Department of Energy (DOE) technical standard (referred to as the Standard) provides guidance for integrating and enhancing worker, public, and environmental protection during facility disposition activities. It provides environment, safety, and health (ES and H) guidance to supplement the project management requirements and associated guidelines contained within DOE O 430.1A, Life-Cycle Asset Management (LCAM), and amplified within the corresponding implementation guides. In addition, the Standard is designed to support an Integrated Safety Management System (ISMS), consistent with the guiding principles and core functions contained in DOE P 450.4, Safety Management System Policy, and discussed in DOE G 450.4-1, Integrated Safety Management System Guide. The ISMS guiding principles represent the fundamental policies that guide the safe accomplishment of work and include: (1) line management responsibility for safety; (2) clear roles and responsibilities; (3) competence commensurate with responsibilities; (4) balanced priorities; (5) identification of safety standards and requirements; (6) hazard controls tailored to work being performed; and (7) operations authorization. This Standard specifically addresses the implementation of the above ISMS principles four through seven, as applied to facility disposition activities.

  4. Preliminary Hazard Analysis for the Remote-Handled Low-Level Waste Disposal Facility

    Lisa Harvego; Mike Lehto

    2010-05-01

    The need for remote handled low level waste (LLW) disposal capability has been identified. A new onsite, remote-handled LLW disposal facility has been identified as the highest ranked alternative for providing continued, uninterrupted remote-handled LLW disposal capability for remote-handled LLW that is generated as part of the nuclear mission of the Idaho National Laboratory and from spent nuclear fuel processing activities at the Naval Reactors Facility. Historically, this type of waste has been disposed of at the Radioactive Waste Management Complex. Disposal of remote-handled LLW in concrete disposal vaults at the Radioactive Waste Management Complex will continue until the facility is full or until it must be closed in preparation for final remediation of the Subsurface Disposal Area (approximately at the end of Fiscal Year 2017). This document supports the conceptual design for the proposed remote-handled LLW disposal facility by providing an initial nuclear facility hazard categorization and by identifying potential hazards for processes associated with onsite handling and disposal of remote-handled LLW.

  5. An analysis of decommissioning costs for the AFRRI TRIGA reactor facility

    A decommissioning cost analysis for the AFRRI TRIGA Reactor Facility was made. AFRRI is not at this time suggesting that the AFRRI TRIGA Reactor Facility be decommissioned. This report was prepared to be in compliance with paragraph 50.33 of Title 10, Code of Federal Regulations which requires the assurance of availability of future decommissioning funding. The planned method of decommissioning is the immediate decontamination of the AFRRI TRIGA Reactor site to allow for restoration of the site to full public access - this is called DECON. The cost of DECON for the AFRRI TRIGA Reactor Facility in 1990 dollars is estimated to be $3,200,000. The anticipated ancillary costs of facility site demobilization and spent fuel shipment is an additional $600,000. Thus the total cost of terminating reactor operations at AFRRI will be about $3,800,000. The primary basis for this cost estimate is a study of the decommissioning costs of a similar reactor facility that was performed by Battelle Pacific Northwest Laboratory (PNL) as provided in USNRC publication NUREG/CR-1756. The data in this study were adapted to reflect the decommissioning requirements of the AFRRI TRIGA. (author)

  6. Neutron activation analysis of reference materials

    The importance is pointed out of neutron activation analysis in the preparation of reference materials, and studies are reported conducted recently by UJV. Instrumental neutron activation analysis has been used in testing homogeneity and in determining 28 elements in newly prepared reference standards of coal fly ash designated ENO, EOP and ECH. For accuracy testing, the same method was used in the analysis of NBS SRM-1633a Trace Elements in Coal Fly Ash and IAEA CRM Soil-5 and RM Soil-7. Radiochemical neutron activation analysis was used in determining Cd, Cu, Mn, Mo, and Zn in biological materials NBS SRM-1577 Bovine Liver, Bowen's Kale and in IAEA RM Milk Powder A-11 and Animal Muscle H-4. In all instances very good precision and accuracy of neutron activation analysis results were shown. (author)

  7. Do provisions to advance chemical facility safety also advance chemical facility security? An analysis of possible synergies

    Hedlund, Frank Huess

    The European Commission has launched a study on the applicability of existing chemical industry safety provisions to enhancing security of chemical facilities covering the situation in 18 EU Member States. This paper reports some preliminary analytical findings regarding the extent to which...... exist at the mitigation level. At the strategic policy level, synergies are obvious. The security of chemical facilities is important. First, facilities with large inventories of toxic materials could be attractive targets for terrorists. The concern is sabotage causing an intentional release that could...

  8. Using uncertainty analysis to guide and evaluate modifications of a boundary layer heat transfer test facility

    A case study is presented on the use of uncertainty analysis to guide the modification of an existing, mature experimental facility with well known characteristics and to insure that the results obtained with the modified facility will meet the new requirements. The existing facility's test surface was composed of 24 individually heated 0.1 m wide test plates. This facility was to be used to conduct a set of experiments to investigate the effect of a step change in surface roughness from a rough to a smooth surface on the heat transfer in the turbulent boundary layer. Of particular interest was the Stanton number distribution immediately downstream of the rough-to-smooth interface. The existing 0.1 m wide test plates could not provide the desired resolution; so, narrower plates were needed in this region. Unfortunately as the plates are made narrower, the uncertainty in Stanton number becomes larger. This paper presents and discusses the use of uncertainty analysis to guide this modification, to set up the experimental procedures and test limits, and to reduce the actual data. Sample Stanton number results are included

  9. Activation analysis in virus research

    Nucleic acids contain various amounts of trace elements. With divalent cations tobacco mosaic virus ribonucleic acid (TMV-RNA) undergoes changes in optical density at 258 mμ in both optical rotation and in sedimentation behaviour. These changes suggested the transformation from a random coil to a more orderly configuration. The connection between trace elements and the biological activity of TMV-RNA has been investigated. However, the relationship between metal ions and the virus infection process has not received much attention. This subject is discussed briefly in this paper. 8 refs

  10. Auditable Safety Analysis and Final Hazard Classification for the 105-N Reactor Zone and 109-N Steam Generator Zone Facility

