WorldWideScience

Sample records for beryllium processing facility

  1. Beryllium chemistry and processing

    CERN Document Server

    Walsh, Kenneth A

    2009-01-01

    This book introduces beryllium; its history, its chemical, mechanical, and physical properties including nuclear properties. The 29 chapters include the mineralogy of beryllium and the preferred global sources of ore bodies. The identification and specifics of the industrial metallurgical processes used to form oxide from the ore and then metal from the oxide are thoroughly described. The special features of beryllium chemistry are introduced, including analytical chemical practices. Beryllium compounds of industrial interest are identified and discussed. Alloying, casting, powder processing, forming, metal removal, joining and other manufacturing processes are covered. The effect of composition and process on the mechanical and physical properties of beryllium alloys assists the reader in material selection. The physical metallurgy chapter brings conformity between chemical and physical metallurgical processing of beryllium, metal, alloys, and compounds. The environmental degradation of beryllium and its all...

  2. Beryllium. Evaluation of beryllium hydroxide industrial processes. Pt. 3

    International Nuclear Information System (INIS)

    Lires, O.A.; Delfino, C.A.; Botbol, J.

    1991-01-01

    This work continues the 'Beryllium' series. It is a historical review of different industrial processes of beryllium hydroxide obtention from beryllium ores. Flowsheats and operative parameters of five plants are provided. These plants (Degussa, Brush Beryllium Co., Beryllium Corp., Murex Ltd., SAPPI) were selected as representative samples of diverse commercial processes in different countries. (Author) [es

  3. Safe waste management practices in beryllium facilities

    International Nuclear Information System (INIS)

    Bhat, P.N.; Soundararajan, S.; Sharma, D.N.

    2012-01-01

    Beryllium, an element with the atomic symbol Be, atomic number 4, has very high stiffness to weight ratio and low density. It has good electrical conductive properties with low coefficient of thermal expansion. These properties make the metal beryllium very useful in varied technological endeavours, However, beryllium is recognised as one of the most toxic metals. Revelation of toxic effects of beryllium resulted in institution of stringent health and safety practices in beryllium handling facilities. The waste generated in such facilities may contain traces of beryllium. Any such waste should be treated as toxic waste and suitable safe waste management practices should be adopted. By instituting appropriate waste management practice and through a meticulously incorporated safety measures and continuous surveillance exercised in such facilities, total safety can be ensured. This paper broadly discusses health hazards posed by beryllium and safe methods of management of beryllium bearing wastes. (author)

  4. Development of retrospective quantitative and qualitative job-exposure matrices for exposures at a beryllium processing facility.

    Science.gov (United States)

    Couch, James R; Petersen, Martin; Rice, Carol; Schubauer-Berigan, Mary K

    2011-05-01

    To construct a job-exposure matrix (JEM) for an Ohio beryllium processing facility between 1953 and 2006 and to evaluate temporal changes in airborne beryllium exposures. Quantitative area- and breathing-zone-based exposure measurements of airborne beryllium were made between 1953 and 2006 and used by plant personnel to estimate daily weighted average (DWA) exposure concentrations for sampled departments and operations. These DWA measurements were used to create a JEM with 18 exposure metrics, which was linked to the plant cohort consisting of 18,568 unique job, department and year combinations. The exposure metrics ranged from quantitative metrics (annual arithmetic/geometric average DWA exposures, maximum DWA and peak exposures) to descriptive qualitative metrics (chemical beryllium species and physical form) to qualitative assignment of exposure to other risk factors (yes/no). Twelve collapsed job titles with long-term consistent industrial hygiene samples were evaluated using regression analysis for time trends in DWA estimates. Annual arithmetic mean DWA estimates (overall plant-wide exposures including administration, non-production, and production estimates) for the data by decade ranged from a high of 1.39 μg/m(3) in the 1950s to a low of 0.33 μg/m(3) in the 2000s. Of the 12 jobs evaluated for temporal trend, the average arithmetic DWA mean was 2.46 μg/m(3) and the average geometric mean DWA was 1.53 μg/m(3). After the DWA calculations were log-transformed, 11 of the 12 had a statistically significant (p < 0.05) decrease in reported exposure over time. The constructed JEM successfully differentiated beryllium exposures across jobs and over time. This is the only quantitative JEM containing exposure estimates (average and peak) for the entire plant history.

  5. Validation of cleaning method for various parts fabricated at a Beryllium facility

    Energy Technology Data Exchange (ETDEWEB)

    Davis, Cynthia M. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-12-15

    This study evaluated and documented a cleaning process that is used to clean parts that are fabricated at a beryllium facility at Los Alamos National Laboratory. The purpose of evaluating this cleaning process was to validate and approve it for future use to assure beryllium surface levels are below the Department of Energy’s release limits without the need to sample all parts leaving the facility. Inhaling or coming in contact with beryllium can cause an immune response that can result in an individual becoming sensitized to beryllium, which can then lead to a disease of the lungs called chronic beryllium disease, and possibly lung cancer. Thirty aluminum and thirty stainless steel parts were fabricated on a lathe in the beryllium facility, as well as thirty-two beryllium parts, for the purpose of testing a parts cleaning method that involved the use of ultrasonic cleaners. A cleaning method was created, documented, validated, and approved, to reduce beryllium contamination.

  6. Beryllium

    International Nuclear Information System (INIS)

    1988-01-01

    In this data sheet the occurrence, ore processing, chemical and physical properties and the uses of beryllium and its alloys is presented. The hazards involved in the use of beryllium and its compounds in the laboratory are discussed with particular reference to its toxicity, carcinogenicity, handling, storage, disposal, fire prevention and the principal health hazards. Further reading is suggested. (UK)

  7. Beryllium

    Science.gov (United States)

    Foley, Nora K.; Jaskula, Brian W.; Piatak, Nadine M.; Schulte, Ruth F.; Schulz, Klaus J.; DeYoung,, John H.; Seal, Robert R.; Bradley, Dwight C.

    2017-12-19

    producing aircraft and satellite structural components that have a high stiffness-to-weight ratio and low surface vibration. Beryllium oxide ceramics are used in a wide range of applications, including missile guidance systems, radar applications, and cell phone transmitters, and they are critical to medical technologies, such as magnetic resonance imaging (MRI) machines, medical lasers, and portable defibrillators.The United States is expected to remain self-sufficient with respect to most of its beryllium requirements, based on information available at the time this chapter was prepared (2013). The United States is one of only three countries that currently process beryllium ores and concentrate them into beryllium products, and these three countries supply most of the rest of the world with these products. Exploration for new deposits in the United States is limited because domestic beryllium production is dominated by a single producer that effectively controls the domestic beryllium market, which is relatively small and specialized, and the market cannot readily accommodate new competition on the raw material supply side.

  8. Hydrodynamic instabilities in beryllium targets for the National Ignition Facility

    Energy Technology Data Exchange (ETDEWEB)

    Yi, S. A., E-mail: austinyi@lanl.gov; Simakov, A. N.; Wilson, D. C.; Olson, R. E.; Kline, J. L.; Batha, S. H. [Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, New Mexico 87545 (United States); Clark, D. S.; Hammel, B. A.; Milovich, J. L.; Salmonson, J. D.; Kozioziemski, B. J. [Lawrence Livermore National Laboratory, P.O. Box 808, Livermore, California 94551 (United States)

    2014-09-15

    Beryllium ablators offer higher ablation velocity, rate, and pressure than their carbon-based counterparts, with the potential to increase the probability of achieving ignition at the National Ignition Facility (NIF) [E. I. Moses et al., Phys. Plasmas 16, 041006 (2009)]. We present here a detailed hydrodynamic stability analysis of low (NIF Revision 6.1) and high adiabat NIF beryllium target designs. Our targets are optimized to fully utilize the advantages of beryllium in order to suppress the growth of hydrodynamic instabilities. This results in an implosion that resists breakup of the capsule, and simultaneously minimizes the amount of ablator material mixed into the fuel. We quantify the improvement in stability of beryllium targets relative to plastic ones, and show that a low adiabat beryllium capsule can be at least as stable at the ablation front as a high adiabat plastic target.

  9. New facility for post irradiation examination of neutron irradiated beryllium

    International Nuclear Information System (INIS)

    Ishitsuka, Etsuo; Kawamura, Hiroshi

    1995-01-01

    Beryllium is expected as a neutron multiplier and plasma facing materials in the fusion reactor, and the neutron irradiation data on properties of beryllium up to 800 degrees C need for the engineering design. The acquisition of data on the tritium behavior, swelling, thermal and mechanical properties are first priority in ITER design. Facility for the post irradiation examination of neutron irradiated beryllium was constructed in the hot laboratory of Japan Materials Testing Reactor to get the engineering design data mentioned above. This facility consist of the four glove boxes, dry air supplier, tritium monitoring and removal system, storage box of neutron irradiated samples. Beryllium handling are restricted by the amount of tritium;7.4 GBq/day and 60 Co;7.4 MBq/day

  10. Beryllium

    International Nuclear Information System (INIS)

    Hansen, N.B.

    1980-01-01

    A method for determination of beryllium in minerals and rocks is described. The method comprises microanalysis and trace analysis. Because of the toxidity of beryllium the method is designed for determination of a hitherto unknown small amount, 1-10 nanogram Be. With the optimal amount for determination, 3 ng Be, the relative error of the method is 10%. The description includes an inventory of chemicals and apparatus, also an example of application of the method on the mineral epididymite. In brief, the sample is melted with sodium carbonate and sodium tetra borate; when required the sample in advance is fumed with hydrogen fluoride and sulphuric acid to evaporate silica. The residuum is dissolved in water and hydrogen chloride, upon which the solution is made to volume. In the Ring oven interfering compounds are masked with EDTA. Beryllium is settled with chrome azurol and ammonia. Beryllium is identified and evaluated in comparison with previously produced standards. (author)

  11. Beryllium

    International Nuclear Information System (INIS)

    Hansen, N.B.

    1979-01-01

    A method for determination of beryllium in minerals and rocks is described. The method comprises microanalysis and trace analysis. Because of the toxidity of beryllium the method is designed for determination of a hitherto unknown small amount, 1-10 nanogram Be. With the optimal amount for determination, 3 ng Be, the relative error of the method is 10%. The description includes an inventory of chemicals and apparatus, also an example of application of the method on the mineral epididymite. In brief, the sample is melted with sodium carbonate and sodium tetra borate; when required the sample in advance is fumed with hydrogen fluoride and sulphuric acid to evaporate silica. The residuum is dissolved in water and hydrogen chloride, upon which the solution is made to volume. In the Ring oven interfering compounds are masked with EDTA. Beryllium is settled with chrome azurol and ammonia. Beryllium is identified and evaluated in comparison with previously produced standards. (author)

  12. Control of beryllium powder at a DOE facility

    International Nuclear Information System (INIS)

    Langner, G.C.; Creek, K.L.; Castro, R.G.

    1997-01-01

    Beryllium is contained in a number of domestic and national defense items. Although many items might contain beryllium in some manner, few people need worry about the adverse effects caused by exposure to beryllium because it is the inhalable form of beryllium that is most toxic. Chronic beryllium disease (CBD), a granulomas and fibrotic lung disease with long latency, can be developed after inhalation exposures to beryllium. It is a progressive, debilitating lung disease. Its occurrence in those exposed to beryllium has been difficult to predict because some people seem to react to low concentration exposures whereas others do not react to high concentration exposures. Onset of the disease frequently occurs between 15 to 20 years after exposure begins. Some people develop the disease after many years of low concentration exposures but others do not develop CBD even though beryllium is shown to be present in lungs and urine. Conclusions based on these experiences are that their is some immunological dependence of developing CBD in about 3--4% of the exposed population, but the exact mechanism involved has not yet been identified. Acute beryllium disease can occur after a single exposure to a concentration of greater than 0.100 mg/m3 (inhalation exposure); it is characterized by the development of chemical pneumoconiosis, a respiratory disease. The acute effect of skin contact is a dermatitis characterized by itching and reddened, elevated, or fluid-accumulated lesions which appear particularly on the exposed surfaces of the body, especially the face, neck, arms, and hands. Small particles of beryllium that enter breaks in the skin can lead to the development of granulomas and/or open sores that do not heal until the beryllium has been removed. Our interest is only airborne beryllium, which is found in areas that machine or produce beryllium

  13. First beryllium capsule implosions on the National Ignition Facility

    Energy Technology Data Exchange (ETDEWEB)

    Kline, J. L.; Yi, S. A.; Simakov, A. N.; Olson, R. E.; Wilson, D. C.; Kyrala, G. A.; Perry, T. S.; Batha, S. H.; Zylstra, A. B. [Los Alamos National Laboratory, Los Alamos, New Mexico 87544 (United States); Dewald, E. L.; Tommasini, R.; Ralph, J. E.; Strozzi, D. J.; MacPhee, A. G.; Callahan, D. A.; Hinkel, D. E.; Hurricane, O. A.; Milovich, J. L.; Rygg, J. R.; Khan, S. F. [Lawrence Livermore National Laboratory, Livermore, California 94550 (United States); and others

    2016-05-15

    The first indirect drive implosion experiments using Beryllium (Be) capsules at the National Ignition Facility confirm the superior ablation properties and elucidate possible Be-ablator issues such as hohlraum filling by ablator material. Since the 1990s, Be has been the preferred Inertial Confinement Fusion (ICF) ablator because of its higher mass ablation rate compared to that of carbon-based ablators. This enables ICF target designs with higher implosion velocities at lower radiation temperatures and improved hydrodynamic stability through greater ablative stabilization. Recent experiments to demonstrate the viability of Be ablator target designs measured the backscattered laser energy, capsule implosion velocity, core implosion shape from self-emission, and in-flight capsule shape from backlit imaging. The laser backscatter is similar to that from comparable plastic (CH) targets under the same hohlraum conditions. Implosion velocity measurements from backlit streaked radiography show that laser energy coupling to the hohlraum wall is comparable to plastic ablators. The measured implosion shape indicates no significant reduction of laser energy from the inner laser cone beams reaching the hohlraum wall as compared with plastic and high-density carbon ablators. These results indicate that the high mass ablation rate for beryllium capsules does not significantly alter hohlraum energetics. In addition, these data, together with data for low fill-density hohlraum performance, indicate that laser power multipliers, required to reconcile simulations with experimental observations, are likely due to our limited understanding of the hohlraum rather than the capsule physics since similar multipliers are needed for both Be and CH capsules as seen in experiments.

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

    Energy Technology Data Exchange (ETDEWEB)

    BECHTEL NEVADA

    2004-07-01

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

  15. Assessment of the beryllium lymphocyte proliferation test using statistical process control.

    Science.gov (United States)

    Cher, Daniel J; Deubner, David C; Kelsh, Michael A; Chapman, Pamela S; Ray, Rose M

    2006-10-01

    Despite more than 20 years of surveillance and epidemiologic studies using the beryllium blood lymphocyte proliferation test (BeBLPT) as a measure of beryllium sensitization (BeS) and as an aid for diagnosing subclinical chronic beryllium disease (CBD), improvements in specific understanding of the inhalation toxicology of CBD have been limited. Although epidemiologic data suggest that BeS and CBD risks vary by process/work activity, it has proven difficult to reach specific conclusions regarding the dose-response relationship between workplace beryllium exposure and BeS or subclinical CBD. One possible reason for this uncertainty could be misclassification of BeS resulting from variation in BeBLPT testing performance. The reliability of the BeBLPT, a biological assay that measures beryllium sensitization, is unknown. To assess the performance of four laboratories that conducted this test, we used data from a medical surveillance program that offered testing for beryllium sensitization with the BeBLPT. The study population was workers exposed to beryllium at various facilities over a 10-year period (1992-2001). Workers with abnormal results were offered diagnostic workups for CBD. Our analyses used a standard statistical technique, statistical process control (SPC), to evaluate test reliability. The study design involved a repeated measures analysis of BeBLPT results generated from the company-wide, longitudinal testing. Analytical methods included use of (1) statistical process control charts that examined temporal patterns of variation for the stimulation index, a measure of cell reactivity to beryllium; (2) correlation analysis that compared prior perceptions of BeBLPT instability to the statistical measures of test variation; and (3) assessment of the variation in the proportion of missing test results and how time periods with more missing data influenced SPC findings. During the period of this study, all laboratories displayed variation in test results that

  16. Experimental studies and modeling of processes of hydrogen isotopes interaction with beryllium

    Energy Technology Data Exchange (ETDEWEB)

    Tazhibaeva, I.L.; Chikhray, Y.V.; Romanenko, O.G.; Klepikov, A.Kh.; Shestakov, V.P.; Kulsartov, T.V. [Science Research Inst. of Experimental and Theoretical Physics of Kazakh State Univ., Almaty (Kazakhstan); Kenzhin, E.A.

    1998-01-01

    The objective of this work was to clarify the surface beryllium oxide influence on hydrogen-beryllium interaction characteristics. Analysis of experimental data and modeling of processes of hydrogen isotopes accumulation, diffusion and release from neutron irradiated beryllium was used to achieve this purpose as well as the investigations of the changes of beryllium surface element composition being treated by H{sup +} and Ar{sup +} plasma glowing discharge. (author)

  17. Beryllium processing technology review for applications in plasma-facing components

    International Nuclear Information System (INIS)

    Castro, R.G.; Jacobson, L.A.; Stanek, P.W.

    1993-07-01

    Materials research and development activities for the International Thermonuclear Experimental Reactor (ITER), i.e., the next generation fusion reactor, are investigating beryllium as the first-wall containment material for the reactor. Important in the selection of beryllium is the ability to process, fabricate and repair beryllium first-wall components using existing technologies. Two issues that will need to be addressed during the engineering design activity will be the bonding of beryllium tiles in high-heat-flux areas of the reactor, and the in situ repair of damaged beryllium tiles. The following review summarizes the current technology associated with welding and joining of beryllium to itself and other materials, and the state-of-the-art in plasma-spray technology as an in situ repair technique for damaged beryllium tiles. In addition, a review of the current status of beryllium technology in the former Soviet Union is also included

  18. Beryllium processing technology review for applications in plasma-facing components

    Energy Technology Data Exchange (ETDEWEB)

    Castro, R.G.; Jacobson, L.A.; Stanek, P.W.

    1993-07-01

    Materials research and development activities for the International Thermonuclear Experimental Reactor (ITER), i.e., the next generation fusion reactor, are investigating beryllium as the first-wall containment material for the reactor. Important in the selection of beryllium is the ability to process, fabricate and repair beryllium first-wall components using existing technologies. Two issues that will need to be addressed during the engineering design activity will be the bonding of beryllium tiles in high-heat-flux areas of the reactor, and the in situ repair of damaged beryllium tiles. The following review summarizes the current technology associated with welding and joining of beryllium to itself and other materials, and the state-of-the-art in plasma-spray technology as an in situ repair technique for damaged beryllium tiles. In addition, a review of the current status of beryllium technology in the former Soviet Union is also included.

  19. Examination of Beryllium Under Intense High Energy Proton Beam at CERN's HiRadMat Facility

    CERN Document Server

    Ammigan, K.; Hurh, P.; Zwaska, R.; Atherton, A.; Caretta, O.; Davenne,T.; Densham, C.; Fitton, M.; Loveridge, P.; O'Dell, J.; Roberts, S.; Kuksenko, V.; Butcher, M.; Calviani, M.; Guinchard, M.; Losito, R.

    2017-01-01

    Beryllium is extensively used in various accelerator beam lines and target facilities as material for beam win- dows, and to a lesser extent, as secondary particle produc- tion targets. With increasing beam intensities of future ac- celerator facilities, it is critical to understand the response of beryllium under extreme conditions to avoid compro- mising particle production efficiency by limiting beam pa- rameters. As a result, the planned experiment at CERN’s HiRadMat facility will take advantage of the test facility’s tunable high intensity proton beam to probe and investigate the damage mechanisms of several grades of beryllium. The test matrix will consist of multiple arrays of thin discs of varying thicknesses as well as cylinders, each exposed to increasing beam intensities. Online instrumentations will acquire real time temperature, strain, and vibration data of the cylinders, while Post-Irradiation-Examination (PIE) of the discs will exploit advanced microstructural characteri- zation and imagin...

  20. Examination of Beryllium Under Intense High Energy Proton Beam at CERN's HiRadMat Facility

    CERN Document Server

    Ammigan, K; Hurh, P; Zwaska, R; Atherton, A; Caretta, O; Davenne, t; Densham, C; Fitton, M; Loveridge, P; O'Dell, J; Roberts, S; Kuksenko, v; Butcher, M; Calviani, M; Guinchard, M; Losito, R

    2015-01-01

    Beryllium is extensively used in various accelerator beam lines and target facilities as material for beam win- dows, and to a lesser extent, as secondary particle produc- tion targets. With increasing beam intensities of future ac- celerator facilities, it is critical to understand the response of beryllium under extreme conditions to avoid compro- mising particle production efficiency by limiting beam pa- rameters. As a result, the planned experiment at CERN’s HiRadMat facility will take advantage of the test facility’s tunable high intensity proton beam to probe and investigate the damage mechanisms of several grades of beryllium. The test matrix will consist of multiple arrays of thin discs of varying thicknesses as well as cylinders, each exposed to increasing beam intensities. Online instrumentations will acquire real time temperature, strain, and vibration data of the cylinders, while Post-Irradiation-Examination (PIE) of the discs will exploit advanced microstructural characteri- zation and imagin...

  1. The effect of processing parameters on plasma sprayed beryllium for fusion applications

    International Nuclear Information System (INIS)

    Castro, R.G.; Stanek, P.W.; Jacobson, L.A.; Cowgill, D.F.; Snead, L.L.

    1993-01-01

    Plasma spraying is being investigated as a potential coating technique for applying thin (0.1--5mm) layers of beryllium on plasma facing surfaces of blanket modules in ITER and also as an in-situ repair technique for repairing eroded beryllium surfaces in high heat flux divertor regions. High density spray deposits (>98% of theoretical density) of beryllium will be required in order to maximize the thermal conductivity of the beryllium coatings. A preliminary investigation was done to determine the effect of various processing parameters (particle size, particle morphology, secondary gas additions and reduced chamber pressure) on the as-deposited density of beryllium. The deposits were made using spherical beryllium feedstock powder which was produced by centrifugal atomization at Los Alamos National Laboratory (LANL). Improvements in the as-deposited densities and deposit efficiencies of the beryllium spray deposits will be discussed along with the corresponding thermal conductivity and outgassing behavior of these deposits

  2. The effect of processing parameters on plasma sprayed beryllium for fusion applications

    International Nuclear Information System (INIS)

    Castro, R.G.; Stanek, P.W.; Jacobson, L.W.; Cowgill, D.F.

    1993-01-01

    Plasma spraying is being investigated as a potential coating technique for applying thin (0.1-5mm) layers of beryllium on plasma facing surfaces of blanket modules in ITER and also as an in-situ repair technique for repairing eroded beryllium surfaces in high heat flux divertor regions. High density spray deposits (>98% of theoretical density) of beryllium will be required in order to maximize the thermal conductivity of the beryllium coatings. A preliminary investigation was done to determine the effect of various processing parameters (particle size, particle morphology, secondary gas additions and reduced chamber pressure) on the as-deposited density of beryllium. The deposits were made using spherical beryllium feedstock powder which was produced by centrifugal atomization at Los Alamos National Laboratory (LANL). Improvements in the as-deposited densities and deposit efficiencies of the beryllium spray deposits will be discussed along with the corresponding thermal conductivity and outgassing behavior of these deposits. (orig.)

  3. X-ray drive of beryllium capsule implosions at the National Ignition Facility

    International Nuclear Information System (INIS)

    Wilson, D C; Yi, S A; Simakov, A N; Kline, J L; Kyrala, G A; Olson, R E; Zylstra, A B; Dewald, E L; Tommasini, R; Ralph, J E; Strozzi, D J; Celliers, P M; Schneider, M B; MacPhee, A G; Callahan, D A; Hurricane, O A; Milovich, J L; Hinkel, D E; Rygg, J R; Rinderknecht, H G

    2016-01-01

    National Ignition Facility experiments with beryllium capsules have followed a path begun with “high-foot” plastic capsule implosions. Three shock timing keyhole targets, one symmetry capsule, a streaked backlit capsule, and a 2D backlit capsule were fielded before the DT layered shot. After backscatter subtraction, laser drive degradation is needed to match observed X-ray drives. VISAR measurements determined drive degradation for the picket, trough, and second pulse. Time dependence of the total Dante flux reflects degradation of the of the third laser pulse. The same drive degradation that matches Dante data for three beryllium shots matches Dante and bangtimes for plastic shots N130501 and N130812. In the picket of both Be and CH hohlraums, calculations over-estimate the x-ray flux > 1.8 keV by ∼100X, while calculating the total flux correctly. In beryllium calculations these X-rays cause an early expansion of the beryllium/fuel interface at ∼3 km/s. VISAR measurements gave only ∼0.3 km/s. The X-ray drive on the Be DT capsule was further degraded by an unplanned decrease of 9% in the total picket flux. This small change caused the fuel adiabat to rise from 1.8 to 2.3. The first NIF beryllium DT implosion achieved 29% of calculated yield, compared to CH capsules with 68% and 21%. (paper)

  4. Defense programs beryllium good practice guide

    International Nuclear Information System (INIS)

    Herr, M.

    1997-07-01

    Within the DOE, it has recently become apparent that some contractor employees who have worked (or are currently working) with and around beryllium have developed chronic beryllium disease (CBD), an occupational granulomatous lung disorder. Respiratory exposure to aerosolized beryllium, in susceptible individuals, causes an immunological reaction that can result in granulomatous scarring of the lung parenchyma, shortness of breath, cough, fatigue, weight loss, and, ultimately, respiratory failure. Beryllium disease was originally identified in the 1940s, largely in the fluorescent light industry. In 1950, the Atomic Energy Commission (AEC) introduced strict exposure standards that generally curtailed both the acute and chronic forms of the disease. Beginning in 1984, with the identification of a CBD case in a DOE contractor worker, there was increased scrutiny of both industrial hygiene practices and individuals in this workforce. To date, over 100 additional cases of beryllium-specific sensitization and/or CBD have been identified. Thus, a disease previously thought to be largely eliminated by the adoption of permissible exposure standards 45 years ago is still a health risk in certain workforces. This good practice guide forms the basis of an acceptable program for controlling workplace exposure to beryllium. It provides (1) Guidance for minimizing worker exposure to beryllium in Defense Programs facilities during all phases of beryllium-related work, including the decontamination and decommissioning (D ampersand D) of facilities. (2) Recommended controls to be applied to the handling of metallic beryllium and beryllium alloys, beryllium oxide, and other beryllium compounds. (3) Recommendations for medical monitoring and surveillance of workers exposed (or potentially exposed) to beryllium, based on the best current understanding of beryllium disease and medical diagnostic tests available. (4) Site-specific safety procedures for all processes of beryllium that is

  5. Low Prevalence of Chronic Beryllium Disease Among Workers at aNuclearWeaponsResearchandDevelopmentFacility

    Science.gov (United States)

    Arjomandi, Mehrdad; Seward, James; Gotway, Michael B.; Nishimura, Stephen; Fulton, George P.; Thundiyil, Josef; King, Talmadge E.; Harber, Philip; Balmes, John R.

    2012-01-01

    Objective To study the prevalence of beryllium sensitization (BeS) and chronic beryllium disease (CBD) in a cohort of workers from a nuclear weapons research and development facility. Methods We evaluated 50 workers with BeS with medical and occupational histories, physical examination, chest imaging with high-resolution computed tomography (N = 49), and pulmonary function testing. Forty of these workers also underwent bronchoscopy for bronchoalveolar lavage and transbronchial biopsies. Results The mean duration of employment at the facility was 18 years and the mean latency (from first possible exposure) to time of evaluation was 32 years. Five of the workers had CBD at the time of evaluation (based on histology or high-resolution computed tomography); three others had evidence of probable CBD. Conclusions These workers with BeS, characterized by a long duration of potential Be exposure and a long latency, had a low prevalence of CBD. PMID:20523233

  6. Low Prevalence of Chronic Beryllium Disease among Workers at a Nuclear Weapons Research and Development Facility

    Energy Technology Data Exchange (ETDEWEB)

    Arjomandi, M; Seward, J P; Gotway, M B; Nishimura, S; Fulton, G P; Thundiyil, J; King, T E; Harber, P; Balmes, J R

    2010-01-11

    To study the prevalence of beryllium sensitization (BeS) and chronic beryllium disease (CBD) in a cohort of workers from a nuclear weapons research and development facility. We evaluated 50 workers with BeS with medical and occupational histories, physical examination, chest imaging with HRCT (N=49), and pulmonary function testing. Forty of these workers also underwent bronchoscopy for bronchoalveolar lavage (BAL) and transbronchial biopsies. The mean duration of employment at the facility was 18 yrs and the mean latency (from first possible exposure) to time of evaluation was 32 yrs. Five of the workers had CBD at the time of evaluation (based on histology or HRCT); three others had evidence of probable CBD. These workers with BeS, characterized by a long duration of potential Be exposure and a long latency, had a low prevalence of CBD.

  7. Offshoots from beryllium development programme

    International Nuclear Information System (INIS)

    Sharma, B.P.; Sinha, P.K.

    1995-01-01

    The paper briefly presents extraction and processing of beryllium metal as practiced in the beryllium facilities at Turbhe, New Bombay. These facilities have been set up to meet the indigenous requirements of the metal in space and nuclear science programmes. As offshoot of this beryllium development programme has been the development of a number of pyro and powder metallurgical equipment. Indigenous development of these pieces of equipment has been a professionally rewarding experience. Efforts are now on to promote these equipment for industrial use. (author). 6 refs., 6 figs., 2 tabs

  8. Plasma spraying of beryllium and beryllium-aluminum-silver alloys

    International Nuclear Information System (INIS)

    Castro, R.G.; Stanek, P.W.; Elliott, K.E.; Jacobson, L.A.

    1994-01-01

    A preliminary investigation on plasma-spraying of beryllium and a beryllium-aluminum-4% silver alloy was done at the Los Alamos National Laboratory's Beryllium Atomization and Thermal Spray Facility (BATSF). Spherical Be and Be-Al-4%Ag powders, which were produced by centrifugal atomization, were used as feedstock material for plasma-spraying. The spherical morphology of the powders allowed for better feeding of fine (<38 μm) powders into the plasma-spray torch. The difference in the as-deposited densities and deposit efficiencies of the two plasma-sprayed powders will be discussed along with the effect of processing parameters on the as-deposited microstructure of the Be-Al-4%Ag. This investigation represents ongoing research to develop and characterize plasma-spraying of beryllium and beryllium-aluminum alloys for magnetic fusion and aerospace applications

  9. Plasma spraying of beryllium and beryllium-aluminum-silver alloys

    International Nuclear Information System (INIS)

    Castro, R.G.; Stanek, P.W.; Elliott, K.E.; Jacobson, L.A.

    1993-01-01

    A preliminary investigation on plasma-spraying of beryllium and a beryllium-aluminum 4% silver alloy was done at the Los Alamos National Laboratory's Beryllium Atomization and Thermal Spray Facility (BATSF). Spherical Be and Be-Al-4%Ag powders, which were produced by centrifugal atomization, were used as feedstock material for plasma-spraying. The spherical morphology of the powders allowed for better feeding of fine (<38 μm) powders into the plasma-spray torch. The difference in the as-deposited densities and deposit efficiencies of the two plasma-sprayed powders will be discussed along with the effect of processing parameters on the as-deposited microstructure of the Be-Al-4%Ag. This investigation represents ongoing research to develop and characterize plasma-spraying of beryllium and beryllium-aluminum alloys for magnetic fusion and aerospace applications

  10. Thermal shock tests with beryllium coupons in the electron beam facility JUDITH

    International Nuclear Information System (INIS)

    Roedig, M.; Duwe, R.; Schuster, J.L.A.

    1995-01-01

    Several grades of American and Russian beryllium have been tested in high heat flux tests by means of an electron beam facility. For safety reasons, major modifications of the facility had to be fulfilled in advance to the tests. The influence of energy densities has been investigated in the range between 1 and 7 MJ/m 2 . In addition the influence of an increasing number of shots at constant energy density has been studied. For all samples, surface profiles have been measured before and after the experiments. Additional information has been gained from scanning electron microscopy, and from metallography

  11. Investigation of beryllium/steam interaction

    Energy Technology Data Exchange (ETDEWEB)

    Chekhonadskikh, A.M.; Vurim, A.D.; Vasilyev, Yu.S.; Pivovarov, O.S. [Inst. of Atomic Energy National Nuclear Center of the Republic of Kazakstan Semipalatinsk (Kazakhstan); Shestakov, V.P.; Tazhibayeva, I.L.

    1998-01-01

    In this report program on investigations of beryllium emissivity and transient processes on overheated beryllium surface attacked by water steam to be carried out in IAE NNC RK within Task S81 TT 2096-07-16 FR. The experimental facility design is elaborated in this Report. (author)

  12. Laser fabrication of beryllium components

    International Nuclear Information System (INIS)

    Hanafee, J.E.; Ramos, T.J.

    1995-08-01

    Working with the beryllium industry on commercial applications and using prototype parts, the authors have found that the use of lasers provides a high-speed, low-cost method of cutting beryllium metal, beryllium alloys, and beryllium-beryllium oxide composites. In addition, they have developed laser welding processes for commercial structural grades of beryllium that do not need a filler metal; i.e., autogenous welds were made in commercial structural grades of beryllium by using lasers

  13. Modeling tritium processes in plasma-facing beryllium

    International Nuclear Information System (INIS)

    Longhurst, G.R.; Anderl, R.A.; Dolan, T.J.; Mulock, M.J.

    1995-01-01

    In this paper we present techniques and recommended parameters for modeling tritium implantation, trapping and release, and permeation, in beryllium-clad structures adjacent to the plasma. Among the features that should be considered are the effects of surface films, the mobility of beryllium through those films, damage caused by ion implantation, especially in regions where pitting may be expected, and bubble formation. Tritium transport parameters recommended are based on fits with experimental data and available theory. Estimates of inventories in ITER using these parameters are also given. 31 refs., 2 figs., 1 tab

  14. Extractive metallurgy of the beryllium

    International Nuclear Information System (INIS)

    Alonso, Neusa; Capocchi, Jose Deodoro Trani

    1995-01-01

    A bibliographic review is performed on the beryllium extractive metallurgy. The work describes the main type of ores and processes applied to the metallic beryllium production, beryllium oxide production using fluoride, sulfide and direct chlorination. The thermodynamic consideration are made on beryllium reduction processes, discussing the viability of the beryllium oxide and hallide reduction processes. Under the technological viewpoint, the Cu-Be alloys main production processes are discussed, and the main toxicity problems related with beryllium are mentioned

  15. Supplementary radiological and beryllium characterization of the facility at 425 Peek Street, Schenectady, New York

    International Nuclear Information System (INIS)

    Foley, R.D.; Allred, J.F.; Carrier, R.F.

    1994-10-01

    At the request of the Office of Naval Reactors through the Office of Remedial Action and Waste Technology, a radiological survey of the Peek Street industrial facility, the adjacent state-owned bike path, and two nearby residential properties was conducted by Oak Ridge National Laboratory (ORNL) in November 1989. The results indicated small isolated areas that exceeded DOE guidelines. These areas totaled approximately 0.2 m 2 of floor area and approximately 3 m 2 of wall area inside the building, and two small areas totaling approximately 5 m 2 outside the building. A small section of one of these areas extended beyond the fence on the east side of the industrial property onto the state-owned property. No residual radioactive material or elevated radiation levels were detected on any portion of the paved section of the bike path or the residential properties adjacent to the site. Because the elevated radiation levels were localized and limited in extent, any credible use scenario, including current use conditions, indicated that no significant radiation exposures would accrue to individuals frequenting the area. Samples were also analyzed for elemental beryllium since that material had formerly been used at the site. In conjunction with the planned remediation at the facility, a supplementary characterization survey was performed to further define the areas containing beryllium in excess of the identified guidelines. Additional radiological characterization of Ra-226, Th-232, and U-238 was also performed in areas that were largely inaccessible prior to the remediation efforts

  16. Defense programs beryllium good practice guide

    Energy Technology Data Exchange (ETDEWEB)

    Herr, M.

    1997-07-01

    Within the DOE, it has recently become apparent that some contractor employees who have worked (or are currently working) with and around beryllium have developed chronic beryllium disease (CBD), an occupational granulomatous lung disorder. Respiratory exposure to aerosolized beryllium, in susceptible individuals, causes an immunological reaction that can result in granulomatous scarring of the lung parenchyma, shortness of breath, cough, fatigue, weight loss, and, ultimately, respiratory failure. Beryllium disease was originally identified in the 1940s, largely in the fluorescent light industry. In 1950, the Atomic Energy Commission (AEC) introduced strict exposure standards that generally curtailed both the acute and chronic forms of the disease. Beginning in 1984, with the identification of a CBD case in a DOE contractor worker, there was increased scrutiny of both industrial hygiene practices and individuals in this workforce. To date, over 100 additional cases of beryllium-specific sensitization and/or CBD have been identified. Thus, a disease previously thought to be largely eliminated by the adoption of permissible exposure standards 45 years ago is still a health risk in certain workforces. This good practice guide forms the basis of an acceptable program for controlling workplace exposure to beryllium. It provides (1) Guidance for minimizing worker exposure to beryllium in Defense Programs facilities during all phases of beryllium-related work, including the decontamination and decommissioning (D&D) of facilities. (2) Recommended controls to be applied to the handling of metallic beryllium and beryllium alloys, beryllium oxide, and other beryllium compounds. (3) Recommendations for medical monitoring and surveillance of workers exposed (or potentially exposed) to beryllium, based on the best current understanding of beryllium disease and medical diagnostic tests available. (4) Site-specific safety procedures for all processes of beryllium that is likely to

  17. Report on Beryllium Strength Experiments Conducted at the TA-55 40 mm Impact Test Facility, Fiscal Year 2017

    Energy Technology Data Exchange (ETDEWEB)

    Anderson, William Wyatt [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Hollowell, Benjamin Charles [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Martinez, Todd P. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Owens, Charles Thomas [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Rivera, Joseph Lee [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-10-10

    A series of experiments is currently in progress at eth 40 mm Impact Test Facility (ITF), located at TA-55, to understand the strength behavior of Beryllium metal at elevated temperature and pressure. In FY 2017, three experiments were conducted as a part of this project.

  18. An investigation of process sensitivity for electron beam evaporation of beryllium

    International Nuclear Information System (INIS)

    Egert, C.M.; Schmoyer, D.D.; Nordin, C.W.; Berry, A.

    1991-01-01

    This paper reports on the process sensitivity of a beryllium coating process investigated using a statistical design of experiments approach. Process sensitivity is a measure of the variation in a given quality characteristic of the coating as a function of the evaporation process parameters. Manufacturing processes which maximize quality while simultaneously minimizing variability are most desirable. Three evaporation process parameters were included in this study: deposition rate, substrate temperature, and run time. A central composite experimental design employing a total of 18 coating runs was used to produce beryllium coatings on aluminum, silicon, fused silica, and beryllium substrates. The quality of the resulting coatings was characterized by scanning electron microscopy, IR spectrophotometry, stylus profilometry, and weight gain (thickness). Analysis of these results allowed the development of functional relationship between the quality characteristics (thickness, reflectance, etc.) and the evaporation process parameters. Process sensitivity for each response was then determined by calculating the gradient of each quality characteristic with respect to all three process parameters. Three dimensional plots were developed of the quality characteristic and its process sensitivity as a function of process parameters. Both quality characteristic and process sensitivity plots will be presented and discussed. For many of the quality characteristics, temperature during deposition was found to be the most sensitive process parameter for the beryllium c-beam evaporation process

  19. Brazing process for beryllium pieces at a temperature lower than 800 deg

    International Nuclear Information System (INIS)

    Cerutti, R.; Flegeau, G.; Haas, C.

    1992-01-01

    This patent describes a brazing process of a beryllium plate with monel rod at a temperature lower than 1073 K under high vacuum. The brazing alloy is composed (% in weight) of silver (58 to 65%), copper (26 to 29%), indium (9 to 13%) and titanium (0 to 2%). (A.B.). 4 refs., 1 fig

  20. Optimized beryllium target design for indirectly driven inertial confinement fusion experiments on the National Ignition Facility

    Energy Technology Data Exchange (ETDEWEB)

    Simakov, Andrei N., E-mail: simakov@lanl.gov; Wilson, Douglas C.; Yi, Sunghwan A.; Kline, John L.; Batha, Steven H. [Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, New Mexico 87545 (United States); Clark, Daniel S.; Milovich, Jose L.; Salmonson, Jay D. [Lawrence Livermore National Laboratory, P.O. Box 808, Livermore, California 94551 (United States)

    2014-02-15

    For indirect drive inertial confinement fusion, Beryllium (Be) ablators offer a number of important advantages as compared with other ablator materials, e.g., plastic and high density carbon. In particular, the low opacity and relatively high density of Be lead to higher rocket efficiencies giving a higher fuel implosion velocity for a given X-ray drive; and to higher ablation velocities providing more ablative stabilization and reducing the effect of hydrodynamic instabilities on the implosion performance. Be ablator advantages provide a larger target design optimization space and can significantly improve the National Ignition Facility (NIF) [J. D. Lindl et al., Phys. Plasmas 11, 339 (2004)] ignition margin. Herein, we summarize the Be advantages, briefly review NIF Be target history, and present a modern, optimized, low adiabat, Revision 6 NIF Be target design. This design takes advantage of knowledge gained from recent NIF experiments, including more realistic levels of laser-plasma energy backscatter, degraded hohlraum-capsule coupling, and the presence of cross-beam energy transfer.

  1. Effect of some types of machining processes on beryllium fatigue strength properties

    International Nuclear Information System (INIS)

    Armbruster, M.

