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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

    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.

  6. 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

  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

    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.

  17. 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

  18. 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.)

  19. 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...

  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. 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

  3. 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.

  4. 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

  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. 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)

  17. 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

  18. 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.

  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. 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.

  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. 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.

  12. 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.

  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. 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.

  19. 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

  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 (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.''

  13. 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.''

  14. 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.''

  15. 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.''

  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

    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

    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.

  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 (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.

  7. 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

  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. 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)

  10. 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.

  11. 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

  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. 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

  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. 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

  20. 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

  1. 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

  2. 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

  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. The remaining activities (including cleaning) will be moved on the 1st July. Minor perturbation in the service might occur. The contact number will ...

  4. 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...

  5. 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

  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. 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...

  9. 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

  10. 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)

  11. 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.

  12. 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

  13. 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

  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. 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

  19. 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

  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. 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)

  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. 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

  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. 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

  11. 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

  12. 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.

  13. 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)

  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. 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)

  19. 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)

  20. 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.

  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. 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.

  4. 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.

  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. 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.

  7. 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.

  8. 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

  9. 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

  10. 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

  11. 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

  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. 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

  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. 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)

  17. 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

  18. 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

  19. 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.

  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. 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

  2. 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

  3. 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

  4. 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

  5. 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)

  6. 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)

  7. 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

  8. 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)

  9. 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

  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

    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

    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.

  13. 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

  14. 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

  15. 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.

  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. 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

  3. 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.

  4. 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.

  5. 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

  6. 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

  7. 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)

  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. 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

  16. 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.

  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. 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

  19. 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...

  20. 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.

  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. 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

  4. 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

  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. 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

  7. 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

  8. 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.

  9. 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...

  10. 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.

  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. 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)

  13. 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)

  14. 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)

  15. 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.

  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. 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

  18. 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...

  19. 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.

  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. 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.

  5. 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

  6. 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.

  7. 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.

  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. 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...