    This document is a graded auditable safety analysis (ASA) and final hazard classification (FHC) for the Reactor/Steam Generator Zone Segment. The Reactor/Steam Generator Zone Segment, part of the N Reactor Complex, that is also known as the Reactor Building and Steam Generator Cells. The installation of the modifications described within to support surveillance and maintenance activities are to be completed by July 1, 1999. The surveillance and maintenance activities addressed within are assumed to continue for the next 15- 20 years, until the initiation of facility D ampersand D (i.e., Interim Safe Storage). The graded ASA in this document is in accordance with EDPI-4.30-01, Rev. 1, Safety Analysis Documentation, (BHI-DE-1) and is consistent with guidance provided by the U.S. Department of Energy. This ASA describes the hazards within the facility and evaluates the adequacy of the measures taken to reduce, control, or mitigate the identified hazards. This document also serves as the FHC for the Reactor/Steam Generator Zone Segment. This FHC is developed through the use of bounding accident analyses that envelope the potential exposures to personnel

  11. Development of environmental sample analysis technique in KAERI. Bulk analysis and establishment of clean laboratory facility (CLASS)

    The development of analytical methods for environmental samples in Korea Atomic Energy Research Institute (KAERI) is discussed. An analysis scheme for environmental samples has been established with an MCICP-MS based bulk analysis with adopting UTEVA resin for chemical separation and a particle analysis using FTTIMS and SIMS. A clean laboratory facility called CLASS (class 100∼ class 1000) was also established in order to prevent any cross contamination of the samples. The amount of U and Pu in the process blank sample prepared in the CLASS facility was estimated as 20 pg and less than 0.005 pg, respectively. The control chart of the analytical performance for the uranium standard sample of 100 ppt (NBL U030) indicated that the analytical performance of KAERI in CLASS is within 5 % of the certified values. (author)

  12. The association between high recreational physical activity and physical activity as a part of daily living in adolescents and availability of local indoor sports facilities and sports clubs

    Niclasen, B.; Petzold, M.; Schnohr, Christina Warrer

    2012-01-01

    Aim: The aim of this study was to examine how vigorous physical activity (recreational physical activity) (VPA) and moderate to vigorous physical activity as a part of daily life (MVPA) is associated with structural characteristics (availability of sports facilities and sports clubs with child...

  13. Neutron activation analysis of geochemical samples

    The present paper will describe the work done at the Technical Research Centre of Finland in developing methods for the large-scale activation analysis of samples for the geochemical prospecting of metals. The geochemical prospecting for uranium started in Finland in 1974 and consequently a manually operated device for the delayed neutron activation analysis of uranium was taken into use. During 1974 9000 samples were analyzed. The small capacity of the analyzer made it necessary to develop a completely automated analyzer which was taken into use in August 1975. Since then 20000-30000 samples have been analyzed annually the annual capacity being about 60000 samples when running seven hours per day. Multielemental instrumental neutron activation analysis is used for the analysis of more than 40 elements. Using instrumental epithermal neutron activation analysis 25-27 elements can be analyzed using one irradiation and 20 min measurement. During 1982 12000 samples were analyzed for mining companies and Geological Survey of Finland. The capacity is 600 samples per week. Besides these two analytical methods the analysis of lanthanoids is an important part of the work. 11 lanthanoids have been analyzed using instrumental neutron activation analysis. Radiochemical separation methods have been developed for several elements to improve the sensitivity of the analysis

  14. Summary of active test of uranium-plutonium co-denitration facility at Rokkasho reprocessing plant

    The aim of this report is to explain and discuss the active test results in the uranium-plutonium (U-Pu) co-denitration facility. We had previously performed the uranium test with depleted uranium from February of 2005 to January of 2006. Then, the active test has been in progress since March of 2006 toward the start of commercial operation. Plutonium nitrate (PuN) and uranium nitrate hexahydrate (UNH) are mixed at the ratio of approximately 1:1 from the non-proliferation viewpoint. The mixed solution is supplied into the denitration dish inside the denitration oven where the solution is denitrated by microwave heating and converted to MOX powder (PuO2-UO3). After denitration, the powder is converted to the product of MOX powder (PuO2-UO2) through some heating processes and stored in temporary canisters. The powder is transferred to the blender, and then filled into powder cans. 3 powder cans are packed into a canister and transferred to storage in the co-denitrated product powder storage building. Confirmation of the denitration ability of the mixed solution and characteristics of the product powder, (1) Stable and continuous operation in the target period, (2) Characteristics of the product powder, (3) Processing ability at each process, (4) Impurities in the product powder. The test results of the last step of the active test of the U-Pu co-denitration facility are presented; (1) Average throughput in 5 days at A and B lines was more than the target value. (2) Mean particle sizes and specific surface areas in MOX powder were within the standards. (3) Each process indicated good result. (4) Impurities in product powder were less than each limitation. (author)

  15. Automation of statistical analysis in the WIPP hazardous waste facility permit for analytical results from characterization

    One goal of characterizing, processing, and shipping waste to the Waste Isolation Pilot Plant (WIPP) is to make all activities as efficient as possible. Data management and repetitive calculations are a critical part of the process that can be automated, thereby increasing the accuracy and rate at which work is completed and reducing costs. This paper presents the tools developed to automate statistical analysis and other calculations required by the WIPP Hazardous Waste Facility Permit (HWFP). Statistical analyses are performed on the analytical results on gas samples from the headspace of waste containers and solid samples from the core of the waste container. The calculations include determining the number of samples, test for the shape of the distribution of the analytical results, mean, standard deviation, upper 90-percent confidence limit of the mean, and the minimum required Waste Acceptance Plan (WAP) sample size. The input data for these calculations are from the batch data reports for headspace gas analytical results and solids analysis, which must also be obtained and collated for proper use. The most challenging component of the statistical analysis, if performed manually, is the determination of the distribution shape; therefore, the distribution testing is typically performed using a certified software tool. All other calculations can be completed manually, with a spreadsheet, custom developed software, and/or certified software tool. Out of the options available, manually performing the calculations or using a spreadsheet are the least desirable. These methods rely heavily on the availability of an expert, such as a statistician, to perform the calculation. These methods are also more open to human error such as transcription or 'cut and paste' errors. A SAS program is in the process of being developed to perform the calculations. Due to the potential size of the data input files and the need to archive the data in an accessible format, the SAS