    1977-01-01

    The aim of this work, which is sponsored by the French D.G.R.S.T., is to determine a machining process giving both the highest tensile strength and the highest fatigue limit to beryllium parts. A comparison is made of the effects of : mechanical machining, electro discharge machining, electro-chemical machining, electrolytical and chemical polishing, glass shot peening, on the mechanical strength and fatigue limits of samples taken from hot pressed and extruded rods and from cast ingot sheets, either notched or not as required. Complementary examinations are performed principally by fractographic study. The results show that for beryllium, electrochemical machining followed by glass shot peening gives the best results; however mechanical machining with electrolytical polishing followed by glass shot peening are also satisfactory. (author)

  2. Technical Basis for PNNL Beryllium Inventory

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, Michelle Lynn

    2014-07-09

    The Department of Energy (DOE) issued Title 10 of the Code of Federal Regulations Part 850, “Chronic Beryllium Disease Prevention Program” (the Beryllium Rule) in 1999 and required full compliance by no later than January 7, 2002. The Beryllium Rule requires the development of a baseline beryllium inventory of the locations of beryllium operations and other locations of potential beryllium contamination at DOE facilities. The baseline beryllium inventory is also required to identify workers exposed or potentially exposed to beryllium at those locations. Prior to DOE issuing 10 CFR 850, Pacific Northwest Nuclear Laboratory (PNNL) had documented the beryllium characterization and worker exposure potential for multiple facilities in compliance with DOE’s 1997 Notice 440.1, “Interim Chronic Beryllium Disease.” After DOE’s issuance of 10 CFR 850, PNNL developed an implementation plan to be compliant by 2002. In 2014, an internal self-assessment (ITS #E-00748) of PNNL’s Chronic Beryllium Disease Prevention Program (CBDPP) identified several deficiencies. One deficiency is that the technical basis for establishing the baseline beryllium inventory when the Beryllium Rule was implemented was either not documented or not retrievable. In addition, the beryllium inventory itself had not been adequately documented and maintained since PNNL established its own CBDPP, separate from Hanford Site’s program. This document reconstructs PNNL’s baseline beryllium inventory as it would have existed when it achieved compliance with the Beryllium Rule in 2001 and provides the technical basis for the baseline beryllium inventory.

  3. Exhaust gas processing facility

    International Nuclear Information System (INIS)

    Terada, Shin-ichi.

    1995-01-01

    The facility of the present invention comprises a radioactive liquid storage vessel, an exhaust gas dehumidifying device for dehumidifying gases exhausted from the vessel and an exhaust gas processing device for reducing radioactive materials in the exhaust gases. A purified gas line is disposed to the radioactive liquid storage vessel for purging exhaust gases generated from the radioactive liquid, then dehumidified and condensed liquid is recovered, and exhaust gases are discharged through an exhaust gas pipe disposed downstream of the exhaust gas processing device. With such procedures, the scale of the exhaust gas processing facility can be reduced and exhaust gases can be processed efficiently. (T.M.)

  4. Measurements of beryllium sputtering yields at JET

    Science.gov (United States)

    Jet-Efda Contributors Stamp, M. F.; Krieger, K.; Brezinsek, S.

    2011-08-01

    The lifetime of the beryllium first wall in ITER will depend on erosion and redeposition processes. The physical sputtering yields for beryllium (both deuterium on beryllium (Be) and Be on Be) are of crucial importance since they drive the erosion process. Literature values of experimental sputtering yields show an order of magnitude variation so predictive modelling of ITER wall lifetimes has large uncertainty. We have reviewed the old beryllium yield experiments on JET and used current beryllium atomic data to produce revised beryllium sputtering yields. These experimental measurements have been compared with a simple physical sputtering model based on TRIM.SP beryllium yield data. Fair agreement is seen for beryllium yields from a clean beryllium limiter. However the yield on a beryllium divertor tile (with C/Be co-deposits) shows poor agreement at low electron temperatures indicating that the effect of the higher sputtering threshold for beryllium carbide is important.

  5. METHOD OF BRAZING BERYLLIUM

    Science.gov (United States)

    Hanks, G.S.; Keil, R.W.

    1963-05-21

    A process is described for brazing beryllium metal parts by coating the beryllium with silver (65- 75 wt%)-aluminum alloy using a lithium fluoride (50 wt%)-lithium chloride flux, and heating the coated joint to a temperature of about 700 un. Concent 85% C for about 10 minutes. (AEC)

  6. Preparation of beryllium hydride

    International Nuclear Information System (INIS)

    Roberts, C.B.

    1975-01-01

    A process is described for preparing beryllium hydride by the direct reaction of beryllium borohydride and aluminum hydride trimethylamine adduct. Volatile by-products and unreacted reactants are readily removed from the product mass by sublimation and/or evaporation. (U.S.)

  7. Effect of transient heating loads on beryllium

    International Nuclear Information System (INIS)

    Kupriyanov, Igor B.; Porezanov, Nicolay P.; Nikolaev, Georgyi N.; Kurbatova, Liudmila A.; Podkovyrov, Vyacheslav L.; Muzichenko, Anatoliy D.; Zhitlukhin, Anatoliy M.; Khimchenko, Leonid N.; Gervash, Alexander A.

    2014-01-01

    Highlights: • We study the effect of transient plasma loads on beryllium erosion and surface microstructure. • Beryllium targets were irradiated by plasma streams with energy of 0.5–1 MJ/m 2 at ∼250 °C. • Under plasma loads 0.5–1 MJ/m 2 cracking of beryllium surface is rather slight. • Under 0.5 MJ/m 2 the mass loss of Be is no more than 0.2 g/m 2 shot and decreasing with shots number. • Under 1 MJ/m 2 maximum mass loss of beryllium was 3.7 g/m 2 shot and decreasing with shots number. - Abstract: Beryllium will be used as a plasma facing material for ITER first wall. It is expected that erosion of beryllium under transient plasma loads such as the edge-localized modes (ELMs) and disruptions will mainly determine a lifetime of ITER first wall. The results of recent experiments with the Russian beryllium of TGP-56FW ITER grade on QSPA-Be plasma gun facility are presented. The Be/CuCrZr mock-ups were exposed to upto 100 shots by deuterium plasma streams with pulse duration of 0.5 ms at ∼250 °C and average heat loads of 0.5 and 1 MJ/m 2 . Experiments were performed at 250 °C. The evolution of surface microstructure and cracks morphology as well as beryllium mass loss are investigated under erosion process

  8. Beryllium production using beryllium fluoride

    International Nuclear Information System (INIS)

    Hubler, Carlos Henrique

    1993-01-01

    This work presents the beryllium production by thermal decomposition of the ammonium beryllium fluoride, followed by magnesium reduction, obtained in the small pilot plant of the Brazilian National Nuclear Energy Commission - Nuclear Engineering Institute

  9. Capsule physics comparison of National Ignition Facility implosion designs using plastic, high density carbon, and beryllium ablators

    Science.gov (United States)

    Clark, D. S.; Kritcher, A. L.; Yi, S. A.; Zylstra, A. B.; Haan, S. W.; Weber, C. R.

    2018-03-01

    Indirect drive implosion experiments on the National Ignition Facility (NIF) [E. I. Moses et al., Phys. Plasmas 16, 041006 (2009)] have now tested three different ablator materials: glow discharge polymer plastic, high density carbon, and beryllium. How do these different ablators compare in current and proposed implosion experiments on NIF? What are the relative advantages and disadvantages of each? This paper compares these different ablator options in capsule-only simulations of current NIF experiments and potential future designs. The simulations compare the impact of the capsule fill tube, support tent, and interface surface roughness for each case, as well as all perturbations in combination. According to the simulations, each ablator is impacted by the various perturbation sources differently, and each material poses unique challenges in the pursuit of ignition on NIF.

  10. Preparation of beryllium hydride

    International Nuclear Information System (INIS)

    Lowrance, B.R.

    1975-01-01

    A process is described for the preparation of beryllium hydride which comprises pyrolyzing, while in solution in a solvent inert under the reaction conditions, with respect to reactants and products and at a temperature in the range of about 100 0 to about 200 0 C, sufficient to result in the formation of beryllium hydride, a di-t-alkyl beryllium etherate wherein each tertiary alkyl radical contains from 4 to 20 carbon atoms. The pyrolysis is carried out under an atmosphere inert under the reaction conditions, with respect to reactants and products. (U.S.)

  11. Corrosion of beryllium

    International Nuclear Information System (INIS)

    Mueller, J.J.; Adolphson, D.R.

    1987-01-01

    The corrosion behavior of beryllium in aqueous and elevated-temperature oxidizing environments has been extensively studied for early-intended use of beryllium in nuclear reactors and in jet and rocket propulsion systems. Since that time, beryllium has been used as a structural material in les corrosive environments. Its primary applications include gyro systems, mirror and reentry vehicle structures, and aircraft brakes. Only a small amount of information has been published that is directly related to the evaluation of beryllium for service in the less severe or normal atmospheric environments associated with these applications. Despite the lack of published data on the corrosion of beryllium in atmospheric environments, much can be deduced about its corrosion behavior from studies of aqueous corrosion and the experiences of fabricators and users in applying, handling, processing, storing, and shipping beryllium components. The methods of corrosion protection implemented to resist water and high-temperature gaseous environments provide useful information on methods that can be applied to protect beryllium for service in future long-term structural applications

  12. Beryllium allergy

    International Nuclear Information System (INIS)

    Schoenherr, S.; Pevny, I.

    1989-12-01

    Beryllium is not only a high potent allergen, but also a fotoallergen and can provoke contact allergic reactions, fotoallergic reactions, granulomatous skin reactions, pulmonary granulomatous diseases and sometimes even systemic diseases. The authors present 9 own cases of a patch test positive beryllium allergy, 7 patients with relevant allergy and 5 patients with an allergic contact stomatitis. (author)

  13. Gaseous waste processing facility

    International Nuclear Information System (INIS)

    Konno, Masanobu; Uchiyama, Yoshio; Suzuki, Kunihiko; Kimura, Masahiro; Kawabe, Ken-ichi.

    1992-01-01

    Gaseous waste recombiners 'A' and 'B' are connected in series and three-way valves are disposed at the upstream and the downstream of the recombiners A and B, and bypass lines are disposed to the recombiners A and B, respectively. An opening/closing controller for the three-way valves is interlocked with a hydrogen densitometer disposed to a hydrogen injection line. Hydrogen gas and oxygen gas generated by radiolysis in the reactor are extracted from a main condenser and caused to flow into a gaseous waste processing system. Gaseous wastes are introduced together with overheated steams to the recombiner A upon injection of hydrogen. Both of the bypass lines of the recombiners A and B are closed, and recombining reaction for the increased hydrogen gas is processed by the recombiners A and B connected in series. In an operation mode not conducting hydrogen injection, it is passed through the bypass line of the recombiner A and processed by the recombiner B. With such procedures, the increase of gaseous wastes due to hydrogen injection can be coped with existent facilities. (I.N.)

  14. Poultry Slaughtering and Processing Facilities

    Data.gov (United States)

    Department of Homeland Security — Agriculture Production Poultry Slaughtering and Processing in the United States This dataset consists of facilities which engage in slaughtering, processing, and/or...

  15. Thermal effects on beryllium mirrors

    International Nuclear Information System (INIS)

    Weinswig, S.

    1989-01-01

    Beryllium is probably the most frequently used material for spaceborne system scan mirrors. Beryllium's properties include lightweightedness, high Young's modulus, high stiffness value, high resonance value. As an optical surface, beryllium is usually nickel plated in order to produce a higher quality surface. This process leads to the beryllium mirror acting like a bimetallic device. The mirror's deformation due to the bimetallic property can possibly degrade the performance of the associated optical system. As large space borne systems are designed and as temperature considerations become more crucial in the instruments, the concern about temporal deformation of the scan mirrors becomes a prime consideration. Therefore, two sets of tests have been conducted in order to ascertain the thermal effects on nickel plated beryllium mirrors. These tests are categorized. The purpose of this paper is to present the values of the bimetallic effect on typical nickel plated beryllium mirrors

  16. Hybrid Orbital and Numerical Grid Representationfor Electronic Continuum Processes: Double Photoionization of Atomic Beryllium

    Energy Technology Data Exchange (ETDEWEB)

    Yip, Frank L; McCurdy, C. William; Rescigno, Thomas N

    2010-04-19

    A general approach for ab initio calculations of electronic continuum processes is described in which the many-electron wave function is expanded using a combination of orbitals at short range and the finite-element discrete variable representation(FEM-DVR) at larger distances. The orbital portion of the basis allows the efficient construction of many-electron configurations in which some of the electrons are bound, but because the orbitals are constructed from an underlying FEM-DVR grid, the calculation of two-electron integrals retains the efficiency of the primitive FEM-DVR approach. As an example, double photoionization of beryllium is treated in a calculation in which the 1s{sup 2} core is frozen. This approach extends the use of exterior complex scaling (ECS) successfully applied to helium and H{sub 2} to calculations with two active electrons on more complicated targets. Integrated, energy-differential and triply-differential cross sections are exhibited, and the results agree well with other theoretical investigations.

  17. Chemistry characterization and samples beryllium process impurity determination; Caracterizacao quimica e determinacao de impurezas de amostras de processo de berilio

    Energy Technology Data Exchange (ETDEWEB)

    Carvalho, Leonel Mathry de

    1992-12-01

    Brazil is the greatest world producer of beryl (3 Be O.Al{sub 2}O{sub 3}.6SiO{sub 2}) and has recently begun to produce beryllium compounds by means of a pilot plant constructed at Governador Valadares city (Minas Gerais - Brazil). The aim of this work was the determination of trace level impurities and macro constituents in the tenth % range to support analytical control process of plant production and characterization of beryllium compounds. The impurities separations and purification process was developed by two steps procedure. The first one using EDTA complexation has separated and reduced some impurities to less than 1 {mu}g/ml level. In the second one it was used a chelating resin (Chelex 100) and the separation efficiency was about 75 to 97 % related with the element tested. High pure berylium oxide standard was obtained from purification of Be(OH)2. The R X fluorescence presented only traces of Cu and Si < 1 % Fe and Mn, Zn, Ca, Al, Na and S were completely removed. The beryllium content was determined by direct atomic emission spectroscopy in argon plasma (Dcp) and compared with classic gravimetric method as Be O. The results were in agreement (49,2+/-0,2 % and 48,3+/-0,1 % respectively) between 95 % of confidence. A low temperature gravimetric method for beryllium determination was also studied using Oxine with microwave furnace. A total of 24 elements including macro and trace level were determined by Dcp and/or spectrophotometric methods. The Be/B separation was studied using anionic resin in poly alcohols medium. A more detailed study of equilibrium conditions is necessary. This work was realized at Laboratorio de Analise Mineral (LAM) of Comissao Nacional de Energia Nuclear - Rio de Janeiro (CNEN). (author)

  18. Beryllium. Health hazards and their control. Pt. 2

    International Nuclear Information System (INIS)

    Lires, O.A.; Delfino, C.A.; Botbol, J.

    1991-01-01

    In this work (continuation of 'Beryllium' series) health hazards, toxic effects, limits of permissible atmospheric contamination and safe exposure to beryllium are described. Guidelines to the design, control operations and hygienic precautions of the working facilities are given. (Author) [es

  19. Production of beryllium oxide of nuclear purity from beryl

    Energy Technology Data Exchange (ETDEWEB)

    Copat, A; Sood, S P

    1984-01-01

    Production of beryllium oxide from beryl by the fluoride process was optimized in this study. Optimum results were obtained using a mixture of sodium hexafluorsilicate and sodium hexafluorferrate as flux and calcinating at 740/sup 0/C for 2 hours. The beryllium concentrate produced was further purified by crystallization as beryllium sulfate to obtain nuclear grade beryllium oxide

  20. Production of beryllium oxide of nuclear purity from beryl

    International Nuclear Information System (INIS)

    Copat, A.; Sood, S.P.

    1983-01-01

    Production of beryllium oxide from beryl by the fluoride process was optimized in this study. Optimum results were obtained using a mixture of sodium hexafluorsilicate and sodium hexafluorferrate as flux and calcinating at 740 0 C for 2 hours. The beryllium concentrate produced was further purified by crystallization as beryllium sulfate to obtain nuclear grade beryllium oxide (Author) [pt

  1. Sintering of beryllium oxide

    International Nuclear Information System (INIS)

    Caillat, R.; Pointud, R.

    1955-01-01

    This study had for origin to find a process permitting to manufacture bricks of beryllium oxide of pure nuclear grade, with a density as elevated as possible and with standardized shape. The sintering under load was the technique kept for the manufacture of the bricks. Because of the important toxicity of the beryllium oxide, the general features for the preliminary study of the sintering, have been determined while using alumina. The obtained results will be able to act as general indication for ulterior studies with sintering under load. (M.B.) [fr

  2. Beryllium poisonings

    International Nuclear Information System (INIS)

    Alibert, S.

    1959-03-01

    This note reports a bibliographical study of beryllium toxicity. Thus, this bibliographical review addresses and outlines aspects and issues like aetiology, cases of acute poisoning (cutaneous manifestations, pulmonary manifestations), chronic poisoning (cutaneous, pulmonary and bone manifestations), excretion and localisation, and prognosis

  3. Industrial Implementation of Environmentally Friendly Nanometal Electroplating Process for Chromium and Copper Beryllium Replacement using Low Cost Pulse Current Power Supplies

    Science.gov (United States)

    2014-09-10

    FINAL REPORT Industrial Implementation of Environmentally Friendly Nanometal Electroplating Process for Chromium and Copper Beryllium...35 3.2 Phase II – Development/Verification that Nanotechnology Based Electroplating Process to Replace EHC/Cu-Be Processes are Compatible With...36 3.3 Phase III – Development of 200kW Power Supply and Compatible Nanostructured Electroplating Processed for Commercialization

  4. Advanced Polymer Processing Facility

    Energy Technology Data Exchange (ETDEWEB)

    Muenchausen, Ross E. [Los Alamos National Laboratory

    2012-07-25

    Some conclusions of this presentation are: (1) Radiation-assisted nanotechnology applications will continue to grow; (2) The APPF will provide a unique focus for radiolytic processing of nanomaterials in support of DOE-DP, other DOE and advanced manufacturing initiatives; (3) {gamma}, X-ray, e-beam and ion beam processing will increasingly be applied for 'green' manufacturing of nanomaterials and nanocomposites; and (4) Biomedical science and engineering may ultimately be the biggest application area for radiation-assisted nanotechnology development.

  5. Reactivity test between beryllium and copper

    International Nuclear Information System (INIS)

    Kawamura, H.; Kato, M.

    1995-01-01

    Beryllium has been expected for using as plasma facing material on ITER. And, copper alloy has been proposed as heat sink material behind plasma facing components. Therefore, both materials must be joined. However, the elementary process of reaction between beryllium and copper alloy does not clear in detail. For example, other authors reported that beryllium reacted with copper at high temperature, but it was not obvious about the generation of reaction products and increasing of the reaction layer. In the present work, from this point, for clarifying the elementary process of reaction between beryllium and copper, the out-of-pile compatibility tests were conducted with diffusion couples of beryllium and copper which were inserted in the capsule filled with high purity helium gas (6N). Annealing temperatures were 300, 400, 500, 600 and 700 degrees C, and annealing periods were 100, 300 and 1000h. Beryllium specimens were hot pressed beryllium, and copper specimens were OFC (Oxygen Free Copper)

  6. Design of plutonium processing facilities

    International Nuclear Information System (INIS)

    Derbyshire, W.; Sills, R.J.

    1982-01-01

    Five considerations for the design of plutonium processing facilities are identified. These are: Toxicity, Radiation, Criticality, Containment and Remote Operation. They are examined with reference to reprocessing spent nuclear fuel and application is detailed both for liquid and dry processes. (author)

  7. Research of beryllium safety issues

    International Nuclear Information System (INIS)

    Longhurst, G.R.; Anderl, R.A.; Dolan, T.J.; Hankins, M.R.; Pawelko, R.J.

    1993-01-01

    Beryllium has been identified as a leading contender for the plasma-facing material in ITER. Its use has some obvious advantages, but there are also a number of safety concerns associated with it. The Idaho National Engineering Laboratory (INEL) has undertaken a number of studies to help resolve some of these issues. One issue is the response of beryllium to neutron irradiation. We have tested samples irradiated in the Advanced Test Reactor (ATR) and are currently preparing to make measurements of the change in mechanical properties of beryllium samples irradiated at elevated temperatures in the Fast Flux Test Facility (FFTF) and the Experimental Breeder Reactor II (EBR-II) at the INEL. Mechanical tests will be conducted at the irradiation temperatures of 375-550 C. Other experiments address permeation and retention of implanted tritium in plasma-sprayed beryllium. In one test the porosity of the material allowed 0.12% of implanted ions and 0.17% of atoms from background gas pressure to pass through the foil with essentially no delay. For comparison, similar tests on fully dense hot-rolled, vacuum melted or sintered powder foils of high purity beryllium showed only 0.001% of implanting ions to pass through the foil, and then only after a delay of several hours. None of the molecular gas appeared to permeate these latter targets. An implication is that plasma-sprayed beryllium may substantially enhance recycling of tritium to the plasma provided it is affixed to a relatively impermeable substrate. (orig.)

  8. The INEL beryllium multiplication experiment

    International Nuclear Information System (INIS)

    Smith, J.R.; King, J.J.

    1991-03-01

    The experiment to measure the multiplication of 14-MeV neutrons in bulk beryllium has been completed. The experiment consists of determining the ratio of 56 Mn activities induced in a large manganese bath by a central 14-MeV neutron source, with and without a beryllium sample surrounding the source. In the manganese bath method a neutron source is placed at the center of a totally-absorbing aqueous solution of MnSo 4 . The capture of neutrons by Mn produces a 56 Mn activity proportional to the emission rate of the source. As applied to the measurement of the multiplication of 14- MeV neutrons in bulk beryllium, the neutron source is a tritium target placed at the end of the drift tube of a small deuteron accelerator. Surrounding the source is a sample chamber. When the sample chamber is empty, the neutrons go directly to the surrounding MnSO 4 solution, and produce a 56 Mn activity proportional to the neutron emission rate. When the chamber contains a beryllium sample, the neutrons first enter the beryllium and multiply through the (n,2n) process. Neutrons escaping from the beryllium enter the bath and produce a 56 Mn activity proportional to the neutron emission rate multiplied by the effective value of the multiplication in bulk beryllium. The ratio of the activities with and without the sample present is proportional to the multiplication value. Detailed calculations of the multiplication and all the systematic effects were made with the Monte Carlo program MCNP, utilizing both the Young and Stewart and the ENDF/B-VI evaluations for beryllium. Both data sets produce multiplication values that are in excellent agreement with the measurements for both raw and corrected values of the multiplication. We conclude that there is not real discrepancy between experimental and calculated values for the multiplication of neutrons in bulk beryllium. 12 figs., 11 tabs., 18 refs

  9. Method for welding beryllium

    Science.gov (United States)

    Dixon, R.D.; Smith, F.M.; O`Leary, R.F.

    1997-04-01

    A method is provided for joining beryllium pieces which comprises: depositing aluminum alloy on at least one beryllium surface; contacting that beryllium surface with at least one other beryllium surface; and welding the aluminum alloy coated beryllium surfaces together. The aluminum alloy may be deposited on the beryllium using gas metal arc welding. The aluminum alloy coated beryllium surfaces may be subjected to elevated temperatures and pressures to reduce porosity before welding the pieces together. The aluminum alloy coated beryllium surfaces may be machined into a desired welding joint configuration before welding. The beryllium may be an alloy of beryllium or a beryllium compound. The aluminum alloy may comprise aluminum and silicon. 9 figs.

  10. Method for welding beryllium

    International Nuclear Information System (INIS)

    Dixon, R.D.; Smith, F.M.; O'Leary, R.F.

    1997-01-01

    A method is provided for joining beryllium pieces which comprises: depositing aluminum alloy on at least one beryllium surface; contacting that beryllium surface with at least one other beryllium surface; and welding the aluminum alloy coated beryllium surfaces together. The aluminum alloy may be deposited on the beryllium using gas metal arc welding. The aluminum alloy coated beryllium surfaces may be subjected to elevated temperatures and pressures to reduce porosity before welding the pieces together. The aluminum alloy coated beryllium surfaces may be machined into a desired welding joint configuration before welding. The beryllium may be an alloy of beryllium or a beryllium compound. The aluminum alloy may comprise aluminum and silicon. 9 figs

  11. Erosion of beryllium under ITER - Relevant transient plasma loads

    Science.gov (United States)

    Kupriyanov, I. B.; Nikolaev, G. N.; Kurbatova, L. A.; Porezanov, N. P.; Podkovyrov, V. L.; Muzichenko, A. D.; Zhitlukhin, A. M.; Gervash, A. A.; Safronov, V. M.

    2015-08-01

    Beryllium will be used as a armor material for the ITER first wall. It is expected that erosion of beryllium under transient plasma loads such as the edge-localized modes (ELMs) and disruptions will mainly determine a lifetime of the ITER first wall. This paper presents the results of recent experiments with the Russian beryllium of TGP-56FW ITER grade on QSPA-Be plasma gun facility. The Be/CuCrZr mock-ups were exposed to up to 100 shots by deuterium plasma streams (5 cm in diameter) with pulse duration of 0.5 ms and heat loads range of 0.2-0.5 MJ/m2 at different temperature of beryllium tiles. The temperature of Be tiles has been maintained about 250 and 500 °C during the experiments. After 10, 40 and 100 shots, the beryllium mass loss/gain under erosion process were investigated as well as evolution of surface microstructure and cracks morphology.

  12. Development of intergranular thermal residual stresses in beryllium during cooling from processing temperatures

    Energy Technology Data Exchange (ETDEWEB)

    Brown, D.W. [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States)], E-mail: dbrown@lanl.gov; Sisneros, T.A.; Clausen, B.; Abeln, S.; Bourke, M.A.M.; Smith, B.G.; Steinzig, M.L.; Tome, C.N.; Vogel, S.C. [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States)

    2009-02-15

    The intergranular thermal residual stresses in texture-free solid polycrystalline beryllium were determined by comparison of crystallographic lattice parameters in solid and powder samples measured by neutron diffraction during cooling from 800 deg. C. The internal stresses are not significantly different from zero >575 deg. C and increase nearly linearly <525 deg. C. At room temperature, the c axis of an average grain is under {approx}200 MPa of compressive internal stress, and the a axis is under 100 MPa of tensile stress. For comparison, the stresses have also been calculated using an Eshelby-type polycrystalline model. The measurements and calculations agree very well when temperature dependence of elastic constants is accounted for, and no plastic relaxation is allowed in the model.

  13. Worker Environment Beryllium Characterization Study

    International Nuclear Information System (INIS)

    2009-01-01

    This report summarizes the conclusion of regular monitoring of occupied buildings at the Nevada Test Site and North Las Vegas facility to determine the extent of beryllium (Be) contamination in accordance with Judgment of Needs 6 of the August 14, 2003, 'Minnema Report.'

  14. Worker Environment Beryllium Characterization Study

    Energy Technology Data Exchange (ETDEWEB)

    NSTec Environment, Safety, Health & Quality

    2009-12-28

    This report summarizes the conclusion of regular monitoring of occupied buildings at the Nevada Test Site and North Las Vegas facility to determine the extent of beryllium (Be) contamination in accordance with Judgment of Needs 6 of the August 14, 2003, “Minnema Report.”

  15. Beryllium electrodeposition on aluminium cathode from chloride melts

    International Nuclear Information System (INIS)

    Nichkov, I.F.; Novikov, E.A.; Serebryakov, G.A.; Kanashin, Yu.P.; Sardyko, G.N.

    1980-01-01

    Cathodic processes during beryllium deposition on liquid and solid aluminium cathodes are investigated. Mixture of sodium, potassium and beryllium chloride melts served as an lectrolyte. Beryllium ion discharge at the expense of alloy formation takes place at more positive potentials than on an indifferent cathode at low current densities ( in the case of liquid aluminium cathode). Metallographic analysis and measurements of microhardness have shown, that the cathodic product includes two phases: beryllium solid solution in aluminium and metallic beryllium. It is concluded, that aluminium-beryllium alloys with high cathodic yield by current can be obtained by the electrolytic method

  16. Sanitary-hygienic and ecological aspects of beryllium production

    Energy Technology Data Exchange (ETDEWEB)

    Dvinskykh, E.M.; Savchuk, V.V.; Sidorov, V.L.; Slobodin, D.B.; Tuzov, Y.V. [Ulba Metallurgical Plant, Ust-Kamenogorsk (Kazakhstan)

    1998-01-01

    The Report describes An organization of sanitary-hygienic and ecological control of beryllium production at Ulba metallurgical plant. It involves: (1) the consideration of main methods for protection of beryllium production personnel from unhealthy effect of beryllium, (2) main kinds of filters, used in gas purification systems at different process areas, (3) data on beryllium monitoring in water, soil, on equipment. This Report also outlines problems connected with designing devices for a rapid analysis of beryllium in air as well as problems of beryllium production on ecological situation in the town. (author)

  17. Erosion of beryllium under ITER – Relevant transient plasma loads

    International Nuclear Information System (INIS)

    Kupriyanov, I.B.; Nikolaev, G.N.; Kurbatova, L.A.; Porezanov, N.P.; Podkovyrov, V.L.; Muzichenko, A.D.; Zhitlukhin, A.M.; Gervash, A.A.; Safronov, V.M.

    2015-01-01

    Highlights: • We study the erosion, mass loss/gain and surface structure evolution of Be/CuCrZr mock-ups, armored with beryllium of TGP-56FW grade after irradiation by deuterium plasma heat load of 0.5 MJ/m 2 at 250 °C and 500 °C. • Beryllium mass loss/erosion under plasma heat load at 250 °C is rather small (no more than 0.2 g/m 2 shot and 0.11 μm/shot, correspondingly, after 40 shots) and tends to decrease with increasing number of shots. • Beryllium mass loss/erosion under plasma heat load at 500 °C is much higher (∼2.3 g/m 2 shot and 1.2 μm/shot, correspondingly, after 10 shot) and tends to decrease with increasing the number of shots (∼0.26 g/m 2 pulse and 0.14 μm/shot, correspondingly, after 100 shot). • Beryllium erosion value derived from the measurements of profile of irradiated surface is much higher than erosion value derived from mass loss data. - Abstract: Beryllium will be used as a armor material for the ITER first wall. It is expected that erosion of beryllium under transient plasma loads such as the edge-localized modes (ELMs) and disruptions will mainly determine a lifetime of the ITER first wall. This paper presents the results of recent experiments with the Russian beryllium of TGP-56FW ITER grade on QSPA-Be plasma gun facility. The Be/CuCrZr mock-ups were exposed to up to 100 shots by deuterium plasma streams (5 cm in diameter) with pulse duration of 0.5 ms and heat loads range of 0.2–0.5 MJ/m 2 at different temperature of beryllium tiles. The temperature of Be tiles has been maintained about 250 and 500 °C during the experiments. After 10, 40 and 100 shots, the beryllium mass loss/gain under erosion process were investigated as well as evolution of surface microstructure and cracks morphology

  18. Erosion of beryllium under ITER – Relevant transient plasma loads

    Energy Technology Data Exchange (ETDEWEB)

    Kupriyanov, I.B., E-mail: igkupr@gmail.com [A.A. Bochvar High Technology Research Institute of Inorganic Materials, Rogova St. 5a, 123060 Moscow (Russian Federation); Nikolaev, G.N.; Kurbatova, L.A.; Porezanov, N.P. [A.A. Bochvar High Technology Research Institute of Inorganic Materials, Rogova St. 5a, 123060 Moscow (Russian Federation); Podkovyrov, V.L.; Muzichenko, A.D.; Zhitlukhin, A.M. [TRINITI, Troitsk, Moscow reg. (Russian Federation); Gervash, A.A. [Efremov Research Institute, S-Peterburg (Russian Federation); Safronov, V.M. [Project Center of ITER, Moscow (Russian Federation)

    2015-08-15

    Highlights: • We study the erosion, mass loss/gain and surface structure evolution of Be/CuCrZr mock-ups, armored with beryllium of TGP-56FW grade after irradiation by deuterium plasma heat load of 0.5 MJ/m{sup 2} at 250 °C and 500 °C. • Beryllium mass loss/erosion under plasma heat load at 250 °C is rather small (no more than 0.2 g/m{sup 2} shot and 0.11 μm/shot, correspondingly, after 40 shots) and tends to decrease with increasing number of shots. • Beryllium mass loss/erosion under plasma heat load at 500 °C is much higher (∼2.3 g/m{sup 2} shot and 1.2 μm/shot, correspondingly, after 10 shot) and tends to decrease with increasing the number of shots (∼0.26 g/m{sup 2} pulse and 0.14 μm/shot, correspondingly, after 100 shot). • Beryllium erosion value derived from the measurements of profile of irradiated surface is much higher than erosion value derived from mass loss data. - Abstract: Beryllium will be used as a armor material for the ITER first wall. It is expected that erosion of beryllium under transient plasma loads such as the edge-localized modes (ELMs) and disruptions will mainly determine a lifetime of the ITER first wall. This paper presents the results of recent experiments with the Russian beryllium of TGP-56FW ITER grade on QSPA-Be plasma gun facility. The Be/CuCrZr mock-ups were exposed to up to 100 shots by deuterium plasma streams (5 cm in diameter) with pulse duration of 0.5 ms and heat loads range of 0.2–0.5 MJ/m{sup 2} at different temperature of beryllium tiles. The temperature of Be tiles has been maintained about 250 and 500 °C during the experiments. After 10, 40 and 100 shots, the beryllium mass loss/gain under erosion process were investigated as well as evolution of surface microstructure and cracks morphology.

  19. BERYLLIUM MEASUREMENT IN COMMERCIALLY AVAILABLE WET WIPES

    Energy Technology Data Exchange (ETDEWEB)

    Youmans-Mcdonald, L.

    2011-02-18

    Analysis for beryllium by fluorescence is now an established method which is used in many government-run laboratories and commercial facilities. This study investigates the use of this technique using commercially available wet wipes. The fluorescence method is widely documented and has been approved as a standard test method by ASTM International and the National Institute for Occupational Safety and Health (NIOSH). The procedure involves dissolution of samples in aqueous ammonium bifluoride solution and then adding a small aliquot to a basic hydroxybenzoquinoline sulfonate fluorescent dye (Berylliant{trademark} Inc. Detection Solution Part No. CH-2) , and measuring the fluorescence. This method is specific to beryllium. This work explores the use of three different commercial wipes spiked with beryllium, as beryllium acetate or as beryllium oxide and subsequent analysis by optical fluorescence. The effect of possible interfering metals such as Fe, Ti and Pu in the wipe medium is also examined.

  20. Beryllium Measurement In Commercially Available Wet Wipes

    International Nuclear Information System (INIS)

    Youmans-Mcdonald, L.

    2011-01-01

    Analysis for beryllium by fluorescence is now an established method which is used in many government-run laboratories and commercial facilities. This study investigates the use of this technique using commercially available wet wipes. The fluorescence method is widely documented and has been approved as a standard test method by ASTM International and the National Institute for Occupational Safety and Health (NIOSH). The procedure involves dissolution of samples in aqueous ammonium bifluoride solution and then adding a small aliquot to a basic hydroxybenzoquinoline sulfonate fluorescent dye (Berylliant(trademark) Inc. Detection Solution Part No. CH-2) , and measuring the fluorescence. This method is specific to beryllium. This work explores the use of three different commercial wipes spiked with beryllium, as beryllium acetate or as beryllium oxide and subsequent analysis by optical fluorescence. The effect of possible interfering metals such as Fe, Ti and Pu in the wipe medium is also examined.

  1. Effect of machining damage on tensile properties of beryllium

    International Nuclear Information System (INIS)

    Hanafee, J.E.

    1976-01-01

    It is well established that damage introduced at the surface of beryllium during machining operations can lower its mechanical properties. Tensile tests were conducted to illustrate this on beryllium presently being used for parts in the W79 program and similar to the new powder-processed beryllium specified for production (tentative specification MEL 76-001319). The objective of this study is to quantitatively illuminate the importance of controlling machining damage in this particular grade of powder-processed beryllium

  2. Food irradiation: Gamma processing facilities

    Energy Technology Data Exchange (ETDEWEB)

    Kunstadt, P [MDS Nordion International, 447 March Road. Kanata, Ontario, K2K148 (Canada)

    1998-12-31

    The number of products being radiation processed is constantly increasing and today include such diverse items as medical disposable, fruits and vegetables, bulk spices, meats, sea foods and waste effluents. Not only do the products differ but also many products, even those within the same groupings, require different minimum and maximum radiation doses. These variations create many different requirements in the irradiator design. The design of Cobalt-60 radiation processing facilities is well established for a number of commercial applications. Installations in over 40 countries, with some in operation since the early 1960s, are testimony to the fact that irradiator design, manufacture, installation and operation is a well established technology. However, in order to design gamma irradiators for the preservation of foods one must recognize those parameters typical to the food irradiation process as well as those systems and methods already well established in the food industry. This paper discusses the basic design concepts for gamma food irradiators. They are most efficient when designed to handle a limited product density range at an established dose. Safety of Cobalt-60 transport, safe facility operation principles and the effect of various processing parameters on economics, will also be discussed. (Author)

  3. Food irradiation: Gamma processing facilities

    International Nuclear Information System (INIS)

    Kunstadt, P.

    1997-01-01

    The number of products being radiation processed is constantly increasing and today include such diverse items as medical disposable, fruits and vegetables, bulk spices, meats, sea foods and waste effluents. Not only do the products differ but also many products, even those within the same groupings, require different minimum and maximum radiation doses. These variations create many different requirements in the irradiator design. The design of Cobalt-60 radiation processing facilities is well established for a number of commercial applications. Installations in over 40 countries, with some in operation since the early 1960s, are testimony to the fact that irradiator design, manufacture, installation and operation is a well established technology. However, in order to design gamma irradiators for the preservation of foods one must recognize those parameters typical to the food irradiation process as well as those systems and methods already well established in the food industry. This paper discusses the basic design concepts for gamma food irradiators. They are most efficient when designed to handle a limited product density range at an established dose. Safety of Cobalt-60 transport, safe facility operation principles and the effect of various processing parameters on economics, will also be discussed. (Author)

  4. Food irradiation: Gamma processing facilities

    Energy Technology Data Exchange (ETDEWEB)

    Kunstadt, P. [MDS Nordion International, 447 March Road. Kanata, Ontario, K2K148 (Canada)

    1997-12-31

    The number of products being radiation processed is constantly increasing and today include such diverse items as medical disposable, fruits and vegetables, bulk spices, meats, sea foods and waste effluents. Not only do the products differ but also many products, even those within the same groupings, require different minimum and maximum radiation doses. These variations create many different requirements in the irradiator design. The design of Cobalt-60 radiation processing facilities is well established for a number of commercial applications. Installations in over 40 countries, with some in operation since the early 1960s, are testimony to the fact that irradiator design, manufacture, installation and operation is a well established technology. However, in order to design gamma irradiators for the preservation of foods one must recognize those parameters typical to the food irradiation process as well as those systems and methods already well established in the food industry. This paper discusses the basic design concepts for gamma food irradiators. They are most efficient when designed to handle a limited product density range at an established dose. Safety of Cobalt-60 transport, safe facility operation principles and the effect of various processing parameters on economics, will also be discussed. (Author)

  5. Defense waste processing facility precipitate hydrolysis process

    International Nuclear Information System (INIS)

    Doherty, J.P.; Eibling, R.E.; Marek, J.C.

    1986-03-01

    Sodium tetraphenylborate and sodium titanate are used to assist in the concentration of soluble radionuclide in the Savannah River Plant's high-level waste. In the Defense Waste Processing Facility, concentrated tetraphenylborate/sodium titanate slurry containing cesium-137, strontium-90 and traces of plutonium from the waste tank farm is hydrolyzed in the Salt Processing Cell forming organic and aqueous phases. The two phases are then separated and the organic phase is decontaminated for incineration outside the DWPF building. The aqueous phase, containing the radionuclides and less than 10% of the original organic, is blended with the insoluble radionuclides in the high-level waste sludge and is fed to the glass melter for vitrification into borosilicate glass. During the Savannah River Laboratory's development of this process, copper (II) was found to act as a catalyst during the hydrolysis reactions, which improved the organic removal and simplified the design of the reactor

  6. The design of a multisource americium-beryllium (Am-Be) neutron irradiation facility using MCNP for the neutronic performance calculation.

    Science.gov (United States)

    Sogbadji, R B M; Abrefah, R G; Nyarko, B J B; Akaho, E H K; Odoi, H C; Attakorah-Birinkorang, S

    2014-08-01

    The americium-beryllium neutron irradiation facility at the National Nuclear Research Institute (NNRI), Ghana, was re-designed with four 20 Ci sources using Monte Carlo N-Particle (MCNP) code to investigate the maximum amount of flux that is produced by the combined sources. The results were compared with a single source Am-Be irradiation facility. The main objective was to enable us to harness the maximum amount of flux for the optimization of neutron activation analysis and to enable smaller sample sized samples to be irradiated. Using MCNP for the design construction and neutronic performance calculation, it was realized that the single-source Am-Be design produced a thermal neutron flux of (1.8±0.0007)×10(6) n/cm(2)s and the four-source Am-Be design produced a thermal neutron flux of (5.4±0.0007)×10(6) n/cm(2)s which is a factor of 3.5 fold increase compared to the single-source Am-Be design. The criticality effective, k(eff), of the single-source and the four-source Am-Be designs were found to be 0.00115±0.0008 and 0.00143±0.0008, respectively. Copyright © 2014 Elsevier Ltd. All rights reserved.

  7. Chronic Beryllium Disease

    Science.gov (United States)

    ... who are exposed to beryllium will not experience health effects. Studies have shown that on average, 1 – 6 percent of exposed workers develop beryllium sensitization, although the rates can be ...

  8. Process improvement of reconversion facilities

    International Nuclear Information System (INIS)

    Park, J. H.; Chang, I. S.; Kim, E. H.; Kim, T. J.; Jeong, K. C.; Woo, M. S.; Hong, S. B.; Choi, J. H.; Chung, W. M.; Lee, K. I.; Hwang, D. S.; Kim, Y. W.; Kim, Y. K.; Choi, C. S.

    1993-01-01

    The project is for the development of recovery and reusing process of ammonium carbonate(AC) which is generated as a waste liquid from the reconversion facilities to reduce the manufacturing cost and the quantity of the waste liquid, and also for the development of the continuous fludized bed reaction process to promote the economics and safeties of the calcination and reduction process. In this second year report, measured the properties of AC solution and analyzed the AC concentration quantitatively. Examined the properties of AUC to investigate the properties of UO 2 powder which was converted from AUC, prepared with AC solution. Designed and installed the 2 tons-U/year pilot plant. Experimented in powder properties to set up the range of operating conditions. Modeled CFB reactor to estimate the conversion of reactor and to analyze the change of fluorine concentration to carry out the defluorination reaction. Experimented out the optimum conditions of the major operating parameters : solid circulation rate, gas velocity, solid holdup and initial inventory in cold bed to get the referential design data for hot bed. (Author)

  9. Some aspects of beryllium disposal in Kazakhstan

    International Nuclear Information System (INIS)

    Shestakov, V.; Chikhray, Y.; Shakhvorostov, Yr.