  16. Analysis Facility infrastructure (TIER3) for ATLAS High Energy physics experiment

    ATLAS project has been asked to define the scope and role of Tier-3 resources (facilities or centres) within the existing ATLAS computing model, activities and facilities. This document attempts to address these questions by describing Tier-3 resources generally, and their relationship to the ATLAS Software and Computing Project. Originally the tiered computing model came out of MONARC (see http://monarc.web.cern.ch/MONARC/) work and was predicated upon the network being a scarce resource. In this model the tiered hierarchy ranged from the Tier-0 (CERN) down to the desktop or workstation (Tier 3). The focus on defining the roles of each tiered component has evolved with the initial emphasis on the Tier-0 (CERN) and Tier-1 (National centres) definition and roles. The various LHC projects, including ATLAS, then evolved the tiered hierarchy to include Tier-2s (Regional centers) as part of their projects. Tier-3s, on the other hand, have (implicitly and sometime explicitly) been defined as whatever an institution could construct to support their Physics goals using institutional and otherwise leveraged resources and therefore have not been considered to be part of the official ATLAS Research Program computing resources nor under their control, meaning there is no formal MOU process to designate sites as Tier-3s and no formal control of the program over the Tier-3 resources. Tier-3s are the responsibility of individual institutions to define, fund, deploy and support. However, having noted this, we must also recognize that Tier-3s must exist and will have implications for how our computing model should support ATLAS physicists. Tier-3 users will want to access data and simulations and will want to enable their Tier-3 resources to support their analysis and simulation work. Tiers 3s are an important resource for physicists to analyze LHC (Large Hadron Collider) data. This document will define how Tier-3s should best interact with the ATLAS computing model, detail the

  17. A software architectural framework specification for neutron activation analysis

    Neutron Activation Analysis (NAA) is a sensitive multi-element nuclear analytical technique that has been routinely applied by research reactor (RR) facilities to environmental, nutritional, health related, geological and geochemical studies. As RR facilities face calls to increase their research output and impact, with existing or reducing budgets, automation of NAA offers a possible solution. However, automation has many challenges, not the least of which is a lack of system architecture standards to establish acceptable mechanisms for the various hardware/software and software/software interactions among data acquisition systems, specialised hardware such as sample changers, sample loaders, and data processing modules. This lack of standardization often results in automation hardware and software being incompatible with existing system components, in a facility looking to automate its NAA operations. This limits the availability of automation to a few RR facilities with adequate budgets or in-house engineering resources. What is needed is a modern open system architecture for NAA, that provides the required set of functionalities. This paper describes such an 'architectural framework' (OpenNAA), and portions of a reference implementation. As an example of the benefits, calculations indicate that applying this architecture to the compilation and QA steps associated with the analysis of 35 elements in 140 samples, with 14 SRM's, can reduce the time required by over 80 %. The adoption of open standards in the nuclear industry has been very successful over the years in promoting interchangeability and maximising the lifetime and output of nuclear measurement systems. OpenNAA will provide similar benefits within the NAA application space, safeguarding user investments in their current system, while providing a solid path for development into the future. (author)

  18. Engineering Evaluation/Cost Analysis for the 100-N Area Ancillary Facilities and Integration Plan

    This document presents the results of an engineering evaluation/cost analysis (EE/CA) that was conducted to evaluate alternatives for addressing final disposition of contaminated buildings and structures in the 100-N Area of the Hanford Site. The Hanford Site is located in southeastern Washington State and is owned by the U.S. Government and operated by the U.S. Department of Energy, Richland Operations Office (RL). In November 1989, the 100 Area of the Hanford Site (as well as the 200, 300, and 1100 Areas) was placed on the U.S. Environmental Protection Agency's National Priorities List (NPL) under the Comprehensive Environmental Response, Compensation, and Liability Act of 1980. The 100 Area NPL includes the 100-N Area, which is in various stages of the remediation process. It has been determined by RL that hazardous substances in the 100-N Area ancillary facilities may present a potential threat to human health or the environment, and that a non-time critical removal action at these facilities is warranted. To help determine the most appropriate action, RL, in cooperation with the Washington State Department of Ecology (Ecology) and the EPA, has prepared this EE/CA. The scope of the evaluation includes the inactive contaminated ancillary facilities in the 100-N Area, the facilities residing in the buffer zone, and the Hanford Generating Plant (HGP) and the solid waste management units (SWMUs) inside HGP support facilities. The 105-N Reactor and 109-N Heat Exchange facilities are excluded from this EE/CA evaluation

  19. Pursuit of improvement in uranium bulk analysis at the clear facility for safeguards environmental samples

    with flexible tube (FIG. 1). By scanning the surface of Texwipe-304 with the top of the tip, the particles are sucked up through the tip and collected on to the filter. Figure 2 shows micrographs of Texwipe-304 surface: (a) new one, (b) particle-swiped one, and (c) particle-recovered one by the Vacuum Suction Method. As seen in the figure, most of the particles were removed from Texwipe-304. Preliminary examination suggested that the particle recovery yield is in acceptable level and that the process blank is low. The method will provide a useful means to improve the reliability in bulk analysis of ultra-trace amount of uranium, even in the case that Texwipe-304 is not replaceable. At the presentation, further results and other activities related to minimize the blank values at the CLEAR facility will be reported. (author)

  20. IFMIF, International Fusion Materials Irradiation Facility conceptual design activity cost report

    This report documents the cost estimate for the International Fusion Materials Irradiation Facility (IFMIF) at the completion of the Conceptual Design Activity (CDA). The estimate corresponds to the design documented in the Final IFMIF CDA Report. In order to effectively involve all the collaborating parties in the development of the estimate, a preparatory meeting was held at Oak Ridge National Laboratory in March 1996 to jointly establish guidelines to insure that the estimate was uniformly prepared while still permitting each country to use customary costing techniques. These guidelines are described in Section 4. A preliminary cost estimate was issued in July 1996 based on the results of the Second Design Integration Meeting, May 20--27, 1996 at JAERI, Tokai, Japan. This document served as the basis for the final costing and review efforts culminating in a final review during the Third IFMIF Design Integration Meeting, October 14--25, 1996, ENEA, Frascati, Italy. The present estimate is a baseline cost estimate which does not apply to a specific site. A revised cost estimate will be prepared following the assignment of both the site and all the facility responsibilities

  1. IFMIF, International Fusion Materials Irradiation Facility conceptual design activity cost report

    Rennich, M.J. [comp.