    2004-01-01

    Historically in Kazakhstan all disposals of used beryllium and beryllium wasted materials were stored and recycled at JSC ''Ulba Metallurgical Plant''. Since Ulba Metallurgical Plant (beside beryllium and tantalum production) is one of the world largest complex producers of fuel for nuclear power plants as well it has possibilities, technologies and experience in processing toxic and radioactive wastes related with those productions. At present time only one operating Kazakhstan research reactors (EWG1M in Kurchatov) contains beryllium made core. The results of current examination of that core allow using it without replacement long time yet (at least for next five-ten years). Nevertheless the problem how to utilize such irradiated beryllium becomes actual issue for Kazakhstan even today. Since Kazakhstan is the member of ITER/DEMO Reactors Projects and is permanently considered as possible provider of huge amount of beryllium for those reactors so that is the reason for starting studies of possibilities of large scale processing/recycling of such disposed irradiated beryllium. It is clear that the Ulba Metallurgical Plant is considered as the best site for it in Kazakhstan. The draft plan how to organize experimental studies of irradiated beryllium disposals in Kazakhstan involving National Nuclear Center, National University (Almaty), JSC ''Ulba Metallurgical Plant'' (Ust-Kamenogorsk) would be presented in this paper as well as proposals to arrange international collaboration in that field through ISTC (International Science Technology Center, Moscow). (author)

  10. Thermal fatigue of beryllium

    International Nuclear Information System (INIS)

    Deksnis, E.; Ciric, D.; Falter, H.

    1995-01-01

    Thermal fatigue life of S65c beryllium castellated to a geometry 6 x 6 x (8-10)mm deep has been tested for steady heat fluxes of 3 MW/m 2 to 5 MW/m 2 and under pulsed heat fluxes (10-20 MW/m 2 ) for which the time averaged heat flux is 5 MW/m 2 . These tests were carried out in the JET neutral beam test facility A test sequence with peak surface temperatures ≤ 600 degrees C produced no visible fatigue cracks. In the second series of tests, with T max ≤ 750 degrees C evidence for fatigue appeared after a minimum of 1350 stress cycles. These fatigue data are discussed in view of the observed lack of thermal fatigue in JET plasma operations with beryllium PFC. JET experience with S65b and S65c is reviewed; recent operations with Φ = 25 MW/m 2 and sustained melting/resolidification are also presented. The need for a failure criterion for finite element analyses of Be PFC lifetimes is discussed

  11. Hydrogen transport behavior of beryllium

    Energy Technology Data Exchange (ETDEWEB)

    Anderl, R.A.; Hankins, M.R.; Longhurst, G.R.; Pawelko, R.J. (Idaho National Engineering Lab., EG and G Idaho, Inc., Idaho Falls, ID (United States)); Macaulay-Newcombe, R.G. (Dept. of Engineering Physics, Univ. Hamilton, ON (Canada))

    1992-12-01

    Beryllium is being evaluated for use as a plasma-facing material in the International Thermonuclear Experimental Reactor (ITER). One concern in the evaluation is the retention and permeation of tritium implanted into the plasma-facing surface. We performed laboratory-scale studies to investigate mechanisms that influence hydrogen transport and retention in beryllium foil specimens of rolled powder metallurgy product and rolled ingot cast beryllium. Specimen characterization was accomplished using scanning electron microscopy. Auger electron spectroscopy, and Rutherford backscattering spectrometry (RBS) techniques. Hydrogen transport was investigated using ion-beam permeation experiments and nuclear reaction analysis (NRA). Results indicate that trapping plays a significant role in permeation, re-emission, and retention, and that surface processes at both upstream and downstream surfaces are also important. (orig.).

  12. Safety handling of beryllium for fusion technology R and D

    International Nuclear Information System (INIS)

    Yoshida, Hiroshi; Okamoto, Makoto; Terai, Takayuki; Odawara, Osamu; Ashibe, Kusuo; Ohara, Atsushi.

    1992-07-01

    Feasibility of beryllium use as a blanket neutron multiplier, first wall and plasma facing material has been studied for the D-T burning experiment reactors such as ITER. Various experimental work of beryllium and its compounds will be performed under the conditions of high temperature and high energy particle exposure simulating fusion reactor conditions. Beryllium is known as a hazardous substance and its handling has been carefully controlled by various health and safe guidances and/or regulations in many countries. Japanese regulations for hazardous substance provide various guidelines on beryllium for the protection of industrial workers and environment. This report was prepared for the safe handling of beryllium in a laboratory scale experiments for fusion technology R and D such as blanket development. Major items in this report are; (1) Brief review of guidances and regulations in USA, UK and Japan. (2) Safe handling and administration manuals at beryllium facilities in INEL, LANL and JET. (3) Conceptual design study of beryllium handling facility for small to mid-scale blanket R and D. (4) Data on beryllium toxicity, example of clinical diagnosis of beryllium disease, and environmental occurence of beryllium. (5) Personnel protection tools of Japanese Industrial Standard for hazardous substance. (author) 61 refs

  13. Beryllium health effects in the era of the beryllium lymphocyte proliferation test.

    Science.gov (United States)

    Maier, L A

    2001-05-01

    The beryllium lymphocyte proliferation test (BeLPT) has revolutionized our approach to the diagnosis, screening, and surveillance of beryllium health effects. Based on the development of a beryllium-specific cell-mediated immune response, the BeLPT has allowed us to define early health effects of beryllium, including beryllium sensitization (BeS), and chronic beryllium disease (CBD) at a subclinical stage. The use of this test as a screening tool has improved our understanding of these health effects. From a number of studies it is apparent that BeS precedes CBD and develops after as little as 9 weeks of beryllium exposure. CBD occurs within 3 months and up to 30 years after initial beryllium exposure. Exposure-response variables have been associated with BeS/CBD, including work as a machinist, chemical or metallurgical operator, laboratory technician, work in ceramics or beryllium metal production, and years of beryllium exposure. Recent studies have found BeS and CBD in workplaces in which the majority of exposures were below the 2 microg/m3 OSHA time-weighted average (TWA). Ideally, the BeLPT would be used in surveillance aimed at defining other risk-related processes, determining exposure variables which predict BeS and CBD, and defining the exposure level below which beryllium health effects do not occur. Unfortunately, the BeLPT can result in false negative tests and still requires an invasive procedure, a bronchoscopy, for the definitive diagnosis of CBD. Thus, research is needed to establish new tests to be used alone or in conjunction with the BeLPT to improve our ability to detect early beryllium health effects.

  14. (Beryllium). Internal Report No. 137, Jan. 15, 1958; Le beryllium

    Energy Technology Data Exchange (ETDEWEB)

    Mouret, P; Rigaud, A

    1959-07-01

    After a brief summary of the physical and chemical properties of beryllium, the various chemical treatments which can be applied to beryllium minerals either directly or after a physical enrichment are discussed. These various treatments give either the hydroxide or beryllium salts, from which either beryllium oxide or metallic beryllium can easily be obtained. The purification, analysis and uses of beryllium are also briefly discussed. (author)

  15. Characterization of plasma sprayed beryllium ITER first wall mockups

    Energy Technology Data Exchange (ETDEWEB)

    Castro, R.G.; Vaidya, R.U.; Hollis, K.J. [Los Alamos National Lab., NM (United States). Material Science and Technology Div.

    1998-01-01

    ITER first wall beryllium mockups, which were fabricated by vacuum plasma spraying the beryllium armor, have survived 3000 thermal fatigue cycles at 1 MW/m{sup 2} without damage during high heat flux testing at the Plasma Materials Test Facility at Sandia National Laboratory in New Mexico. The thermal and mechanical properties of the plasma sprayed beryllium armor have been characterized. Results are reported on the chemical composition of the beryllium armor in the as-deposited condition, the through thickness and normal to the through thickness thermal conductivity and thermal expansion, the four-point bend flexure strength and edge-notch fracture toughness of the beryllium armor, the bond strength between the beryllium armor and the underlying heat sink material, and ultrasonic C-scans of the Be/heat sink interface. (author)

  16. Characterization of Plasma Sprayed Beryllium ITER First Wall Mockups

    International Nuclear Information System (INIS)

    Castro, Richard G.; Vaidya, Rajendra U.; Hollis, Kendall J.

    1997-10-01

    ITER first wall beryllium mockups, which were fabricated by vacuum plasma spraying the beryllium armor, have survived 3000 thermal fatigue cycles at 1 MW/sq m without damage during high heat flux testing at the Plasma Materials Test Facility at Sandia National Laboratory in New Mexico. The thermal and mechanical properties of the plasma sprayed beryllium armor have been characterized. Results are reported on the chemical composition of the beryllium armor in the as-deposited condition, the through thickness and normal to the through thickness thermal conductivity and thermal expansion, the four-point bend flexure strength and edge-notch fracture toughness of the beryllium armor, the bond strength between the beryllium armor and the underlying heat sink material, and ultrasonic C-scans of the Be/heat sink interface

  17. Fluorimetric method for determination of Beryllium

    International Nuclear Information System (INIS)

    Sparacino, N.; Sabbioneda, S.

    1996-10-01

    The old fluorimetric method for the determination of Beryllium, based essentially on the fluorescence of the Beryllium-Morine complex in a strongly alkaline solution, is still competitive and stands the comparison with more modern methods or at least three reasons: in the presence of solid or gaseous samples (powders), the times necessary to finalize an analytic determination are comparable since the stage of the process which lasts the longest is the mineralization of the solid particles containing Beryllium, the cost of a good fluorimeter is by far Inferior to the cost, e. g., of an Emission Spectrophotometer provided with ICP torch and magnets for exploiting the Zeeman effect and of an Atomic absorption Spectrophotometer provided with Graphite furnace; it is possible to determine, fluorimetrically, rather small Beryllium levels (about 30 ng of Beryllium/sample), this potentiality is more than sufficient to guarantee the respect of all the work safety and hygiene rules now in force. The study which is the subject of this publication is designed to the analysis procedure which allows one to reach good results in the determination of Beryllium, chiefly through the control and measurement of the interference effect due to the presence of some metals which might accompany the environmental samples of workshops and laboratories where Beryllium is handled, either at the pure state or in its alloys. The results obtained satisfactorily point out the merits and limits of this analytic procedure

  18. Experimental Simulation of Beryllium Armour Damage Under ITER-like Intense Transient Plasma Loads

    Energy Technology Data Exchange (ETDEWEB)

    Kupriyanov, I.; Basaleev, E.; Nikolaev, G.; Kurbatova, L., E-mail: igkupr@gmail.com [A.A. Bochvar High Technology Research Institute of Inorganic Material, Moscow (Russian Federation); Podkovyrov, V.; Zhitlukhin, A. [SRC RF TRINITI, Troitsk (Russian Federation); Khimchenko, L. L. [Project Centre of ITER, Moscow (Russian Federation)

    2012-09-15

    Full text: Beryllium will be used as a plasma facing material in the next generation of tokamaks such as ITER. During plasma operation in ITER, the plasma facing materials and components will be suffered by different kinds of loading which may affect their surface or their joint to the heat sink. In addition to quasi-stationary loadings which are caused by the normal cycling operation, the plasma facing components and materials may also be exposed to the intense short transient loads like disruptions, ELMs. All these events may lead to beryllium surface melting, cracking, evaporation and erosion. It is expected that the erosion of beryllium under transient plasma loads such as ELMs and disruptions will mainly determine a lifetime of ITER first wall. To obtain the experimental data for the evaluation of the beryllium armor lifetime and dust production under ITER-relevant transient loads, the advanced plasma gun QSPA-Be facility has been constructed in Bochvar Institute. This paper presents recent results of the experiments with Russian beryllium of TGP-56FW ITER grade. The mock-ups of a special design armored with two beryllium targets (80 x 80 x 10 mm{sup 3}) were tested by hydrogen plasma streams (5 cm in diameter) with pulse duration of 0.5 ms and heat load of 0.5 and 1.0 MJ/m{sup 2}. Experiments were performed at RT temperature. The evolution of surface microstructure and profile, cracks morphology and mass loss/gain under erosion process on the beryllium surface exposed to up to 250 shots will be presented and discussed. (author)

  19. Springfield Processing Plant (SPP) Facility Information

    Energy Technology Data Exchange (ETDEWEB)

    Leach, Janice; Torres, Teresa M.

    2012-10-01

    The Springfield Processing Plant is a hypothetical facility. It has been constructed for use in training workshops. Information is provided about the facility and its surroundings, particularly security-related aspects such as target identification, threat data, entry control, and response force data.

  20. Overview - Defense Waste Processing Facility Operating Experience

    International Nuclear Information System (INIS)

    Norton, M.R.

    2002-01-01

    The Savannah River Site's Defense Waste Processing Facility (DWPF) near Aiken, SC is the world's largest radioactive waste vitrification facility. Radioactive operations began in March 1996 and over 1,000 canisters have been produced. This paper presents an overview of the DWPF process and a summary of recent facility operations and process improvements. These process improvements include efforts to extend the life of the DWPF melter, projects to increase facility throughput, initiatives to reduce the quantity of wastewater generated, improved remote decontamination capabilities, and improvements to remote canyon equipment to extend equipment life span. This paper also includes a review of a melt rate improvement program conducted by Savannah River Technology Center personnel. This program involved identifying the factors that impacted melt rate, conducting small scale testing of proposed process changes and developing a cost effective implementation plan

  1. SRS Process Facility Significance Fire Frequency

    Energy Technology Data Exchange (ETDEWEB)

    Sarrack, A.G. [Westinghouse Savannah River Company, AIKEN, SC (United States)

    1995-10-01

    This report documents the method and assumptions of a study performed to determine a site generic process facility significant fire initiator frequency and explains the proper way this value should be used.

  2. Advanced Materials Growth and Processing Facility

    Data.gov (United States)

    Federal Laboratory Consortium — This most extensive of U.S. Army materials growth and processing facilities houses seven dedicated, state-of-the-art, molecular beam epitaxy and three metal organic...

  3. SRS Process Facility Significance Fire Frequency

    International Nuclear Information System (INIS)

    Sarrack, A.G.

    1995-10-01

    This report documents the method and assumptions of a study performed to determine a site generic process facility significant fire initiator frequency and explains the proper way this value should be used

  4. Fabrication of Separator Demonstration Facility process vessel

    International Nuclear Information System (INIS)

    Oberst, E.F.

    1985-01-01

    The process vessel system is the central element in the Separator Development Facility (SDF). It houses the two major process components, i.e., the laser-beam folding optics and the separators pods. This major subsystem is the critical-path procurement for the SDF project. Details of the vaious parts of the process vessel are given

  5. Beryllium poisonings; Les intoxications par le beryllium

    Energy Technology Data Exchange (ETDEWEB)

    Alibert, S.

    1959-03-15

    This note reports a bibliographical study of beryllium toxicity. Thus, this bibliographical review addresses and outlines aspects and issues like aetiology, cases of acute poisoning (cutaneous manifestations, pulmonary manifestations), chronic poisoning (cutaneous, pulmonary and bone manifestations), excretion and localisation, and prognosis.

  6. Processing facility for metal waste

    International Nuclear Information System (INIS)

    Awano, Toshihiko; Kataoka, Yoshitsune.

    1998-01-01

    Each steps of temporarily storing materials to be reduced in the volume to a storage vessel, transferring them to a weighing machine by a conveyor, weighing them by a weighing machine, drying them by a drying means, packing them in containing canisters, sealing and welding them, carrying out the containing canisters after sealing are conducted independently respectively or optionally simultaneously in parallel. Accordingly, isolation from peripheral circumstances is ensured, and improvement of working efficiency, ensuring of safety and simplification of structure of processing devices can be attained. (T.M.)

  7. Study beryllium microplastic deformation

    International Nuclear Information System (INIS)

    Papirov, I.I.; Ivantsov, V.I.; Nikolaenko, A.A.; Shokurov, V.S.; Tuzov, Yu.V.

    2015-01-01

    Microplastic flow characteristics systematically studied for different varieties beryllium. In isostatically pressed beryllium it decreased with increasing particle size of the powder, increasing temperature and increasing the pressing metal purity. High initial values of the limit microelasticity and microflow in some cases are due a high level of internal stresses of thermal origin and over time it can relax slowly. During long-term storage of beryllium materials with high initial resistance values microplastic deformation microflow limit and microflow stress markedly reduced, due mainly to the relaxation of thermal microstrain

  8. Beryllium R and D for fusion applications

    International Nuclear Information System (INIS)

    Scaffidi-Argentina, F.; Longhurst, G.R.; Shestakov, V.; Kawamura, H.

    2000-01-01

    Beryllium is one of the primary candidates as both plasma-facing material (PFM) and neutron multiplier in the next-step fusion reactors. Both sintered-product blocks and pebbles are considered in fusion reactor designs. Beryllium evaporated on carbon tiles has also been used in Joint European Torus (JET) and may be considered for other designs. Future efforts are directed toward the pebble form of beryllium. Research and evaluations of data are underway to determine the most attractive material processing approaches in terms of fabrication cost and quality; technical issues associated with heat transfer; thermal, mechanical and irradiation stability; safety and tritium release. Beryllium plasma-facing components will require periodic repair or replacement, therefore disposal or recycling of activated and tritiated beryllium will also be a concern. Beryllium as a component of the molten salt, Flibe is also being considered in novel approaches to the plasma-structure interface. This paper deals with the main issues related to the use of Be in a fusion reactor as both neutron multiplier and first wall material. These issues include potential reactions with steam during accidents and the health and environmental aspects of its use, reprocessing and reuse, or disposal

  9. Beryllium R and D for fusion applications

    Energy Technology Data Exchange (ETDEWEB)

    Scaffidi-Argentina, F. E-mail: francesco.scaffidi@iket.fzk.de; Longhurst, G.R.; Shestakov, V.; Kawamura, H

    2000-11-01

    Beryllium is one of the primary candidates as both plasma-facing material (PFM) and neutron multiplier in the next-step fusion reactors. Both sintered-product blocks and pebbles are considered in fusion reactor designs. Beryllium evaporated on carbon tiles has also been used in Joint European Torus (JET) and may be considered for other designs. Future efforts are directed toward the pebble form of beryllium. Research and evaluations of data are underway to determine the most attractive material processing approaches in terms of fabrication cost and quality; technical issues associated with heat transfer; thermal, mechanical and irradiation stability; safety and tritium release. Beryllium plasma-facing components will require periodic repair or replacement, therefore disposal or recycling of activated and tritiated beryllium will also be a concern. Beryllium as a component of the molten salt, Flibe is also being considered in novel approaches to the plasma-structure interface. This paper deals with the main issues related to the use of Be in a fusion reactor as both neutron multiplier and first wall material. These issues include potential reactions with steam during accidents and the health and environmental aspects of its use, reprocessing and reuse, or disposal.

  10. Tritium release from neutron irradiated beryllium pebbles

    Energy Technology Data Exchange (ETDEWEB)

    Scaffidi-Argentina, F.; Werle, H. [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany). Inst. fuer Neutronenphysik und Reactortechnik

    1998-01-01

    One of the most important open issues related to beryllium for fusion applications refers to the kinetics of the tritium release as a function of neutron fluence and temperature. The EXOTIC-7 as well as the `Beryllium` experiments carried out in the HFR reactor in Petten are considered as the most detailed and significant tests for investigating the beryllium response under neutron irradiation. This paper reviews the present status of beryllium post-irradiation examinations performed at the Forschungszentrum Karlsruhe with samples from the above mentioned irradiation experiments, trying to elucidate the tritium release controlling processes. In agreement with previous studies it has been found that release starts at about 500-550degC and achieves a maximum at about 700-750degC. The observed release at about 500-550degC is probably due to tritium escaping from chemical traps, while the maximum release at about 700-750degC is due to tritium escaping from physical traps. The consequences of a direct contact between beryllium and ceramics during irradiation, causing tritium implanting in a surface layer of beryllium up to a depth of about 40 mm and leading to an additional inventory which is usually several times larger than the neutron-produced one, are also presented and the effects on the tritium release are discussed. (author)

  11. Outline of the Chemical Processing Facility (CPF)

    International Nuclear Information System (INIS)

    Arita, Katsuhiko

    1978-01-01

    Concerning the Chemical Processing Facility (CPF), a high level radioactive material research facility, to be installed in Tokai Works of Power Reactor and Nuclear Fuel Development Corporation (PNC), the detailed design and the governmental safety inspection were finished. The construction has been already started, and it will be completed in 1980. Under the national policy of establishing a nuclear fuel cycle, PNC is now carrying out the development of its downstream technology. The objects of the Chemical Processing Facility are the researches of the treatment techniques of high level radioactive liquid wastes from fuel reprocessing and of the reprocessing of fast reactor fuel. The following matters are described: purpose of the CPF, i.e. fast reactor fuel reprocessing and high-level liquid waste treatment; construction of the CPF, i.e. buildings, cells and an exhaust stack; test systems, i.e. fuel reprocessing and liquid waste vitrification; and facility safety. (Mori, K.)

  12. The Rocky Flats Environmental Technology Site beryllium characterization project

    International Nuclear Information System (INIS)

    Morrell, D.M.; Miller, J.R.; Allen, D.F.

    1999-01-01

    A site beryllium characterization project was completed at the Rocky Flats Environmental Technology Site (RFETS) in 1997. Information from historical reviews, previous sampling surveys, and a new sampling survey were used to establish a more comprehensive understanding of the locations and levels of beryllium contamination in 35 buildings. A feature of the sampling strategy was to test if process knowledge was a good predictor of where beryllium contamination could be found. Results revealed that this technique was effective at identifying where surface contamination levels might exceed the RFETS smear control level but that it was not effective in identifying where low concentrations of beryllium might be found

  13. Cognitive facilities of governance of transformations processes

    Directory of Open Access Journals (Sweden)

    A. V. Reshetnichenko

    2014-03-01

    For example, each of levels of organization of the both realized and subconscious, facilities of cognition includes the dependent numerical, voice, coloured and concept facilities correlative. As for the system of the realized and subconscious facilities of transformations, their basis is made by the ascending and descending forms of organization of motion of matter, energy, information and organization of elements of life. Fixed in basis of research of mul’timodal’na logician allowed to expose dialectical nature of mechanisms of bifurcations, synthesis, freymuvannya and clusterizations as main condition of forming on principle of new control system by processes development of man, state and society, on the way of mastering of space.

  14. Beryllium and copper-beryllium alloys; Beryllium und Kupfer-Beryllium-Legierungen

    Energy Technology Data Exchange (ETDEWEB)

    Nagel, Nikolaus [Materion Brush GmbH, Stuttgart (Germany). Operation and Quality/EH and S

    2017-02-15

    The light metal beryllium is a comparatively rare element, which today is primarily derived from bertrandite. It is mainly used as pure metal or in the form of copper-beryllium alloys, e.g., in automotive industry, aerospace, and electrical components. The wide range of applications is mainly attributed to the extremely high rigidity/density ratio. An overview of the history of the metal, its production, and recycling as well as the properties of CuBe alloys are given.

  15. The immunotoxicity of beryllium

    International Nuclear Information System (INIS)

    Reeves, A.L.

    1983-01-01

    In the disease berylliosis, granulomatous hypersensitivity is the specific immune response to tissue contact with a poorly soluble particle of beryllium compound, mediated through the accumulation and proliferation of reticuloendothelial cells. A review is given of the work accomplished since the 1950's and particularly since the 1970's to elucidate the nature and consequences of this response to beryllium and its compounds. (U.K.)

  16. Preparation of beryllium hydride

    International Nuclear Information System (INIS)

    Bergeron, C.R.; Baker, R.W.

    1975-01-01

    Beryllium hydride of high bulk density, suitable for use as a component of high-energy fuels, is prepared by the pyrolysis, in solution in an inert solvent, of a ditertiary-alkyl beryllium. An agitator introduces mechanical energy into the reaction system, during the pyrolysis, at the rate of 0.002 to 0.30 horsepower per gallon of reaction mixture. (U.S.)

  17. Image processing technology for nuclear facilities

    International Nuclear Information System (INIS)

    Lee, Jong Min; Lee, Yong Beom; Kim, Woong Ki; Park, Soon Young

    1993-05-01

    Digital image processing technique is being actively studied since microprocessors and semiconductor memory devices have been developed in 1960's. Now image processing board for personal computer as well as image processing system for workstation is developed and widely applied to medical science, military, remote inspection, and nuclear industry. Image processing technology which provides computer system with vision ability not only recognizes nonobvious information but processes large information and therefore this technique is applied to various fields like remote measurement, object recognition and decision in adverse environment, and analysis of X-ray penetration image in nuclear facilities. In this report, various applications of image processing to nuclear facilities are examined, and image processing techniques are also analysed with the view of proposing the ideas for future applications. (Author)

  18. Tritium behavior in ITER beryllium

    International Nuclear Information System (INIS)

    Longhurst, G.R.

    1990-10-01

    The beryllium neutron multiplier in the ITER breeding blanket will generate tritium through transmutations. That tritium constitutes a safety hazard. Experiments evaluating tritium storage and release mechanisms have shown that most of the tritium comes out in a burst during thermal ramping. A small fraction of retained tritium is released by thermally activated processes. Analysis of recent experimental data shows that most of the tritium resides in helium bubbles. That tritium is released when the bubbles undergo swelling sufficient to develop porosity that connects with the surface. That appears to occur when swelling reaches about 10--15%. Other tritium appears to be stored chemically at oxide inclusions, probably as Be(OT) 2 . That component is released by thermal activation. There is considerable variation in published values for tritium diffusion through the beryllium and solubility in it. Data from experiments using highly irradiated beryllium from the Idaho National Engineering Laboratory showed diffusivity generally in line with the most commonly accepted values for fully dense material. Lower density material, planned for use in the ITER blanket may have very short diffusion times because of the open structure. The beryllium multiplier of the ITER breeding blanket was analyzed for tritium release characteristics using temperature and helium production figures at the midplane generated in support of the ITER Summer Workshop, 1990 in Garching. Ordinary operation, either in Physics or Technology phases, should not result in the release of tritium trapped in the helium bubbles. Temperature excursions above 600 degree C result in large-scale release of that tritium. 29 refs., 10 figs., 3 tabs

  19. Material Processing Facility - Skylab Experiment M512

    Science.gov (United States)

    1972-01-01

    This chart details Skylab's Materials Processing Facility experiment (M512). This facility, located in the Multiple Docking Adapter, was developed for Skylab and accommodated 14 different experiments that were carried out during the three marned missions. The abilities to melt and mix without the contaminating effects of containers, to suppress thermal convection and buoyancy in fluids, and to take advantage of electrostatic and magnetic forces and otherwise masked by gravitation opened the way to new knowledge of material properties and processes. This beginning would ultimately lead to the production of valuable new materials for use on Earth.

  20. The structure, properties and performance of plasma-sprayed beryllium for fusion applications

    International Nuclear Information System (INIS)

    Castro, R.G.; Stanek, P.W.; Elliott, K.E.

    1995-01-01

    Plasma-spray technology is under investigation as a method for producing high thermal conductivity beryllium coatings for use in magnetic fusion applications. Recent investigations have focused on optimizing the plasma-spray process for depositing beryllium coatings on damaged beryllium surfaces. Of particular interest has been optimizing the processing parameters to maximize the through-thickness thermal conductivity of the beryllium coatings. Experimental results will be reported on the use of secondary H 2 gas additions to improve the melting of the beryllium powder and transferred-arc cleaning to improve the bonding between the beryllium coatings and the underlying surface. Information will also be presented on thermal fatigue tests which were done on beryllium coated ISX-B beryllium limiter tiles using 10 sec cycle times with 60 sec cooldowns and an International Thermonuclear Experimental Reactor (ITER) relevant divertor heat flux slightly in excess of 5 MW/m 2

  1. Environmental information document defense waste processing facility

    International Nuclear Information System (INIS)

    1981-07-01

    This report documents the impact analysis of a proposed Defense Waste Processing Facility (DWPF) for immobilizing high-level waste currently being stored on an interim basis at the Savannah River Plant (SRP). The DWPF will process the waste into a form suitable for shipment to and disposal in a federal repository. The DWPF will convert the high-level waste into: a leach-resistant form containing above 99.9% of all the radioactivity, and a residue of slightly contaminated salt. The document describes the SRP site and environs, including population, land and water uses; surface and subsurface soils and waters; meteorology; and ecology. A conceptual integrated facility for concurrently producing glass waste and saltcrete is described, and the environmental effects of constructing and operating the facility are presented. Alternative sites and waste disposal options are addressed. Also environmental consultations and permits are discussed

  2. Defense Waste Processing Facility, Savannah River Plant

    International Nuclear Information System (INIS)

    After 10 years of research, development, and testing, the US Department of Energy is building a new facility which will prepare high-level radioactive waste for permanent disposal. The Defense Waste Processing Facility, known as the DWPF, will be the first production-scale facility of its kind in the United States. In the DWPF, high-level waste produced by defense activities at the Savannah River Plant will be processed into a solid form, borosilicate glass, suitable for permanent off-site geologic disposal. With construction beginning in the fall of 1983, the DWPT is scheduled to be operational in 1989. By 2005, the DWPF will have immobilized the backlog of high-level waste which has been accumulating in storage tanks at the Savannah River Plant since 1954. Canisters of the immobilized waste will then be ready for permanent disposal deep under the ground, safely isolated from the environment

  3. Laser welding of a beryllium/tantalum collimator

    International Nuclear Information System (INIS)

    Lingenfelter, A.C.; Anglin, C.D.

    1985-01-01

    This report describes the methods utilized in the fabrication of a collimator from 0.001 inch thick beryllium and tantalum foil. The laser welding process proved to be an acceptable method for joining the beryllium in a standing edge joint configuration

  4. Chemical process safety at fuel cycle facilities

    International Nuclear Information System (INIS)

    Ayres, D.A.

    1997-08-01

    This NUREG provides broad guidance on chemical safety issues relevant to fuel cycle facilities. It describes an approach acceptable to the NRC staff, with examples that are not exhaustive, for addressing chemical process safety in the safe storage, handling, and processing of licensed nuclear material. It expounds to license holders and applicants a general philosophy of the role of chemical process safety with respect to NRC-licensed materials; sets forth the basic information needed to properly evaluate chemical process safety; and describes plausible methods of identifying and evaluating chemical hazards and assessing the adequacy of the chemical safety of the proposed equipment and facilities. Examples of equipment and methods commonly used to prevent and/or mitigate the consequences of chemical incidents are discussed in this document

  5. Defense Waste Processing Facility prototypic analytical laboratory

    International Nuclear Information System (INIS)

    Policke, T.A.; Bryant, M.F.; Spencer, R.B.

    1991-01-01

    The Defense Waste Processing Technology (DWPT) Analytical Laboratory is a relatively new laboratory facility at the Savannah River Site (SRS). It is a non-regulated, non-radioactive laboratory whose mission is to support research and development (R ampersand D) and waste treatment operations by providing analytical and experimental services in a way that is safe, efficient, and produces quality results in a timely manner so that R ampersand D personnel can provide quality technical data and operations personnel can efficiently operate waste treatment facilities. The modules are sample receiving, chromatography I, chromatography II, wet chemistry and carbon, sample preparation, and spectroscopy

  6. The new MAW scrap processing facility

    International Nuclear Information System (INIS)

    Kueppers, L.

    1994-01-01

    The shielded bunker for heat-generating waste attached to the MAW scrap processing cell will be modified and extended to comprise several MAW scrap processing cells of enhanced throughput capacity, and a new building to serve as an airlock and port for acceptance of large shipping casks (shipping cask airlock, TBS). The new facility is to process scrap from decommissioned nuclear installations, and in addition radwaste accrued at operating plants of utilities. This will allow efficient and steady use of the new MAW scrap processing facility. The planning activities for modification and extension are based on close coordination between KfK and the GNS mbH, in order to put structural dimensioning and capacity planning on a realistic basis in line with expected amounts of radwaste from operating nuclear installations of utilities. The paper indicates the currently available waste amount assessments covering solid radwaste (MAW) from the decommissioning of the WAK, MZFR, and KNK II, and existing waste amounts consisting of core internals of German nuclear power plant. The figures show that the MAW scrap processing facility will have to process an overall bulk of about 1100 Mg of solid waste over the next ten years to come. (orig./HP) [de

  7. Materials evaluation for a transuranic processing facility

    International Nuclear Information System (INIS)

    Barker, S.A.; Schwenk, E.B.; Divine, J.R.

    1990-11-01

    The Westinghouse Hanford Company, with the assistance of the Pacific Northwest Laboratory, is developing a transuranium extraction process for preheating double-shell tank wastes at the Hanford Site to reduce the volume of transuranic waste being sent to a repository. The bench- scale transuranium extraction process development is reaching a stage where a pilot plant design has begun for the construction of a facility in the existing B Plant. Because of the potential corrosivity of neutralized cladding removal waste process streams, existing embedded piping alloys in B Plant are being evaluated and ''new'' alloys are being selected for the full-scale plant screening corrosion tests. Once the waste is acidified with HNO 3 , some of the process streams that are high in F - and low in Al and zr can produce corrosion rates exceeding 30,000 mil/yr in austenitic alloys. Initial results results are reported concerning the applicability of existing plant materials to withstand expected process solutions and conditions to help determine the feasibility of locating the plant at the selected facility. In addition, process changes are presented that should make the process solutions less corrosive to the existing materials. Experimental work confirms that Hastelloy B is unsatisfactory for the expected process solutions; type 304L, 347 and 309S stainless steels are satisfactory for service at room temperature and 60 degrees C, if process stream complexing is performed. Inconel 625 was satisfactory for all solutions. 17 refs., 5 figs., 8 tabs

  8. Mechanical performance of irradiated beryllium pebbles

    Energy Technology Data Exchange (ETDEWEB)

    Scaffidi-Argentina, F.; Dalle-Donne, M.; Werle, H. [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany). Inst. fuer Neutronenphysik und Reaktortechnik

    1998-01-01

    For the Helium Cooled Pebble Bed (HCPB) Blanket, which is one of the two reference concepts studied within the European Fusion Technology Programme, the neutron multiplier consists of a mixed bed of about 2 and 0.1-0.2 mm diameter beryllium pebbles. Beryllium has no structural function in the blanket, however microstructural and mechanical properties are important, as they might influence the material behavior under neutron irradiation. The EXOTIC-7 as well as the `Beryllium` experiments carried out in the HFR reactor in Petten are considered as the most detailed and significant tests for investigating it. This paper reviews the present status of beryllium post-irradiation examinations performed at the Forschungszentrum Karlsruhe with samples from these irradiation experiments, emphasizing the effects of irradiation of essential material properties and trying to elucidate the processes controlling the property changes. The microstructure, the porosity distribution, the impurity content, the behavior under compression loads and the compatibility of the beryllium pebbles with lithium orthosilicate (Li{sub 4}SiO{sub 4}) during the in-pile irradiation are presented and critically discussed. Qualitative information on ductility and creep obtained by hardness-type measurements are also supplied. (author)

  9. CNAEM waste processing and storage facility

    International Nuclear Information System (INIS)

    Osmanlioglu, A.E.; Kahraman, A.; Altunkaya, M.

    1998-01-01

    Radioactive waste in Turkey is generated from various applications. Radioactive waste management activities are carried out in a facility at Cekmece Nuclear Research and Training Center (CNAEM). This facility has been assigned to take all low-level radioactive wastes generated by nuclear applications in Turkey. The wastes are generated from research and nuclear applications mainly in medicine, biology, agriculture, quality control in metal processing and construction industries. These wastes are classified as low- level radioactive wastes and their activities are up to 10 -3 Ci/m 3 (except spent sealed sources). Chemical treatment and cementation of liquid radwaste, segregation and compaction of solid wastes and conditioning of spent sources are the main processing activities of this facility. A.so, analyses, registration, quality control and interim storage of conditioned low-level wastes are the other related activities of this facility. Conditioned wastes are stored in an interim storage building. All waste management activities, which have been carried out in CNAEM, are generally described in this paper. (author)

  10. Beryllium coating on Inconel tiles

    International Nuclear Information System (INIS)

    Bailescu, V.; Burcea, G.; Lungu, C.P.; Mustata, I.; Lungu, A.M.; Rubel, M.; Coad, J.P.; Matthews, G.; Pedrick, L.; Handley, R.

    2007-01-01

    Full text of publication follows: The Joint European Torus (JET) is a large experimental nuclear fusion device. Its aim is to confine and study the behaviour of plasma in conditions and dimensions approaching those required for a fusion reactor. The plasma is created in the toroidal shaped vacuum vessel of the machine in which it is confined by magnetic fields. In preparation for ITER a new ITER-like Wall (ILW) will be installed on Joint European Torus (JET), a wall not having any carbon facing the plasma [1]. In places Inconel tiles are to be installed, these tiles shall be coated with Beryllium. MEdC represented by the National Institute for Laser, Plasma and Radiation Physics, Magurele, Bucharest and in direct cooperation with Nuclear Fuel Plant Pitesti started to coat Inconel tiles with 8 μm of Beryllium in accordance with the requirements of technical specification and fit for installation in the JET machine. This contribution provides an overview of the principles of manufacturing processes using thermal evaporation method in vacuum and the properties of the prepared coatings. The optimization of the manufacturing process (layer thickness, structure and purity) has been carried out on Inconel substrates (polished and sand blasted) The results of the optimization process and analysis (SEM, TEM, XRD, Auger, RBS, AFM) of the coatings will be presented. Reference [1] Takeshi Hirai, H. Maier, M. Rubel, Ph. Mertens, R. Neu, O. Neubauer, E. Gauthier, J. Likonen, C. Lungu, G. Maddaluno, G. F. Matthews, R. Mitteau, G. Piazza, V. Philipps, B. Riccardi, C. Ruset, I. Uytdenhouwen, R and D on full tungsten divertor and beryllium wall for JET TIER-like Wall Project, 24. Symposium on Fusion Technology - 11-15 September 2006 -Warsaw, Poland. (authors)

  11. Waste minimization at a plutonium processing facility

    International Nuclear Information System (INIS)

    Pillay, K.K.S.

    1995-01-01

    As part of Los Alamos National Laboratory's (LANL) mission to reduce the nuclear danger throughout the world, the plutonium processing facility at LANL maintains expertise and skills in nuclear weapons technologies as well as leadership in all peaceful applications of plutonium technologies, including fuel fabrication for terrestrial and space reactors and heat sources and thermoelectric generators for space missions. Another near-term challenge resulted from two safety assessments performed by the Defense Nuclear Facilities Safety Board and the U.S. Department of Energy during the past two years. These assessments have necessitated the processing and stabilization of plutonium contained in tons of residues so that they can be stored safely for an indefinite period. This report describes waste streams and approaches to waste reduction of plutonium management

  12. Gas processing at DOE nuclear facilities

    Energy Technology Data Exchange (ETDEWEB)

    Jacox, J.

    1995-02-01

    The term {open_quotes}Gas Processing{close_quotes} has many possible meanings and understandings. In this paper, and panel, we will be using it to generally mean the treatment of gas by methods other than those common to HVAC and Nuclear Air Treatment. This is only a working guideline not a rigorous definition. Whether a rigorous definition is desirable, or even possible is a question for some other forum. Here we will be discussing the practical aspects of what {open_quotes}Gas Processing{close_quotes} includes and how existing Codes, Standards and industry experience can, and should, apply to DOE and NRC Licensed facilities. A major impediment to use of the best engineering and technology in many nuclear facilities is the administrative mandate that only systems and equipment that meet specified {open_quotes}nuclear{close_quotes} documents are permissible. This paper will highlight some of the limitations created by this approach.

  13. Consenting process for radiation facilities. V. 4

    International Nuclear Information System (INIS)

    2011-03-01

    Safety codes and standards are formulated on the basis of nationally and internationally accepted safety criteria for design, construction and operation of specific equipment, systems, structures and components of nuclear and radiation facilities. Safety, codes establish the objectives and set requirements that shall be fulfilled to provide adequate assurance for safety. Safety codes establish the objectives and set requirements that shall be fulfilled to provide adequate assurance for safety. Safety guides elaborate various requirements and furnish approaches for their implementation. Safety manuals deal with specific topics and contain detailed scientific and technical information on the subject. These documents are prepared by experts in the relevant fields and are extensively reviewed by advisory committees of the Atomic Energy Regulatory Board (AERB) before they are published. The documents are revised when necessary, in the light of experience and feedback from users as well as new developments in the field. AERB issued a safety code on Regulation of Nuclear and Radiation Facilities (AERB/SC/G) to spell out the requirements/obligations to be met by a nuclear or radiation facility for the issue of regulatory consent at every stage. This safety guide apprises the details of the regulatory requirements for setting up the radiation facility such as consenting process, the stages requiring consent, wherever applicable documents to be submitted and the nature and extent of review. The guide also gives information on methods of review and assessment adopted by AERB

  14. Consenting process for radiation facilities. V. 3

    International Nuclear Information System (INIS)

    2011-03-01

    Safety codes and standards are formulated on the basis of nationally and internationally accepted safety criteria for design, construction and operation of specific equipment, systems, structures and components of nuclear and radiation facilities. Safety, codes establish the objectives and set requirements that shall be fulfilled to provide adequate assurance for safety. Safety codes establish the objectives and set requirements that shall be fulfilled to provide adequate assurance for safety. Safety guides elaborate various requirements and furnish approaches for their implementation. Safety manuals deal with specific topics and contain detailed scientific and technical information on the subject. These documents are prepared by experts in the relevant fields and are extensively reviewed by advisory committees of the Atomic Energy Regulatory Board (AERB) before they are published. The documents are revised when necessary, in the light of experience and feedback from users as well as new developments in the field. AERB issued a safety code on Regulation of Nuclear and Radiation Facilities (AERB/SC/G) to spell out the requirements/obligations to be met by a nuclear or radiation facility for the issue of regulatory consent at every stage. This safety guide apprises the details of the regulatory requirements for setting up the radiation facility such as consenting process, the stages requiring consent, wherever applicable documents to be submitted and the nature and extent of review. The guide also gives information on methods of review and assessment adopted by AERB

  15. Consenting process for radiation facilities. V. 1

    International Nuclear Information System (INIS)

    2011-03-01

    Safety codes and standards are formulated on the basis of nationally and internationally accepted safety criteria for design, construction and operation of specific equipment, systems, structures and components of nuclear and radiation facilities. Safety, codes establish the objectives and set requirements that shall be fulfilled to provide adequate assurance for safety. Safety codes establish the objectives and set requirements that shall be fulfilled to provide adequate assurance for safety. Safety guides elaborate various requirements and furnish approaches for their implementation. Safety manuals deal with specific topics and contain detailed scientific and technical information on the subject. These documents are prepared by experts in the relevant fields and are extensively reviewed by advisory committees of the Atomic Energy Regulatory Board (AERB) before they are published. The documents are revised when necessary, in the light of experience and feedback from users as well as new developments in the field. AERB issued a safety code on Regulation of Nuclear and Radiation Facilities (AERB/SC/G) to spell out the requirements/obligations to be met by a nuclear or radiation facility for the issue of regulatory consent at every stage. This safety guide apprises the details of the regulatory requirements for setting up the radiation facility such as consenting process, the stages requiring consent, wherever applicable documents to be submitted and the nature and extent of review. The guide also gives information on methods of review and assessment adopted by AERB

  16. Defense Waste Processing Facility Process Simulation Package Life Cycle

    International Nuclear Information System (INIS)

    Reuter, K.

    1991-01-01

    The Defense Waste Processing Facility (DWPF) will be used to immobilize high level liquid radioactive waste into safe, stable, and manageable solid form. The complexity and classification of the facility requires that a performance based operator training to satisfy Department of Energy orders and guidelines. A major portion of the training program will be the application and utilization of Process Simulation Packages to assist in training the Control Room Operators on the fluctionality of the process and the application of the Distribution Control System (DCS) in operating and managing the DWPF process. The packages are being developed by the DWPF Computer and Information Systems Simulation Group. This paper will describe the DWPF Process Simulation Package Life Cycle. The areas of package scope, development, validation, and configuration management will be reviewed and discussed in detail

  17. Assessment of the feasibility and advantages of beryllium recycling

    International Nuclear Information System (INIS)

    Druyts, F.; Braet, J.; Ooms, L.

    2006-01-01

    This paper proposes a generic route for the recycling of beryllium from fusion reactors, based on critical issues associated with beryllium pebbles after their service life in the HCPB breeding blanket. These critical issues are the high tritium inventory, the presence of long-lived radionuclides (among which transuranics due to traces of uranium in the base metal), and the chemical toxicity of beryllium. On the basis of the chemical and radiochemical characteristics of the neutron irradiated beryllium pebbles, we describe a possible recycling route. The first step is the detritiation of the material. This can be achieved by heating the pebbles to 800 o C under an argon flow. The argon gas avoids oxidation of the beryllium, and at the proposed temperature the tritium inventory is readily released from the pebbles. In a second step, the released tritium can be oxidised on a copper oxide bed to produce tritiated water, which is consistent with the current international strategy to convert all kinds of tritiated waste into tritiated water, which can subsequently be treated in a water detritiation plant. Removal of radionuclides from the beryllium pebbles may be achieved by several types of chloride processes. The first step is to pass chlorine gas (in an argon flow) over the pebbles, thus yielding volatile BeCl 2 . This beryllium chloride can then be purified by fractional distillation. As a small fraction of the beryllium chloride contains the long-lived 10 Be isotope, 10BeCl 2 has to be separated from 9BeCl 2 , which could be achieved by centrifugal techniques. The product can then be reduced to obtain high-purity metallic beryllium. Two candidate reduction methods were identified: fused salt electrolysis and thermal decomposition. Both these methods require laboratory parametric studies to maximise the yield and achieve a high purity metal, before either process can be upgraded to a larger scale. The eventual product of the chloride reduction process must be a high

  18. Exposure and genetics increase risk of beryllium sensitisation and chronic beryllium disease in the nuclear weapons industry.

    Science.gov (United States)

    Van Dyke, Michael V; Martyny, John W; Mroz, Margaret M; Silveira, Lori J; Strand, Matt; Cragle, Donna L; Tankersley, William G; Wells, Susan M; Newman, Lee S; Maier, Lisa A

    2011-11-01

    Beryllium sensitisation (BeS) and chronic beryllium disease (CBD) are caused by exposure to beryllium with susceptibility affected by at least one well-studied genetic host factor, a glutamic acid residue at position 69 (E69) of the HLA-DPβ chain (DPβE69). However, the nature of the relationship between exposure and carriage of the DPβE69 genotype has not been well studied. The goal of this study was to determine the relationship between DPβE69 and exposure in BeS and CBD. Current and former workers (n=181) from a US nuclear weapons production facility, the Y-12 National Security Complex (Oak Ridge, Tennessee, USA), were enrolled in a case-control study including 35 individuals with BeS and 19 with CBD. HLA-DPB1 genotypes were determined by PCR-SSP. Beryllium exposures were assessed through worker interviews and industrial hygiene assessment of work tasks. After removing the confounding effect of potential beryllium exposure at another facility, multivariate models showed a sixfold (OR 6.06, 95% CI 1.96 to 18.7) increased odds for BeS and CBD combined among DPβE69 carriers and a fourfold (OR 3.98, 95% CI 1.43 to 11.0) increased odds for those exposed over an assigned lifetime-weighted average exposure of 0.1 μg/m(3). Those with both risk factors had higher increased odds (OR 24.1, 95% CI 4.77 to 122). DPβE69 carriage and high exposure to beryllium appear to contribute individually to the development of BeS and CBD. Among workers at a beryllium-using facility, the magnitude of risk associated with either elevated beryllium exposure or carriage of DPβE69 alone appears to be similar.

  19. Microstructure Analysis on Beryllium Reflector Blocks of Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Suk Hoon; Jang, Jin Sung; Jeong, Yong Hwan; Han, Chang Hee; Jung, Yang Il; Kim, Tae Kyu [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Choi, Yong Seok; Oh, Kyu Hwan [Seoul National University, Seoul (Korea, Republic of)

    2012-05-15

    A pure beryllium has a very low mass absorption coefficient: it has been used as the reflector element material in research reactors. The lifetime of beryllium reflector elements usually determined by the swelling: the swelling leads to dimensional change in the reflector frame, which results in bending or cracking of the parts. The mechanical interference in between parts should be avoided; the anisotropy of beryllium also needs to be considered. A beryllium has hexagonal close-pack (HCP) crystal structure, which is inherently anisotropic. It has virtually no ductility in one direction. There are two main aspects in the manufacturing of beryllium which will affect its isotropy, and those are the powder morphology and the consolidation process. Powder metallurgy permits the material to be produced in isotropic and fine-grained form, which overcomes the crystal structure problem by distributing loads in low ductility oriented grains to high ductility oriented grains. There are three representative consolidating methods to make beryllium reflector blocks. Traditionally, most powder-derived grades of beryllium have been consolidated by vacuum hot-pressing (VHP). A column of loose beryllium powder is compacted under vacuum by the pressure of the opposed upper and lower punches, bringing the billet to final density. The VHP process is directional in nature: it contributes to the anisotropy of the material properties. Another consolidating method for beryllium powder is hot isostatic pressing (HIPing), which will enhance its isotropy. During HIPing, The argon gas exerts pressure uniformly in all directions on the can containing the beryllium powder. The HIP process is effective to improve the isotropy of the resulting material as well as refinement of grain sizes. The last consolidating method is hot extrusion (HE). A roughly close packed beryllium is subjected to severe plastic defomation, the grains are refined and the tensile strength is enhanced. Since the material

  20. Radiation processing facilities and services in Malaysia

    International Nuclear Information System (INIS)

    Zulkafli Ghazali

    2007-01-01

    It is envisaged that radiation processing will continue to play an important role towards the progress and development of industry in Malaysia. Malaysian Government will continue to play an active role to support R and D in this field by providing the necessary infrastructure, facility, trained manpower and research funds. Additional e-beam accelerator is planned to be installed at Nuclear Malaysia in 2007. The medium energy electron beam accelerator (1 MeV, 50 mA) will be mainly use to evaluate the commercial viability for treating aqueous products such as wastewater. (author)

  1. Irradiated Beryllium Disposal Workshop, Idaho Falls, ID, May 29-30, 2002

    Energy Technology Data Exchange (ETDEWEB)

    Longhurst, Glen Reed; Anderson, Gail; Mullen, Carlan K; West, William Howard

    2002-07-01

    In 2001, while performing routine radioactive decay heat rate calculations for beryllium reflector blocks for the Advanced Test Reactor (ATR), it became evident that there may be sufficient concentrations of transuranic isotopes to require classification of this irradiated beryllium as transuranic waste. Measurements on samples from ATR reflector blocks and further calculations confirmed that for reflector blocks and outer shim control cylinders now in the ATR canal, transuranic activities are about five times the threshold for classification. That situation implies that there is no apparent disposal pathway for this material. The problem is not unique to the ATR. The High Flux Isotope Reactor at Oak Ridge National Laboratory, the Missouri University Research Reactor at Columbia, Missouri and other reactors abroad must also deal with this issue. A workshop was held in Idaho Falls Idaho on May 29-30, 2002 to acquaint stakeholders with these findings and consider a path forward in resolving the issues attendant to disposition of irradiated material. Among the findings from this workshop were (1) there is a real potential for the US to be dependent on foreign sources for metallic beryllium within about a decade; (2) there is a need for a national policy on beryllium utilization and disposition and for a beryllium coordinating committee to be assembled to provide guidance on that policy; (3) it appears it will be difficult to dispose of this material at the Waste Isolation Pilot Plant (WIPP) near Carlsbad, New Mexico due to issues of Defense classification, facility radioactivity inventory limits, and transportation to WIPP; (4) there is a need for a funded DOE program to seek resolution of these issues including research on processing techniques that may make this waste acceptable in an existing disposal pathway or allow for its recycle.

  2. 15 CFR 923.13 - Energy facility planning process.

    Science.gov (United States)

    2010-01-01

    ... facility planning process. The management program must contain a planning process for energy facilities... 15 Commerce and Foreign Trade 3 2010-01-01 2010-01-01 false Energy facility planning process. 923... affected public and private parties will be involved in the planning process. [61 FR 33806, June 28, 1996...

  3. Status of beryllium study for fusion in RF

    International Nuclear Information System (INIS)

    Khomutov, A.M.; Kupriyanov, I.B.; Markushkin, Yu.E.; Gervash, A.; Kolbasov, B.N.

    2004-01-01

    The main directions of research activities in the field of beryllium application science and technology carried out in Russia during 2001-2003 have been reviewed. The main results of these investigations have been highlighted. First wall and port-limier. The investigation on the actively cooled components with beryllium cladding is under progress objecting on the clarification of their ultimate thermo cycling capabilities. The study of behavior of bulk beryllium and the boundary region of the contact with the cooling structure under the intensive thermo cycling loading and neutron irradiation have been the object of consideration in particular. The works on the optimization and modification of industrial fabrication processes for commercial scaled production of beryllium tile were also under way. The influence of neutron irradiation. The new experimental data on the nuclear properties of several Russian beryllium grades has been obtained. The samples have been subjected to the high neutron dozes. The influence of low temperature (70-200degree C) neutron irradiation on the thermal conductivity has been examined in particular. The interrelations of the helium inventory and temperature of neutron irradiation with tritium release out of irradiated beryllium samples have been analyzed. The beryllium associated safety questions. The experiments on the modeling of normal working conditions and conditions imitating the plasma disruption events in ITER performance scenario have been continued. The new experimental information on the coefficient of pulverization of beryllium and the accumulation of deuterium in beryllium under the action of proton beam has been collected. The dependence of the reaction rate constant for the beryllium oxidation by the water vapor for different conditions has been analyzed. The compact, porous and powder beryllium samples have been tested at the wide range of temperature, pressure and duration of reaction with water vapor. The calculating

  4. The analysis of beryllium-copper diffusion joint after HHF test

    International Nuclear Information System (INIS)

    Guiniatouline, R.N.; Mazul, I.V.; Rubkin, S.Y.

    1995-01-01

    The development of beryllium-copper joints which can withstand to relevant ITER divertor conditions is one of the important tasks at present time. One of the main problem for beryllium-copperjoints, is the inter-metallic layers, the strength and life time of joints significantly depends from the width and contents of the intermetallic layers. The objective of this work is to study the diffusion joint of TGP-56 beryllium to OFHC copper after thermal response and thermocyclic tests with beryllium-copper mockup. The BEY test were performed at e-beam facility (EBTS, SNLA). The following methods were used for analyses: the roentgenographic analysis; X-ray spectrum analysis; the fracture graphic analysis. During the investigation the followed studies were done: the analysis of diffusion boundary Be-Cu, which was obtained at the crossection of one of the tiles, the analysis of the debonded surfaces of a few beryllium tiles and corresponding copper parts; the analysis of upper surface of one of the tiles after HHF tests. The results of this work have showed that: the joint roentgenographic and elements analyses indicated the following phases in the diffusion zone: Cu 2 Be (∼170 μm), CuBe (∼30μm), CuBe 2 (∼1 μm) and solid solution of copper in beryllium. The phases Cu 2 Be, CuBe and solid solution of copper in beryllium were indicated using quantitative microanalysis and phases CuBe, CuBe 2 , Cu, Be - by roentgenographic analysis; the source of fracture (initial crack) is located in the central part of the tiles, the crack caused by the influence of residual stresses during cooling of a mock-up after fabrication and developed under the conditions of slow elastic-plastic growing during the process of thermal fatigue testing. The analysis gives the important data about joint's quality and also may be used for any type of joints and its comparison for ITER applications

  5. Beryllium. Its minerals. Pt. 1

    International Nuclear Information System (INIS)

    Lires, O.A.; Delfino, C.A.; Botbol, J.

    1990-01-01

    With this work a series of reports begins, under the generic name 'Beryllium', related to several aspects of beryllium technology. The target is to update, with critical sense, current bibliographic material in order to be used in further applications. Some of the most important beryllium ores, the Argentine emplacement of their deposits and world occurrence are described. Argentine and world production, resources and reserves are indicated here as well. (Author) [es

  6. JET-ISX-B beryllium limiter experiment safety analysis report and operational safety requirements

    International Nuclear Information System (INIS)

    Edmonds, P.H.

    1985-09-01

    An experiment to evaluate the suitability of beryllium as a limiter material has been completed on the ISX-B tokamak. The experiment consisted of two phases: (1) the initial operation and characterization in the ISX experiment, and a period of continued operation to the specified surface fluence (10 22 atoms/cm 2 ) of hydrogen ions; and (2) the disassembly, decontamination, or disposal of the ISX facility. During these two phases of the project, the possibility existed for beryllium and/or beryllium oxide powder to be produced inside the vacuum vessel. Beryllium dust is a highly toxic material, and extensive precautions are required to prevent the release of the beryllium into the experimental work area and to prevent the contamination of personnel working on the device. Details of the health hazards associated with beryllium and the appropriate precautions are presented. Also described in appendixes to this report are the various operational safety requirements for the project

  7. Beryllium Metal Supply Options

    Science.gov (United States)

    1989-01-01

    Carcinogen Assessment Group (U.S. Environmental Protec- tion Agency, 1987). The National Institute of Occupational Safety and Health is currently examining...Epidemiology of beryllium intox- ication. Arch. Ind. Hyg. Occup . Med. 4:123-151. U. S. Environmental Protection Agency 1987. Health Assessment Document...1957 to 1961. He rejoined the Bureau of Mines in 1961 as the aluminum and bauxite commcdity specialist. In 1973 he became chief of the Division of

  8. Beryllium and zirconium

    International Nuclear Information System (INIS)

    Salesse, Marc

    1959-01-01

    Pure beryllium and zirconium, both isolated at about the same date but more than a century ago remained practically unused for eighty years. Fifteen years ago they were released from this state of inactivity by atomic energy, which made them into current metal a with an annual production which runs into tens of tons for the one and thousands for the other. The reasons for this promotion promise well for the future of the two metals, which moreover will probably find additional uses in other branches of industry. The attraction of beryllium and zirconium for atomic energy is easily explained. The curve of figure 1 gives the price per gram of uranium-235 as a function of enrichment: this price increases by about a factor of 3 on passing from natural uranium (0, 7 percent 235 U) to almost pure uranium-235. Because of their tow capture cross-section beryllium and zirconium make it possible, or at least easier, to use natural uranium and they thus enjoy an advantage the extent of which must be calculated for each reactor or fuel element project, but which is generally considerable. It will be seen later that this advantage should be based on figures which are even more favourable that would appear from the simple ratio 3 of the price of pure uranium- 235 contained in natural uranium. Reprint of a paper published in 'Industries Atomiques' - n. 1-2, 1959

  9. Characterization of shocked beryllium

    Directory of Open Access Journals (Sweden)

    Papin P.A.

    2012-08-01

    Full Text Available While numerous studies have investigated the low-strain-rate constitutive response of beryllium, the combined influence of high strain rate and temperature on the mechanical behavior and microstructure of beryllium has received limited attention over the last 40 years. In the current work, high strain rate tests were conducted using both explosive drive and a gas gun to accelerate the material. Prior studies have focused on tensile loading behavior, or limited conditions of dynamic strain rate and/or temperature. Two constitutive strength (plasticity models, the Preston-Tonks-Wallace (PTW and Mechanical Threshold Stress (MTS models, were calibrated using common quasi-static and Hopkinson bar data. However, simulations with the two models give noticeably different results when compared with the measured experimental wave profiles. The experimental results indicate that, even if fractured by the initial shock loading, the Be remains sufficiently intact to support a shear stress following partial release and subsequent shock re-loading. Additional “arrested” drive shots were designed and tested to minimize the reflected tensile pulse in the sample. These tests were done to both validate the model and to put large shock induced compressive loads into the beryllium sample.

  10. 9 CFR 590.540 - Spray process drying facilities.

    Science.gov (United States)

    2010-01-01

    ... 9 Animals and Animal Products 2 2010-01-01 2010-01-01 false Spray process drying facilities. 590.540 Section 590.540 Animals and Animal Products FOOD SAFETY AND INSPECTION SERVICE, DEPARTMENT OF..., Processing, and Facility Requirements § 590.540 Spray process drying facilities. (a) Driers shall be of a...

  11. Status of beryllium development for fusion applications

    International Nuclear Information System (INIS)

    Billone, M.C.; Donne, M.D.; Macaulay-Newcombe, R.G.

    1994-05-01

    Beryllium is a leading candidate material for the neutron multiplier of tritium breeding blankets and the plasma facing component of first wall and divertor systems. Depending on the application, the fabrication methods proposed include hot-pressing, hot-isostatic-pressing, cold isostatic pressing/sintering, rotary electrode processing and plasma spraying. Product forms include blocks, tubes, pebbles, tiles and coatings. While, in general, beryllium is not a leading structural material candidate, its mechanical performance, as well its performance with regard to sputtering, heat transport, tritium retention/release, helium-induced swelling and chemical compatibility, is an important consideration in first-wall/blanket design. Differential expansion within the beryllium causes internal stresses which may result in cracking, thereby affecting the heat transport and barrier performance of the material. Overall deformation can result in loading of neighboring structural material. Thus, in assessing the performance of beryllium for fusion applications, it is important to have a good database in all of these performance areas, as well as a set of properties correlations and models for the purpose of interpolation/extrapolation

  12. Defense Waste Processing Facility Recycle Stream Evaporation

    International Nuclear Information System (INIS)

    STONE, MICHAEL

    2006-01-01

    The Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS) stabilizes high level radioactive waste (HLW) by vitrification of the waste slurries. DWPF currently produces approximately five gallons of dilute recycle for each gallon of waste vitrified. This recycle stream is currently sent to the HLW tank farm at SRS where it is processed through the HLW evaporators with the concentrate eventually sent back to the DWPF for stabilization. Limitations of the HLW evaporators and storage space constraints in the tank farm have the potential to impact the operation of the DWPF and could limit the rate that HLW is stabilized. After an evaluation of various alternatives, installation of a dedicated evaporator for the DWPF recycle stream was selected for further evaluation. The recycle stream consists primarily of process condensates from the pretreatment and vitrification processes. Other recycle streams consist of process samples, sample line flushes, sump flushes, and cleaning solutions from the decontamination and filter dissolution processes. The condensate from the vitrification process contains some species, such as sulfate, that are not appreciably volatile at low temperature and could accumulate in the system if 100% of the evaporator concentrate was returned to DWPF. These species are currently removed as required by solids washing in the tank farm. The cleaning solutions are much higher in solids content than the other streams and are generated 5-6 times per year. The proposed evaporator would be required to concentrate the recycle stream by a factor of 30 to allow the concentrate to be recycled directly to the DWPF process, with a purge stream sent to the tank farm as required to prevent buildup of sulfate and similar species in the process. The overheads are required to meet stringent constraints to allow the condensate to be sent directly to an effluent treatment plant. The proposed evaporator would nearly de-couple the DWPF process from the

  13. Defense waste processing facility startup progress report

    International Nuclear Information System (INIS)

    Iverson, D.C.; Elder, H.H.

    1992-01-01

    The Savannah River Site (SRS) has been operating a nuclear fuel cycle since the 1950's to produce nuclear materials in support of the national defense effort. About 83 million gallons of high level waste produced since operation began have been consolidated into 33 million gallons by evaporation at the waste tank farm. The Department of Energy has authorized the construction of the Defense Waste Processing Facility (DWPF) to immobilize the waste as a durable borosilicate glass contained in stainless steel canisters, prior to emplacement in a federal repository. The DWPF is now mechanically complete and undergoing commissioning and run-in activities. Cold startup testing using simulated non-radioactive feeds is scheduled to begin in November 1992 with radioactive operation scheduled to begin in May 1994. While technical issues have been identified which can potentially affect DWPF operation, they are not expected to negatively impact the start of non-radioactive startup testing

  14. Operational experience of gamma radiation processing facility

    International Nuclear Information System (INIS)

    Patel, Nilesh

    2014-01-01

    Universal lSO-MED is now proud to announce an extension of its irradiation service for low-dose applications specifically in agriculture commodities, food and healthcare applications with the start of Gujarat Agro Radiation Processing Facility at Village: Bavla, Ahmedabad (A Government Enterprise) Operated, Maintained and Managed by Universal Medicap Ltd. Availability of hygienic, safe and nutritious food commodities is essential for any sustainable human development. Food stability is an important element of economic stability and self-reliance of a nation. Though the need to preserve food has been felt by the mankind since the time immemorial, it is even stronger in today's context. The rising population and increasing gap between demand and supply, agro-climatic conditions, in adequate post-harvest practices, seasonal nature of produce and long distances between production and consumption centers underscore the need to device improved conservation and preservation strategies

  15. Beryllium. Beryllium oxide, obtention and properties. Pt.4

    International Nuclear Information System (INIS)

    Lires, O.A.; Delfino, C.A.; Botbol, J.

    1991-01-01

    As a continuation of the 'Beryllium' series this work reviews several methods of high purity beryllia production. Diverse methods of obtention and purification from different beryllium compounds are described. Some chemical, mechanical and electrical properties related with beryllia obtention methods are summarized. (Author) [es

  16. Beryllium brazing considerations in CANDU fuel bundle manufacture

    International Nuclear Information System (INIS)

    Harmsen, J.; Pant, A.; Lewis, B.J.; Thompson, W.T.

    2010-01-01

    'Full text:' Appendages of CANDU fuel bundle elements are currently joined to zircaloy sheaths by vacuum beryllium brazing. Ongoing environmental and workplace concerns about beryllium combined with the continuous efforts by Cameco Fuel Manufacturing in its improvement process, initiated this study to find a substitute for pure beryllium. The presentation will review the necessary functionality of brazing alloy components and short list a series of alloys with the potential to duplicate the performance of pure beryllium. Modifications to current manufacturing processes based on in-plant testing will be discussed in relation to the use of these alloys. The presentation will conclude with a summary of the progress to date and further testing expected to be necessary.

  17. Status of beryllium development for fusion applications

    International Nuclear Information System (INIS)

    Billone, M.C.; Macaulay-Newcombe, R.G.

    1995-01-01

    Beryllium is a leading candidate material for the neutron multiplier of tritium breeding blankets and the plasma-facing component of first-wall and divertor systems. Depending on the application, the fabrication methods proposed include hot-pressing, hot-isostatic-pressing, cold-isostatic-pressing/sintering, rotary electrode processing and plasma spraying. Product forms include blocks, tubes, pebbles, tiles and coatings. While, in general, beryllium is not a leading structural material candidate, its mechanical performance, as well as its performance with regard to sputtering, heat transport, tritium retention/release, helium-induced swelling and chemical compatibility, is an important consideration in first-wall/blanket design. Differential expansion within the beryllium causes internal stresses which may result in cracking, thereby affecting the heat transport and barrier performance of the material. Overall deformation can result in loading of neighboring structural material. Thus, in assessing the performance of beryllium for fusion applications, it is important to have a good database in all of these performance areas, as well as a set of properties correlations and models for the purpose of interpolation/extrapolation.In this current work, the range of anticipated fusion operating conditions is reviewed. The thermal, mechanical, chemical compatibility, tritium retention/release, and helium retention/swelling databases are then reviewed for fabrication methods and fusion operating conditions of interest. Properties correlations and uncertainty ranges are also discussed. In the case of the more complex phenomena of tritium retention/release and helium-induced swelling, fundamental mechanisms and models are reviewed in more detail. Areas in which additional data are needed are highlighted, along with some trends which suggest ways of optimizing the performance of beryllium for fusion applications. (orig.)

  18. Radiation protection at radioisotope processing facilities

    International Nuclear Information System (INIS)

    Hillier, L.R.; Decaire, R.

    2002-01-01

    MDS Inc. is Canada's largest diversified health and life sciences company and provides health care services and products to prevent, diagnose and treat disease. MDS Nordion Inc. is a subsidiary of MDS Inc. and is located in Ottawa, Ontario. It provides much of the world's supply of radioisotopes used in nuclear medicine primarily to diagnose, but also to treat disease. MDS Nordion is composed of three major production divisions at its Ottawa location and serves customers in three major markets. These are primarily: radioisotopes used in nuclear medicine (Nuclear Medicine Division), radiation processing for sterilization of medical equipment and supplies, and food (Ion Technologies Division), and teletherapy equipment used in cancer treatment (Therapy Systems Division). MDS Nordion supplies customers in over 100 countries, exporting more than 95 percent of its product processed in Canada. Every year, 15 to 20 million diagnostic imaging tests are carried out in hospitals around the world, using radioisotopes supplied by MDS Nordion. In addition, 150 to 200 million cubic feet (that's enough to cover an entire CFL field - including the end zones - stacked over half a kilometer high) of single use medical products are sterilized using MDS Nordion supplied equipment. MDS Nordion receives medical isotopes from AECL, Chalk River Laboratories and processes the material to purify and quantify the radioisotope product. Sealed sources, comprised of cobalt 60, are supplied from CANDU reactors. Production processes include ventilated shielded cells with remote manipulators, gloveboxes and fumehoods, to effectively control the safety of the workplace and the environment, and to prevent contamination of the products. The facilities are highly regulated by the Canadian Nuclear Safety Commission (CNSC) for safety and environmental protection. Products are also regulated by Health Canada and the US-Food and Drug Administration (FDA). (author)

  19. Operational experience with the JET beryllium evaporators in the J1W test bed

    International Nuclear Information System (INIS)

    Peacock, A.T.; Dietz, K.J.; Israel, G.; Jensen, H.S.; Johnson, A.; Pick, M.A.; Saibene, G.; Sartori, R.

    1989-01-01

    Four beryllium evaporators were fitted onto the JET vessel during March 1989. These evaporators are planned to give the first introduction of beryllium into the JET machine to study the effect of using beryllium as a first wall material. Over 200 hours operational experience with such an evaporator had been gained on a test bed facility in which the evaporation rate, radial evaporant distribution and head operating temperature had been determined. The results obtained on this facility with two different heat materials, sintered S-65B and vacuum cast beryllium are described. The test vessel has also been used to conduct beryllium wall pumping experiments using the ''Langmuir effect''. The initial results of these experiments will be described. (author)

  20. Method for fabricating beryllium structures

    Science.gov (United States)

    Hovis, Jr., Victor M.; Northcutt, Jr., Walter G.

    1977-01-01

    Thin-walled beryllium structures are prepared by plasma spraying a mixture of beryllium powder and about 2500 to 4000 ppm silicon powder onto a suitable substrate, removing the plasma-sprayed body from the substrate and placing it in a sizing die having a coefficient of thermal expansion similar to that of the beryllium, exposing the plasma-sprayed body to a moist atmosphere, outgassing the plasma-sprayed body, and then sintering the plasma-sprayed body in an inert atmosphere to form a dense, low-porosity beryllium structure of the desired thin-wall configuration. The addition of the silicon and the exposure of the plasma-sprayed body to the moist atmosphere greatly facilitate the preparation of the beryllium structure while minimizing the heretofore deleterious problems due to grain growth and grain orientation.

  1. A graded approach to safety documentation at processing facilities

    International Nuclear Information System (INIS)

    Cowen, M.L.

    1992-01-01

    Westinghouse Savannah River Company (WSRC) has over 40 major Safety Analysis Reports (SARs) in preparation for non-reactor facilities. These facilities include nuclear material production facilities, waste management facilities, support laboratories and environmental remediation facilities. The SARs for these various projects encompass hazard levels from High to Low, and mission times from startup, through operation, to shutdown. All of these efforts are competing for scarce resources, and therefore some mechanism is required for balancing the documentation requirements. Three of the key variables useful for the decision making process are Depth of Safety Analysis, Urgency of Safety Analysis, and Resource Availability. This report discusses safety documentation at processing facilities

  2. Doped beryllium lanthanate crystals

    International Nuclear Information System (INIS)

    1974-01-01

    Monocrystals of doped beryllium lanthanate, Be 2 Lasub(2-2x)Zsub(2x)O 5 --where Z may be any rare earth, but preferably neodymium, and x may have values between 0.001 and 0.2, but preferably between 0.007 and 0.015-- are recommended as laser hosts. They are softer and may be grown at a lower temperature than Y 3 A1 5 O 12 :Nd (YAG:Nd). Their chemical composition and preparation are described. An example of an optically pumped laser apparatus with this type of monocrystal as laser host is presented

  3. Spectrochemical determination of beryllium and lithium in stream sediments

    International Nuclear Information System (INIS)

    Gallimore, D.L.; Hues, A.D.; Palmer, B.A.; Cox, L.E.; Simi, O.R.; Bieniewski, T.M.; Steinhaus, D.W.

    1979-11-01

    A spectrochemical method was developed to analyze 200 or more samples of stream sediments per day for beryllium and lithium. One part of ground stream sediment is mixed with two parts graphite-SiO 2 buffer, packed into a graphite electrode, and excited in a direct-current arc. The resulting emission goes to a 3.4-m, direct-reading, Ebert spectrograph. A desk-top computer system is used to record and process the signals, and to report the beryllium and lithium concentrations. The limits of detection are 0.2 μg/g for beryllium and 0.5 μg/g for lithium. For analyses of prepared reference materials, the relative standard deviations were 16% for determining 0.2 to 100 μg/g of beryllium and 15% for determining 0.5 to 500 μg/g of lithium. A correction is made for vanadium interference

  4. Pinellas Plant facts. [Products, processes, laboratory facilities

    Energy Technology Data Exchange (ETDEWEB)

    1986-09-01

    This plant was built in 1956 in response to a need for the manufacture of neutron generators, a principal component in nuclear weapons. The neutron generators consist of a miniaturized linear ion accelerator assembled with the pulsed electrical power supplies required for its operation. The ion accelerator, or neutron tube, requires ultra clean, high vacuum technology: hermetic seals between glass, ceramic, glass-ceramic, and metal materials: plus high voltage generation and measurement technology. The existence of these capabilities at the Pinellas Plant has led directly to the assignment of the lightning arrester connector, specialty capacitor, vacuum switch, and crystal resonator. Active and reserve batteries and the radioisotopically-powered thermoelectric generator draw on the materials measurement and controls technologies which are required to ensure neutron generator life. A product development and production capability in alumina ceramics, cermet (electrical) feedthroughs, and glass ceramics has become a specialty of the plant; the laboratories monitor the materials and processes used by the plant's commercial suppliers of ferroelectric ceramics. In addition to the manufacturing facility, a production development capability is maintained at the Pinellas Plant.

  5. Corrosion of beryllium oxide

    International Nuclear Information System (INIS)

    Elston, J.; Caillat, R.

    1958-01-01

    Data are reported on the volatilization rate of beryllium oxide in moist air depending on temperature and water vapour concentration. They are concerned with powder samples or sintered shapes of various densities. For sintered samples, the volatilization rate is very low under the following conditions: - temperature: 1300 deg. C, - water vapour concentration in moist air: 25 g/m 3 , - flow rate: 12 I/hour corresponding to a speed of 40 m/hour on the surface of the sample. For calcinated powders (1300 deg. C), grain growth has been observed under a stream of moist air at 1100 deg. C. For instance, grain size changes from 0,5 to at least 2 microns after 500 hours of exposure at this temperature. Furthermore, results data are reported on corrosion of sintered beryllium oxide in pressurized water. At 250 deg. C, under a pressure of 40 kg/cm 2 water is very slightly corrosive; however, internal strains are revealed. Finally, some features on the corrosion in liquid sodium are exposed. (author) [fr

  6. Waste Receiving and Processing Facility (WRAP) Drawing List

    International Nuclear Information System (INIS)

    WEIDERT, J.R.

    1999-01-01

    This supporting document delineates the process of identification, categorization, and/or classification of the WRAP facility drawings used to support facility operations and maintenance. This document provides a listing of those essential or safety related drawings which have been identified to date. All other WRAP facility drawings have been classified as general

  7. Thermal cycling tests of actively cooled beryllium copper joints

    Energy Technology Data Exchange (ETDEWEB)

    Roedig, M.; Duwe, R.; Linke, J.; Schuster, A.; Wiechers, B. [Forschungszentrum Juelich GmbH (Germany)

    1998-01-01

    Screening tests (steady state heating) and thermal fatigue tests with several kinds of beryllium-copper joints have been performed in an electron beam facility. Joining techniques under investigation were brazing with silver containing and silver-free braze materials, hot isostatic pressing (HIP) and diffusion bonding (hot pressing). Best thermal fatigue performance was found for the brazed samples. (author)

  8. 9 CFR 590.546 - Albumen flake process drying facilities.

    Science.gov (United States)

    2010-01-01

    ... 9 Animals and Animal Products 2 2010-01-01 2010-01-01 false Albumen flake process drying facilities. 590.546 Section 590.546 Animals and Animal Products FOOD SAFETY AND INSPECTION SERVICE... INSPECTION ACT) Sanitary, Processing, and Facility Requirements § 590.546 Albumen flake process drying...

  9. Physicochemical characteristics of aerosol particles generated during the milling of beryllium silicate ores: implications for risk assessment.

    Science.gov (United States)

    Stefaniak, Aleksandr B; Chipera, Steve J; Day, Gregory A; Sabey, Phil; Dickerson, Robert M; Sbarra, Deborah C; Duling, Mathew G; Lawrence, Robert B; Stanton, Marcia L; Scripsick, Ronald C

    2008-01-01

    Inhalation of beryllium dusts generated during milling of ores and cutting of beryl-containing gemstones is associated with development of beryllium sensitization and low prevalence of chronic beryllium disease (CBD). Inhalation of beryllium aerosols generated during primary beryllium production and machining of the metal, alloys, and ceramics are associated with sensitization and high rates of CBD, despite similar airborne beryllium mass concentrations among these industries. Understanding the physicochemical properties of exposure aerosols may help to understand the differential immunopathologic mechanisms of sensitization and CBD and lead to more biologically relevant exposure standards. Properties of aerosols generated during the industrial milling of bertrandite and beryl ores were evaluated. Airborne beryllium mass concentrations among work areas ranged from 0.001 microg/m(3) (beryl ore grinding) to 2.1 microg/m(3) (beryl ore crushing). Respirable mass fractions of airborne beryllium-containing particles were 80% in high-energy input areas (beryl melting, beryl grinding). Particle specific surface area decreased with processing from feedstock ores to drumming final product beryllium hydroxide. Among work areas, beryllium was identified in three crystalline forms: beryl, poorly crystalline beryllium oxide, and beryllium hydroxide. In comparison to aerosols generated by high-CBD risk primary production processes, aerosol particles encountered during milling had similar mass concentrations, generally lower number concentrations and surface area, and contained no identifiable highly crystalline beryllium oxide. One possible explanation for the apparent low prevalence of CBD among workers exposed to beryllium mineral dusts may be that characteristics of the exposure material do not contribute to the development of lung burdens sufficient for progression from sensitization to CBD. In comparison to high-CBD risk exposures where the chemical nature of aerosol

  10. 5. IEA International workshop on beryllium technology for fusion. Book of abstracts

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-07-01

    The collection includes the abstracts of reports presented to the 5-th IEA international workshop on beryllium technology for fusion. The themes of reports are as follows: status of beryllium technology for fusion in Russia; manufacturing and testing of Be armoured first wall mock-up for ITER; development of the process of diffusion welding of metals stainless steel-copper-beryllium into a single composite; some features of beryllium-laser beam interaction; the effect of irradiation dose on tritium and helium release from neutron irradiated beryllium; thermal properties of neutron irradiated Be{sub 12}Ti. The results of investigating the mechanical properties variation and swelling of beryllium under high temperature neutron irradiation are presented.

  11. 5. IEA International workshop on beryllium technology for fusion. Book of abstracts

    International Nuclear Information System (INIS)

    2001-01-01

    The collection includes the abstracts of reports presented to the 5-th IEA international workshop on beryllium technology for fusion. The themes of reports are as follows: status of beryllium technology for fusion in Russia; manufacturing and testing of Be armoured first wall mock-up for ITER; development of the process of diffusion welding of metals stainless steel-copper-beryllium into a single composite; some features of beryllium-laser beam interaction; the effect of irradiation dose on tritium and helium release from neutron irradiated beryllium; thermal properties of neutron irradiated Be 12 Ti. The results of investigating the mechanical properties variation and swelling of beryllium under high temperature neutron irradiation are presented [ru

  12. Status of beryllium R and D in Japan

    International Nuclear Information System (INIS)

    Kawamura, H.; Ishida, K.

    2004-01-01

    Recently, several R and D program of beryllium for fusion are being promoted in Japan and community of beryllium study is growing up. In the R and D area of beryllium for solid breeding blanket, major subjects are beryllide application for prototype reactor, lifetime evaluation of neutron multiplier, impurity effect of beryllium and recycling of irradiated beryllium. Especially, the study of beryllide application has significant progress in these two years. The basic properties such as tritium inventory, oxidation behavior, steam interaction for stoichiometric Be 12 Ti fabricated by HIP (Hot Isostatic Pressing) have been studied and some advantages against beryllium were made clear. For manufacturing technology development, phase diagram and ductility improvement have been studied. And, Be 12 Ti pebbles with the improved microstructure were successfully fabricated by Rotating Electrode Process. In order to enhance the R and D activities, the R and D network consisted of industries, universities and laboratories in all Japan have been organized. Many collaboration and information exchange strongly promotes the R and D and some projects for commercial application have been launched form these activities. Also international collaborative project such as IEA and ISTC have been launched or planned. Recent result of R and D in Japan is described on this paper. (author)

  13. Beryllium for first wall, limiter and divertor - a literature survey

    International Nuclear Information System (INIS)

    Schuster, A.; Smid, I.; Kny, E.

    1994-01-01

    A survey of the topical literature on beryllium as material for plasma interactive components in future fusion devices is given. The radiation damage which can be expected as a result of the neutron irradiation from ignited tokamak plasma is discussed. The response to high heat fluxes and simulation experiments in different test facilities are referred. Another focus will be on the material properties literature data, on joining techniques and on compatibility with other materials. The performance of a beryllium coated first wall at JET is reported. Some relevant literature on other candidate materials for plasma interactive components shall be considered

  14. Sintering of beryllium oxide; Frittage de l'oxyde de beryllium

    Energy Technology Data Exchange (ETDEWEB)

    Caillat, R; Pointud, R [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1955-07-01

    This study had for origin to find a process permitting to manufacture bricks of beryllium oxide of pure nuclear grade, with a density as elevated as possible and with standardized shape. The sintering under load was the technique kept for the manufacture of the bricks. Because of the important toxicity of the beryllium oxide, the general features for the preliminary study of the sintering, have been determined while using alumina. The obtained results will be able to act as general indication for ulterior studies with sintering under load. (M.B.) [French] Cette etude a eu pour origine la recherche d'un procede permettant de fabriquer industriellement des briques d'oxyde de beryllium nucleaireraent pures, de densite aussi elevee que possible et de forme standardisee. Le frittage sous charge fut la technique retenue pour la fabrication des briques. En raison de la grande toxicite de l'oxyde de beryllium, les caracteristiques generales du frittage, pour l'etude preliminaire, ont ete determine en utilisant de l'alumine. Les resultats obtenus pourront servir d'indication generale pour des etudes ulterieurs avec frittage sous charge. (M.B.)

  15. Sampling for Beryllium Surface Contamination using Wet, Dry and Alcohol Wipe Sampling

    Energy Technology Data Exchange (ETDEWEB)

    Kerr, Kent [Central Missouri State Univ., Warrensburg, MO (United States)

    2004-12-01

    This research project was conducted at the National Nuclear Security Administration's Kansas City Plant, operated by Honeywell Federal Manufacturing and Technologies, in conjunction with the Safety Sciences Department of Central Missouri State University, to compare relative removal efficiencies of three wipe sampling techniques currently used at Department of Energy facilities. Efficiencies of removal of beryllium contamination from typical painted surfaces were tested by wipe sampling with dry Whatman 42 filter paper, with water-moistened (Ghost Wipe) materials, and by methanol-moistened wipes. Test plates were prepared using 100 mm X 15 mm Pyrex Petri dishes with interior surfaces spray painted with a bond coat primer. To achieve uniform deposition over the test plate surface, 10 ml aliquots of solution containing 1 beryllium and 0.1 ml of metal working fluid were transferred to the test plates and subsequently evaporated. Metal working fluid was added to simulate the slight oiliness common on surfaces in metal working shops where fugitive oil mist accumulates over time. Sixteen test plates for each wipe method (dry, water, and methanol) were processed and sampled using a modification of wiping patterns recommended by OSHA Method 125G. Laboratory and statistical analysis showed that methanol-moistened wipe sampling removed significantly more (about twice as much) beryllium/oil-film surface contamination as water-moistened wipes (p< 0.001), which removed significantly more (about twice as much) residue as dry wipes (p <0.001). Evidence for beryllium sensitization via skin exposure argues in favor of wipe sampling with wetting agents that provide enhanced residue removal efficiency.

  16. Progress in production and characterization of RF beryllium for ITER fist wall

    International Nuclear Information System (INIS)

    Kupriyanov, I.B.; Nikolaev, G.N.; Khomutov, A.M.; Gervash, A.; Romanov, P.V.

    2007-01-01

    Full text of publication follows: Beryllium was chosen as the armor material for the Fist Wall and Port Limiter of ITER. As results of ITER EDA activities the reference beryllium grade S-65C (USA, Brush Wellman) and the DshG-200 grade (RF, Bochvar Institute) were selected as materials for armor of ITER FW. The main criterion of a choice of given grades was their highest resistance against High Heat Flux loading shown by results of thermal fatigue/shock test experiments carried out earlier in SNL (USA) and FZ Juelich (Germany). Recently in RF the new Be grade TGP-56FW has been proposed, which is a modified analogue of DshG-200 grade with improved properties, and also a low-waste manufacturing process for production of Be tiles for armor of ITER FW has been developed. In comparison with DshG-200 grade the main advantage of TGP-56FW grade consists in higher level of strength properties and plasticity as well as higher uniformity of physical-mechanical properties in transverse and longitudinal directions relatively pressing axis. In this paper the results of RF R and D activities in production and characterization of TGP- 56FW beryllium grade for FW armor are presented. The results of HHF tests (thermal fatigue/disruption simulation) on TSEFEY-M testing facility for full-scale beryllium tiles produced from TGP-56FW grade are also reported. A new in-pile integrated test experiment for Be mock-ups that combines thermal cycling (2x10 4 cycles at 0.5 MW/m 2 ) and neutron irradiation (about 0.6 dpa) is discussed.(authors)

  17. 10 CFR 70.64 - Requirements for new facilities or new processes at existing facilities.