    1996-12-01

    This report documents the cost estimate for the International Fusion Materials Irradiation Facility (IFMIF) at the completion of the Conceptual Design Activity (CDA). The estimate corresponds to the design documented in the Final IFMIF CDA Report. In order to effectively involve all the collaborating parties in the development of the estimate, a preparatory meeting was held at Oak Ridge National Laboratory in March 1996 to jointly establish guidelines to insure that the estimate was uniformly prepared while still permitting each country to use customary costing techniques. These guidelines are described in Section 4. A preliminary cost estimate was issued in July 1996 based on the results of the Second Design Integration Meeting, May 20--27, 1996 at JAERI, Tokai, Japan. This document served as the basis for the final costing and review efforts culminating in a final review during the Third IFMIF Design Integration Meeting, October 14--25, 1996, ENEA, Frascati, Italy. The present estimate is a baseline cost estimate which does not apply to a specific site. A revised cost estimate will be prepared following the assignment of both the site and all the facility responsibilities.

  2. Review of twenty-five years' activities of Co-60 irradiation facility in Tokai Research Establishment

    Since the Co-60 irradiation facility was established in Tokai Research Establishment in 1958, it has been used for many irradiation services in a wide field in responce to the various needs both in JAERI and from the outside of JAERI. During this period the reconstructions and repairs have been performed in many respects of the equipments and the building, and the irradiation sources were newly added in several times. The present report is a history of the activity of the Co-60 irradiation facility for 25 years, describing 1) the maintenance and the reformations of the equipments associated with the cave such as a reinforcement of the cave shielding, the renewal of the shielding windows and the reformation of the source storage, 2) the specifications and the assembly workes of the sources, 3) the summary of the main apparatuses used in the experiments, 4) the extention and the reconstructions of the building, 5) the historical changes of the themes of experiments performed by the users. (author)

  3. Through-Life Management of Built Facilities: Towards a Framework for Analysis

    Koskela, Lauri; Siriwardena, Mohan; Rooke, John

    2008-01-01

    Although built facilities are required to cater to changing requirements over time, effective through life management is absent as an in-process activity from most large scale procurements. Through a review of key literature, several approaches which address aspects relevant to through-life management are discussed, and an attempt is made to create a unified view framework of understanding of what constitutes through-life management. Furthermore, an initial diagnostic style checklist is provi...

  4. The Influence of Urban Land-Use and Public Transport Facilities on Active Commuting in Wellington, New Zealand: Active Transport Forecasting Using the WILUTE Model

    Joreintje Dingena Mackenbach; Edward Randal; Pengjun Zhao; Philippa Howden-Chapman

    2016-01-01

    Physical activity has numerous physical and mental health benefits, and active commuting (walking or cycling to work) can help meet physical activity recommendations. This study investigated socioeconomic differences in active commuting, and assessed the impact of urban land-use and public transport policies on active commuting in the Wellington region in New Zealand. We combined data from the New Zealand Household Travel Survey and GIS data on land-use and public transport facilities with th...

  5. Tritium Facility effluent measurements: statistical analysis for 52 periods in 1981

    A statistical comparison has been made of Tritium Facility effluents for 52 periods in 1981, as measured by continuous sampling from two ventilation exhaust stacks. Pulse-counter data, recorded on 15-minute intervals, have been summed over approximately weekly periods and compared with adsorption apparatus data. A by-product of the analysis is a calibration of pulse counters in terms of adsorption data which are believed to be more reliable

  6. Proposed outline of safety analysis report for facilities for geologic isolation of radioactive wastes

    This report is concerned with formulating the appropriate and necessary contents for a Safety Analysis Report (SAR) for safe, long-term radioactive waste isolation in federal repositories. The material is presented as a guide rather than as an outline of a SAR. Site characteristics, design criteria, facility design, operational systems and components, radiation protection and operational safety, long-term waste isolation, conduct of operations, technical specifications, and quality assurance are covered. Recommendations are given for further research studies

  7. 200 Area Treated Effluent Disposal Facility (TEDF) Effluent Sampling and Analysis Plan

    BROWN, M.J.

    2000-05-18

    This Sampling and Analysis Plan (SAP) has been developed to comply with effluent monitoring requirements at the 200 Area Treated Effluent Disposal Facility (TEDF), as stated in Washington State Waste Discharge Permit No. ST 4502 (Ecology 2000). This permit, issued by the Washington State Department of Ecology (Ecology) under the authority of Chapter 90.48 Revised Code of Washington (RCW) and Washington Administrative Code (WAC) Chapter 173-216, is an April 2000 renewal of the original permit issued on April 1995.

  8. Fermilab Central Computing Facility: Energy conservation report and mechanical systems design optimization and cost analysis study

    Krstulovich, S.F.

    1986-11-12

    This report is developed as part of the Fermilab Central Computing Facility Project Title II Design Documentation Update under the provisions of DOE Document 6430.1, Chapter XIII-21, Section 14, paragraph a. As such, it concentrates primarily on HVAC mechanical systems design optimization and cost analysis and should be considered as a supplement to the Title I Design Report date March 1986 wherein energy related issues are discussed pertaining to building envelope and orientation as well as electrical systems design.

  9. Fracture mechanics analysis of a high-pressure hydrogen facility compressor

    Vroman, G. A.

    1974-01-01

    The investigation and analysis of a high-pressure hydrogen facility compressor is chronicled, and a life prediction based on fracture mechanics is presented. Crack growth rates in SA 105 Gr II steel are developed for the condition of sustained loading, using a hypothesis of hydrogen embrittlement associated with plastic zone reverse yielding. The resultant formula is compared with test data obtained from laboratory specimens.