    Science.gov (United States)

    2010-01-01

    ... postulated accidents that could lead to loss of safety functions. (5) Chemical protection. The design must... 10 Energy 2 2010-01-01 2010-01-01 false Requirements for new facilities or new processes at... Critical Mass of Special Nuclear Material § 70.64 Requirements for new facilities or new processes at...

  18. 40 CFR 52.279 - Food processing facilities.

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 3 2010-07-01 2010-07-01 false Food processing facilities. 52.279 Section 52.279 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) AIR PROGRAMS (CONTINUED) APPROVAL AND PROMULGATION OF IMPLEMENTATION PLANS California § 52.279 Food processing facilities. (a) The following regulations are disapproved...

  19. Fracture toughness of irradiated beryllium

    International Nuclear Information System (INIS)

    Beeston, J.M.

    1978-01-01

    The fracture toughness of nuclear grade hot-pressed beryllium upon irradiation to fluences of 3.5 to 5.0 x 10 21 n/cm 2 , E greater than 1 MeV, was determined. Procedures and data relating to a round-robin test contributing to a standard ASTM method for unirradiated beryllium are discussed in connection with the testing of irradiated specimens. A porous grade of beryllium was also irradiated and tested, thereby enabling some discrimination between the models for describing the fracture toughness behavior of porous beryllium. The fracture toughness of unirradiated 2 percent BeO nuclear grade beryllium was 12.0 MPa m/sup 1 / 2 /, which was reduced 60 percent upon irradiation at 339 K and testing at 295 K. The fracture toughness of a porous grade of beryllium was 13.1 MPa m/sup 1 / 2 /, which was reduced 68 percent upon irradiation and testing at the same conditions. Reasons for the reduction in fracture toughness upon irradiation are discussed

  20. Beryllium technology workshop, Clearwater Beach, Florida, November 20, 1991

    International Nuclear Information System (INIS)

    Longhurst, G.R.

    1991-12-01

    This report discusses the following topics: beryllium in the ITER blanket; mechanical testing of irradiated beryllium; tritium release measurements on irradiated beryllium; beryllium needs for plasma-facing components; thermal conductivity of plasma sprayed beryllium; beryllium research at the INEL; Japanese beryllium research activities for in-pile mockup tests on ITER; a study of beryllium bonding of copper alloy; new production technologies; thermophysical properties of a new ingot metallurgy beryllium product line; implications of beryllium:steam interactions in fusion reactors; and a test program for irradiation embrittlement of beryllium at JET

  1. Beryllium technology workshop, Clearwater Beach, Florida, November 20, 1991

    Energy Technology Data Exchange (ETDEWEB)

    Longhurst, G.R.

    1991-12-01

    This report discusses the following topics: beryllium in the ITER blanket; mechanical testing of irradiated beryllium; tritium release measurements on irradiated beryllium; beryllium needs for plasma-facing components; thermal conductivity of plasma sprayed beryllium; beryllium research at the INEL; Japanese beryllium research activities for in-pile mockup tests on ITER; a study of beryllium bonding of copper alloy; new production technologies; thermophysical properties of a new ingot metallurgy beryllium product line; implications of beryllium:steam interactions in fusion reactors; and a test program for irradiation embrittlement of beryllium at JET.

  2. Considerations about the licensing process of special nuclear industrial facilities

    Energy Technology Data Exchange (ETDEWEB)

    Talarico, M.A., E-mail: talaricomarco@hotmail.com [Marinha do Brasil, Rio de Janeiro, RJ (Brazil). Coordenacao do Porgrama de Submarino com Propulsao Nuclear; Melo, P.F. Frutuoso e [Coordenacao dos Programas de Pos-Graduacao em Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear

    2015-07-01

    This paper brings a discussion about the challenges involved in the development of a new kind of nuclear facility in Brazil, a naval base for nuclear submarines, with attention to the licensing process and considerations about the risk-informed decision making application to the licensing process. Initially, a model of such a naval base, called in this work, special industrial facility, is proposed, with its systems and respective sets of basic requirements, in order to make it possible the accomplishment of the special industrial facility support function to the nuclear submarine. A discussion about current challenges to overcome in this project is presented: the challenges due to the new characteristics of this type of nuclear facility; existence of several interfaces between the special industrial facilities systems and nuclear submarine systems in design activities; lack of specific regulation in Brazil to allow the licensing process of special industrial facilities by the nuclear safety authority; and comments about the lack of information from reference nuclear facilities, as is the case with nuclear power reactors (for example, the German Grafenrheinfeld nuclear plant is the reference plant for the Brazilian Angra 2 nuclear plant). Finally, in view of these challenges, an analysis method of special industrial facility operational scenarios to assist the licensing process is proposed. Also, considerations about the application of risk-informed decision making to the special industrial facility activity and licensing process in Brazil are presented. (author)

  3. Considerations about the licensing process of special nuclear industrial facilities

    International Nuclear Information System (INIS)

    Talarico, M.A.; Melo, P.F. Frutuoso e

    2015-01-01

    This paper brings a discussion about the challenges involved in the development of a new kind of nuclear facility in Brazil, a naval base for nuclear submarines, with attention to the licensing process and considerations about the risk-informed decision making application to the licensing process. Initially, a model of such a naval base, called in this work, special industrial facility, is proposed, with its systems and respective sets of basic requirements, in order to make it possible the accomplishment of the special industrial facility support function to the nuclear submarine. A discussion about current challenges to overcome in this project is presented: the challenges due to the new characteristics of this type of nuclear facility; existence of several interfaces between the special industrial facilities systems and nuclear submarine systems in design activities; lack of specific regulation in Brazil to allow the licensing process of special industrial facilities by the nuclear safety authority; and comments about the lack of information from reference nuclear facilities, as is the case with nuclear power reactors (for example, the German Grafenrheinfeld nuclear plant is the reference plant for the Brazilian Angra 2 nuclear plant). Finally, in view of these challenges, an analysis method of special industrial facility operational scenarios to assist the licensing process is proposed. Also, considerations about the application of risk-informed decision making to the special industrial facility activity and licensing process in Brazil are presented. (author)

  4. Beryllium R and D for blanket application

    Energy Technology Data Exchange (ETDEWEB)

    Dalle Donne, M.; Scaffidi-Argentina, F. [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany). Inst. fuer Neutronenphysik und Reaktortechnik; Longhurst, G.R. [Idaho National Engineering Lab., Idaho Falls (United States); Kawamura, H. [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

    1998-10-01

    The paper describes the main problems and the R and D for the beryllium to be used as neutron multiplier in blankets. As the four ITER partners propose to use beryllium in the form of pebbles for their DEMO relevant blankets (only the Russians consider the porous beryllium option as an alternative) and the ITER breeding blanket will use beryllium pebbles as well, the paper is mainly based on beryllium pebbles. Also the work on the chemical reactivity of fully dense and porous beryllium in contact with water steam is described, due to the safety importance of this point. (orig.) 29 refs.

  5. Beryllium R and D for blanket application

    International Nuclear Information System (INIS)

    Dalle Donne, M.; Scaffidi-Argentina, F.; Kawamura, H.

    1998-01-01

    The paper describes the main problems and the R and D for the beryllium to be used as neutron multiplier in blankets. As the four ITER partners propose to use beryllium in the form of pebbles for their DEMO relevant blankets (only the Russians consider the porous beryllium option as an alternative) and the ITER breeding blanket will use beryllium pebbles as well, the paper is mainly based on beryllium pebbles. Also the work on the chemical reactivity of fully dense and porous beryllium in contact with water steam is described, due to the safety importance of this point. (orig.)

  6. Beryllium R&D for blanket application

    Science.gov (United States)

    Donne, M. Dalle; Longhurst, G. R.; Kawamura, H.; Scaffidi-Argentina, F.

    1998-10-01

    The paper describes the main problems and the R&D for the beryllium to be used as neutron multiplier in blankets. As the four ITER partners propose to use beryllium in the form of pebbles for their DEMO relevant blankets (only the Russians consider the porous beryllium option as an alternative) and the ITER breeding blanket will use beryllium pebbles as well, the paper is mainly based on beryllium pebbles. Also the work on the chemical reactivity of fully dense and porous beryllium in contact with water steam is described, due to the safety importance of this point.

  7. 304 Concretion facility closure plan

    International Nuclear Information System (INIS)

    1990-04-01

    The Hanford Site, located northwest of Richland, Washington, houses reactors, chemical-separation systems, and related facilities used for the production of special nuclear materials. The 300 Area of the Hanford Site contains reactor fuel manufacturing facilities and several research and development laboratories. Recyclable scrap uranium Zircaloy-2 and copper silicon alloy, uranium-titanium alloy, beryllium/Zircaloy-2 alloy, and Zircaloy-2 chips and fines were secured in concrete billets in the 304 Concretion Facility, located in the 300 Area. The beryllium/Zircaloy-2 alloy and Zircaloy-2 chips and fines are designated as low-level radioactive mixed waste (LLRMW) with the characteristic of ignitability. The concretion process reduced the ignitability of the fines and chips for safe storage and shipment. This process has been discontinued and the 304 Concretion Facility is now undergoing closure as defined in the Resource Conservation and Recovery Act of 1976 (RCRA). This closure plan presents a description of the facility, the history of materials and wastes managed, and the procedures that will be followed to close the 304 Concretion Facility (304 Facility). Clean closure of the 304 Facility is the proposed method for closure of the facility. Justification for this proposal is presented. 15 refs., 22 figs., 4 tabs

  8. Automation in a material processing/storage facility

    International Nuclear Information System (INIS)

    Peterson, K.; Gordon, J.

    1997-01-01

    The Savannah River Site (SRS) is currently developing a new facility, the Actinide Packaging and Storage Facility (APSF), to process and store legacy materials from the United States nuclear stockpile. A variety of materials, with a variety of properties, packaging and handling/storage requirements, will be processed and stored at the facility. Since these materials are hazardous and radioactive, automation will be used to minimize worker exposure. Other benefits derived from automation of the facility include increased throughput capacity and enhanced security. The diversity of materials and packaging geometries to be handled poses challenges to the automation of facility processes. In addition, the nature of the materials to be processed underscores the need for safety, reliability and serviceability. The application of automation in this facility must, therefore, be accomplished in a rational and disciplined manner to satisfy the strict operational requirements of the facility. Among the functions to be automated are the transport of containers between process and storage areas via an Automatic Guided Vehicle (AGV), and various processes in the Shipping Package Unpackaging (SPU) area, the Accountability Measurements (AM) area, the Special Isotope Storage (SIS) vault and the Special Nuclear Materials (SNM) vault. Other areas of the facility are also being automated, but are outside the scope of this paper

  9. The Valduc waste incineration facility starts operations (iris process)

    International Nuclear Information System (INIS)

    Chateauvieux, H.; Guiberteuau, P.; Longuet, T.; Lannaud, J.; Lorich, M.

    1998-01-01

    In the operation of its facilities the Valduc Research Center produces alpha-contaminated solid waste and thus decided to build an incineration facility to treat the most contaminated combustible waste. The process selected for waste incineration is the IRIS process developed by the CEA at the Marcoule Nuclear Research Center. The Valduc Center asked SGN to build the incineration facility. The facility was commissioned in late 1996, and inactive waste incineration campaigns were run in 1997. The operator conducted tests with calibrated radioactive sources to qualify the systems for measuring holdup of active material from outside the equipment. Chlorinated waste incineration test runs were performed using the phosphatizing process developed by the Marcoule Research Center. Inspections performed after these incineration runs revealed the complete absence of corrosion in the equipment. Active commissioning of the facility is scheduled for mid-1998. The Valduc incinerator is the first industrial application of the IRIS process. (author)

  10. Saltstone studies using the scaled continuous processing facility

    Energy Technology Data Exchange (ETDEWEB)

    Fowley, M. D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Cozzi, A. D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Hansen, E. K. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-08-01

    The Savannah River National Laboratory (SRNL) has supported the Saltstone Facility since its conception with bench-scale laboratory experiments, mid-scale testing at vendor facilities, and consultations and testing at the Saltstone Facility. There have been minimal opportunities for the measurement of rheological properties of the grout slurry at the Saltstone Production Facility (SPF); thus, the Scaled Continuous Processing Facility (SCPF), constructed to provide processing data related to mixing, transfer, and other operations conducted in the SPF, is the most representative process data for determining the expected rheological properties in the SPF. These results can be used to verify the laboratory scale experiments that support the SPF using conventional mixing processes that appropriately represent the shear imparted to the slurry in the SPF.

  11. Defense Waste Processing Facility radioactive operations -- Part 2, Glass making

    International Nuclear Information System (INIS)

    Carter, J.T.; Rueter, K.J.; Ray, J.W.; Hodoh, O.

    1996-01-01

    The Savannah River Site's Defense Waste Processing Facility (DWPF) near Aiken, SC is the nation's first and world's largest vitrification facility. Following a ten year construction period and nearly 3 year non-radioactive test program, the DWPF began radioactive operations in March, 1996. The results of the first 8 months of radioactive operations are presented. Topics include facility production from waste preparation batching to canister filling

  12. An experimental facility for microwave induced plasma processing of materials

    International Nuclear Information System (INIS)

    Patil, D.S.; Ramachandran, K.; Bhide, A.L.; Venkatramani, N.

    1997-01-01

    Microwave induced plasma processing offers many advantages over conventional processes. However this technology is in the development stage. This report gives a detailed information about a microwave plasma processing facility (2.45 GHz, 700 W) set up in the Laser and Plasma Technology Division. The equipment details and the results obtained on deposition of diamond like carbon (DLC) thin films and surface modification of polymer PET (polyethylene terephthalate) using this facility are given in this report. (author)

  13. Control of DWPF [Defense Waste Processing Facility] melter feed composition

    International Nuclear Information System (INIS)

    Edwards, R.E. Jr.; Brown, K.G.; Postles, R.L.

    1990-01-01

    The Defense Waste Processing Facility will be used to immobilize Savannah River Site high-level waste into a stable borosilicate glass for disposal in a geologic repository. Proper control of the melter feed composition in this facility is essential to the production of glass which meets product durability constraints dictated by repository regulations and facility processing constraints dictated by melter design. A technique has been developed which utilizes glass property models to determine acceptable processing regions based on the multiple constraints imposed on the glass product and to display these regions graphically. This system along with the batch simulation of the process is being used to form the basis for the statistical process control system for the facility. 13 refs., 3 figs., 1 tab

  14. Modelisation and distribution of neutron flux in radium-beryllium source (226Ra-Be)

    Science.gov (United States)

    Didi, Abdessamad; Dadouch, Ahmed; Jai, Otman

    2017-09-01

    Using the Monte Carlo N-Particle code (MCNP-6), to analyze the thermal, epithermal and fast neutron fluxes, of 3 millicuries of radium-beryllium, for determine the qualitative and quantitative of many materials, using method of neutron activation analysis. Radium-beryllium source of neutron is established to practical work and research in nuclear field. The main objective of this work was to enable us harness the profile flux of radium-beryllium irradiation, this theoretical study permits to discuss the design of the optimal irradiation and performance for increased the facility research and education of nuclear physics.

  15. Investigations of the ternary system beryllium-carbon-tungsten and analyses of beryllium on carbon surfaces

    International Nuclear Information System (INIS)

    Kost, Florian

    2009-01-01

    Beryllium, carbon and tungsten are planned to be used as first wall materials in the future fusion reactor ITER. The aim of this work is a characterization of mixed material formation induced by thermal load. To this end, model systems (layers) were prepared and investigated, which give insight into the basic physical and chemical concepts. Before investigating ternary systems, the first step was to analyze the binary systems Be/C and Be/W (bottom-up approach), where the differences between the substrates PG (pyrolytic graphite) and HOPG (highly oriented pyrolytic graphite) were of special interest. Particularly X-ray photoelectron spectroscopy (XPS), low energy ion scattering (ISS) and Rutherford backscattering spectroscopy (RBS) were used as analysis methods. Beryllium evaporated on carbon shows an island growth mode, whereas a closed layer can be assumed for layer thicknesses above 0.7 nm. Annealing of the Be/C system induces Be 2 C island formation for T≥770 K. At high temperatures (T≥1170 K), beryllium carbide dissociates, resulting in (metallic) beryllium desorption. For HOPG, carbide formation starts at higher temperatures compared to PG. Activation energies for the diffusion processes were determined by analyzing the decreasing beryllium amount versus annealing time. Surface morphologies were characterized using angle-resolved XPS (ARXPS) and atomic force microscopy (AFM). Experiments were performed to study processes in the Be/W system in the temperature range from 570 to 1270 K. Be 2 W formation starts at 670 K, a complete loss of Be 2 W is observed at 1170 K due to dissociation (and subsequent beryllium desorption). Regarding ternary systems, particularly Be/C/W and C/Be/W were investigated, attaching importance to layer thickness (reservoir) variations. At room temperature, Be 2 C, W 2 C, WC and Be 2 W formation at the respective interfaces was observed. Further Be 2 C is forming with increasing annealing temperatures. Depending on the layer

  16. The nature of beryllium disease

    International Nuclear Information System (INIS)

    Williams, W.J.

    1977-01-01

    The increasing use of beryllium in modern industry poses a continuing health hazard with a real risk of producing incapacitating disease and even death. Beryllium and its salts are very toxic, even in small doses and may produce lesions in any organ. The majority of cases follow inhalation and may cause either acute or chronic lung disease. Acute pulmonary disease is a form of chemical pneumonitis while the chronic disease is characterised by the production of granulomas and fibrosis. The skin may be affected with the finding of dermatitis, acute or chronic ulceration. Other organs commonly involved include the liver and kidneys. The pathology of beryllium disease is not specific and diagnosis depends on satisfying the following criteria - history of exposure, consistent clinical, radiographic and pathological finding, presence of beryllium in tissue/fluid and evidence of hypersensitivity. Recent development of 'in vitro' tests of hypersensitivity may prove of value in both diagnosis and prevention of disease. Beryllium disease responds to steroid therapy but the only sure treatment is avoidance of exposure. (author)

  17. Westinghouse integrated cementation facility. Smart process automation minimizing secondary waste

    International Nuclear Information System (INIS)

    Fehrmann, H.; Jacobs, T.; Aign, J.

    2015-01-01

    The Westinghouse Cementation Facility described in this paper is an example for a typical standardized turnkey project in the area of waste management. The facility is able to handle NPP waste such as evaporator concentrates, spent resins and filter cartridges. The facility scope covers all equipment required for a fully integrated system including all required auxiliary equipment for hydraulic, pneumatic and electric control system. The control system is based on actual PLC technology and the process is highly automated. The equipment is designed to be remotely operated, under radiation exposure conditions. 4 cementation facilities have been built for new CPR-1000 nuclear power stations in China

  18. 304 Concretion Facility Closure Plan

    International Nuclear Information System (INIS)

    1991-10-01

    The Hanford Site, located northwest of Richland, Washington, houses reactors, chemical-separation systems, and related facilities used for the production of special nuclear materials. The 300 Area of the Hanford Site contains reactor fuel manufacturing facilities and several research and development laboratories. Recyclable scrap uranium with Zircaloy-2 and copper silicon allo , uranium-titanium alloy, beryllium/Zircaloy-2 alloy, and Zircaloy-2 chips and fines were secured in concrete billets (7.5-gal containers) in the 304 Concretion Facility (304 Facility), located in the 300 Area. The beryllium/Zircaloy-2 alloy and Zircaloy-2 chips and fines are designated as low-level radioactive mixed waste (LLRMW) with the characteristic of ignitability. The concretion process reduced the ignitability of the fines and chips for safe storage and shipment. This process has been discontinued and the 304 Concretion Facility is now undergoing closure as defined in the Resource Conservation and Recovery Act of 1976 (RCRA) and the Washington Administrative Code (WAC) Dangerous Waste Regulations, WAC 173-303-040 (Ecology 1991). This closure plan presents a description of the facility, the history of materials and wastes managed, and the procedures that will be followed to close the 304 Facility. The strategy for closure of the 304 Facility is presented in Section 6.0

  19. Characterization of Beryllium Windows for Coherent X-ray Optics

    International Nuclear Information System (INIS)

    Goto, Shunji; Yabashi, Makina; Tamasaku, Kenji; Ishikawa, Tetsuya

    2007-01-01

    Beryllium foils fabricated by several processes were characterized using spatially coherent x rays at 1-km beamline of SPring-8. By thickness dependence of bright x-ray spot density due to Fresnel diffraction from several-micron deficiencies, we found that speckles (bright x-ray spots) were due to voids with densities 103-104 mm-3 in powder foils and ingot foils. Compared with powder and ingot foils, a polished physical-vapor-deposited (PVD) beryllium foil gave highly uniform beams with no speckles. The PVD process eliminates the internal voids in principle and the PVD foil is the best for coherent x-ray applications

  20. Technical issues and solutions on ITER first wall beryllium application. Industrial viewpoint

    International Nuclear Information System (INIS)

    Iwadachi, T.; Uda, M.; Ito, M.; Miyakawa, M.; Ibuki, M.

    2004-01-01

    Beryllium is selected as reference armor material of ITER primary first wall and is joined to the copper alloy heat sink such as CuCrZr or Dispersion Strengthened Copper (DSCu) Various joining technologies have been successfully developed and the manufacturing possibilities of large size first wall panels with beryllium armor has been demonstrated. Based on such results, further technical improvement is needed to reduce manufacturing cost and ensure the reliability of joining in actual size first wall. The technical issues to optimize the fabrication process of beryllium attachment were shown in this paper from an industrial point of view. Determination of the optimum size and the surface qualities of beryllium tiles are important issues in term of the material specification to ensure joining reliability and to reduce cost. The consolidation method and the finish machining methods of beryllium tiles are also critical in terms of material cost. These items should be determined by paying concern to the accommodation of the joining methods. The selections of slitting methods for attached beryllium have a great influence on fabrication cost. In the actual fabrication of beryllium attachment, safety provisions for exposure to beryllium in working environment and the recycling of the waste from the fabrication processes will be concerned sufficiently. (author)

  1. High-strength beryllium block

    International Nuclear Information System (INIS)

    Pinto, N.P.; Keith, G.H.

    1977-01-01

    Beryllium billets hot isopressed using fine powder of high purity have exceptionally attractive properties; average tensile ultimate, 0.2% offset yield strength and elongation are 590 MPa, 430 MPa and 4.0% respectively. Properties are attributed to the fine grain size (about 4.0 μm average diameter) and the relatively low levels of BeO present as fine, well-dispersed particles. Dynamic properties, e.g., fracture toughness, are similar to those of standard grade, high-purity beryllium. The modulus of beryllium is retained to very high stress levels, and the microyield stress or precision elastic limit is higher than for other grades, including instrument grades. Limited data for billets made from normal-purity fine powders show similar room temperature properties. (author)

  2. Preparation of copper-beryllium alloys from Indian beryl

    International Nuclear Information System (INIS)

    Paul, C.M.; Sharma, B.P.; Subba Rao, K.S.; Rajadhyaksha, M.G.; Sundaram, C.V.

    1975-01-01

    The report presents the results of laboratory scale investigations on the preparation of copper-beryllium and aluminium-beryllium master alloys starting from Indian beryl and adopting the fluoride process. The flow-sheet involves : (1) conversion of the Be-values in beryl into water soluble sodium beryllium fluoride (2) preparation of beryllium hydroxide by alkali treatment of aqueous Na 2 BeF 4 (3) conversion of Be(OH) 2 to (NH 4 ) 2 BeF 4 by treatment with NH 4 HF 2 (4) thermal decomposition of (NH 4 ) 2 BeF 4 to BeF 2 and (5) magnesium reduction of BeF 2 (with the addition of copper/aluminium) to obtain beryllium alloys. The method has been successfully employed for the preparation of Cu-Be master alloys containing about 8% Be and free of Mg on a 200 gm scale. An overall Be-recovery of about 80% has been achieved. Al-8% Be master alloys have also been prepared by this method. Toxicity and health hazards associated with Be are discussed and the steps taken to ensure safe handling of Be are described. (author)

  3. Preparation of copper-beryllium alloys from Indian beryl

    International Nuclear Information System (INIS)

    Paul, C.M.; Sharma, B.P.; Subba Rao, K.S.; Rajadhyaksha, M.G.; Sundaram, C.V.

    1975-01-01

    The paper presents the results of laboratory-scale investigations on the preparation of copper-beryllium and aluminium beryllium master alloys starting from Indian beryl and adopting the fluoride process. The flowsheet involves: (1) conversion of the Be-values in beryl into water soluble sodium beryllium fluoride, (2) preparation of beryllium hydroxide by alkali treatment of aqueous Na 2 BeF 4 (3) conversion of Be(OH) 2 to (NH 4 ) 2 BeF 4 by treatment with NH 4 HF 2 (4) thermal decomposition of (NH 4 ) 2 BeF 4 to BeF 2 and (5) magnesium reduction of BeF 2 (without/with) the addition of copper/aluminium to obtain beryllium metal/alloys. The method has been successfully employed for the preparation of Cu-Be master alloys containing about 8% Be and free of Mg on a 200 gm scale. A1-80% Be master alloys have also been prepared by this method. Toxicity and health hazards associated with Be are discussed and the steps taken to ensure safe handling of Be are described. (author)

  4. Structure/property relationships in multipass GMA welding of beryllium.

    Energy Technology Data Exchange (ETDEWEB)

    Hochanadel, P. W. (Patrick W.); Hults, W. L. (William L.); Thoma, D. J. (Dan J.); Dave, V. R. (Vivek R.); Kelly, A. M. (Anna Marie); Pappin, P. A. (Pallas A.); Cola, M. J. (Mark J.); Burgardt, P. (Paul)

    2001-01-01

    Beryllium is an interesting metal that has a strength to weight ratio six times that of steel. Because of its unique mechanical properties, beryllium is used in aerospace applications such as satellites. In addition, beryllium is also used in x-ray windows because it is nearly transparent to x-rays. Joining of beryllium has been studied for decades (Ref.l). Typically joining processes include braze-welding (either with gas tungsten arc or gas metal arc), soldering, brazing, and electron beam welding. Cracking which resulted from electron beam welding was recently studied to provide structure/property relationships in autogenous welds (Ref. 2). Braze-welding utilizes a welding arc to melt filler, and only a small amount of base metal is melted and incorporated into the weld pool. Very little has been done to characterize the braze-weld in terms of the structure/property relationships, especially with reference to multipass welding. Thus, this investigation was undertaken to evaluate the effects of multiple passes on microstructure, weld metal composition, and resulting material properties for beryllium welded with aluminum-silicon filler metal.

  5. Low-level radioactive waste from rare metals processing facilities

    International Nuclear Information System (INIS)

    Eng, J.; Hendricks, D.W.; Feldman, J.; Giardina, P.A.

    1980-01-01

    This paper reviews the situations at the existing Teledyne Wah Chang Co., Inc. located at Albany, Oregon, and the former Carborundum Corp./Amax Specialty Metals, Inc., facilities located at Parkersburg, West Virginia, and Akron, New York, in order to show the extent of the radioactivity problem at rare metals processing facilities and the need to identify for radiological review other rare metal and rare earth processing sites

  6. SAFARI-1 research reactor beryllium reflector element replacement, management and relocation

    Energy Technology Data Exchange (ETDEWEB)

    Kock, Marisa De; Vlok, Jwh; Steynberg, B J [South Africa Atomic Energy Corporation (Necsa) (South Africa)

    2012-03-15

    decommissioned beryllium elements were also measured to investigate the occurrence of ageing mechanisms and their consequences, and to determine the location of the element's vulnerable areas. This supports the Beryllium Management Programme in terms of an optimal shuffling scheme to mitigate the effects of ageing. Currently, the decommissioned beryllium reflector elements are stored in a storage rack within the reactor pool. Due to limited space it is necessary to relocate these elements. A proposed relocation strategy based on existing literature is discussed. This strategy covers the requirements, relocation process and types of storage to be considered. (author)

  7. SAFARI-1 research reactor beryllium reflector element replacement, management and relocation

    International Nuclear Information System (INIS)

    Kock, Marisa De; Vlok, Jwh; Steynberg, B.J.

    2012-01-01

    decommissioned beryllium elements were also measured to investigate the occurrence of ageing mechanisms and their consequences, and to determine the location of the element's vulnerable areas. This supports the Beryllium Management Programme in terms of an optimal shuffling scheme to mitigate the effects of ageing. Currently, the decommissioned beryllium reflector elements are stored in a storage rack within the reactor pool. Due to limited space it is necessary to relocate these elements. A proposed relocation strategy based on existing literature is discussed. This strategy covers the requirements, relocation process and types of storage to be considered. (author)

  8. ASPECTS OF QMS IN RADIATION PROCESSING FACILITIES

    Directory of Open Access Journals (Sweden)

    LUNGU Ion Bogdan

    2016-05-01

    there are numerous factors that influence the effectiveness of service process, the objective is to analyze few important criteria like traceability, maintenance, and control of nonconform products and how it influence the desired results. The expected results show the necessity and utility of such an assessment in order to achieve improved results and implicitly, the increase in customer satisfaction.

  9. Capabilities for processing shipping casks at spent fuel storage facilities

    International Nuclear Information System (INIS)

    Baker, W.H.; Arnett, L.M.

    1978-01-01

    Spent fuel is received at a storage facility in heavily shielded casks transported either by rail or truck. The casks are inspected, cooled, emptied, decontaminated, and reshipped. The spent fuel is transferred to storage. The number of locations or space inside the building provided to perform each function in cask processing will determine the rate at which the facility can process shipping casks and transfer spent fuel to storage. Because of the high cost of construction of licensed spent fuel handling and storage facilities and the difficulty in retrofitting, it is desirable to correctly specify the space required. In this paper, the size of the cask handling facilities is specified as a function of rate at which spent fuel is received for storage. The minimum number of handling locations to achieve a given throughput of shipping casks has been determined by computer simulation of the process. The simulation program uses a Monte Carlo technique in which a large number of casks are received at a facility with a fixed number of handling locations in each process area. As a cask enters a handling location, the time to process the cask at that location is selected at random from the distribution of process time. Shipping cask handling times are based on experience at the General Electric Storage Facility, Morris, Illinois. Shipping cask capacity is based on the most recent survey available of the expected capability of reactors to handle existing rail or truck casks

  10. Gas bubble network formation in irradiated beryllium pebbles monitored by X-Ray micro-tomography

    Energy Technology Data Exchange (ETDEWEB)

    Bolier, E; Ferrero, C. [Forschungszentrum Karlsruhe, Zimer 203, Gebaeude 451, Abteilung HVT-TL (Germany); Moslang, A. [Forschungszentrum Karlsruhe GmbH, FZK, Karlsruhe (Germany); Pieritz, R.A. [CNRS, Lab. de Glaciologie et Geophysique de l' Environnement, 38 - Saint Martin d' Heres (France)

    2007-07-01

    Full text of publication follows: The efficient and safe operation of helium cooled ceramic breeder blankets requires among others an efficient tritium release during operation at blanket relevant temperatures. In the past out-of-pile thermal desorption studies on low temperature neutron irradiated beryllium have shown that tritium and helium release peaks occur together. This phenomenon can be interpreted in terms of growth and coalescence of helium bubbles and tritium that either is trapped inside the helium bubbles in form of T{sub 2} molecules or in their strain field. With increasing temperature the bubble density and size at grain interfaces increase together with the probability of interconnected porosities and channel formation to the outer surface, leading to simultaneous helium and tritium release peaks in TDS. For a reliable prediction of gas release up to end-of-life conditions at blanket relevant temperatures, knowledge of the dynamics of bubble growth and coalescence as well as the 3D distribution of bubble network formation is indispensable. Such data could also be used to experimentally validate any future model predictions of tritium and helium release rates. A high resolution computer aided micro-tomography (CMT) setup has been developed at the European Synchrotron Radiation Facility which allowed reconstructing 3-D images of beryllium pebbles without damaging them. By postprocessing the data a 3D rendering of inner surfaces and of interconnected channel networks can be obtained, thus allowing the identification of open porosities in neutron irradiated and tempered beryllium pebbles. In our case Beryllium pebbles of 2 mm diameter had been neutron irradiated in the 'Beryllium' experiment at 770 K with 1.24 x 10{sup 25} nxm{sup -2} resulting in 480 appm He and 12 appm Tritium. After annealing at 1500 K CMT was performed on the pebbles with 4.9 and 1.4 {mu}m voxel resolution, respectively, followed by morphological and topological post-processing

  11. Cause of pitting in beryllium

    International Nuclear Information System (INIS)

    Kershaw, R.P.

    1982-01-01

    Light microscopy, bare-film radiography, secondary ion mass spectroscopy, electron microprobe and physical testing were used to examine beryllium specimens exhibiting a stratified, pitted, pattern after chemical milling. The objective was to find the cause of this pattern. Specimens were found to have voids in excess of density specification allowances. These voids are attributed, at least in part, to the sublimation of beryllium fluoride during the vacuum hot pressing operation. The origin of the pattern is attributed to these voids and etching out of fines and associated impurities. Hot isostatic pressing with a subsequent heat treatment close residual porosity and dispersed impurities enough to correct the problem

  12. Bioenvironmental Engineering Guide to Beryllium

    Science.gov (United States)

    2017-07-26

    Dermal contact with beryllium can result in dermatitis resembling first- or second-degree burns and skin granulomas [7]. Beryllium dust, fume...minute short-term exposure limit (STEL) of 2.0 µg/m3 [§1910.1024(c)(2) & §1926.1124(c)(2)], and added provisions to prevent skin contact [§1910.1024(b...document you want more information, contact the Environmental, Safety, and Occupational Health (ESOH) Service Center at DSN 798-3764, 1-888-232-ESOH (3764

  13. Ninth Processing Campaign in the Waste Calcining Facility

    International Nuclear Information System (INIS)

    Childs, K.F.; Donovan, R.I.; Swenson, M.C.

    1982-04-01

    This report discusses the Ninth (and final) Processing Campaign at the Waste Calcining Facility. Several processing interruptions were experienced during this campaign and the emphasis of this report is on process and equipment performance with operating problems and corrective actions discussed in detail

  14. Design Criteria for Process Wastewater Pretreatment Facilities

    Science.gov (United States)

    1988-05-01

    Stripping Column H13 ’Re Purpose: The purpose of this report, is to provide design criteria for pretreatment needs for ’ I. INTRODUCTION ’". discharge of...which a portion of the vessel is filled with packing. Packing materials vary from corrugated steel to bundles of fibers (Langdon et al., 1972) to beds...concentration(s) using Table 20. Wastewater treatability studies should be considered as a process-screening tool for all wastewater streams for

  15. An Application of Business Process Management to Health Care Facilities.

    Science.gov (United States)

    Hassan, Mohsen M D

    The purpose of this article is to help health care facility managers and personnel identify significant elements of their facilities to address, and steps and actions to follow, when applying business process management to them. The ABPMP (Association of Business Process Management Professionals) life-cycle model of business process management is adopted, and steps from Lean, business process reengineering, and Six Sigma, and actions from operations management are presented to implement it. Managers of health care facilities can find in business process management a more comprehensive approach to improving their facilities than Lean, Six Sigma, business process reengineering, and ad hoc approaches that does not conflict with them because many of their elements can be included under its umbrella. Furthermore, the suggested application of business process management can guide and relieve them from selecting among these approaches, as well as provide them with specific steps and actions that they can follow. This article fills a gap in the literature by presenting a much needed comprehensive application of business process management to health care facilities that has specific steps and actions for implementation.

  16. Inhalation hazards from machining beryllium metal

    International Nuclear Information System (INIS)

    Hoover, M.D.; Finch, G.L.; Mewhinney, J.A.; Eidson, A.F.

    1987-01-01

    Beryllium metal has special nuclear and structural properties that make it useful for applications in fission and fusion reactor designs. Unfortunately, concerns for its toxicity have made designers wary of using beryllium metal. The work being reported here was undertaken to characterize the aerosols produced by two very common operations performed during preparation or modification of components for use in reactor systems: sawing and milling of beryllium metal. The study also covered beryllium metal alloys to allow comparison. Information from this study is to enable better assessments of the risk of using beryllium metal in reactor designs

  17. Comparative thermal cyclic test of different beryllium grades previously subjected to simulated disruption loads

    International Nuclear Information System (INIS)

    Gervash, A.; Giniyatulin, R.; Mazul, I.

    1999-01-01

    Considering beryllium as plasma facing armour this paper presents recent results obtained in Russia. A special process of joining beryllium to a Cu-alloy material structure is described and recent results of thermal cycling tests of such joints are presented. Summarizing the results, the authors show that a Cu-alloy heat sink structure armoured with beryllium can survive high heat fluxes (≥10 MW/m 2 ) during 1000 heating/cooling cycles without serious damage to the armour material and its joint. The principal feasibility of thermal cycling of beryllium grades and their joints directly in the core of a nuclear reactor is demonstrated and the main results of this test are presented. The paper also describes the thermal cycling of different beryllium grades having cracks initiated by previously applied high heat loads simulating plasma disruptions. (orig.)

  18. Opportunities for Process Monitoring Techniques at Delayed Access Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Curtis, Michael M.; Gitau, Ernest TN; Johnson, Shirley J.; Schanfein, Mark; Toomey, Christopher

    2013-09-20

    Except for specific cases where the International Atomic Energy Agency (IAEA) maintains a continuous presence at a facility (such as the Japanese Rokkasho Reprocessing Plant), there is always a period of time or delay between the moment a State is notified or aware of an upcoming inspection, and the time the inspector actually enters the material balance area or facility. Termed by the authors as “delayed access,” this period of time between inspection notice and inspector entrance to a facility poses a concern. Delayed access also has the potential to reduce the effectiveness of measures applied as part of the Safeguards Approach for a facility (such as short-notice inspections). This report investigates the feasibility of using process monitoring to address safeguards challenges posed by delayed access at a subset of facility types.

  19. Usability Briefing - a process model for healthcare facilities

    DEFF Research Database (Denmark)

    Fronczek-Munter, Aneta

    2014-01-01

    Background: In complex buildings with many types of users it can be difficult to satisfy the numerous, often contradictory requirements. Research in usability mostly focuses on evaluating products or facilities with users, after they were built. This paper is part of a PhD project “Usability...... with various users/stakeholders, using creative boundary objects at workshops.  Practical Implications: The research results have relevance to researchers, client organisations, facility managers and architects planning new complex facilities.  Research limitations: The proposed model is theoretical and needs...... briefing for hospitals”, where methods for capturing user needs and experiences at hospital facilities are investigated in order to feed into design processes and satisfy the users’ needs and maximise the effectiveness of facilities. Purpose: This paper introduces the concept of usability briefing...

  20. Method for removal of beryllium contamination from an article

    Science.gov (United States)

    Simandl, Ronald F.; Hollenbeck, Scott M.

    2012-12-25

    A method of removal of beryllium contamination from an article is disclosed. The method typically involves dissolving polyisobutylene in a solvent such as hexane to form a tackifier solution, soaking the substrate in the tackifier to produce a preform, and then drying the preform to produce the cleaning medium. The cleaning media are typically used dry, without any liquid cleaning agent to rub the surface of the article and remove the beryllium contamination below a non-detect level. In some embodiments no detectible residue is transferred from the cleaning wipe to the article as a result of the cleaning process.

  1. Sampling and analysis plan for assessment of beryllium in soils surrounding TA-40 building 15

    Energy Technology Data Exchange (ETDEWEB)

    Ruedig, Elizabeth [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-12-19

    Technical Area (TA) 40 Building 15 (40-15) is an active firing site at Los Alamos National Laboratory. The weapons facility operations (WFO) group plans to build an enclosure over the site in 2017, so that test shots may be conducted year-round. The enclosure project is described in PRID 16P-0209. 40-15 is listed on LANL OSH-ISH’s beryllium inventory, which reflects the potential for beryllium in/on soils and building surfaces at 40-15. Some areas in and around 40-15 have previously been sampled for beryllium, but past sampling efforts did not achieve complete spatial coverage of the area. This Sampling and Analysis Plan (SAP) investigates the area surrounding 40-15 via 9 deep (≥1-ft.) soil samples and 11 shallow (6-in.) soil samples. These samples will fill the spatial data gaps for beryllium at 40-15, and will be used to support OSH-ISH’s final determination of 40-15’s beryllium registry status. This SAP has been prepared by the Environmental Health Physics program in consultation with the Industrial Hygiene program. Industrial Hygiene is the owner of LANL’s beryllium program, and will make a final determination with regard to the regulatory status of beryllium at 40-15.

  2. Optimization of beryllium for fusion blanket applications

    International Nuclear Information System (INIS)

    Billone, M.C.

    1993-01-01

    The primary function of beryllium in a fusion reactor blanket is neutron multiplication to enhance tritium breeding. However, because heat, tritium and helium will be generated in and/or transported through beryllium and because the beryllium is in contact with other blanket materials, the thermal, mechanical, tritium/helium and compatibility properties of beryllium are important in blanket design. In particular, tritium retention during normal operation and release during overheating events are safety concerns. Accommodating beryllium thermal expansion and helium-induced swelling are important issues in ensuring adequate lifetime of the structural components adjacent to the beryllium. Likewise, chemical/metallurgical interactions between beryllium and structural components need to be considered in lifetime analysis. Under accident conditions the chemical interaction between beryllium and coolant and breeding materials may also become important. The performance of beryllium in fusion blanket applications depends on fabrication variables and operational parameters. First the properties database is reviewed to determine the state of knowledge of beryllium performance as a function of these variables. Several design calculations are then performed to indicate ranges of fabrication and operation variables that lead to optimum beryllium performance. Finally, areas for database expansion and improvement are highlighted based on the properties survey and the design sensitivity studies

  3. Status of RF beryllium characterization for ITER Fist Wall

    Science.gov (United States)

    Kupriyanov, I. B.; Nikolaev, G. N.; Roedig, M.; Gervash, A. А.; Linke, I. J.; Kurbatova, L. A.; Perevalov, S. I.; Giniyatulin, R. N.

    2011-10-01

    The status of RF R&D activities in production and characterization of TGP-56FW beryllium grade is presented. The results of metallographic studies of microstructure and cracks morphology are reported for full-scale Be tiles (56 × 56 × 10 mm) subjected to VDE simulation tests in TSEFEY-M testing facility (VDE-10 MJ/m 2 during 0.1 s, 1 shot ) and following low cycle thermal fatigue tests (500 thermal cycles at 1.5 MW/m 2). First results of plasma disruption tests ( E = 1.2-5 MJ/m 2, 5 ms), which were obtained during the realization of Thermal Shock/VDE Qualification program of RF beryllium in JUDITH-1 facility, are also discussed.

  4. Status of RF beryllium characterization for ITER Fist Wall

    International Nuclear Information System (INIS)

    Kupriyanov, I.B.; Nikolaev, G.N.; Roedig, M.; Gervash, A.A.; Linke, I.J.; Kurbatova, L.A.; Perevalov, S.I.; Giniyatulin, R.N.

    2011-01-01

    The status of RF R and D activities in production and characterization of TGP-56FW beryllium grade is presented. The results of metallographic studies of microstructure and cracks morphology are reported for full-scale Be tiles (56 x 56 x 10 mm) subjected to VDE simulation tests in TSEFEY-M testing facility (VDE-10 MJ/m 2 during 0.1 s, 1 shot) and following low cycle thermal fatigue tests (500 thermal cycles at 1.5 MW/m 2 ). First results of plasma disruption tests (E = 1.2-5 MJ/m 2 , 5 ms), which were obtained during the realization of Thermal Shock/VDE Qualification program of RF beryllium in JUDITH-1 facility, are also discussed.

  5. Galvanic corrosion of beryllium welds

    International Nuclear Information System (INIS)

    Hill, M.A.; Butt, D.P.; Lillard, R.S.

    1997-01-01

    Beryllium is difficult to weld because it is highly susceptible to cracking. The most commonly used filler metal in beryllium welds is Al-12 wt.% Si. Beryllium has been successfully welded using Al-Si filler metal with more than 30 wt.% Al. This filler creates an aluminum-rich fusion zone with a low melting point that tends to backfill cracks. Drawbacks to adding a filler metal include a reduction in service temperature, a lowering of the tensile strength of the weld, and the possibility for galvanic corrosion to occur at the weld. To evaluate the degree of interaction between Be and Al-Si in an actual weld, sections from a mock beryllium weldment were exposed to 0.1 M Cl - solution. Results indicate that the galvanic couple between Be and the Al-Si weld material results in the cathodic protection of the weld and of the anodic dissolution of the bulk Be material. While the cathodic protection of Al is generally inefficient, the high anodic dissolution rate of the bulk Be during pitting corrosion combined with the insulating properties of the Be oxide afford some protection of the Al-Si weld material. Although dissolution of the Be precipitate in the weld material does occur, no corrosion of the Al-Si matrix was observed

  6. The defense waste processing facility: A status report

    International Nuclear Information System (INIS)

    Cowan, S.P.; Fulmer, D.C.

    1987-01-01

    The Defense Waste Processing Fascility (DWPF) will be the nation's first production scale facility for immobilizing high-level waste for disposal. It will also be the largest facility of its kind in the world. The technology, design, and construction efforts are on schedule for ''hot'' operation in fiscal year 1990. This paper provides a status report on the DWPF technology, design, and construction, and describes some of the challenges that have arisen during design and construction

  7. Defense waste processing facility radioactive operations. Part 1 - operating experience

    International Nuclear Information System (INIS)

    Little, D.B.; Gee, J.T.; Barnes, W.M.

    1997-01-01

    The Savannah River Site's Defense Waste Processing Facility (DWPF) near Aiken, SC is the nation's first and the world's largest vitrification facility. Following a ten year construction program and a 3 year non-radioactive test program, DWPF began radioactive operations in March 1996. This paper presents the results of the first 9 months of radioactive operations. Topics include: operations of the remote processing equipment reliability, and decontamination facilities for the remote processing equipment. Key equipment discussed includes process pumps, telerobotic manipulators, infrared camera, Holledge trademark level gauges and in-cell (remote) cranes. Information is presented regarding equipment at the conclusion of the DWPF test program it also discussed, with special emphasis on agitator blades and cooling/heating coil wear. 3 refs., 4 figs

  8. Defense Waste Processing Facility -- Radioactive operations -- Part 3 -- Remote operations

    International Nuclear Information System (INIS)

    Barnes, W.M.; Kerley, W.D.; Hughes, P.D.

    1997-01-01

    The Savannah River Site's Defense Waste Processing Facility (DWPF) near Aiken, South Carolina is the nation's first and world's largest vitrification facility. Following a ten year construction period and nearly three years of non-radioactive testing, the DWPF began radioactive operations in March 1996. Radioactive glass is poured from the joule heated melter into the stainless steel canisters. The canisters are then temporarily sealed, decontaminated, resistance welded for final closure, and transported to an interim storage facility. All of these operations are conducted remotely with equipment specially designed for these processes. This paper reviews canister processing during the first nine months of radioactive operations at DWPF. The fundamental design consideration for DWPF remote canister processing and handling equipment are discussed as well as interim canister storage

  9. NASA Construction of Facilities Validation Processes - Total Building Commissioning (TBCx)

    Science.gov (United States)

    Hoover, Jay C.

    2004-01-01

    Key Atributes include: Total Quality Management (TQM) System that looks at all phases of a project. A team process that spans boundaries. A Commissioning Authority to lead the process. Commissioning requirements in contracts. Independent design review to verify compliance with Facility Project Requirements (FPR). Formal written Commissioning Plan with Documented Results. Functional performance testing (FPT) against the requirements document.

  10. Innovation process and innovativeness of facility management organizations

    NARCIS (Netherlands)

    Mudrak, T.; Wagenberg, van A.F.; Wubben, E.F.M.

    2005-01-01

    Purpose - The innovation patterns and processes in facility management (FM) organizations are crucial for the development of FM as a discipline, but they are not yet fully explored and understood. This paper aims to clarify FM innovation from the perspective of innovation processes and the

  11. Process control and dosimetry in a multipurpose irradiation facility

    Science.gov (United States)

    Cabalfin, E. G.; Lanuza, L. G.; Solomon, H. M.

    1999-08-01

    Availability of the multipurpose irradiation facility at the Philippine Nuclear Research Institute has encouraged several local industries to use gamma radiation for sterilization or decontamination of various products. Prior to routine processing, dose distribution studies are undertaken for each product and product geometry. During routine irradiation, dosimeters are placed at the minimum and maximum dose positions of a process load.

  12. Facility effluent monitoring plan for the Waste Receiving and Processing Facility Module 1

    International Nuclear Information System (INIS)

    Lewis, C.J.

    1995-10-01

    A facility effluent monitoring plan is required by the US Department of Energy in Order 5400.1 for any operations that involve hazardous materials and radioactive substances that could impact employee or public safety or the environment. This document is prepared using the specific guidelines identified in A Guide for Preparing Hanford Site Facility Effluent Monitoring Plans, WHC-EP-0438. This facility effluent monitoring plan assesses effluent monitoring systems and evaluates whether they are adequate to ensure the public health and safety as specified in applicable federal state, and local requirements. This facility effluent monitoring plan shall ensure lonq-range integrity of the effluent monitoring systems by requiring an update whenever a new process or operation introduces new hazardous materials or significant radioactive materials. This document must be reviewed annually even if there are no operational changes, and it must be updated as a minimum every three years

  13. Facility Effluent Monitoring Plan for the 325 Radiochemical Processing Laboratory

    International Nuclear Information System (INIS)

    Shields, K.D.; Ballinger, M.Y.

    1999-03-01

    This Facility Effluent Monitoring Plan (FEMP) has been prepared for the 325 Building Radiochemical Processing Laboratory (RPL) at the Pacific Northwest National Laboratory (PNNL) to meet the requirements in DOE Order 5400.1, ''General Environmental Protection Programs.'' This FEMP has been prepared for the RPL primarily because it has a ''major'' (potential to emit >0.1 mrem/yr) emission point for radionuclide air emissions according to the annual National Emission Standards for Hazardous Air Pollutants (NESHAP) assessment performed. This section summarizes the airborne and liquid effluents and the inventory based NESHAP assessment for the facility. The complete monitoring plan includes characterization of effluent streams, monitoring/sampling design criteria, a description of the monitoring systems and sample analysis, and quality assurance requirements. The RPL at PNNL houses radiochemistry research, radioanalytical service, radiochemical process development, and hazardous and radioactive mixed waste treatment activities. The laboratories and specialized facilities enable work ranging from that with nonradioactive materials to work with picogram to kilogram quantities of fissionable materials and up to megacurie quantities of other radionuclides. The special facilities within the building include two shielded hot-cell areas that provide for process development or analytical chemistry work with highly radioactive materials and a waste treatment facility for processing hazardous, mixed radioactive, low-level radioactive, and transuranic wastes generated by PNNL activities

  14. Facility Effluent Monitoring Plan for the 325 Radiochemical Processing Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Shields, K.D.; Ballinger, M.Y.

    1999-04-02

    This Facility Effluent Monitoring Plan (FEMP) has been prepared for the 325 Building Radiochemical Processing Laboratory (RPL) at the Pacific Northwest National Laboratory (PNNL) to meet the requirements in DOE Order 5400.1, ''General Environmental Protection Programs.'' This FEMP has been prepared for the RPL primarily because it has a ''major'' (potential to emit >0.1 mrem/yr) emission point for radionuclide air emissions according to the annual National Emission Standards for Hazardous Air Pollutants (NESHAP) assessment performed. This section summarizes the airborne and liquid effluents and the inventory based NESHAP assessment for the facility. The complete monitoring plan includes characterization of effluent streams, monitoring/sampling design criteria, a description of the monitoring systems and sample analysis, and quality assurance requirements. The RPL at PNNL houses radiochemistry research, radioanalytical service, radiochemical process development, and hazardous and radioactive mixed waste treatment activities. The laboratories and specialized facilities enable work ranging from that with nonradioactive materials to work with picogram to kilogram quantities of fissionable materials and up to megacurie quantities of other radionuclides. The special facilities within the building include two shielded hot-cell areas that provide for process development or analytical chemistry work with highly radioactive materials and a waste treatment facility for processing hazardous, mixed radioactive, low-level radioactive, and transuranic wastes generated by PNNL activities.

  15. Design ampersand construction innovations of the defense waste processing facility

    International Nuclear Information System (INIS)

    McKibben, J.M.; Pair, C.R.; Bethmann, H.K.

    1990-01-01

    Construction of the Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS) is essentially complete. The facility is designed to convert high-level radioactive waste, now contained in large steel tanks as aqueous salts and sludge, into solid borosilicate glass in stainless steel canisters. All processing of the radioactive material and operations in a radioactive environment will be done remotely. The stringent requirements dictated by remote operation and new approaches to the glassification process led to the development of a number of first-of-a-kind pieces of equipment, new construction fabrication and erection techniques, and new applications of old techniques. The design features and construction methods used in the vitrification building and its equipment were to accomplish the objective of providing a state-of-the-art vitrification facility. 3 refs., 10 figs

  16. OVERVIEW OF BERYLLIUM SAMPLING AND ANALYSIS

    Energy Technology Data Exchange (ETDEWEB)

    Brisson, M

    2009-04-01

    Because of its unique properties as a lightweight metal with high tensile strength, beryllium is widely used in applications including cell phones, golf clubs, aerospace, and nuclear weapons. Beryllium is also encountered in industries such as aluminium manufacturing, and in environmental remediation projects. Workplace exposure to beryllium particulates is a growing concern, as exposure to minute quantities of anthropogenic forms of beryllium may lead to sensitization and to chronic beryllium disease, which can be fatal and for which no cure is currently known. Furthermore, there is no known exposure-response relationship with which to establish a 'safe' maximum level of beryllium exposure. As a result, the current trend is toward ever lower occupational exposure limits, which in turn make exposure assessment, both in terms of sampling and analysis, more challenging. The problems are exacerbated by difficulties in sample preparation for refractory forms of beryllium, such as beryllium oxide, and by indications that some beryllium forms may be more toxic than others. This chapter provides an overview of sources and uses of beryllium, health risks, and occupational exposure limits. It also provides a general overview of sampling, analysis, and data evaluation issues that will be explored in greater depth in the remaining chapters. The goal of this book is to provide a comprehensive resource to aid personnel in a wide variety of disciplines in selecting sampling and analysis methods that will facilitate informed decision-making in workplace and environmental settings.

  17. The dominance of thermocicling treatments conditions on microplastic of spheral ground up powder beryllium

    International Nuclear Information System (INIS)

    Ivantsov, V.I.

    2003-01-01

    The dominance of thermocicling treatments conditions on structure,electrical resistance,microplastic characteristics and behavior in the microdeformation area of beryllium was investigated. It is established, that TC causes considerable structural changes in beryllium at the upper thermocicling temperature exceeding 523 K. Attached to upper TC temperature less then 523 K it is observed hardening of metal, conditioned by processes of twining and relaxation of micro stresses in overstrained micro volumes. The temperature range of TC determinate the stage evolution of the flow and the deformation hardening at different stages defined by changes in the beryllium structure caused by processing

  18. Mechanisms of hydrogen retention in metallic beryllium and beryllium oxide and properties of ion-induced beryllium nitride

    International Nuclear Information System (INIS)

    Oberkofler, Martin

    2011-01-01

    In the framework of this thesis laboratory experiments on atomically clean beryllium surfaces were performed. They aim at a basic understanding of the mechanisms occurring upon interaction of a fusion plasma with a beryllium first wall. The retention and the temperature dependent release of implanted deuterium ions are investigated. An atomistic description is developed through simulations and through the comparison with calculations based on density functional theory. The results of these investigations are compared to the behaviour of hydrogen upon implantation into thermally grown beryllium oxide layers. Furthermore, beryllium nitride is produced by implantation of nitrogen into metallic beryllium and its properties are investigated. The results are interpreted with regard to the use of beryllium in a fusion reactor. (orig.)

  19. Waste Receiving and Processing (WRAP) facility engineering study

    International Nuclear Information System (INIS)

    Christie, M.A.; Cammann, J.W.; McBeath, R.S.; Rode, H.H.

    1985-01-01

    A new Hanford waste management facility, the Waste Receiving and Processing (WRAP) facility (planned to be operational by FY 1994) will receive, inspect, process, and repackage contact-handled transuranic (CH-TRU) contaminated solid wastes. The wastes will be certified according to the waste acceptance criteria for disposal at the Waste Isolation Pilot Plant (WIPP) geologic repository in southeast New Mexico. Three alternatives which could cost effectively be applied to certify Hanford CH-TRU waste to the WIPP Waste Acceptance Criteria (WIPP-WAC) have been examined in this updated engineering study. The alternatives differed primarily in the reference processing systems used to transform nonconforming waste into an acceptable, certified waste form. It is recommended to include the alternative of shredding and immobilizing nonconforming wastes in cement (shred/grout processing) in the WRAP facility. Preliminary capital costs for WRAP in mid-point-of-construction (FY 1991) dollars were estimated at $45 million for new construction and $37 million for modification and installation in an existing Hanford surplus facility (231-Z Building). Operating, shipping, and decommissioning costs in FY 1986 dollars were estimated at $126 million, based on a 23-y WRAP life cycle (1994 to 2017). During this period, the WRAP facility will receive an estimated 38,000 m 3 (1.3 million ft 3 ) of solid CH-TRU waste. The study recommends pilot-scale testing and evaluation of the processing systems planned for WRAP and advises further investigation of the 231-Z Building as an alternative to new facility construction

  20. Development and experimental study of beryllium window for ITER radial X-ray camera

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Zhaoxi [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Jin, Guangxu [Materion Brush (United States); Chen, Kaiyun; Chen, Yebin; Song, Yuntao [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Hu, Liqun, E-mail: lqhu@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Niu, Luying; Sheng, Xiuli; Cheng, Yong; Lu, Kun [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China)

    2013-12-15

    Highlights: • The thickness of the beryllium foil is chosen as 80 μm to guarantee its safety under high pressure differential in accident events. • Using low purity of beryllium as the transition material, the effect of thermal stress caused by diffusion bonding process can be reduced. • Sealing ring and honeycomb-like supports are designed and used in the mechanical clamped beryllium window to enhance its sealing and safety performance. • The beryllium windows have good performance under severe working conditions like high temperature baking, vibration or impact load. -- Abstract: Radial X-ray camera (RXC) is a diagnostic device planned to be installed in the ITER Equatorial Port no. 12. Beryllium window will be installed between the inner and outer camera of RXC, which severs as the transmission photocathode substrate and also the vacuum isolation component. In this paper the design and manufacture process of two types of beryllium windows were introduced. Although 50 μm thickness of beryllium foil is the best choice, the 80 μm one with X-ray threshold of 1.34 keV was selected for safety consideration. Using the intermediate layer (low purity of beryllium) between the beryllium foil and the stainless steel base flange is an effective strategy to limit the welding thermal deformation and thermal stress of the thin foil caused by bonding between different materials. By using ANSYS software, the feasibility of the aperture design was analyzed and validated. Metal sealing ring was applied in the mechanical clamped beryllium window for its good stability under high temperature and neutron radiation. Although both of the hollow metal sealing ring with 0.03 mm silver coating and the pure silver sealing ring can satisfy the sealing requirement, the later one was chosen to produce the final product. Two hours 240 °C high temperature baking test, two hours 3.3 Hz vibration test and fatigue test were performed on the two types of beryllium windows. Based on the

  1. Copper-beryllium alloys for technical applications

    International Nuclear Information System (INIS)

    Heller, W.

    1976-01-01

    Data of physical properties are compiled for the most commonly used copper-beryllium alloys (CuBe 2, CuBe 1.7, CuCoBe, and CuCoAgBe), with emphasis on their temperature dependence and their variation with particular annealing and hardening treatments. The purpose is to provide a reference source and to indicate the versatility of these materials with respect to other copper alloys and to pure copper. The special features of CuBe alloys include high mechanical strength with reasonably high electrical conductivity, as well as good wear and corrosion resistance. For example, CuBe 2 has a yield strength of up to 1200 N/mm 2 , about three times that of pure copper, whilst the electrical conductivity of CuCoBe can be as high as 28 MS/m, nearly half that of pure copper. Typical applications are springs and electrical contacts. The importance of a proper heat treatment is discussed in some detail, notably the metallurgy and effects of low-temperature annealing (precipitation-hardening). A chapter on manufacturing processes covers machining, brazing, welding, and cleaning. This is followed by some remarks on safety precautions against beryllium poisoning. CuBe alloys are commercially available in the form of wires, strips, rods, and bars. Typical dimensions, specifications, a brief cost estimate, and addresses of suppliers are listed. (Author)

  2. Belgian research on fusion beryllium waste

    International Nuclear Information System (INIS)

    Druyts, F.; Mallants, D.; Sillen, X.; Iseghem, P. Van

    2004-01-01

    Future fusion power plants will generate important quantities of neutron irradiated beryllium. Although recycling is the preferred management option for this waste, this may not be technically feasible for all of the beryllium, because of its radiological characteristics. Therefore, at SCK·CEN, we initiated a research programme aimed at studying aspects of the disposal of fusion beryllium, including waste characterisation, waste acceptance criteria, conditioning methods, and performance assessment. One of the main issues to be resolved is the development of fusion-specific waste acceptance criteria for surface or deep geological disposal, in particular with regard to the tritium content. In case disposal is the only solution, critical nuclides can be immobilised by conditioning the waste. As a first approach to immobilising beryllium waste, we investigated the vitrification of beryllium. Corrosion tests were performed on both metallic and vitrified beryllium to provide source data for performance assessment. Finally, a first step in performance assessment was undertaken. (author)

  3. Beryllium minerals - demand strong for miniaturisation

    International Nuclear Information System (INIS)

    Griffiths, J.

    1985-01-01

    Beryllium is an essential constituent of over 40 minerals of which two are exploited commercially. Beryl is largely produced in the USSR and China and bertrandite in the U.S.A. Phenacite, from Canada, is also under investigation. The largest extraction plant for the recovery of beryllium in the western world is in Utah, U.S.A. and the company also produces beryllium oxide used in the manufacture of ceramics widely used in the electronics industry and for refractory articles. Beryllium-copper alloys in strip, rod and tube form are produced in the U.S.A., Germany and the U.K. Beryllium ceramics are important because of their high thermal conductivity, electrical insulation, strength and rigidity. The alloys, used as electric connectors, microswitch contacts are important for their high suitability for miniaturisation. The future growth potential for the beryllium industry is in the automotive industries in Europe and Japan. (U.K.)

  4. Method for hot pressing beryllium oxide articles

    Science.gov (United States)

    Ballard, Ambrose H.; Godfrey, Jr., Thomas G.; Mowery, Erb H.

    1988-01-01

    The hot pressing of beryllium oxide powder into high density compacts with little or no density gradients is achieved by employing a homogeneous blend of beryllium oxide powder with a lithium oxide sintering agent. The lithium oxide sintering agent is uniformly dispersed throughout the beryllium oxide powder by mixing lithium hydroxide in an aqueous solution with beryllium oxide powder. The lithium hydroxide is converted in situ to lithium carbonate by contacting or flooding the beryllium oxide-lithium hydroxide blend with a stream of carbon dioxide. The lithium carbonate is converted to lithium oxide while remaining fixed to the beryllium oxide particles during the hot pressing step to assure uniform density throughout the compact.

  5. Experimental results of beryllium exposed to intense high energy proton beam pulses

    CERN Document Server

    Ammigan, K; Hurh, P; Zwaska, R; Butcher, M; Guinchard, M; Calviani, M; Losito, R; Roberts, S; Kuksenko, V; Atherton, A; Caretta, O; Davenne, T; Densham, C; Fitton, M; Loveridge, J; O'Dell, J

    2017-01-01

    Beryllium is extensively used in various accelerator beam lines and target facilities as a material for beam windows, and to a lesser extent, as secondary particle production targets. With increasing beam intensities of future accelerator facilities, it is critical to understand the response of beryllium under extreme conditions to reliably operate these components as well as avoid compromising particle production efficiency by limiting beam parameters. As a result, an exploratory experiment at CERN’s HiRadMat facility was carried out to take advantage of the test facility’s tunable high intensity proton beam to probe and investigate the damage mechanisms of several beryllium grades. The test matrix consisted of multiple arrays of thin discs of varying thicknesses as well as cylinders, each exposed to increasing beam intensities. This paper outlines the experimental measurements, as well as findings from Post-Irradiation-Examination (PIE) work where different imaging techniques were used to analyze and co...

  6. Mechanical properties of irradiated beryllium

    International Nuclear Information System (INIS)

    Beeston, J.M.; Longhurst, G.R.; Wallace, R.S.

    1992-01-01

    Beryllium is planned for use as a neutron multiplier in the tritium breeding blanket of the International Thermonuclear Experimental Reactor (ITER). After fabricating samples of beryllium at densities varying from 80 to 100% of the theoretical density, we conducted a series of experiments to measure the effect of neutron irradiation on mechanical properties, especially strength and ductility. Samples were irradiated in the Advanced Test Reactor (ATR) to a neutron fluence of 2.6 x 10 25 n/m 2 (E > MeV) at an irradiation temperature of 75deg C. These samples were subsequently compression-tested at room temperature, and the results were compared with similar tests on unirradiated specimens. We found that the irradiation increased the strength by approximately four times and reduced the ductility to approximately one fourth. Failure was generally ductile, but the 80% dense irradiated samples failed in brittle fracture with significant generation of fine particles and release of small quantities of tritium. (orig.)

  7. Mechanical properties of irradiated beryllium

    Science.gov (United States)

    Beeston, J. M.; Longhurst, G. R.; Wallace, R. S.; Abeln, S. P.

    1992-10-01

    Beryllium is planned for use as a neutron multiplier in the tritium breeding blanket of the International Thermonuclear Experimental Reactor (ITER). After fabricating samples of beryllium at densities varying from 80 to 100% of the theoretical density, we conducted a series of experiments to measure the effect of neutron irradiation on mechanical properties, especially strength and ductility. Samples were irradiated in the Advanced Test Reactor (ATR) to a neutron fluence of 2.6 × 10 25 n/m 2 ( E > 1 MeV) at an irradiation temperature of 75°C. These samples were subsequently compression-tested at room temperature, and the results were compared with similar tests on unirradiated specimens. We found that the irradiation increased the strength by approximately four times and reduced the ductility to approximately one fourth. Failure was generally ductile, but the 80% dense irradiated samples failed in brittle fracture with significant generation of fine particles and release of small quantities of tritium.

  8. MEASUREMENTS OF THE PROPERTIES OF BERYLLIUM FOIL

    International Nuclear Information System (INIS)

    ZHAO, Y.; WANG, H.

    2000-01-01

    The electrical conductivity of beryllium at radio frequency (800 MHz) and liquid nitrogen temperature were investigated and measured. This summary addresses a collection of beryllium properties in the literature, an analysis of the anomalous skin effect, the test model, the experimental setup and improvements, MAFIA simulations, the measurement results and data analyses. The final results show that the conductivity of beryllium is not as good as indicated by the handbook, yet very close to copper at liquid nitrogen temperature

  9. New audio applications of beryllium metal

    International Nuclear Information System (INIS)

    Sato, M.

    1977-01-01

    The major applications of beryllium metal in the field of audio appliances are for the vibrating cones for the two types of speakers 'TWITTER' for high range sound and 'SQUAWKER' for mid range sound, and also for beryllium cantilever tube assembled in stereo cartridge. These new applications are based on the characteristic property of beryllium having high ratio of modulus of elasticity to specific gravity. The production of these audio parts is described, and the audio response is shown. (author)

  10. Thermal expansion of beryllium oxide

    International Nuclear Information System (INIS)

    Solodukhin, A.V.; Kruzhalov, A.V.; Mazurenko, V.G.; Maslov, V.A.; Medvedev, V.A.; Polupanova, T.I.

    1987-01-01

    Precise measurements of temperature dependence of the coefficient of linear expansion in the 22-320 K temperature range on beryllium oxide monocrystals are conducted. A model of thermal expansion is suggested; the range of temperature dependence minimum of the coefficient of thermal expansion is well described within the frames of this model. The results of the experiment may be used for investigation of thermal stresses in crystals

  11. Intoxication experiments with beryllium sulphate

    International Nuclear Information System (INIS)

    Bucurescu, I.; Stan, T.

    1990-01-01

    The changes in the particular number of animals in two groups of 40 rats each subjected to intoxication experiments with beryllium sulphate was investigated. The two investigations had very different characteristics. In the case of chronic intoxication there was a marked lethality over given time intervals. In the case of subacute intoxication the number of animals decreased with time. It was found empirically that this change can be described by an exponential relationship which lends itself to statistical interpretation. (author)

  12. Thermodynamic properties of beryllium hydroxide

    International Nuclear Information System (INIS)

    Baur, A.; Lecocq, A.

    1964-01-01

    The study of the hydro-thermal decomposition of beryllium hydroxide has made it possible to determine the free energy of formation and the entropy. The results obtained are in good agreement with the theoretical values calculated from the solubility product of this substance. They give furthermore the possibility of acquiring a better understanding of the BeO-H 2 O-Be (OH) 2 system between 20 and 1500 C. (authors) [fr

  13. Hanford Site Beryllium Program: Past, Present, and Future - 12428

    Energy Technology Data Exchange (ETDEWEB)

    Fisher, Mark [CH2M Hill Plateau Remediation Company, Richland, Washington 99354 (United States); Garcia, Pete [U.S. Department of Energy - Richland Office, Richland, Washington 99352 (United States); Goeckner, Julie [U.S. Department of Energy - HQ, EMCBC, Cincinnati, Ohio 45202 (United States); Millikin, Emily [Washington Closure Hanford, Richland, Washington 99354 (United States); Stoner, Mike [Mission Support Alliance, Richland, Washington 99354 (United States)

    2012-07-01

    The U.S. Department of Energy (DOE) has a long history of beryllium use because of the element's broad application to many nuclear operations and processes. At the Hanford Site beryllium alloy was used to fabricate parts for reactors, including fuel rods for the N-Reactor during plutonium production. Because of continued confirmed cases of chronic beryllium disease (CBD), and data suggesting CBD occurs at exposures to low-level concentrations, the DOE decided to issue a rule to further protect federal and contractor workers from hazards associated with exposure to beryllium. When the beryllium rule was issued in 1999, each of the Hanford Site contractors developed a Chronic Beryllium Disease Prevention Program (CBDPP) and initial site wide beryllium inventories. A new site-wide CBDPP, applicable to all Hanford contractors, was issued in May, 2009. In the spring of 2010 the DOE Headquarters Office of Health, Safety, and Security (HSS) conducted an independent inspection to evaluate the status of implementation of the Hanford Site Chronic Beryllium Disease Prevention Program (CBDPP). The report identified four Findings and 12 cross-cutting Opportunities for Improvement (OFIs). A corrective action plan (CAP) was developed to address the Findings and crosscutting OFIs. The DOE directed affected site contractors to identify dedicated resources to participate in development of the CAP, along with involving stakeholders. The CAP included general and contractor-specific recommendations. Following initiation of actions to implement the approved CAP, it became apparent that additional definition of product deliverables was necessary to assure that expectations were adequately addressed and CAP actions could be closed. Consequently, a supplement to the original CAP was prepared and transmitted to DOE-HQ for approval. Development of the supplemental CAP was an eight month effort. From the onset a core group of CAP development members were identified to develop a mechanism

  14. Thermal Properties of Beryllium Metal

    International Nuclear Information System (INIS)

    Cho, Tae Won; Baek, Je Kyun; Jeong, Gwan Yoon; Kim, Ji Hyeon; Sohn, Dong Seong

    2013-01-01

    It is known that the presence of as-fabricated porosity largely affect thermal conductivity of beryllium. Therefore, in this paper we will suggest a new thermal conductivity equation which consider volume fraction and discuss how this can be applied to irradiation induced degradation of thermal conductivity later. This study was performed to develop a new correlation of thermal conductivity of Beryllium. Although there are many factors like BeO contents, impurity level, grain size, and porosity, we assumed porosity will be the dominant factor for thermal conductivity. Therefore, a new correlation which consider volume fraction by applying Maxwell-Eucken equation is developed and this is consistent to some degrees. However, increasing impurity level and decreasing grain size will decrease thermal conductivity. Therefore, we need to consider their effects although we assume BeO contents, impurity, and grain size do not make noticeable effects in the future. Furthermore, thermal conductivity degradation by neutron irradiation should be considered afterward. There are two main factors for the thermal conductivity degradation: the one is defects formed by neutron collisions and the other is helium generated by transmutation of Be. It is known that they make a considerable degradation of conductivity. Beryllium is known there are considerable volume increases by helium accumulation. Therefore, we anticipate our suggested model can be applicable if it has been developed furthermore considering irradiation induced swelling

  15. (Beryllium). Internal Report No. 137, Jan. 15, 1958

    International Nuclear Information System (INIS)

    Mouret, P.; Rigaud, A.

    1959-01-01

    After a brief summary of the physical and chemical properties of beryllium, the various chemical treatments which can be applied to beryllium minerals either directly or after a physical enrichment are discussed. These various treatments give either the hydroxide or beryllium salts, from which either beryllium oxide or metallic beryllium can easily be obtained. The purification, analysis and uses of beryllium are also briefly discussed. (author)

  16. Structural design considerations for a radwaste processing facility

    International Nuclear Information System (INIS)

    Foelber, S.C.; Sabbe, M.A.

    1985-01-01

    The structural engineer needs to consider several criteria when designing a radioactive-waste processing facility in order to properly balance the requirements of safety and economy. This paper addresses the design criteria and structural design of a vitrification building and the special equipment and supports associated with remote process operations. In addition, approaches to construction, and the role of scale models to aid in engineering design and construction are discussed. 5 figures

  17. Evaluation of mercury in the liquid waste processing facilities

    Energy Technology Data Exchange (ETDEWEB)

    Jain, Vijay [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Shah, Hasmukh [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Occhipinti, John E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Wilmarth, William R. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Edwards, Richard E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-08-13

    This report provides a summary of Phase I activities conducted to support an Integrated Evaluation of Mercury in Liquid Waste System (LWS) Processing Facilities. Phase I activities included a review and assessment of the liquid waste inventory and chemical processing behavior of mercury using a system by system review methodology approach. Gaps in understanding mercury behavior as well as action items from the structured reviews are being tracked. 64% of the gaps and actions have been resolved.

  18. 324 Facility B-Cell quality process plan

    International Nuclear Information System (INIS)

    Carlson, J.L.

    1998-01-01

    This report documents the quality process plan for the restart of a hot cell in the B Plant, originally a bismuth phosphate processing facility, but later converted to a waste fractionation plant. B-Cell is currently being cleaned out and deactivated. TPA Milestone M-89-02 dictates that all mixed waste and equipment be removed from B-Cell by 5/31/1999. This report describes the major activities that remain for completion of the TPA milestone

  19. Beryllium dust generation resulting from plasma bombardment

    International Nuclear Information System (INIS)

    Doerner, R.; Mays, C.

    1997-01-01

    The beryllium dust resulting from erosion of beryllium samples subjected to plasma bombardment has been measured in PISCES-B. Loose surface dust was found to be uniformly distributed throughout the device and accounts for 3% of the eroded material. A size distribution measurement of the loose surface dust shows an increasing number of particles with decreasing diameter. Beryllium coatings on surfaces with a line of sight view of the target interaction region account for an additional 33% of the eroded beryllium material. Flaking of these surface layers is observed and is thought to play a significant role in dust generation inside the vacuum vessel. (orig.)

  20. Microplasticity in hot-pressed beryllium

    International Nuclear Information System (INIS)

    Plane, D.C.; Bonfield, W.

    1977-01-01

    Closed hysteresis loops measured in the microstrain region of hot pressed, commercially pure, polycrystalline beryllium are correlated with a dislocation - impurity atom, energy dissipating mechanism. (author)

  1. Beryllium thin films for resistor applications

    Science.gov (United States)

    Fiet, O.

    1972-01-01

    Beryllium thin films have a protective oxidation resistant property at high temperature and high recrystallization temperature. However, the experimental film has very low temperature coefficient of resistance.

  2. Defense waste processing facility project at the Savannah River Plant

    International Nuclear Information System (INIS)

    Baxter, R.G.; Maher, R.; Mellen, J.B.; Shafranek, L.F.; Stevens, W.R. III.

    1984-01-01

    The Du Pont Company is building for the Department of Energy a facility to vitrify high-level waste at the Savannah River Plant near Aiken, South Carolina. The Defense Waste Processing Facility (DWPF) will solidify existing and future radioactive wastes produced by defense activities at the site. At the present time engineering and design are 45% complete, the site has been cleared, and startup is expected in 1989. This paper will describe project status as well as features of the design. 9 figures

  3. Considerations in setting up and planning a graft processing facility.

    Science.gov (United States)

    Koh, Mickey B C

    2017-12-01

    The graft processing facility forms one of the core components of a clinical haematopoietic stem cell transplant program. The quality of a graft is instrumental in leading to consistent and reproducible outcomes of engraftment and other parameters. As such, meticulous planning and consideration is required and will include core elements including physical design and clinical correlates. The successful running of such a facility depends on an overarching quality program and adherence to local and international regulatory guidelines. Copyright © 2017 King Faisal Specialist Hospital & Research Centre. Published by Elsevier B.V. All rights reserved.

  4. Solid radioactive waste processing facility of the NPP Leningrad

    International Nuclear Information System (INIS)

    Weichard, Swetlana

    2008-01-01

    On behalf of the Russian Company Rosenergoatom NUKEM Technologies GmbH is planning and constructing a complete facility for the processing of solid low- and medium-active radioactive wastes. The NPP Leningrad comprises 4 units of RBMK-1000 reactors, the plant life has been extended by 15 years, the first unit is to be decommissioned in 2018. The construction of four new units is planned. NUKEM is in charge of planning, manufacture, construction and startup of the following facilities: sorting, internal transport, combustion and waste gas cleaning, emission surveillance, compacting, packaging and radiological measurement.

  5. Remote process connectors for the new waste calcining facility

    International Nuclear Information System (INIS)

    Jacobs, R.T.; Carter, J.A.; Hohback, A.C.

    1978-01-01

    The remote process connectors developed, used, and tested at the Remote Maintenance Development Facility are described. These connectors, including the three-bolt kinematic-graphite flange and watertight electrical connectors, are assembled on master jigs (holding-welding fixture) to form interchangeable pump and valve loop assemblies. These assemblies, with their guide-in platforms, make possible a method of performing remote maintenance at the New Waste Calcining Facility which is a departure from methods that until now have been the standard of the industry

  6. Technical evaluation of proposed Ukrainian Central Radioactive Waste Processing Facility

    International Nuclear Information System (INIS)

    Gates, R.; Glukhov, A.; Markowski, F.

    1996-06-01

    This technical report is a comprehensive evaluation of the proposal by the Ukrainian State Committee on Nuclear Power Utilization to create a central facility for radioactive waste (not spent fuel) processing. The central facility is intended to process liquid and solid radioactive wastes generated from all of the Ukrainian nuclear power plants and the waste generated as a result of Chernobyl 1, 2 and 3 decommissioning efforts. In addition, this report provides general information on the quantity and total activity of radioactive waste in the 30-km Zone and the Sarcophagus from the Chernobyl accident. Processing options are described that may ultimately be used in the long-term disposal of selected 30-km Zone and Sarcophagus wastes. A detailed report on the issues concerning the construction of a Ukrainian Central Radioactive Waste Processing Facility (CRWPF) from the Ukrainian Scientific Research and Design institute for Industrial Technology was obtained and incorporated into this report. This report outlines various processing options, their associated costs and construction schedules, which can be applied to solving the operating and decommissioning radioactive waste management problems in Ukraine. The costs and schedules are best estimates based upon the most current US industry practice and vendor information. This report focuses primarily on the handling and processing of what is defined in the US as low-level radioactive wastes

  7. Facility Effluent Monitoring Plan for the Waste Receiving and Processing (WRAP) Facility

    Energy Technology Data Exchange (ETDEWEB)

    DAVIS, W.E.

    2000-03-08

    A facility effluent monitoring plan is required by the U.S. Department of Energy in Order 5400.1 for any operations that involve hazardous materials and radioactive substances that could impact employee public safety, or the environment. This facility effluent monitoring plan assesses effluent monitoring systems and evaluates whether these systems are adequate to ensure the public health and safety as specified in applicable federal, state, and local requirements. This facility effluent monitoring plan ensures long-range integrity of the effluent monitoring systems by requiring an update whenever a new process or operation introduces new hazardous materials or significant radioactive materials. This document must be reviewed annually even if there are no operational changes, and must be updated, as a minimum, every 3 years.

  8. Facility Effluent Monitoring Plan for the Waste Receiving and Processing (WRAP) Facility

    International Nuclear Information System (INIS)

    DAVIS, W.E.

    2000-01-01

    A facility effluent monitoring plan is required by the U.S. Department of Energy in Order 5400.1 for any operations that involve hazardous materials and radioactive substances that could impact employee public safety, or the environment. This facility effluent monitoring plan assesses effluent monitoring systems and evaluates whether these systems are adequate to ensure the public health and safety as specified in applicable federal, state, and local requirements. This facility effluent monitoring plan ensures long-range integrity of the effluent monitoring systems by requiring an update whenever a new process or operation introduces new hazardous materials or significant radioactive materials. This document must be reviewed annually even if there are no operational changes, and must be updated, as a minimum, every 3 years

  9. Data triggered data processing at the Mirror Fusion Test Facility

    International Nuclear Information System (INIS)

    Jackson, R.J.; Balch, T.R.; Preckshot, G.G.

    1986-01-01

    A primary characteristic of most batch systems is that the input data files must exist before jobs are scheduled. On the Mirror Fusion Test Facility (MFTF-B) at Lawrence Livermore National Laboratory the authors schedule jobs to process experimental data to be collected during a five minute shot cycle. The data driven processing system emulates a coarsely granular data flow architecture. Processing jobs are scheduled before the experimental data is collected. Processing jobs ''fire'', or execute, as input data becomes available. Similar to UNIX ''pipes'', data produced by upstream processing nodes may be used as inputs by following nodes. Users, working on the networked SUN workstations, specify data processing templates which define processes and their data dependencies. Data specifications indicate the source of data; actual associations with specific data instantiations are made when the jobs are scheduled. The authors report here on details of diagnostic data processing and their experiences

  10. Overhead remote handling systems for the process facility modifications project

    International Nuclear Information System (INIS)

    Wiesener, R.W.; Grover, D.L.

    1987-01-01

    Each of the cells in the process facility modifications (PFM) project complex is provided with a variety of general purpose remote handling equipment including bridge cranes, monorail hoist, bridge-mounted electromechanical manipulator (EMM) and an overhead robot used for high efficiency particulate air (HEPA) filter changeout. This equipment supplements master-slave manipulators (MSMs) located throughout the complex to provide an overall remote handling system capability. The overhead handling equipment is used for fuel and waste material handling operations throughout the process cells. The system also provides the capability for remote replacement of all in-cell process equipment which may fail or be replaced for upgrading during the lifetime of the facility

  11. APET methodology for Defense Waste Processing Facility: Mode C operation

    International Nuclear Information System (INIS)

    Taylor, R.P. Jr.; Massey, W.M.

    1995-04-01

    Safe operation of SRS facilities continues to be the highest priority of the Savannah River Site (SRS). One of these facilities, the Defense Waste Processing Facility or DWPF, is currently undergoing cold chemical runs to verify the design and construction preparatory to hot startup in 1995. The DWPFF is a facility designed to convert the waste currently stored in tanks at the 200-Area tank farm into a form that is suitable for long term storage in engineered surface facilities and, ultimately, geologic isolation. As a part of the program to ensure safe operation of the DWPF, a probabilistic Safety Assessment of the DWPF has been completed. The results of this analysis are incorporated into the Safety Analysis Report (SAR) for DWPF. The usual practice in preparation of Safety Analysis Reports is to include only a conservative analysis of certain design basis accidents. A major part of a Probabilistic Safety Assessment is the development and quantification of an Accident Progression Event Tree or APET. The APET provides a probabilistic representation of potential sequences along which an accident may progress. The methodology used to determine the risk of operation of the DWPF borrows heavily from methods applied to the Probabilistic Safety Assessment of SRS reactors and to some commercial reactors. This report describes the Accident Progression Event Tree developed for the Probabilistic Safety Assessment of the DWPF

  12. Beryllium mock-ups development and ultrasonic testing for ITER divertor conditions

    International Nuclear Information System (INIS)

    Barabash, V.R.; Bykov, V.A.; Giniyatulin, R.N.; Gervash, A.A.; Gurieva, T.M.; Egorov, K.E.; Komarov, V.L.; Korolkov, M.D.; Mazul, I.V.; Gitarsky, L.S.; Strulia, I.L.; Sizenev, V.S.; Pronyakin, V.T.

    1995-01-01

    At the present time beryllium is considered as the most suitable armour material for the ITER divertor application. Different types of Be-divertor mock-up construction are compared in the report. Two different technologies of beryllium tiles joining to a heat sink body are analysed: high temperature brazing and thermodiffusion bonding. The comparative analysis of different constructions has been performed on the basis of 2-D finite element calculation for temperatures and stresses. The main parameters and diagnostic capabilities of electron beam facility for HHF testing of beryllium mock-ups are described. The first results of HHF tests of ''beryllium-copper saddle-MAGT tube'' and ''beryllium-copper plate-SS body'' mock-ups are presented. The reasons of the damages during the HHF are analysed. The technique of ultrasonic testing of the thermodifussion bonding and brazing quality for beryllium-copper joints is presented. The recorded results are prepared in the form of ultrasound grams. The testing results are compared with the metallographic analysis. (orig.)