  10. Descriptions of reference LWR facilities for analysis of nuclear fuel cycles. Appendixes

    The appendixes present the calculations that were used to derive the release factors discussed for each fuel cycle facility in Volume I. Appendix A presents release factor calculations for a surface mine, underground mine, milling facility, conversion facility, diffusion enrichment facility, fuel fabrication facility, PWR, BWR, and reprocessing facility. Appendix B contains additional release factors calculated for a BWR, PWR, and a reprocessing facility. Appendix C presents release factors for a UO2 fuel fabrication facility

  11. Evolution of Interactive Analysis Facilities: from NAF to NAF 2.0

    Haupt, Andreas; Kemp, Yves; Nowak, Friederike

    2014-06-01

    In 2007, the National Analysis Facility (NAF) was set up within the framework of the Helmholtz Alliance "Physics at the Terascale", and is located at DESY. Its purpose was and is the provision of an analysis infrastructure for up-to-date research in Germany, complementing the Grid by offering a interactive access to the data. It has been well received within the physics community, and has proven to be a highly successful concept. We will review experiences with the original NAF, and discuss both the resulting motivation and constraints for the transition to an evolved model. We call this new facility the NAF 2.0. We will present a new setup including its building blocks and user handling, and give an overview of the current status. The integration of new communities has broadened the range of the analysis facility beyond its primary focus on LHC and ILC experiments. To finish, an outlook on further developments like the adoption of new technologies will be given.

  12. Development and Commissioning of an External Beam Facility in the Union College Ion Beam Analysis Laboratory

    Yoskowitz, Joshua; Clark, Morgan; Labrake, Scott; Vineyard, Michael

    2015-10-01

    We have developed an external beam facility for the 1.1-MV tandem Pelletron accelerator in the Union College Ion Beam Analysis Laboratory. The beam is extracted from an aluminum pipe through a 1 / 4 ' ' diameter window with a 7.5- μm thick Kapton foil. This external beam facility allows us to perform ion beam analysis on samples that cannot be put under vacuum, including wet samples and samples too large to fit into the scattering chamber. We have commissioned the new facility by performing proton induced X-ray emission (PIXE) analysis of several samples of environmental interest. These include samples of artificial turf, running tracks, and a human tooth with an amalgam filling. A 1.7-MeV external proton beam was incident on the samples positioned 2 cm from the window. The resulting X-rays were measured using a silicon drift detector and were analyzed using GUPIX software to determine the concentrations of elements in the samples. The results on the human tooth indicate that while significant concentrations of Hg, Ag, and Sn are present in the amalgam filling, only trace amounts of Hg appear to have leached into the tooth. The artificial turf and running tracks show rather large concentrations of a broad range of elements and trace amounts of Pb in the turf infill.

  13. Applications of neutron activation analysis in industry

    Neutron activation analysis technique is discussed in brief. This technique is used for quality control of raw materials, process materials and finished products, as well as activities in research and development for the improvement of the products and new products. The uses of this technique in several experienced industries are mentioned (author)

  14. Lightweight scheduling of elastic analysis containers in a competitive cloud environment: a Docked Analysis Facility for ALICE

    Berzano, D.; Blomer, J.; Buncic, P.; Charalampidis, I.; Ganis, G.; Meusel, R.

    2015-12-01

    During the last years, several Grid computing centres chose virtualization as a better way to manage diverse use cases with self-consistent environments on the same bare infrastructure. The maturity of control interfaces (such as OpenNebula and OpenStack) opened the possibility to easily change the amount of resources assigned to each use case by simply turning on and off virtual machines. Some of those private clouds use, in production, copies of the Virtual Analysis Facility, a fully virtualized and self-contained batch analysis cluster capable of expanding and shrinking automatically upon need: however, resources starvation occurs frequently as expansion has to compete with other virtual machines running long-living batch jobs. Such batch nodes cannot relinquish their resources in a timely fashion: the more jobs they run, the longer it takes to drain them and shut off, and making one-job virtual machines introduces a non-negligible virtualization overhead. By improving several components of the Virtual Analysis Facility we have realized an experimental “Docked” Analysis Facility for ALICE, which leverages containers instead of virtual machines for providing performance and security isolation. We will present the techniques we have used to address practical problems, such as software provisioning through CVMFS, as well as our considerations on the maturity of containers for High Performance Computing. As the abstraction layer is thinner, our Docked Analysis Facilities may feature a more fine-grained sizing, down to single-job node containers: we will show how this approach will positively impact automatic cluster resizing by deploying lightweight pilot containers instead of replacing central queue polls.

  15. New studies in forensic neutron activation analysis

    Three recently completed studies in forensic neutron activation analysis are reported: a study of 0.22-caliber rimfire cartridge primers, a large-scale study of shotgun pellets, and a new 5-element procedure for the analysis of bullet-lead and shotgun-pellet samples. (author) 12 refs

  16. New studies in forensic neutron activation analysis

    Earlier studies in forensic neutron activation analysis are being extended in This Laboratory. Three of these new studies are reported here: 1) a study of 0.22-caliber rimfire cartridge primers, 2) a large-scale study of shotgun pellets, and 3) a new 5-element procedure for the analysis of bullet-lead and shotgun-pellet samples. (author)

  17. A program for activation analysis data processing

    An ALGOL program for activation analysis data handling is presented. The program may be used either for single channel spectrometry data or for multichannel spectrometry. The calculation of instrumental error and of analysis standard deviation is carried out. The outliers are tested, and the regression line diagram with the related observations are plotted by the program. (author)

  18. The dry storage cask in interim storage facility and safeguards activity

    The Japan Atomic Power Company (JAPC) is preparing for interim storage of spent fuel at Recyclable-Fuel Storage Center (RFSC) in Aomori Prefecture. Metallic dry casks are employed to contain the spent fuel and to serve for about 50 years in RFSC. Metallic dry casks have already been used for spent fuel dry storage at Tokai No.2 power station. But, RFSC is not exactly the same as the dry storage facility in Tokai No.2 power station, for example, casks are transported out side of the reactor site and RFSC has no fuel handling system. Therefore, additional implementation of safeguards is necessary. This report introduces the design and handling of metallic dry casks for RFSC and the currently developing status of the safeguards activity such as containment and surveillance for the fuel loading at the power station, the cask receipt and storage at RFSC, etc. (author)