  13. Defense Waste Processing Facility staged operations: environmental information document

    International Nuclear Information System (INIS)

    1981-11-01

    Environmental information is presented relating to a staged version of the proposed Defense Waste Processing Facility (DWPF) at the Savannah River Plant. The information is intended to provide the basis for an Environmental Impact Statement. In either the integral or the staged design, the DWPF will convert the high-level waste currently stored in tanks into: a leach-resistant form containing about 99.9% of all the radioactivity, and a residual, slightly contaminated salt, which is disposed of as saltcrete. In the first stage of the staged version, the insoluble sludge portion of the waste and the long lived radionuclides contained therein will be vitrified. The waste glass will be sealed in canisters and stored onsite until shipped to a Federal repository. In the second stage, the supernate portion of the waste will be decontaminated by ion exchange. The recovered radionuclides will be transferred to the Stage 1 facility, and mixed with the sludge feed before vitrification. The residual, slightly contaminated salt solution will be mixed with Portland cement to form a concrete product (saltcrete) which will be buried onsite in an engineered landfill. This document describes the conceptual facilities and processes for producing glass waste and decontaminated salt. The environmental effects of facility construction, normal operations, and accidents are then presented. Descriptions of site and environs, alternative sites and waste disposal options, and environmental consultations and permits are given in the base Environmental Information Document

  14. Hanford Site Treated Effluent Disposal Facility process flow sheet

    International Nuclear Information System (INIS)

    Bendixsen, R.B.

    1993-04-01

    This report presents a novel method of using precipitation, destruction and recycle factors to prepare a process flow sheet. The 300 Area Treated Effluent Disposal Facility (TEDF) will treat process sewer waste water from the 300 Area of the Hanford Site, located near Richland, Washington, and discharge a permittable effluent flow into the Columbia River. When completed and operating, the TEDF effluent water flow will meet or exceed water quality standards for the 300 Area process sewer effluents. A preliminary safety analysis document (PSAD), a preconstruction requirement, needed a process flow sheet detailing the concentrations of radionuclides, inorganics and organics throughout the process, including the effluents, and providing estimates of stream flow quantities, activities, composition, and properties (i.e. temperature, pressure, specific gravity, pH and heat transfer rates). As the facility begins to operate, data from process samples can be used to provide better estimates of the factors, the factors can be entered into the flow sheet and the flow sheet will estimate more accurate steady state concentrations for the components. This report shows how the factors were developed and how they were used in developing a flow sheet to estimate component concentrations for the process flows. The report concludes with how TEDF sample data can improve the ability of the flow sheet to accurately predict concentrations of components in the process

  15. Calculation of MUF for the Pyro-processing Facility

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Yon Hong; Kim, Woo Jin; Han, Jae Jun; Chang, Sun Young; Hwang, Yong Soo

    2016-05-15

    The IAEA safeguards system is divided into DIQ (Design Information Questionnaire), nuclear material accountancy, and additional measure such as C/S (Containment and Surveillance). As the detailed requirements for judging the diversion of nuclear materials, the IAEA suggests SQ (Significant Quantity) about SNM (Special Nuclear Materials), such as U and Pu, and the timeliness goal of detection about the diversion of nuclear materials. To operate facilities, it is required to accomplish these goals. In particular, in the case of the treatment facilities of spent nuclear fuel that has a high Pu content, it is very important to meet the requirements to judge the diversion of nuclear materials. However, given that item counting is impossible in bulk facilities, MUF (Material Unaccounted For) occurs inevitably in the process of nuclear material accountancy. Therefore, to meet the requirements, it is necessary to make evaluation in advance. To reduce such a MUF, the effects on a total MUF were analyzed. As a result, the error arising in a particular process such as U/TRU ingot and Porous Pellets was significant. Therefore, it is necessary to reduce measurement error in the process. MUF is one of requirements to judge the diversion of nuclear materials, and the requirement should be met. Nevertheless, it is required to come up with additional measures to prevent the exclusive use and reduce MUF, such as containment, surveillance, or multi-channel based processing design.

  16. Continuous Material Balance Reconciliation for a Modern Plutonium Processing Facility

    International Nuclear Information System (INIS)

    CLARK, THOMASG.

    2004-01-01

    This paper describes a safeguards approach that can be deployed at any modern plutonium processing facility to increase the level of safeguards assurance and significantly reduce the impact of safeguards on process operations. One of the most perplexing problems facing the designers of plutonium processing facilities is the constraint placed upon the limit of error of the inventory difference (LEID). The current DOE manual constrains the LEID for Category I and II material balance areas to 2 per cent of active inventory up to a Category II quantity of the material being processed. For 239Pu a Category II quantity is two kilograms. Due to the large material throughput anticipated for some of the modern plutonium facilities, the required LEID cannot be achieved reliably during a nominal two month inventory period, even by using state-of-the-science non-destructive assay (NDA) methods. The most cost-effective and least disruptive solution appears to be increasing the frequency of material balance closure and thus reducing the throughput being measured during each inventory period. Current inventory accounting practices and systems can already provide the book inventory values at any point in time. However, closing the material balance with measured values has typically required the process to be cleaned out, and in-process materials packaged and measured. This process requires one to two weeks of facility down time every two months for each inventory, thus significantly reducing productivity. To provide a solution to this problem, a non-traditional approach is proposed that will include using in-line instruments to provide measurement of the process materials on a near real-time basis. A new software component will be developed that will operate with the standard LANMAS application to provide the running material balance reconciliation, including the calculation of the inventory difference and variance propagation. The combined measurement system and software

  17. Beryllium and growth. II. The effect of beryllium on plant growth

    Energy Technology Data Exchange (ETDEWEB)

    Hoagland, M B

    1952-01-01

    Experiments were undertaken to determine whether beryllium could replace magnesium in a growing organism. This was stimulated by the several known growth effects of beryllium in animals and by the fact that beryllium apparently competes with magnesium for animal alkaline phosphatases. The following findings are noted: (1) beryllium can reduce the magnesium requirement of plants by some 60% within a certain range of magnesium deficiency. (2) The residual obligatory magnesium requirements is probably accounted for by chlorophyll since beryllium appears to have no primary effect on chlorophyll or chlorophyll production. (3) The pH of the nutrient solution is critical: at acid pH's, beryllium is highly toxic, and growth increase due to beryllium only appears at initial pH's above 11.2, although this initial pH rapidly falls to neutrality during the experimental period. 22 references, 4 figures, 1 table.

  18. Directory of gamma processing facilities in Member States

    International Nuclear Information System (INIS)

    2004-02-01

    Ionizing radiation can modify physical, chemical and biological properties of materials. This characteristic of radiation was recognised very soon after the discovery of radioactivity. At present, the principal applications concern sterilisation of health care products, food irradiation and materials modification for polymers. Besides naturally occurring radioactive isotopes, artificial ones were produced using cyclotrons. A significant impetus, however, was given to the radiation processing industry with the advent of nuclear reactors, which were used to produce radioisotopes. Gamma ray emitters like cobalt-60 became popular radiation sources for medical and industrial applications. Many gamma ray irradiators have been built and it is estimated that less than 200 are currently in operation all over the world. In recent times, the use of electron accelerators as a radiation source (sometimes equipped with X ray converter) is increasing. However, gamma irradiators are difficult to replace, especially in the case of non-uniform and high-density products. The International Atomic Energy Agency (IAEA) has several programmes related to industrial irradiation applications for processing of various products including those related to health care, pharmaceuticals, food and polymers, and applications associated with plant design, dosimetry and safety. Through the technical co-operation programme, the IAEA supports these activities in developing countries and helps them to build local capacity to implement various industrial applications of radiation processing. The IAEA also organises and conducts training courses and workshops, provides individual training to personnel, and sends experts to the radiation facilities in Member States where help is needed. All these activities can be carried out much more efficiently and effectively if there were a comprehensive directory of radiation facilities operating in Member States. Also, such a compilation would be a valuable tool for

  19. Facilities for studying the double beta decay processes

    International Nuclear Information System (INIS)

    Zdesenko, Yu.G.

    1980-01-01

    Modern state, tendencies and perspectiVes of the development of experimental installations to study double β-decay are treated. The main peculiarities of direct recognition and full experiments on the study of double β-decay are considered. A simple ratio is obtained from statistical considerations which connects the life time limits of the nuclei with the facility parameters to conduct direct recognition experiments. Possibilities of different detectors are evaluated on the basis of the ratio. Requirements for the modern technique for complete investigation of double β-decay are formulated and two designs of facilities meeting the requirements are considered. It is shown that the facility with proportional chambers is more perspective. On the basis of the analysis of the facility development to study double β-decay, conclusion is made that the final and unambiguous proof of the existence of double β-decay process can be obtained only directly in the experiments with immediate recording of the decay acts. Possibilities of the existing and developed facilities to conduct recognition (direct) experiments are such, that with their help life time limits as to neutronless double β-decay at the level of 10 21 -10 22 years can be established. Counters on the basis of the condensed noble gases, semiconductor detectors made of TeCd, scintillators of big volume are the most perspective detectors. To conduct complete experiments it is necessary to develop a facility with sensitivity sufficient for the detection of two-neutrino double β-activeness when Tsub(1/2)=10sup(21) years [ru

  20. Radioactive waste package assay facility. Volume 3. Data processing

    International Nuclear Information System (INIS)

    Creamer, S.C.; Lalies, A.A.; Wise, M.O.

    1992-01-01

    This report, in three volumes, covers the work carried out by Taylor Woodrow Construction Ltd, and two major sub-contractors: Harwell Laboratory (AEA Technology) and Siemens Plessey Controls Ltd, on the development of a radioactive waste package assay facility, for cemented 500 litre intermediate level waste drums. Volume 3, describes the work carried out by Siemens Plessey Controls Ltd on the data-processing aspects of an integrated waste assay facility. It introduces the need for a mathematical model of the assay process and develops a deterministic model which could be tested using Harwell experimental data. Relevant nuclear reactions are identified. Full implementation of the model was not possible within the scope of the Harwell experimental work, although calculations suggested that the model behaved as predicted by theory. 34 figs., 52 refs., 5 tabs

  1. Remote viewing of melter interior Defense Waste Processing Facility

    International Nuclear Information System (INIS)

    Heckendorn, F.M. II.

    1986-01-01

    A remote system has been developed and demonstrated for continuous reviewing of the interior of a glass melter, which is used to vitrify highly radioactive waste. The system is currently being implemented with the Defense Waste Processing Facility (DWPF) now under construction at the Savannah River Plant (SRP). The environment in which the borescope/TV unit is implemented combines high temperature, high ionizing radiation, low light, spattering, deposition, and remote maintenance

  2. Master slave manipulator maintenance at the Defense Waste Processing Facility

    International Nuclear Information System (INIS)

    Lethco, A.J.; Beasley, K.M.

    1991-01-01

    Equipment has been developed and tested to provide transport, installation, removal, decontamination, and repair for the master slave manipulators that are required for thirty-five discrete work locations in the 221-S Vitrification Building of the Defense Waste Processing Facility at the Westinghouse Savannah River Company. This specialized equipment provides a standardized scheme for work locations at different elevations with two types of manipulators

  3. Beryllium in aircraft brakes - a summary

    International Nuclear Information System (INIS)

    Zenczak, S.

    1977-01-01

    Beryllium has been in use in aircraft brakes for ten years. During the original design phases of the several aircraft programs using beryllium a number of problems requiring solution confronted the designers. In actual service the solution to these problems performed much better than had been anticipated. A summary is presented. (author)

  4. ICT diagnostic method of beryllium welding quality

    International Nuclear Information System (INIS)

    Sun Lingxia; Wei Kentang; Ye Yunchang

    2002-01-01

    To avoid the interference of high density material for the quality assay of beryllium welding line, a slice by slice scanning method was proposed based upon the research results of the Industrial Computerized Tomography (ICT) diagnostics for weld penetration, weld width, off-centered deviation and weld defects of beryllium-ring welding seam with high density material inside

  5. Modeling of hydrogen interactions with beryllium

    Energy Technology Data Exchange (ETDEWEB)

    Longhurst, G.R. [Lockheed Martin Idaho Technologies Co., Idaho Falls, ID (United States)

    1998-01-01

    In this paper, improved mathematical models are developed for hydrogen interactions with beryllium. This includes the saturation effect observed for high-flux implantation of ions from plasmas and retention of tritium produced from neutronic transmutations in beryllium. Use of the models developed is justified by showing how they can replicated experimental data using the TMAP4 tritium transport code. (author)

  6. Managing health effects of beryllium exposure

    National Research Council Canada - National Science Library

    Committee on Beryllium Alloy Exposures; Committee on Toxicology; National Research Council; Division on Earth and Life Studies; National Research Council

    2008-01-01

    ... to its occurrence in exposed people. Despite reduced workplace exposure, chronic beryllium disease continues to occur. In addition, beryllium has been classified as a likely human carcinogen by several agencies, such as the International Agency for Research on Cancer, the National Toxicology Program, and the U.S. Environmental Protection Agency. Thos...

  7. Structure investigations of some beryllium materials

    Energy Technology Data Exchange (ETDEWEB)

    Faeldt, I; Lagerberg, G

    1960-05-15

    Metallographic structure, microhardness and texture have been studied on various types of beryllium metal including hot pressed powder, a rolled strip and an extruded tube It was found that beryllium exhibits its highest hardness in directions perpendicular to the basal plane. Good ideas of the prevailing textures were obtained with an ordinary X-ray diffractometer.

  8. Structure investigations of some beryllium materials

    International Nuclear Information System (INIS)

    Faeldt, I.; Lagerberg, G.

    1960-05-01

    Metallographic structure, microhardness and texture have been studied on various types of beryllium metal including hot pressed powder, a rolled strip and an extruded tube It was found that beryllium exhibits its highest hardness in directions perpendicular to the basal plane. Good ideas of the prevailing textures were obtained with an ordinary X-ray diffractometer

  9. Electron-beam fusion welding of beryllium

    International Nuclear Information System (INIS)

    Campbell, R.P.; Dixon, R.D.; Liby, A.L.

    1978-01-01

    Ingot-sheet beryllium (Be) having three different chemistries and three different thicknesses was fusion-welded by the electron-beam process. Several different preheats were used to obtain 100% penetration and crack-free welds. Cracking susceptability was found to be related to aluminum (Al) content; the higher Al-content material was most susceptable. However, adequate preheat allowed full penetration and crack-free welds to be made in all materials tested. The effect of a post-weld heat treatment on the mechanical properties of these compositions was also determined. The heat treatment produced no significant effect on the ultimate tensile strength. However, the yield strength was decreased and the ductility was increased. These changes are attributed to the formation of AlFeBe 4 and FeBe 11

  10. The status of beryllium technology for fusion

    Energy Technology Data Exchange (ETDEWEB)

    Scaffidi-Argentina, F.; Longhurst, G.R. E-mail: gx1@inel.gov; Shestakov, V.; Kawamura, H

    2000-12-01

    Beryllium was used for a number of years in the Joint European Torus (JET), and it is planned to be used extensively on the lower heat-flux surfaces of the reduced technical objective/reduced cost international thermonuclear experimental reactor (RTO/RC ITER). It has been included in various forms in a number of tritium breeding blanket designs. There are technical advantages but also a number of safety issues associated with the use of beryllium. Research in a variety of technical areas in recent years has revealed interesting issues concerning the use of beryllium in fusion. Progress in this research has been presented at a series of International Workshops on Beryllium Technology for Fusion. The most recent workshop was held in Karlsruhe, Germany on 15-17 September 1999. In this paper, a summary of findings presented there and their implications for the use of beryllium in the development of fusion reactors are presented.

  11. The status of beryllium technology for fusion

    International Nuclear Information System (INIS)

    Scaffidi-Argentina, F.; Longhurst, G.R.; Shestakov, V.; Kawamura, H.

    2000-01-01

    Beryllium was used for a number of years in the Joint European Torus (JET), and it is planned to be used extensively on the lower heat-flux surfaces of the reduced technical objective/reduced cost international thermonuclear experimental reactor (RTO/RC ITER). It has been included in various forms in a number of tritium breeding blanket designs. There are technical advantages but also a number of safety issues associated with the use of beryllium. Research in a variety of technical areas in recent years has revealed interesting issues concerning the use of beryllium in fusion. Progress in this research has been presented at a series of International Workshops on Beryllium Technology for Fusion. The most recent workshop was held in Karlsruhe, Germany on 15-17 September 1999. In this paper, a summary of findings presented there and their implications for the use of beryllium in the development of fusion reactors are presented

  12. Assessment of LANL beryllium waste management documentation

    International Nuclear Information System (INIS)

    Danna, J.G.; Jennrich, E.A.; Lund, D.M.; Davis, K.D.; Hoevemeyer, S.S.

    1991-04-01

    The objective of this report is to determine present status of the preparation and implementation of the various high priority documents required to properly manage the beryllium waste generated at the Laboratory. The documents being assessed are: Waste Acceptance Criteria, Waste Characterization Plan, Waste Certification Plan, Waste Acceptance Procedures, Waste Characterization Procedures, Waste Certification Procedures, Waste Training Procedures and Waste Recordkeeping Procedures. Beryllium is regulated (as a dust) under 40 CFR 261.33 as ''Discarded commercial chemical products, off specification species, container residues and spill residues thereof.'' Beryllium is also identified in the 3rd thirds ruling of June 1, 1990 as being restricted from land disposal (as a dust). The beryllium waste generated at the Laboratory is handled separately because beryllium has been identified as a highly toxic carcinogenic material

  13. Beryllium for fusion application - recent results

    International Nuclear Information System (INIS)

    Khomutov, A.; Barabash, V.; Chakin, V.; Chernov, V.; Davydov, D.; Gorokhov, V.; Kawamura, H.; Kolbasov, B.; Kupriyanov, I.; Longhurst, G.; Scaffidi-Argentina, F.; Shestakov, V.

    2002-01-01

    The main issues for the application of beryllium in fusion reactors are analyzed taking into account the latest results since the ICFRM-9 (Colorado, USA, October 1999) and presented at 5th IEA Be Workshop (10-12 October 2001, Moscow Russia). Considerable progress has been made recently in understanding the problems connected with the selection of the beryllium grades for different applications, characterization of the beryllium at relevant operational conditions (irradiation effects, thermal fatigue, etc.), and development of required manufacturing technologies. The key remaining problems related to the application of beryllium as an armour in near-term fusion reactors (e.g. ITER) are discussed. The features of the application of beryllium and beryllides as a neutron multiplier in the breeder blanket for power reactors (e.g. DEMO) in pebble-bed form are described

  14. Beryllium for fusion application - recent results

    Science.gov (United States)

    Khomutov, A.; Barabash, V.; Chakin, V.; Chernov, V.; Davydov, D.; Gorokhov, V.; Kawamura, H.; Kolbasov, B.; Kupriyanov, I.; Longhurst, G.; Scaffidi-Argentina, F.; Shestakov, V.

    2002-12-01

    The main issues for the application of beryllium in fusion reactors are analyzed taking into account the latest results since the ICFRM-9 (Colorado, USA, October 1999) and presented at 5th IEA Be Workshop (10-12 October 2001, Moscow Russia). Considerable progress has been made recently in understanding the problems connected with the selection of the beryllium grades for different applications, characterization of the beryllium at relevant operational conditions (irradiation effects, thermal fatigue, etc.), and development of required manufacturing technologies. The key remaining problems related to the application of beryllium as an armour in near-term fusion reactors (e.g. ITER) are discussed. The features of the application of beryllium and beryllides as a neutron multiplier in the breeder blanket for power reactors (e.g. DEMO) in pebble-bed form are described.

  15. The nature of innovation processes in Facility Management services

    DEFF Research Database (Denmark)

    Nardelli, Giulia

    Purpose: This work investigates the dynamics of interaction between stakeholders of Facilities Management (FM) innovation and improvement processes. The aim is to understand how the complex value chain of FM services influences innovation processes within this field. Theory: This study combines...... theories on innovation in services with research focused on the empirical field of FM. More specifically, the analytical framework for this study applies the differentiation between reactive and proactive innovation processes by Toivonen and Tuominen (2009) to the value chain identified by Coenen...... has a threefold impact on the nature of innovation processes within this field. Firstly, end-users of FM services are usually not involved in innovation processes, although they might sometimes play a role as initial drivers. Secondly, FM services are intangible but more easily reproducible than other...

  16. Postirradiation examination of beryllium pebbles

    International Nuclear Information System (INIS)

    Gelles, D.S.

    1998-01-01

    Postirradiation examinations of COBRA-1A beryllium pebbles irradiated in the EBR-II fast reactor at neutron fluences which generated 2700--3700 appm helium have been performed. Measurements included density change, optical microscopy, scanning electron microscopy, and transmission electron microscopy. The major change in microstructure is development of unusually shaped helium bubbles forming as highly non-equiaxed thin platelet-like cavities on the basal plane. Measurement of the swelling due to cavity formation was in good agreement with density change measurements

  17. Overview of planning process at FFTF [Fast Flux Test Facility

    International Nuclear Information System (INIS)

    Gadeken, A.D.

    1986-03-01

    The planning process at the Fast Flux Test Facility (FFTF) is controlled through a hierarchy of documents ranging from a ten-year strategic plan to a weekly schedule. Within the hierarchy are a Near-Term (three-year) Operating Plan, a Cycle (six-month) Plan, and an Outage/Operating Phase Schedule. Coordination of the planning process is accomplished by a dedicated preparation team that also provides an overview of the formal planning timetable which identifies key action items required to be completed before an outage/operating phase can begin

  18. Process component inventory in a large commercial reprocessing facility

    International Nuclear Information System (INIS)

    Canty, M.J.; Berliner, A.; Spannagel, G.

    1983-01-01

    Using a computer simulation program, the equilibrium operation of the Pu-extraction and purification processes of a reference commercial reprocessing facility was investigated. Particular attention was given to the long-term net fluctuations of Pu inventories in hard-to-measure components such as the solvent extraction contractors. Comparing the variance of these inventories with the measurement variance for Pu contained in feed, analysis and buffer tanks, it was concluded that direct or indirect periodic estimation of contactor inventories would not contribute significantly to improving the quality of closed material balances over the process MBA

  19. Occupational exposure to beryllium in French enterprises: a survey of airborne exposure and surface levels.

    Science.gov (United States)

    Vincent, Raymond; Catani, Jacques; Créau, Yvon; Frocaut, Anne-Marie; Good, Andrée; Goutet, Pierre; Hou, Alain; Leray, Fabrice; André-Lesage, Marie-Ange; Soyez, Alain

    2009-06-01

    An assessment survey of occupational exposure to beryllium (Be) was conducted in France between late 2004 and the end of 2006. Exposure estimates were based on the analytical results of samples collected from workplace air and from work surfaces in 95 facilities belonging to 37 sectors of activity. The results of this study indicated airborne Be concentrations in excess of the occupational exposure limit value of 2 microg m(-3) recommended in France. Metallurgy and electronic component manufacturing represented the activities and occupations where workers had the highest arithmetic mean exposures to Be. Surface contamination levels were also high and frequently exceeded thresholds recommended by different bodies. These results should prompt the development of prevention programmes that include Be substitution, process control and surface decontamination, in conjunction with suitable medical surveillance.

  20. Afterheat usage from cooling facilities in ORC processes

    International Nuclear Information System (INIS)

    Theede, Florian; Luke, Andrea

    2016-01-01

    In the course of the reduction of climate warming an energy-efficient lay-out of processes is necessary. A possibility for the efficiency increasement is the usage of afterheat currents for instance in ORC processes. Connected with the limitation of refrigerants with high greenhouse potential it comes to the increased application of transcritical cooling facilities with carbon dioxide (CO_2) as refrigerant. By the high pressures after the compression arise here new afterheat sources on a temperature level of about 100 C. An alternative for the simple back-cooling or the heating support and drinking-water heating represents the current production in an ORC process. Great challenges for the lay-out of such an ORC process are the selection of the working fluid as well as the lay-out of the heat exchangers. Established refrigerants in the low-temperature like R245fa for ORC facilities will be in forseeable future no more available. For the study of the possible replacement by alternative refrigerants a simulation model has been developed. By means of this model different refrigerants are analyzed regarding their performance and simultaneously the effects on process and other components studied. The results show that in the temperature range two hydrofluoroolefines R1233zd[E] and R1234ze[Z] as well as the hadron carbon butane can thermodynamically form an alternative.

  1. The sodium process facility at Argonne National Laboratory - West

    International Nuclear Information System (INIS)

    Michelbacher, J.A.; Henslee, S.P.; McDermott, M.D.; Price, J.R.; Rosenberg, K.E.; Wells, P.B.

    1997-01-01

    Argonne National Laboratory - West (ANL-W) has approximately 680,000 liters (180,000 gallons) of raw sodium stored in facilities on site. As mandated by the State of Idaho and the United States Department of Energy (DOE), this sodium must be transformed into a stable condition for land disposal. To comply with this mandate, ANL-W designed and built the Sodium Process Facility (SPF) for the processing of this sodium into a dry, sodium carbonate powder. The major portion of the sodium stored at ANL-W is radioactively contaminated. The SPF was designed to react elemental sodium to sodium carbonate through two-stages involving caustic process and carbonate process steps. The sodium is first reacted to sodium hydroxide in the caustic process step. The caustic process step involves the injection of sodium into a nickel reaction vessel filled with a 50 wt% solution of sodium hydroxide. Water is also injected, controlling the boiling point of the solution. In the carbonate process, the sodium hydroxide is reacted with carbon dioxide to form sodium carbonate. This dry powder, similar in consistency to baking soda, is a waste form acceptable for burial in the State of Idaho as a non-hazardous, radioactive waste. The caustic process was originally designed and built in the 1980s for reacting the 290,000 liters (77,000 gallons) of primary sodium from the Fermi-1 Reactor to sodium hydroxide. The hydroxide was slated to be used to neutralize acid products from the PUREX process at the Hanford site. However, changes in the DOE mission precluded the need for hydroxide and the caustic process was never operated. With the shutdown of the Experimental Breeder Reactor-II (EBR-II), the necessity for a facility to react sodium was identified. In order to comply with Resource Conservation and Recovery Act (RCRA) requirements, the sodium had to be converted into a waste form acceptable for disposal in a Sub-Title D low-level radioactive waste disposal facility. Sodium hydroxide is a RCRA

  2. Accident Fault Trees for Defense Waste Processing Facility

    Energy Technology Data Exchange (ETDEWEB)

    Sarrack, A.G.

    1999-06-22

    The purpose of this report is to document fault tree analyses which have been completed for the Defense Waste Processing Facility (DWPF) safety analysis. Logic models for equipment failures and human error combinations that could lead to flammable gas explosions in various process tanks, or failure of critical support systems were developed for internal initiating events and for earthquakes. These fault trees provide frequency estimates for support systems failures and accidents that could lead to radioactive and hazardous chemical releases both on-site and off-site. Top event frequency results from these fault trees will be used in further APET analyses to calculate accident risk associated with DWPF facility operations. This report lists and explains important underlying assumptions, provides references for failure data sources, and briefly describes the fault tree method used. Specific commitments from DWPF to provide new procedural/administrative controls or system design changes are listed in the ''Facility Commitments'' section. The purpose of the ''Assumptions'' section is to clarify the basis for fault tree modeling, and is not necessarily a list of items required to be protected by Technical Safety Requirements (TSRs).

  3. A commercial multipurpose radiation processing facility for Hawaii

    International Nuclear Information System (INIS)

    Welt, M.A.

    1985-01-01

    The State of Hawaii offers a unique challenge for the designer of an economically feasible radiation processing system. Based on the prevailing agricultural export requirements, the radiation facility must be capable for handling a variety of bulky fruit and vegetable products for insect disinfestation purposes and, yet, provide proper economies for the users of the facility. A capability must exist for irradiating other types of products requiring higher doses, e.g., fish and shellfish products for shelf-life extension, which might require a dose approximately eight times higher than the disinfestation dose, or even medical product or a food sterilization dose, which would be approximately twelve times higher than the required shelf-life extension dose. The Radiation Technology Model RT 4l0l-4048 radiation processing facility provides the necessary versatility and operational reliability to meet the challenge. The technical features and economic analyses demonstrate the advantages of this computer-operated pallet irradiation system. Actual performance data from the Radiation Technology subsidiary operations in West Memphis, Arkasas, and Burlilngton, North Carolina, are presented along with photographs of the proposed system for Hawaii

  4. Analytical methods and laboratory facility for the Defense Waste Processing Facility

    International Nuclear Information System (INIS)

    Coleman, C.J.; Dewberry, R.A.; Lethco, A.J.; Denard, C.D.

    1985-01-01

    This paper describes the analytical methods, instruments, and laboratory that will support vitrification of defense waste. The Defense Waste Processing Facility (DWPF) is now being constructed at Savannah River Plant (SRP). Beginning in 1989, SRP high-level defense waste will be immobilized in borosilicate glass for disposal in a federal repository. The DWPF will contain an analytical laboratory for performing process control analyses. Additional analyses will be performed for process history and process diagnostics. The DWPF analytical facility will consist of a large shielded sampling cell, three shielded analytical cells, a laboratory for instrumental analysis and chemical separations, and a counting room. Special instrumentation is being designed for use in the analytical cells, including microwave drying/dissolution apparatus, and remote pipetting devices. The instrumentation laboratory will contain inductively coupled plasma, atomic absorption, Moessbauer spectrometers, a carbon analyzer, and ion chromatography equipment. Counting equipment will include intrinsic germanium detectors, scintillation counters, Phoswich alpha, beta, gamma detectors, and a low-energy photon detector

  5. Technical issues for beryllium use in fusion blanket applications

    International Nuclear Information System (INIS)

    McCarville, T.J.; Berwald, D.H.; Wolfer, W.; Fulton, F.J.; Lee, J.D.; Maninger, R.C.; Moir, R.W.; Beeston, J.M.; Miller, L.G.

    1985-01-01

    Beryllium is an excellent non-fissioning neutron multiplier for fusion breeder and fusion electric blanket applications. This report is a compilation of information related to the use of beryllium with primary emphasis on the fusion breeder application. Beryllium resources, production, fabrication, properties, radiation damage and activation are discussed. A new theoretical model for beryllium swelling is presented

  6. The Sodium Process Facility at Argonne National Laboratory-West

    International Nuclear Information System (INIS)

    Michelbacher, J.A.; Henslee, S.P.; McDermott, M.D.; Price, J.R.; Rosenberg, K.E.; Wells, P.B.

    1998-01-01

    Argonne National Laboratory-West (ANL-W) has approximately 680,000 liters of raw sodium stored in facilities on site. As mandated by the State of Idaho and the US Department of Energy (DOE), this sodium must be transformed into a stable condition for land disposal. To comply with this mandate, ANL-W designed and built the Sodium Process Facility (SPF) for the processing of this sodium into a dry, sodium carbonate powder. The major portion of the sodium stored at ANL-W is radioactively contaminated. The sodium will be processed in three separate and distinct campaigns: the 290,000 liters of Fermi-1 primary sodium, the 50,000 liters of the Experimental Breeder Reactor-II (EBR-II) secondary sodium, and the 330,000 liters of the EBR-II primary sodium. The Fermi-1 and the EBR-II secondary sodium contain only low-level of radiation, while the EBR-II primary sodium has radiation levels up to 0.5 mSv (50 mrem) per hour at 1 meter. The EBR-II primary sodium will be processed last, allowing the operating experience to be gained with the less radioactive sodium prior to reacting the most radioactive sodium. The sodium carbonate will be disposed of in 270 liter barrels, four to a pallet. These barrels are square in cross-section, allowing for maximum utilization of the space on a pallet, minimizing the required landfill space required for disposal

  7. The Sodium Process Facility at Argonne National Laboratory-West

    Energy Technology Data Exchange (ETDEWEB)

    Michelbacher, J.A.; Henslee, S.P. McDermott, M.D.; Price, J.R.; Rosenberg, K.E.; Wells, P.B.

    1998-07-01

    Argonne National Laboratory-West (ANL-W) has approximately 680,000 liters of raw sodium stored in facilities on site. As mandated by the State of Idaho and the US Department of Energy (DOE), this sodium must be transformed into a stable condition for land disposal. To comply with this mandate, ANL-W designed and built the Sodium Process Facility (SPF) for the processing of this sodium into a dry, sodium carbonate powder. The major portion of the sodium stored at ANL-W is radioactively contaminated. The sodium will be processed in three separate and distinct campaigns: the 290,000 liters of Fermi-1 primary sodium, the 50,000 liters of the Experimental Breeder Reactor-II (EBR-II) secondary sodium, and the 330,000 liters of the EBR-II primary sodium. The Fermi-1 and the EBR-II secondary sodium contain only low-level of radiation, while the EBR-II primary sodium has radiation levels up to 0.5 mSv (50 mrem) per hour at 1 meter. The EBR-II primary sodium will be processed last, allowing the operating experience to be gained with the less radioactive sodium prior to reacting the most radioactive sodium. The sodium carbonate will be disposed of in 270 liter barrels, four to a pallet. These barrels are square in cross-section, allowing for maximum utilization of the space on a pallet, minimizing the required landfill space required for disposal.

  8. Manufacturing Demonstration Facility: Roll-to-Roll Processing

    Energy Technology Data Exchange (ETDEWEB)

    Datskos, Panos G [ORNL; Joshi, Pooran C [ORNL; List III, Frederick Alyious [ORNL; Duty, Chad E [ORNL; Armstrong, Beth L [ORNL; Ivanov, Ilia N [ORNL; Jacobs, Christopher B [ORNL; Graham, David E [ORNL; Moon, Ji Won [ORNL

    2015-08-01

    This Manufacturing Demonstration Facility (MDF)e roll-to-roll processing effort described in this report provided an excellent opportunity to investigate a number of advanced manufacturing approaches to achieve a path for low cost devices and sensors. Critical to this effort is the ability to deposit thin films at low temperatures using nanomaterials derived from nanofermentation. The overarching goal of this project was to develop roll-to-roll manufacturing processes of thin film deposition on low-cost flexible substrates for electronics and sensor applications. This project utilized ORNL s unique Pulse Thermal Processing (PTP) technologies coupled with non-vacuum low temperature deposition techniques, ORNL s clean room facility, slot dye coating, drop casting, spin coating, screen printing and several other equipment including a Dimatix ink jet printer and a large-scale Kyocera ink jet printer. The roll-to-roll processing project had three main tasks: 1) develop and demonstrate zinc-Zn based opto-electronic sensors using low cost nanoparticulate structures manufactured in a related MDF Project using nanofermentation techniques, 2) evaluate the use of silver based conductive inks developed by project partner NovaCentrix for electronic device fabrication, and 3) demonstrate a suite of low cost printed sensors developed using non-vacuum deposition techniques which involved the integration of metal and semiconductor layers to establish a diverse sensor platform technology.

  9. Process control and dosimetry in a multipurpose irradiation facility

    International Nuclear Information System (INIS)

    Cabalfin, E.G.; Lanuza, L.G.; Solomon, H.M.

    1998-01-01

    Complete text of publication follows. To introduce and demonstrate radiation processing to the local industries, the Philippine Nuclear Research Institute (PNRI) with the technical assistance of the International Atomic Energy Agency (IAEA) has set up a pilot scale multipurpose gamma irradiation facility. Though on a limited scale, this has led to the commercial radiation sterilization and decontamination of various products, such as empty aluminum tubes, empty gelatin capsules, spices and fresh onions. Process control in this facility involves dose measurement to ensure that the products receive the required dose to get the desired beneficial effect. Prior to routine processing, dose distribution studies to determine the locations of minimum and maximum absorbed dose are undertaken for each product and product-source geometry. The product loading pattern, which meets the required dose uniformity ratio and which gives the optimum amount of product per loading is then chosen. During routine irradiation, dosimeters are placed at the minimum and maximum absorbed dose positions of a process load. If locations of minimum or maximum dose are not readily accessible, dosimeters are placed at reference positions. The relationship of the absorbed dose at these reference positions with the absorbed dose at the minimum or maximum position is established beforehand. Fricke and ethanol chlorobenzene (ECB) dosimeters are used to measure absorbed dose. PNRI participates in the International Dose Assurance Service (IDAS) of IAEA. Results show that absorbed dose as measured by alanine agreed with ECB within 5%, while that from Fricke agreed to within 2%

  10. Process control and dosimetry in a multipurpose irradiation facility

    Energy Technology Data Exchange (ETDEWEB)

    Cabalfin, E G; Lanuza, L G; Solomon, H M [Philippine Nuclear Research Institute, Diliman, Quezon City (Philippines)

    1999-12-31

    Complete text of publication follows. To introduce and demonstrate radiation processing to the local industries, the Philippine Nuclear Research Institute (PNRI) with the technical assistance of the International Atomic Energy Agency (IAEA) has set up a pilot scale multipurpose gamma irradiation facility. Though on a limited scale, this has led to the commercial radiation sterilization and decontamination of various products, such as empty aluminum tubes, empty gelatin capsules, spices and fresh onions. Process control in this facility involves dose measurement to ensure that the products receive the required dose to get the desired beneficial effect. Prior to routine processing, dose distribution studies to determine the locations of minimum and maximum absorbed dose are undertaken for each product and product-source geometry. The product loading pattern, which meets the required dose uniformity ratio and which gives the optimum amount of product per loading is then chosen. During routine irradiation, dosimeters are placed at the minimum and maximum absorbed dose positions of a process load. If locations of minimum or maximum dose are not readily accessible, dosimeters are placed at reference positions. The relationship of the absorbed dose at these reference positions with the absorbed dose at the minimum or maximum position is established beforehand. Fricke and ethanol chlorobenzene (ECB) dosimeters are used to measure absorbed dose. PNRI participates in the International Dose Assurance Service (IDAS) of IAEA. Results show that absorbed dose as measured by alanine agreed with ECB within 5%, while that from Fricke agreed to within 2%.

  11. Design of facilities for processing pyrophoric radioactive material

    International Nuclear Information System (INIS)

    Bristow, H.A.S.; Hunter, S.D.

    1976-01-01

    The safe processing of large quantities of plutonium-bearing material poses difficult problems the solution of which sometimes involves conflicting requirements. The difficulties are increased when plutonium of a high burnup is used and the position becomes considerably more complicated when the chemical nature of the material being handled is such that it is pyrophoric. This paper describes the design principles and methods used to establish a facility capable of manufacturing large quantities of mixed plutonium/uranium carbide. The facility which included process stages such as milling, granulation, pellet pressing, furnacing and pin filling, was largely a conversion of an existing processing line. The paper treats the major plant hazards individually and indicates the methods used to counter them, outlining the main design principles employed and describing their application to selected items of equipment. Examples of the problems encountered with typical items of equipment are discussed. Some guide-lines are listed which should be of general value to designers and developers working on equipment for processing plutonium-bearing solids. The methods described have been successfully employed to provide a plant for the manufacture of mixed plutonium/uranium carbide on a scale of many hundreds of kilograms with no serious incident.(author)

  12. Mechanical properties of irradiated beryllium

    Energy Technology Data Exchange (ETDEWEB)

    Beeston, J.M.; Longhurst, G.R.; Wallace, R.S. (EG and G Idaho, Inc., Idaho Falls, ID (United States). Idaho National Engineering Lab.); Abeln, S.P. (EG and G Rocky Flats, Inc., Golden, CO (United States))

    1992-10-01

    Beryllium is planned for use as a neutron multiplier in the tritium breeding blanket of the International Thermonuclear Experimental Reactor (ITER). After fabricating samples of beryllium at densities varying from 80 to 100% of the theoretical density, we conducted a series of experiments to measure the effect of neutron irradiation on mechanical properties, especially strength and ductility. Samples were irradiated in the Advanced Test Reactor (ATR) to a neutron fluence of 2.6 x 10[sup 25] n/m[sup 2] (E > MeV) at an irradiation temperature of 75deg C. These samples were subsequently compression-tested at room temperature, and the results were compared with similar tests on unirradiated specimens. We found that the irradiation increased the strength by approximately four times and reduced the ductility to approximately one fourth. Failure was generally ductile, but the 80% dense irradiated samples failed in brittle fracture with significant generation of fine particles and release of small quantities of tritium. (orig.).

  13. Electron microscope study of irradiated beryllium oxide

    International Nuclear Information System (INIS)

    Bisson, A.A.

    1965-06-01

    The beryllium oxide is studied first by fractography, before and after irradiation, using sintered samples. The fractures are examined under different aspects. The higher density sintered samples, with transgranular fractures are the most interesting for a microscopic study. It is possible to mark the difference between the 'pores' left by the sintering process and the 'bubbles' of gases that can be produced by former thermal treatments. After irradiation, the grain boundaries are very much weakened. By annealing, it is possible to observe the evolution of the gases produced by the reaction (n, 2n) and (n. α) and gathered on the grain boundaries. The irradiated beryllium oxide is afterwards studied by transmission. For that, a simple method has been used: little chips of the crushed material are examined. Clusters of point defects produced by neutrons are thus detected in crystals irradiated at the three following doses: 6 x 10 19 , 9 x 10 19 and 2 x 10 20 n f cm -2 at a temperature below 100 deg. C. For the irradiation at 6 x 10 19 n f cm -2 , the defects are merely visible, but at 2 x l0 20 n f cm -2 the crystals an crowded with clusters and the Kikuchi lines have disappeared from the micro-diffraction diagrams. The evolution of the clusters into dislocation loops is studied by a series of annealings. The activation energy (0,37 eV) calculated from the annealing curves suggests that it must be interstitials that condense into dislocation loops. Samples irradiated at high temperatures (650, 900 and 1100 deg. C) are also studied. In those specimens the size of the loops is not the same as the equilibrium size obtained after out of pile annealing at the same temperature. Those former loops are more specifically studied and their Burgers vector is determined by micro-diffraction. (author) [fr

  14. Investigations of the ternary system beryllium-carbon-tungsten and analyses of beryllium on carbon surfaces; Untersuchung des ternaeren Systems Beryllium-Kohlenstoff-Wolfram und Betrachtungen von Beryllium auf Kohlenstoffoberflaechen

    Energy Technology Data Exchange (ETDEWEB)

    Kost, Florian

    2009-05-25

    Beryllium, carbon and tungsten are planned to be used as first wall materials in the future fusion reactor ITER. The aim of this work is a characterization of mixed material formation induced by thermal load. To this end, model systems (layers) were prepared and investigated, which give insight into the basic physical and chemical concepts. Before investigating ternary systems, the first step was to analyze the binary systems Be/C and Be/W (bottom-up approach), where the differences between the substrates PG (pyrolytic graphite) and HOPG (highly oriented pyrolytic graphite) were of special interest. Particularly X-ray photoelectron spectroscopy (XPS), low energy ion scattering (ISS) and Rutherford backscattering spectroscopy (RBS) were used as analysis methods. Beryllium evaporated on carbon shows an island growth mode, whereas a closed layer can be assumed for layer thicknesses above 0.7 nm. Annealing of the Be/C system induces Be{sub 2}C island formation for T{>=}770 K. At high temperatures (T{>=}1170 K), beryllium carbide dissociates, resulting in (metallic) beryllium desorption. For HOPG, carbide formation starts at higher temperatures compared to PG. Activation energies for the diffusion processes were determined by analyzing the decreasing beryllium amount versus annealing time. Surface morphologies were characterized using angle-resolved XPS (ARXPS) and atomic force microscopy (AFM). Experiments were performed to study processes in the Be/W system in the temperature range from 570 to 1270 K. Be{sub 2}W formation starts at 670 K, a complete loss of Be{sub 2}W is observed at 1170 K due to dissociation (and subsequent beryllium desorption). Regarding ternary systems, particularly Be/C/W and C/Be/W were investigated, attaching importance to layer thickness (reservoir) variations. At room temperature, Be{sub 2}C, W{sub 2}C, WC and Be{sub 2}W formation at the respective interfaces was observed. Further Be{sub 2}C is forming with increasing annealing temperatures

  15. Multiscale modelling of hydrogen behaviour on beryllium (0001 surface

    Directory of Open Access Journals (Sweden)

    Ch. Stihl

    2016-12-01

    Full Text Available Beryllium is proposed to be a neutron multiplier and plasma facing material in future fusion devices. Therefore, it is crucial to acquire an understanding of the microscopic mechanisms of tritium accumulation and release as a result of transmutation processes that Be undergoes under neutron irradiation. A multiscale simulation of ad- and desorption of hydrogen isotopes on the beryllium (0001 surface is developed. It consists of ab initio calculations of certain H adsorption configurations, a suitable cluster expansion approximating the energies of arbitrary configurations, and a kinetic Monte Carlo method for dynamic simulations of adsorption and desorption. The processes implemented in the kinetic Monte Carlo simulation are deduced from further ab initio calculations comprising both, static relaxation as well as molecular dynamics runs. The simulation is used to reproduce experimental data and the results are compared and discussed. Based on the observed results, proposals for a refined model are made.