  19. Search for reaction-in-flight neutrons using thulium activation at the National Ignition Facility

    Grim, Gary; Rundberg, Robert; Tonchev, Anton; Fowler, Malcolm; Wilhelmy, Jerry; Archuleta, Tom; Bionta, Richard; Boswell, Mitzi; Gostic, Julie; Griego, Jeff; Knittel, Kenn; Klein, Andi; Moody, Ken; Shaughnessy, Dawn; Wilde, Carl; Yeamans, Charles

    2013-10-01

    We report on measurements of reaction-in-flight (RIF) neutrons at the National Ignition Facility. RIF neutrons are produced in cryogenically layered implision by up-scattered deuterium, or tritium ions that undergo subsequent fusion reactions. The rate of RIF neutron production is proportional to the fuel areal density (| | R) and ion-stopping length in the dense fuel assembly. Thus, RIF neutrons provide information on charge particle stopping in a strongly coupled plasma, where perturbative modeling breaks down. To measure RIF neutrons, a set of thulium activation foils was placed 50 cm from layered cryogenic implosions at the NIF. The reaction 169Tm(n,3n)167Tm has a neutron kinetic energy threshold of 14.96 MeV. We will present results from initial experiments performed during the spring of 2013. Prepared by LANL under Contract DE-AC-52-06-NA25396, TSPA, LA-UR-13-22085.

  20. Facile synthesis of PtCu nanowires with enhanced electrocatalytic activity

    Wei Hong[1,2; Jin Wang[1,3; Erkang Wang[1,2

    2015-01-01

    Using Te nanowires as a sacrificial template, We developed a facile wet-chemical method for the synthesis of bimetallic PtCu nanowires. The as-prepared PtCu nanowires possess a porous structure and high aspect ratio. Transmission electron microscopy, X-ray diffraction, energy dispersive spectroscopy, energy dispersive X-ray spectrum elemental mapping, inductively coupled plasma- mass spectroscopy, and X-ray photoelectron spectroscopy (XPS) measurement techniques are used to analyze the structure and composition of the as-prepared nanowires. The XPS results verify that the incorporation of Cu led to the modified electronic state of Pt. Electrocatalytic results prove that the as-prepared nanowires present superior activity for methanol and ethanol electrooxidation in an alkaline solution.

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

    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-time critical removal

  2. High level process shielded line (CBP) and high level analysis shielded line (CBA): two of the newest facilities of ATALANTE facility

    The two newest facilities in the ATALANTE complex, a high-level shielded process line (CBP) and high-level shielded analysis line (CBA), are described and their work programs detailed, notably the dissolution in CBP of 15 kg of spent fuel to demonstrate the technological feasibility of partitioning the minor actinides. The analytical support role of CBA is also discussed. (authors)

  3. Evaluation of business activity using financial analysis

    Kučerová, Martina

    2012-01-01

    Czech University of Life Sciences Prague Faculty of Economics and Management Department of Trade and Accounting Abstract of Diploma Thesis Evaluation of business activity using financial analysis Martina Kučerová © 2012 CULS in Prague Summary: The aim of this thesis is to apply methods of financial analysis on chosen business for its first three accounting periods 2009, 2010 and 2011. The thesis is further analysis of my bachelor thesis on to...

  4. Analysis of flow fields, temperatures and ruthenium transport in the test facility

    Ruthenium transport experiments were conducted at VTT during years 2002- 2006. Experiments gave information about ruthenium behaviour in air ingress accident conditions. This study complements those experiments with an analysis of the flows and thermal fields in the test system. Temperature profiles were measured at the walls of the experimental facility. Computational fluid dynamics (CFD) simulations used the measured profiles and provided predictions of flows and temperatures inside the furnace. Ruthenium transport was also modelled with CFD. Thermal characterisation of the reactor demonstrated that buoyancy has a significant role during the cooling after the furnace. A hypothesis of the dominant role of RuO2 and RuO3 condensation on reactor walls gave simulation results that are in accordance with radiotracer measurements of deposition in experiments conducted with furnace at 1500K. Actually, RuO3 does not condensate, but it thermal decomposes to RuO2. This does not seem to have effect on result. Particle formation around the furnace exit could be detected from the comparison of modelling results with the measured profiles. In several other experiments ruthenium behaviour is dominated by other issues. These are related to the complex ruthenium chemistry that includes various surface reactions. Thermal equilibrium indicates significant gaseous RuO4 concentration around 1300 K. It seems that seed particles decreased the catalytic decomposition activity of RuO4 to RuO2 around this temperature pushing the gas concentration towards the equilibrium, and further give rise to gaseous RuO4 transport to low temperatures. At higher temperature increasing mass flow rate of RuO2 particles is likely to catalyse (decomposition) reaction of RuO4 to RuO2. (au)

  5. Risk analysis of the aqueous fast reactor fuel cycle facility in the conceptual design stage

    This paper describes the radioactive release risk of the advanced aqueous reprocessing and fabrication facility for the fast reactor fuel cycle. Because this advanced facility is still in the conceptual design stage, the risk analysis aims at grasping the entire risk comprehensively as simple as possible. As a potential hazard, it was shown that the main process in the reprocessing and fuel fabrication facilities involved only an order of 10-3 of radioactivity in the single reactor core of large scale. Abnormal phenomena related to radioactive solution that can cause radioactive release from the facility to the environmental atmosphere in a large quantity were identified as follows: in-vessel boiling caused by loss of cooling system, a leak and fire of inflammable organic solvent in a cell, in-vessel boiling due to criticality accident, an explosion. Simplified estimation about the quantitative risk of radioactive release showed that in-vessel boiling due to loss of cooling system had the largest contribution to the non-volatile radioactive substance release in a large quantity and that criticality accidents initiated from incomplete extraction stripping of Pu nuclides were dominant in the release risk of radioactive iodine and noble gas with a short-half-life. (author)

  6. Chemical Hygiene Plan for Onsite Measurement and Sample Shipping Facility Activities

    This chemical hygiene plan presents the requirements established to ensure the protection of employee health while performing work in mobile laboratories, the sample shipping facility, and at the onsite radiological counting facility. This document presents the measures to be taken to promote safe work practices and to minimize worker exposure to hazardous chemicals. Specific hazardous chemicals present in the mobile laboratories, the sample shipping facility, and in the radiological counting facility are presented in Appendices A through G

  7. Do provisions to advance chemical facility safety also advance chemical facility security? An analysis of possible synergies

    Hedlund, Frank Huess

    2012-01-01

    The European Commission has launched a study on the applicability of existing chemical industry safety provisions to enhancing security of chemical facilities covering the situation in 18 EU Member States. This paper reports some preliminary analytical findings regarding the extent to which...... Infrastructures (ECI Directive) addresses facility security but does not cover the chemical sector. Chemical facility safety at EU level is addressed by way of the Seveso-II Directive. Preliminary estimates by the chemical industry suggest that perhaps 80% of the existing safety measures under Seveso-II would...... existing provisions that have been put into existence to advance safety objectives due to synergy effects could be expected advance security objectives as well. The paper provides a conceptual definition of safety and security and presents a framework of their essential components. Key differences are...

  8. Safeguarding uranium enrichment facilities. Review and analysis of the status of safeguards technology for uranium enrichment facilities

    The objective of this paper is to examine critically the diversion potential at uranium enrichment facilities and to outline a basic safeguards strategy which counters all identified hazards as completely as possible yet with a minimum of non-essential redundancy. Where existing technology does not appear to be adequate for effective safeguards, the limitations are examined, and suggestions for further R and D effort are made. Parts of this report are generally applicable to all currently known enrichment processes, while other parts are specifically directed toward facilities based on the gas centrifuge process. It is hoped that additional sections discussing a safeguards strategy for gas diffusion facilities can be added later. It should be emphasized that this is a technical report, and does not reflect any legal positions. The safeguards strategy and subsequent inspection procedures are intended as guidelines, not as negotiating positions

  9. Activation analysis in science and technics

    Physical bases of the method of neutron activation analysis are described. Reactions of element nuclei with neutrons, gamma quanta and charged particles, ways of radioactivity measurement, gamma spectrometry in particular, have been described. The method errors, as well as element determination sensitivity in samples of different composition are considered, perspectives of activation analysis development being reflected. The practical use of the method in the studies on solid-state physics, during the analysis of semiconductors and pure materials, in criminalistics, agriculture, the environmnental control in geology and biology, is shown

  10. Defense In-Depth Accident Analysis Evaluation of Tritium Facility Bldgs. 232-H, 233-H, and 234-H

    'The primary purpose of this report is to document a Defense-in-Depth (DID) accident analysis evaluation for Department of Energy (DOE) Savannah River Site (SRS) Tritium Facility Buildings 232-H, 233-H, and 234-H. The purpose of a DID evaluation is to provide a more realistic view of facility radiological risks to the offsite public than the bounding deterministic analysis documented in the Safety Analysis Report, which credits only Safety Class items in the offsite dose evaluation.'

  11. Evaluation of methods for seismic analysis of nuclear fuel reprocessing and fabrication facilities

    Methods of seismic analysis for critical structures and equipment in nuclear fuel reprocessing plants (NFRPs) and mixed oxide fuel fabrication plants (MOFFPs) are evaluated. The purpose of this series of reports is to provide the NRC with a technical basis for assessing seismic analysis methods and for writing regulatory guides in which methods ensuring the safe design of nuclear fuel cycle facilities are recommended. The present report evaluates methods of analyzing buried pipes and wells, sloshing effects in large pools, earth dams, multiply supported equipment, pile foundations, and soil-structure interactions

  12. Thermohydraulic analysis of the IAEA standard problem test on the PMK-NHV facility

    International Atomic Energy Agency (IAEA) has supported a standard test problem simulating small break loss of coolant accident on the test facility PMH-NHV in Budapest. The present pretest analysis of that transient was done using the computer code RELAP4/MOD6. The results were compared to the measurements data and to data of 19 other laboratories around the world that have performed the same analysis. The correspondence of the results to the measured data is reasonable. There are bigger discrepancies, which in turn influence other variables. (author)

  13. Safety Analysis (SA) of the decontamination facility, Building 419, at the Lawrence Livermore National Laboratory

    This safety analysis was performed for the Manager, Plant Services at LLNL and fulfills the requirements of DOE Order 5481.1. The analysis was based on field inspections, document review, computer calculations, and extensive input from Waste Management personnel. It was concluded that the maximum quantities of radioactive materials that safety procedures allow to be handled in this building do not pose undue risks on- or off-site even in postulated severe accidents. Risk from the various hazards at this facility vary from low to moderate as specified in DOE Order 5481.1. Recommendations are made for improvements that will reduce risks even further

  14. Software systems for processing and analysis at the NOVA high-energy laser facility

    A typical laser interaction experiment at the NOVA high-energy laser facility produces in excess of 20 Mbytes of digitized data. Extensive processing and analysis of this raw data from a wide variety of instruments is necessary to produce results that can be readily used to interpret the experiment. Using VAX-based computer hardware, software systems have been set up to convert the digitized instrument output to physics quantities describing the experiment. A relational data-base management system is used to coordinate all levels of processing and analysis. Software development emphasizes structured design, flexibility, automation, and ease of use

  15. Spent nuclear fuel project, Cold Vacuum Drying Facility human factors engineering (HFE) analysis: Results and findings

    This report presents the background, methodology, and findings of a human factors engineering (HFE) analysis performed in May, 1998, of the Spent Nuclear Fuels (SNF) Project Cold Vacuum Drying Facility (CVDF), to support its Preliminary Safety Analysis Report (PSAR), in responding to the requirements of Department of Energy (DOE) Order 5480.23 (DOE 1992a) and drafted to DOE-STD-3009-94 format. This HFE analysis focused on general environment, physical and computer workstations, and handling devices involved in or directly supporting the technical operations of the facility. This report makes no attempt to interpret or evaluate the safety significance of the HFE analysis findings. The HFE findings presented in this report, along with the results of the CVDF PSAR Chapter 3, Hazards and Accident Analyses, provide the technical basis for preparing the CVDF PSAR Chapter 13, Human Factors Engineering, including interpretation and disposition of findings. The findings presented in this report allow the PSAR Chapter 13 to fully respond to HFE requirements established in DOE Order 5480.23. DOE 5480.23, Nuclear Safety Analysis Reports, Section 8b(3)(n) and Attachment 1, Section-M, require that HFE be analyzed in the PSAR for the adequacy of the current design and planned construction for internal and external communications, operational aids, instrumentation and controls, environmental factors such as heat, light, and noise and that an assessment of human performance under abnormal and emergency conditions be performed (DOE 1992a)

  16. Overview of NORM and activities by a NORM licensed permanent decontamination and waste processing facility

    Mirro, G.A. [Growth Resources, Inc., Lafayette, LA (United States)

    1997-02-01

    This paper presents an overview of issues related to handling NORM materials, and provides a description of a facility designed for the processing of NORM contaminated equipment. With regard to handling NORM materials the author discusses sources of NORM, problems, regulations and disposal options, potential hazards, safety equipment, and issues related to personnel protection. For the facility, the author discusses: description of the permanent facility; the operations of the facility; the license it has for handling specific radioactive material; operating and safety procedures; decontamination facilities on site; NORM waste processing capabilities; and offsite NORM services which are available.

  17. Automated absolute activation analysis with californium-252 sources

    MacMurdo, K.W.; Bowman, W.W.

    1978-09-01

    A 100-mg /sup 252/Cf neutron activation analysis facility is used routinely at the Savannah River Laboratory for multielement analysis of many solid and liquid samples. An absolute analysis technique converts counting data directly to elemental concentration without the use of classical comparative standards and flux monitors. With the totally automated pneumatic sample transfer system, cyclic irradiation-decay-count regimes can be pre-selected for up to 40 samples, and samples can be analyzed with the facility unattended. An automatic data control system starts and stops a high-resolution gamma-ray spectrometer and/or a delayed-neutron detector; the system also stores data and controls output modes. Gamma ray data are reduced by three main programs in the IBM 360/195 computer: the 4096-channel spectrum and pertinent experimental timing, counting, and sample data are stored on magnetic tape; the spectrum is then reduced to a list of significant photopeak energies, integrated areas, and their associated statistical errors; and the third program assigns gamma ray photopeaks to the appropriate neutron activation product(s) by comparing photopeak energies to tabulated gamma ray energies. Photopeak areas are then converted to elemental concentration by using experimental timing and sample data, calculated elemental neutron capture rates, absolute detector efficiencies, and absolute spectroscopic decay data. Calculational procedures have been developed so that fissile material can be analyzed by cyclic neutron activation and delayed-neutron counting procedures. These calculations are based on a 6 half-life group model of delayed neutron emission; calculations include corrections for delayed neutron interference from /sup 17/O. Detection sensitivities of < or = 400 ppB for natural uranium and 8 ppB (< or = 0.5 (nCi/g)) for /sup 239/Pu were demonstrated with 15-g samples at a throughput of up to 140 per day. Over 40 elements can be detected at the sub-ppM level.

  18. Automated absolute activation analysis with californium-252 sources

    A 100-mg 252Cf neutron activation analysis facility is used routinely at the Savannah River Laboratory for multielement analysis of many solid and liquid samples. An absolute analysis technique converts counting data directly to elemental concentration without the use of classical comparative standards and flux monitors. With the totally automated pneumatic sample transfer system, cyclic irradiation-decay-count regimes can be pre-selected for up to 40 samples, and samples can be analyzed with the facility unattended. An automatic data control system starts and stops a high-resolution gamma-ray spectrometer and/or a delayed-neutron detector; the system also stores data and controls output modes. Gamma ray data are reduced by three main programs in the IBM 360/195 computer: the 4096-channel spectrum and pertinent experimental timing, counting, and sample data are stored on magnetic tape; the spectrum is then reduced to a list of significant photopeak energies, integrated areas, and their associated statistical errors; and the third program assigns gamma ray photopeaks to the appropriate neutron activation product(s) by comparing photopeak energies to tabulated gamma ray energies. Photopeak areas are then converted to elemental concentration by using experimental timing and sample data, calculated elemental neutron capture rates, absolute detector efficiencies, and absolute spectroscopic decay data. Calculational procedures have been developed so that fissile material can be analyzed by cyclic neutron activation and delayed-neutron counting procedures. These calculations are based on a 6 half-life group model of delayed neutron emission; calculations include corrections for delayed neutron interference from 17O. Detection sensitivities of 239Pu were demonstrated with 15-g samples at a throughput of up to 140 per day. Over 40 elements can be detected at the sub-ppM level

  19. WHC-SD-W252-FHA-001, Rev. 0: Preliminary fire hazard analysis for Phase II Liquid Effluent Treatment and Disposal Facility, Project W-252

    A Fire Hazards Analysis was performed to assess the risk from fire and other related perils and the capability of the facility to withstand these hazards. This analysis will be used to support design of the facility

  20. Pretest analysis of containment studies facility model for simulated loss of coolant accident conditions

    An experimental facility called Containment Studies Facility (CSF) has been constructed at Bhabha Atomic Research Centre (BARC), Trombay for the purpose of research and development in the area of nuclear reactor containment thermal hydraulics. The facility consists of reinforced concrete containment structural model and a Primary Heat Transport Model (PHTM) vessel. The containment model is approximately 1:250 volumetrically scaled down model of a 220 MWe Indian Pressurized Heavy Water Reactor (IPHWR) containment system and the PHTM represents the primary heat transport system of the prototype reactor. The PHTM with a pressure vessel and associated pump and piping system is designed for simulating the Loss of Coolant Accident (LOCA) or Main Steam Line Break (MSLB) conditions within the containment model. As part of CSF project thermal hydraulic analysis, a pretest analysis was carried out for simulated LOCA conditions. Blow down mass and energy discharge data were obtained using Relap/MOD3.2 code for different blow down conditions and were used as inputs to CONTRAN code for simulating LOCA or main steam line break (MSLB) conditions in the containment model. Pressure and temperature transients in the CSF model for different blow down conditions and a number of parametric studies were conducted to assess the influence of a large number of thermodynamic and geometrical parameters which are known to affect the transients and alter the peak pressure and temperature values. (author)