  16. The regulatory process for the decommissioning of nuclear facilities

    International Nuclear Information System (INIS)

    1990-01-01

    The objective of this publication is to provide general guidance to Member States for regulating the decommissioning of nuclear facilities within the established nuclear regulatory framework. The Guide should also be useful to those responsible for, or interested in, the decommissioning of nuclear facilities. The Guide describes in general terms the process to be used in regulating decommissioning and the considerations to be applied in the development of decommissioning regulations and guides. It also delineates the responsibilities of the regulatory body and the licensee in decommissioning. The provisions of this Guide are intended to apply to all facilities within the nuclear fuel cycle and larger industrial installations using long lived radionuclides. For smaller installations, however, less extensive planning and less complex regulatory control systems should be acceptable. The Guide deals primarily with decommissioning after planned shutdown. Most provisions, however, are also applicable to decommissioning after an abnormal event, once cleanup operations have been terminated. The decommissioning planning in this case must take account of the abnormal event. 28 refs, 1 fig

  17. Research of flaw assessment methods for beryllium reflector elements

    International Nuclear Information System (INIS)

    Shibata, Akira; Ito, Masayasu; Takemoto, Noriyuki; Tanimoto, Masataka; Tsuchiya, Kunihiko; Nakatsuka, Masafumi; Ohara, Hiroshi; Kodama, Mitsuhiro

    2012-02-01

    Reflector elements made from metal beryllium is widely used as neutron reflectors to increase neutron flux in test reactors. When beryllium reflector elements are irradiated by neutron, bending of reflector elements caused by swelling occurs, and beryllium reflector elements must be replaced in several years. In this report, literature search and investigation for non-destructive inspection of Beryllium and experiments for Preliminary inspection to establish post irradiation examination method for research of characteristics of metal beryllium under neutron irradiation were reported. (author)

  18. Preliminary proposal for a beryllium technology program for fusion applications

    International Nuclear Information System (INIS)

    1985-02-01

    The program was designed to provide the answers to the critical issues of beryllium technology needed in fusion blanket designs. The four tasks are as follows: (1) Beryllium property measurements needed for fusion data base. (2) Beryllium stress relaxation and creep measurements for lifetime modelling calculations. (3) Simplified recycle technique development for irradiated beryllium. (4) Beryllium neutron multiplier measurements using manganese bath absolute calibration techniques

  19. Preliminary technical data summary defense waste processing facility stage 2

    International Nuclear Information System (INIS)

    1980-12-01

    This Preliminary Technical Data Summary presents the technical basis for design of Stage 2 of the Staged Defense Waste Processing Facility (DWPF). Process changes incorporated in the staged DWPF relative to the Alternative DWPF described in PTDS No. 3 (DPSTD-77-13-3) are the result of ongoing research and development and are aimed at reducing initial capital investment and developing a process to efficiently immobilize the radionuclides in Savannah River Plant (SRP) high-level liquid waste. The radionuclides in SRP waste are present in sludge that has settled to the bottom of waste storage tanks and in crystallized salt and salt solution (supernate). Stage 1 of the DWPF receives washed, aluminum dissolved sludge from the waste tank farms and immobilizes it in a borosilicate glass matrix. The supernate is retained in the waste tank farms until completion of Stage 2 of the DWPF at which time it is filtered and decontaminated by ion exchange in the Stage 2 facility. The decontaminated supernate is concentrated by evaporation and mixed with cement for burial. The radioactivity removed from the supernate is fixed in borosilicate glass along with the sludge. This document gives flowsheets, material and curie balances, material and curie balance bases, and other technical data for design of Stage 2 of the DWPF. Stage 1 technical data are presented in DPSTD-80-38

  20. Supplemental environmental impact statement - defense waste processing facility

    International Nuclear Information System (INIS)

    1994-11-01

    This document supplements the Final Environmental Impact Statement (EIS) DOE Issued in 1982 (DOE/EIS-0082) to construct and operate the Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS), a major DOE installation in southwestern South Carolina. That EIS supported the decision to construct and operate the DWPF to immobilize high-level waste generated as a result of nuclear materials processing at SRS. The DWPF would use a vitrification process to incorporate the radioactive waste into borosilicate glass and seal it in stainless steel canisters for eventual disposal at a permanent geologic repository. The DWPF is now mostly constructed and nearly ready for full operation. However, DOE has made design changes to the DWPF since the 1982 EIS to improve efficiency and safety of the facility. Each of these modifications was subjected to appropriate NEPA review. The purpose of this Supplemental EIS is to assist DOE in deciding whether and how to proceed with operation of the DWPF as modified since 1982 while ensuring appropriate consideration of potential environmental effects. In this document, DOE assesses the potential environmental impacts of completing and operating the DWPF in light of these design changes, examines the impact of alternatives, and identifies potential actions to be taken to reduce adverse impacts. Evaluations of impacts on water quality, air quality, ecological systems, land use, geologic resources, cultural resources, socioeconomics, and health and safety of onsite workers and the public are included in the assessment

  1. Materials, Processes, and Facile Manufacturing for Bioresorbable Electronics: A Review.

    Science.gov (United States)

    Yu, Xiaowei; Shou, Wan; Mahajan, Bikram K; Huang, Xian; Pan, Heng

    2018-05-07

    Bioresorbable electronics refer to a new class of advanced electronics that can completely dissolve or disintegrate with environmentally and biologically benign byproducts in water and biofluids. They have provided a solution to the growing electronic waste problem with applications in temporary usage of electronics such as implantable devices and environmental sensors. Bioresorbable materials such as biodegradable polymers, dissolvable conductors, semiconductors, and dielectrics are extensively studied, enabling massive progress of bioresorbable electronic devices. Processing and patterning of these materials are predominantly relying on vacuum-based fabrication methods so far. However, for the purpose of commercialization, nonvacuum, low-cost, and facile manufacturing/printing approaches are the need of the hour. Bioresorbable electronic materials are generally more chemically reactive than conventional electronic materials, which require particular attention in developing the low-cost manufacturing processes in ambient environment. This review focuses on material reactivity, ink availability, printability, and process compatibility for facile manufacturing of bioresorbable electronics. © 2018 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  2. Recommended design correlations for S-65 beryllium

    International Nuclear Information System (INIS)

    Billone, M.C.

    1995-01-01

    The properties of tritium and helium behavior in irradiated beryllium are reviewed, along with the thermal-mechanical properties needed for ITER design analysis. Correlations are developed to describe the performance of beryllium in a fusion reactor environment. While this paper focuses on the use of beryllium as a plasma-facing component (PFC) material, the correlations presented here can also be used to describe the performance of beryllium as a neutron multiplier for a tritium breeding blanket. The performance properties for beryllium are subdivided into two categories: properties which do not change with irradiation damage to the bulk of the material; and properties which are degraded by neutron irradiation. The irradiation-independent properties described within are: thermal conductivity, specific heat capacity, thermal expansion, and elastic constants. Irradiation-dependent properties include: yield strength, ultimate tensile strength, plastic tangent modulus, uniform and total tensile elongation, thermal and irradiation-induced creep strength, He-induced swelling and tritium retention/release. The approach taken in developing properties correlations is to describe the behavior of dense, pressed S-65 beryllium -- the material chosen for ITER PFC application -- as a function of temperature. As there are essentially no data on the performance of porous and/or irradiated S-65 beryllium, the degradation of properties with as-fabricated porosity and irradiation are determined from the broad data base on S-200F, as well as other types and grades, and applied to S-65 beryllium by scaling factors. The resulting correlations can be used for Be produced by vacuum hot pressing (VHP) and cold-pressing (CP)/sintering(S)/hot-isostatic-pressing (HIP). The performance of plasma-sprayed beryllium is discussed but not quantified

  3. Safeguards System for the Advanced Spent Fuel Conditioning Process Facility

    International Nuclear Information System (INIS)

    Kim, Ho-dong; Lee, T.H.; Yoon, J.S.; Park, S.W; Lee, S.Y.; Li, T.K.; Menlove, H.; Miller, M.C.; Tolba, A.; Zarucki, R.; Shawky, S.; Kamya, S.

    2007-01-01

    The advanced spent fuel conditioning process (ACP) which is a part of a pyro-processing has been under development at Korean Atomic Energy Research Institute (KAERI) since 1997 to tackle the problem of an accumulation of spent fuel. The concept is to convert spent oxide fuel into a metallic form in a high temperature molten salt in order to reduce the heat energy, volume, and radioactivity of a spent fuel. Since the inactive tests of the ACP have been successfully implemented to confirm the validity of the electrolytic reduction technology, a lab-scale hot test will be undertaken in a couple of years to validate the concept. For this purpose, the KAERI has built the ACP Facility (ACPF) at the basement of the Irradiated Material Examination Facility (IMEF) of KAERI, which already has a reserved hot-cell area. Through the bilateral arrangement between US Department of Energy (DOE) and Korean Ministry of Science and Technology (MOST) for safeguards R and D, the KAERI has developed elements of safeguards system for the ACPF in cooperation with the Los Alamos National Laboratory (LANL). The reference safeguards design conditions and equipment were established for the ACPF. The ACPF safeguards system has many unique design specifications because of the particular characteristics of the pyro-process materials and the restrictions during a facility operation. For the material accounting system, a set of remote operation and maintenance concepts has been introduced for a non-destructive assay (NDA) system. The IAEA has proposed a safeguards approach to the ACPF for the different operational phases. Safeguards measures at the ACPF will be implemented during all operational phases which include a 'Cold Test', a 'Hot Test' and at the end of a 'Hot test'. Optimization of the IAEA's inspection efforts was addressed by designing an effective safeguards approach that relies on, inter alia, remote monitoring using cameras, installed NDA instrumentation, gate monitors and seals

  4. Beryllium and growth. III. The effect of beryllium on plant phosphatase

    Energy Technology Data Exchange (ETDEWEB)

    Hoagland, M B

    1952-01-01

    The purpose of the investigations was to correlate the apparent ability of beryllium to substitute for magnesium in plant growth with a specific biochemical effect of the metal. Through association with earlier work on beryllium inhibition of animal alkaline phosphatase, a study was made of the effect of beryllium and other metals upon the activity of a phosphatase derived from tomato leaves. Although only indirect evidence is available that this enzyme system was magnesium-activated, beryllium was found to inhibit reversibly the splitting of GP and ATP. Other metals were also found to be inhibitory but the ATP-ase inhibition - and especially the ratio of P split from GP to P split from ATP - was higher for beryllium than for any other metal studied. The significance of this finding in relation to energy metabolism, growth, and beryllium toxicity is discussed. 12 references, 5 figures, 2 tables.

  5. Spectrographic measurement of beryllium in the atmosphere; Dosage spectrographique du beryllium dans l'atmosphere

    Energy Technology Data Exchange (ETDEWEB)

    Artaud, J; Cittanova, J [Commissariat a l' Energie Atomique, Service d' Analyses et Recherches Chimiques Appliquees, Saclay (France). Centre d' Etudes Nucleaires; Crehange, G; Frequelin, S [Commissariat a l' Energie Atomique, Dir. des Applications Militaires, Service Chimie, Saclay (France). Centre d' Etudes Nucleaires; Baudin, G [Commissariat a l' Energie Atomique, Grenoble (France). Centre d' Etudes Nucleaires

    1961-07-01

    We describe here a method for the spectrographic determination of beryllium on filters which is valid for amounts varying between 0,01 and 30 {mu}g of beryllium and which is independent of the nature of the beryllium compound involved. This is a flux method (graphite-lithium carbonate mixture), the excitation being by a direct current arc. (author) [French] Nous decrivons ici, une methode de dosage spectrographique de beryllium sur filtre, valable pour des teneurs comprises entre 0,01 et 30 {mu}g de beryllium et independante de la nature du compose de beryllium a doser. C'est une methode de 'flux' (melange graphite-carbonate de lithium) l'excitation etant un arc a courant continu. (auteur)

  6. Development of Spectrophotometric Process Monitors for Aqueous Reprocessing Facilities

    International Nuclear Information System (INIS)

    Smith, N.; Krebs, J.; Hebden, A.

    2015-01-01

    The safeguards envelope of an aqueous reprocessing plant can be extended beyond traditional measures to include surveillance of the process chemistry itself. By observing the concentration of accountable species in solution directly, a measure of real time accountancy can be applied. Of equal importance, select information on the process chemistry can be determined that will allow the operator and inspectors to verify that the process is operating as intended. One of the process monitors that can be incorporated is molecular spectroscopy, such as UV-Visible absorption spectroscopy. Argonne National Laboratory has developed a process monitoring system that can be tailored to meet the specific chemistry requirements of a variety of processes. The Argonne Spectroscopic Process monitoring system (ASP) is composed of commercial-off-the-shelf (COTS) spectroscopic hardware, custom manufactured sample handling components (to meet end user requirements) and the custom Plutonium and Uranium Measurement and Acquisition System (PUMAS) software. Two versions of the system have been deployed at the Savannah River Site's H-Canyon facility, tailored for high and low concentration streams. (author)

  7. Metallurgical viewpoints on the brittleness of beryllium

    Energy Technology Data Exchange (ETDEWEB)

    Lagerberg, G

    1960-02-15

    At present the development and use of beryllium metal for structural applications is severely hampered by its brittleness. Reasons for this lack of ductility are reviewed in discussing the deformation behaviour of beryllium in relation to other hexagonal metals. The ease of fracturing in beryllium is assumed to be a consequence of a limited number of deformation modes in combination with high deformation resistance. Models for the nucleation of fracture are suggested. The relation of ductility to elastic constants as well as to grain size, texture and alloying additions is discussed.

  8. THE IDAHO NATIONAL LABORATORY BERYLLIUM TECHNOLOGY UPDATE

    International Nuclear Information System (INIS)

    Glen R. Longhurst

    2007-01-01

    A Beryllium Technology Update meeting was held at the Idaho National Laboratory on July 18, 2007. Participants came from the U.S., Japan, and Russia. There were two main objectives of this meeting. One was a discussion of current technologies for beryllium in fission reactors, particularly the Advanced Test Reactor and the Japan Materials Test Reactor, and prospects for material availability in the coming years. The second objective of the meeting was a discussion of a project of the International Science and Technology Center regarding treatment of irradiated beryllium for disposal. This paper highlights discussions held during that meeting and major conclusions reached

  9. Metallurgical viewpoints on the brittleness of beryllium

    International Nuclear Information System (INIS)

    Lagerberg, G.

    1960-02-01

    At present the development and use of beryllium metal for structural applications is severely hampered by its brittleness. Reasons for this lack of ductility are reviewed in discussing the deformation behaviour of beryllium in relation to other hexagonal metals. The ease of fracturing in beryllium is assumed to be a consequence of a limited number of deformation modes in combination with high deformation resistance. Models for the nucleation of fracture are suggested. The relation of ductility to elastic constants as well as to grain size, texture and alloying additions is discussed

  10. The Status of Beryllium Research for Fusion in the United States

    International Nuclear Information System (INIS)

    Glen R. Longhurst

    2003-01-01

    Use of beryllium in fusion reactors has been considered for neutron multiplication in breeding blankets and as an oxygen getter for plasma-facing surfaces. Previous beryllium research for fusion in the United States included issues of interest to fission (swelling and changes in mechanical and thermal properties) as well as interactions with plasmas and hydrogen isotopes and methods of fabrication. When the United States formally withdrew its participation in the International Thermonuclear Experimental Reactor (ITER) program, much of this effort was terminated. The focus in the U.S. has been mainly on toxic effects of beryllium and on industrial hygiene and health-related issues. Work continued at the INEEL and elsewhere on beryllium-containing molten salts. This activity is part of the JUPITER II Agreement. Plasma spray of ITER first wall samples at Los Alamos National Laboratory has been performed under the European Fusion Development Agreement. Effects of irradiation on beryllium structure are being studied at Oak Ridge National Laboratory. Numerical and phenomenological models are being developed and applied to better understand important processes and to assist with design. Presently, studies are underway at the University of California Los Angeles to investigate thermo-mechanical characteristics of beryllium pebble beds, similar to research being carried out at Forschungszentrum Karlsruhe and elsewhere. Additional work, not funded by the fusion program, has dealt with issues of disposal, and recycling

  11. Quality control through dosimetry at a contract radiation processing facility

    International Nuclear Information System (INIS)

    Du Plessis, T.A.; Roediger, A.H.A.

    1985-01-01

    Reliable dosimetry procedures constitute a very important part of process control and quality assurance at a contract gamma radiation processing facility that caters for a large variety of different radiation applications. The choice, calibration and routine intercalibration of the dosimetry systems employed form the basis of a sound dosimetry policy in radiation processing. With the dosimetric procedures established, detailed dosimetric mapping of the irradiator upon commissioning (and whenever source modifications take place) is carried out to determine the radiation processing characteristics and peformance of the plant. Having established the irradiator parameters, routine dosimetry procedures, being part of the overall quality control measures, are employed. In addition to routine dosimetry, independent monitoring of routine dosimetry is performed on a bi-monthly basis and the results indicate a variation of better than 3%. On an annaul basis the dosimetry systems are intercalibrated through at least one primary standard dosimetry laboratory and to date a variation of better than 5% has been experienced. The company also participates in the Pilot Dose Assurance Service of the International Atomic Energy Agency, using the alanine/ESR dosimetry system. Routine calibration of the instrumentation employed is carried out on a regular basis. Detailed permanent records are compiled on all dosimetric and instrumentation calibrations, and the routine dosimetry employed at the plant. Certificates indicating the measured absorbed radiation doses are issued on request and in many cases are used for the dosimetric release of sterilized medical and pharmaceutical products. These procedures, used by Iso-Ster at its industrial gamma radiation facility, as well as the experience built up over a number of years using radiation dosimetry for process control and quality assurance are discussed. (author)

  12. Measurement of neutron yield by 62 MeV proton beam on a thick Beryllium target

    International Nuclear Information System (INIS)

    Alba, R; Cosentino, G; Zoppo, A Del; Pietro, A Di; Figuera, P; Finocchiaro, P; Maiolino, C; Santonocito, D; Schillaci, M; Barbagallo, M; Colonna, N; Boccaccio, P; Esposito, J; Celentano, A; Osipenko, M; Ricco, G; Ripani, M; Viberti, C M; Kostyukov, A

    2013-01-01

    In the framework of research on IVth generation reactors and high intensity neutron sources a low-power prototype neutron amplifier was recently proposed by INFN. It is based on a low-energy, high current proton cyclotron, whose beam, impinging on a thick Beryllium converter, produces a fast neutron spectrum. The world database on the neutron yield from thick Beryllium target in the 70 MeV proton energy domain is rather scarce. The new measurement was performed at LNS, covering a wide angular range from 0 to 150 degrees and an almost complete neutron energy interval. In this contribution the preliminary data are discussed together with the proposed ADS facility.

  13. Development of the NBS beryllium isotopic standard reference material

    International Nuclear Information System (INIS)

    Inn, K.G.W; Fassett, J.D.; Coursey, B.M.; Walker, R.L.; Raman, S.

    1987-01-01

    The National Bureau of Standards, in conjunction with the Oak Ridge National Laboratory and the Accelerator Mass Spectrometry community, is in the process of developing a beryllium isotopic solution Standard Reference Material. The master 10 Be/ 9 Be solution was characterized isotopically by resonance-ionization and secondary-ion mass-spectrometric-based techniques, and radioactivity measurements were by liquid scintillation counting. The master solution was gravimetrically diluted with 9 Be to a final 10 Be/ 9 Be atomic ratio of 3 x 10 -11 . The preliminary data indicate a half life for 10 Be of 1.3 million years, and AMS measurements are within 10% of the known beryllium isotopic ratio

  14. An Automated 476 MHz RF Cavity Processing Facility at SLAC

    CERN Document Server

    McIntosh, P; Schwarz, H

    2003-01-01

    The 476 MHz accelerating cavities currently used at SLAC are those installed on the PEP-II B-Factory collider accelerator. They are designed to operate at a maximum accelerating voltage of 1 MV and are routinely utilized on PEP-II at voltages up to 750 kV. During the summer of 2003, SPEAR3 will undergo a substantial upgrade, part of which will be to replace the existing 358.54 MHz RF system with essentially a PEP-II high energy ring (HER) RF station operating at 476.3 MHz and 3.2 MV (or 800 kV/cavity). Prior to installation, cavity RF processing is required to prepare them for use. A dedicated high power test facility is employed at SLAC to provide the capability of conditioning each cavity up to the required accelerating voltage. An automated LabVIEW based interface controls and monitors various cavity and test stand parameters, increasing the RF fields accordingly such that stable operation is finally achieved. This paper describes the high power RF cavity processing facility, highlighting the features of t...

  15. Radiation Monitoring System in Advanced Spent Fuel Conditioning Process Facility

    Energy Technology Data Exchange (ETDEWEB)

    You, Gil Sung; Kook, D. H.; Choung, W. M.; Ku, J. H.; Cho, I. J.; You, G. S.; Kwon, K. C.; Lee, W. K.; Lee, E. P

    2006-09-15

    The Advanced spent fuel Conditioning Process is under development for effective management of spent fuel by converting UO{sub 2} into U-metal. For demonstration of this process, {alpha}-{gamma} type new hot cell was built in the IMEF basement . To secure against radiation hazard, this facility needs radiation monitoring system which will observe the entire operating area before the hot cell and service area at back of it. This system consists of 7 parts; Area Monitor for {gamma}-ray, Room Air Monitor for particulate and iodine in both area, Hot cell Monitor for hot cell inside high radiation and rear door interlock, Duct Monitor for particulate of outlet ventilation, Iodine Monitor for iodine of outlet duct, CCTV for watching workers and material movement, Server for management of whole monitoring system. After installation and test of this, radiation monitoring system will be expected to assist the successful ACP demonstration.

  16. Materials evaluation programs at the Defense Waste Processing Facility

    International Nuclear Information System (INIS)

    Gee, J.T.; Iverson, D.C.; Bickford, D.F.

    1992-01-01

    The Savannah River Site (SRS) has been operating a nuclear fuel cycle since the 1950s to produce nuclear materials in support of the national defense effort. About 83 million gallons of high-level waste produced since operations began has been consolidated by evaporation into 33 million gallons at the waste tank farm. The Department of Energy authorized the construction of the Defense Waste Processing Facility (DWPF), the function of which is to immobilize the waste as a durable borosilicate glass contained in stainless steel canisters prior to the placement of the canisters in a federal repository. The DWPF is now mechanically complete and is undergoing commissioning and run-in activities. A brief description of the DWPF process is provided

  17. Radiation Monitoring System in Advanced Spent Fuel Conditioning Process Facility

    International Nuclear Information System (INIS)

    You, Gil Sung; Kook, D. H.; Choung, W. M.; Ku, J. H.; Cho, I. J.; You, G. S.; Kwon, K. C.; Lee, W. K.; Lee, E. P.

    2006-09-01

    The Advanced spent fuel Conditioning Process is under development for effective management of spent fuel by converting UO 2 into U-metal. For demonstration of this process, α-γ type new hot cell was built in the IMEF basement . To secure against radiation hazard, this facility needs radiation monitoring system which will observe the entire operating area before the hot cell and service area at back of it. This system consists of 7 parts; Area Monitor for γ-ray, Room Air Monitor for particulate and iodine in both area, Hot cell Monitor for hot cell inside high radiation and rear door interlock, Duct Monitor for particulate of outlet ventilation, Iodine Monitor for iodine of outlet duct, CCTV for watching workers and material movement, Server for management of whole monitoring system. After installation and test of this, radiation monitoring system will be expected to assist the successful ACP demonstration

  18. Combined aging of beryllium bronze

    International Nuclear Information System (INIS)

    Duraev, P.P.; Kaplun, Yu.A.; Pastukhova, Zh.P.; Rakhshtadt, A.G.

    1986-01-01

    This article evaluates the possibility of increasing the resistance of beryllium bronze to small plastic deformations as a result of the application of stepped aging under stress. Low-temperature aging under conditions of bending under a stress of about 100 MPa was applied to alloy BrBNT1, 9Mg at 150, 180, and 210 0 C, high-temperature aging at 300 and 340 0 C under stress and without stress. As a result of applying stepped aging under stress, the elastic limit of the alloy BrBNT1, 9Mg was raised to 900 MPa. Stepped aging under stress has a substantial effect on the relaxation stability of the alloy. The procedure suggested in the article for aging may be used efficiently for treating elastic elements made of other brands of bronze as well

  19. The adhesive bonding of beryllium structural components

    International Nuclear Information System (INIS)

    Fullerton-Batten, R.C.

    1977-01-01

    Where service conditions permit, adhesive bonding is a highly recommendable, reliable means of joining beryllium structural parts. Several important programs have successfully used adhesive bonding for joining structural and non-structural beryllium components. Adhesive bonding minimizes stress concentrations associated with other joining techniques and considerably improves fatigue resistance. In addition, no degradation of base metal properties occur. In many instances, structural joints can be fabricated more cheaply by adhesive bonding or in combination with adhesive bonding than by any other method used alone. An evaluation program on structural adhesive bonding of beryllium sheet components is described. A suitable surface pretreatment for beryllium adherends prior to bonding is given. Tensile shear strength and fatigue properties of FM 1000 and FM 123-5 adhesive bonded joints are reviewed and compared with data obtained from riveted joints of similar geometry. (author)

  20. Chronic Beryllium Disease Prevention Program Report

    Energy Technology Data Exchange (ETDEWEB)

    Lee, S

    2012-03-29

    This document describes how Lawrence Livermore National Laboratory (LLNL) meets the requirements and management practices of federal regulation 10 CFR 850, 'Chronic Beryllium Disease Prevention Program (CBDPP).' This revision of the LLNL CBDPP incorporates clarification and editorial changes based on lessons learned from employee discussions, observations and reviews of Department of Energy (DOE) Complex and commercial industry beryllium (Be) safety programs. The information is used to strengthen beryllium safety practices at LLNL, particularly in the areas of: (1) Management of small parts and components; and (2) Communication of program status to employees. Future changes to LLNL beryllium activities and on-going operating experience will be incorporated into the program as described in Section S, 'Performance Feedback.'

  1. Beryllium-aluminum alloys for investment castings

    International Nuclear Information System (INIS)

    Nachtrab, W.T.; Levoy, N.

    1997-01-01

    Beryllium-aluminum alloys containing greater than 60 wt % beryllium are very favorable materials for applications requiring light weight and high stiffness. However, when produced by traditional powder metallurgical methods, these alloys are expensive and have limited applications. To reduce the cost of making beryllium-aluminum components, Nuclear Metals Inc. (NMI) and Lockheed Martin Electronics and Missiles have recently developed a family of patented beryllium-aluminum alloys that can be investment cast. Designated Beralcast, the alloys can achieve substantial weight savings because of their high specific strength and stiffness. In some cases, weight has been reduced by up to 50% over aluminum investment casting. Beralcast is now being used to make thin wall precision investment castings for several advanced aerospace applications, such as the RAH-66 Comanche helicopter and F-22 jet fighter. This article discusses alloy compositions, properties, casting method, and the effects of cobalt additions on strength

  2. Experiments on tritium behavior in beryllium, (2)

    International Nuclear Information System (INIS)

    Ishitsuka, Etsuo; Kawamura, Hiroshi; Nakata, Hirokatsu; Sugai, Hiroyuki; Tanase, Masakazu.

    1990-02-01

    Beryllium has been used as the neutron reflector of material testing reactor and as the neutron multiplier for the fusion reactor lately. To study the tritium behavior in beryllium, we conducted the experiments, i.e., tritium release by recoil or diffusion by using the hot-pressed beryllium which had been produced both tritium and helium by neutron irradiation. From our experiments, we found that (1) amount of tritium production per one cycle irradiation (lasting 22 days) of JMTR is 10 mCi/g, (2) amount of tritium per surface area of hot-pressed beryllium released by recoil is 4 μCi/cm 2 , (3) diffusion coefficient of tritium in a temperature range of 800 ∼1180degC can be expressed with the following equation; D = 8.7 x 10 4 exp(-2.9x10 5 /R/T) cm 2 /s. (author)

  3. Design Methodology of Process Layout considering Various Equipment Types for Large scale Pyro processing Facility

    International Nuclear Information System (INIS)

    Yu, Seung Nam; Lee, Jong Kwang; Lee, Hyo Jik

    2016-01-01

    At present, each item of process equipment required for integrated processing is being examined, based on experience acquired during the Pyropocess Integrated Inactive Demonstration Facility (PRIDE) project, and considering the requirements and desired performance enhancement of KAPF as a new facility beyond PRIDE. Essentially, KAPF will be required to handle hazardous materials such as spent nuclear fuel, which must be processed in an isolated and shielded area separate from the operator location. Moreover, an inert-gas atmosphere must be maintained, because of the radiation and deliquescence of the materials. KAPF must also achieve the goal of significantly increased yearly production beyond that of the previous facility; therefore, several parts of the production line must be automated. This article presents the method considered for the conceptual design of both the production line and the overall layout of the KAPF process equipment. This study has proposed a design methodology that can be utilized as a preliminary step for the design of a hot-cell-type, large-scale facility, in which the various types of processing equipment operated by the remote handling system are integrated. The proposed methodology applies to part of the overall design procedure and contains various weaknesses. However, if the designer is required to maximize the efficiency of the installed material-handling system while considering operation restrictions and maintenance conditions, this kind of design process can accommodate the essential components that must be employed simultaneously in a general hot-cell system

  4. Beryllium concentration in pharyngeal tonsils in children

    Directory of Open Access Journals (Sweden)

    Ewa Nogaj

    2014-06-01

    Full Text Available Power plant dust is believed to be the main source of the increased presence of the element beryllium in the environment which has been detected in the atmospheric air, surface waters, groundwater, soil, food, and cigarette smoke. In humans, beryllium absorption occurs mainly via the respiratory system. The pharyngeal tonsils are located on the roof of the nasopharynx and are in direct contact with dust particles in inhaled air. As a result, the concentration levels of beryllium in the pharyngeal tonsils are likely to be a good indicator of concentration levels in the air. The presented study had two primary aims: to investigate the beryllium concentration in pharyngeal tonsils in children living in southern Poland, and the appropriate reference range for this element in children’s pharyngeal tonsils. Pharyngeal tonsils were extracted from a total of 379 children (age 2–17 years, mean 6.2 ± 2.7 years living in southern Poland. Tonsil samples were mineralized in a closed cycle in a pressure mineralizer PDS 6, using 65% spectrally pure nitric acid. Beryllium concentration was determined using the ICP-AES method with a Perkin Elmer Optima 5300DVTM. The software Statistica v. 9 was used for the statistical analysis. It was found that girls had a significantly greater beryllium concentration in their pharyngeal tonsils than boys. Beryllium concentration varies greatly, mostly according to the place of residence. Based on the study results, the reference value for beryllium in pharyngeal tonsils of children is recommended to be determined at 0.02–0.04 µg/g.

  5. Waste receiving and processing facility module 1, detailed design report

    International Nuclear Information System (INIS)

    1993-10-01

    WRAP 1 baseline documents which guided the technical development of the Title design included: (a) A/E Statement of Work (SOW) Revision 4C: This DOE-RL contractual document specified the workscope, deliverables, schedule, method of performance and reference criteria for the Title design preparation. (b) Functional Design Criteria (FDC) Revision 1: This DOE-RL technical criteria document specified the overall operational criteria for the facility. The document was a Revision 0 at the beginning of the design and advanced to Revision 1 during the tenure of the Title design. (c) Supplemental Design Requirements Document (SDRD) Revision 3: This baseline criteria document prepared by WHC for DOE-RL augments the FDC by providing further definition of the process, operational safety, and facility requirements to the A/E for guidance in preparing the design. The document was at a very preliminary stage at the onset of Title design and was revised in concert with the results of the engineering studies that were performed to resolve the numerous technical issues that the project faced when Title I was initiated, as well as, by requirements established during the course of the Title II design

  6. [Design of an HACCP program for a cocoa processing facility].

    Science.gov (United States)

    López D'Sola, Patrizia; Sandia, María Gabriela; Bou Rached, Lizet; Hernández Serrano, Pilar

    2012-12-01

    The HACCP plan is a food safety management tool used to control physical, chemical and biological hazards associated to food processing through all the processing chain. The aim of this work is to design a HACCP Plan for a Venezuelan cocoa processing facility.The production of safe food products requires that the HACCP system be built upon a solid foundation of prerequisite programs such as Good Manufacturing Practices (GMP) and Sanitation Standard Operating Procedures (SSOP). The existence and effectiveness of these prerequisite programs were previously assessed.Good Agriculture Practices (GAP) audit to cocoa nibs suppliers were performed. To develop the HACCP plan, the five preliminary tasks and the seven HACCP principles were accomplished according to Codex Alimentarius procedures. Three Critical Control Points (CCP) were identified using a decision tree: winnowing (control of ochratoxin A), roasting (Salmonella control) and metallic particles detection. For each CCP, Critical limits were established, the Monitoring procedures, Corrective actions, Procedures for Verification and Documentation concerning all procedures and records appropriate to these principles and their application was established. To implement and maintain a HACCP plan for this processing plant is suggested. Recently OchratoxinA (OTA) has been related to cocoa beans. Although the shell separation from the nib has been reported as an effective measure to control this chemical hazard, ochratoxin prevalence study in cocoa beans produced in the country is recommended, and validate the winnowing step as well

  7. Beryllium armored mockups for fusion high flux application

    International Nuclear Information System (INIS)

    Gervash, A.; Giniyatulin, R.; Mazul, I.; Watson, R.

    1998-01-01

    One of the main requirements to use Be as a candidate for plasma facing component in ITER is providing a reliable joint between Be and Cu-alloy heat sink structure. In this work authors present the results of recent activity on this way. To create Be/CuCrZr joints the unique fast e-beam brazing technology was developed in Russia. The numbers of Be/CuCrZr mock-ups were manufactured in Efremov Institute by fast e-beam brazing using Cu-Sn-In-Ni brazing alloy. These mock-ups were tested by Sandia Laboratory at the EBTS electron beam facility. The goals of the tests were to define the allowable dimensions of the armour tiles for the heat loads of more than 10 MW/m 2 , to find the limit of bond strength for the Be/CuCrZr joint and response to heat loads and to estimate the life time of the brazed tiles by thermo-cyclic testing. The screening and thermal fatigue results are presented. With the aim to check the applicability of developed fast brazing process to DS-Cu alloy (GlidCop) the actively cooled mock-up produced by fast brazing Be onto GlidCop structure was manufactured and tested in Efremov Institute, main results are discussed. Summarizing testing data authors show that the CuCrZr or DS-Cu heat sink armored with beryllium can survive high heat fluxes (≥10 MW/m 2 ) during thousand heating/cooling cycles without serious damaging in the armour material and its joint. (author)

  8. Fracture testing and performance of beryllium copper alloy C17510

    International Nuclear Information System (INIS)

    Murray, H.A.; Zatz, I.J.

    1994-05-01

    When a literature search and discussion with manufacturers revealed that there was virtually no existing data related to the fracture properties and behavior of copper beryllium alloy C17510, a series of test programs was undertaken to ascertain this information for several variations in material processing and chemistry. These variations in C17510 were primarily optimized for combinations of strength and conductivity. While originally intended for use as cyclically loaded high-field, high-strength conductors in fusion energy research, material testing of C17510 has indicated that it is an attractive and economical alternative for a host of other structural, mechanical and electrical applications. ASTM tests performed on three variations of C17510 alloys included both J-integral and plane strain fracture toughness testing and fatigue crack growth rate tests, as well as verifying tensile, hardness, Charpy, and other well defined mechanical properties. Fracture testing was performed at both room and liquid nitrogen temperatures, which bound the thermal environment anticipated for the fusion components being designed. Fatigue crack propagation stress ratios ranged from nominal zero to minus one at each temperature. In order to confirm the test results, duplicate and independent test programs were awarded to separate facilities with appropriate test experience, whenever possible. The primary goal of the test program, to determine and bound the fracture toughness and Paris constants for C17510,was accomplished. In addition, a wealth of information was accumulated pertaining to crack growth characteristics, effects of directionality and potential testing pitfalls. The paper discusses the test program and its findings in detail

  9. Establishing a central waste processing and storage facility in Ghana

    International Nuclear Information System (INIS)

    Glover, E.T.; Fletcher, J.J.; Darko, E.O.

    2001-01-01

    Radioactive waste and spent sealed sources in Ghana are generated from various nuclear applications - diagnostic and therapeutic procedures in medicine, measurement and processing techniques in industry, irradiation techniques for food preservation and sterilization of medical products and a research reactor for research and teaching. Statistics available indicate that over 15 institutions in Ghana are authorized to handle radiation sources. At present radioactive waste and spent sealed sources are collected and stored in the interim facility without conditioning. With the increasing use of radioactive sources in the industry, medicine for diagnostic and therapeutic purpose and research and teaching, the volume of waste is expected to increase. The radioactive waste expected include spent ion exchange resins from the nuclear reactor water purification system, incompactible solid waste from mechanical filter, liquid and organic waste and spent sealed sources. It is estimated that four 200L drums will be needed annually to condition the waste to be generated. The National Radioactive Waste Management Centre (NRWMC) was therefore established to carry radioactive waste safety operations in Ghana and research to ensure that each waste type is managed in the most appropriate manner. Its main task includes development and establishment of the radioactive waste management infrastructure with a capacity considering the future nuclear technology development in Ghana. The first phase covers the establishment of administrative structure, development of basic regulations and construction of the radioactive waste processing and storage facility. The Ghana Radioactive Waste Management regulation has been presented to the Parliament of Ghana for consideration. The initial draft was reviewed by the RPB. A 3-day national seminar on the Understanding and Implementation of the Regulation on Radioactive Waste Management in Ghana was held to discuss and educate the general public on the

  10. The Defense Waste Processing Facility, from vision to reality

    International Nuclear Information System (INIS)

    Randall, C.T.

    2000-01-01

    When the Savannah River Plant began operation in the early 1950's producing nuclear materials for the National defense, liquid, highly radioactive waste was generated as a by-product. Since that time the waste has been stored in large, carbon steel tanks that are buried underground. In 1960 one of the tanks developed a leak, and before recovery measures could be taken, about 25-gallons of radioactive salt solution had overflowed the secondary liner and seeped into the soil surrounding the tank. Significant improvements to the tanks were made, but constant surveillance was still required. Thus, the opinion began forming that storage of the mobile, highly radioactive waste in tanks was not a responsible long-term practice. So in the late 1960's the Savannah River Laboratory began research to find a suitable long-term solution to the waste disposal problem. Several alternative waste forms were evaluated, and in 1972 the first Savannah River waste was vitrified on a laboratory scale. By the mid-1970's, the DuPont Company, prime contractor at the Savannah River Plant, began to develop a vision of constructing America's first vitrification plant to immobilize the high level radioactive waste in borosilicate glass. This vision was later championed by DuPont in the form of a vitrification plant called the Defense Waste Processing Facility (DWPF). Today, the DWPF processes Savannah River High Level Waste sludge turning it into a solid, durable waste form of borosilicate glass. The DWPF is the world's largest vitrification facility. It was brought to reality through over 25-years of research and 13-years of careful construction, tests, and reviews at a cost of approximately $3 billion dollars

  11. Beryllium satellite thrust cone design, manufacture and test

    International Nuclear Information System (INIS)

    Schneiter, H.; Chandler, D.

    1977-01-01

    Pre-formed beryllium sheet material has been used in the design, manufacturing and test of a satellite thrust cone structure. Adhesive bonding was used for attachment of aluminium flanges and conical segment lap strips. Difficulties in beryllium structure design such as incompatibilities with aluminium and handling problems are discussed. Testing to optimize beryllium-beryllium and beryllium-aluminium adhesive bonds is described. The completed thrust cone assembly has been subjected to static load testing and the results are presented. A summary of the relative merits of the use of beryllium in satellite structures is given with recommendations for future users. (author)

  12. Process Technical Basis Documentation Diagram for a solid-waste processing facility

    International Nuclear Information System (INIS)

    Benar, C.J.; Petersen, C.A.

    1994-02-01

    The Process Technical Basis Documentation Diagram is for a solid-waste processing facility that could be designed to treat, package, and certify contact-handled mixed low-level waste for permanent disposal. The treatment processes include stabilization using cementitious materials and immobilization using a polymer material. The Diagram identifies several engineering/demonstration activities that would confirm the process selection and process design. An independent peer review was conducted at the request of Westinghouse Hanford Company to determine the technical adequacy of the technical approach for waste form development. The peer review panel provided comments and identified documents that it felt were needed in the Diagram as precedence for Title I design. The Diagram is a visual tool to identify traceable documentation of key activities, including those documents suggested by the peer review, and to show how they relate to each other. The Diagram is divided into three sections: (1) the Facility section, which contains documents pertaining to the facility design, (2) the Process Demonstration section, which contains documents pertaining to the process engineering/demonstration work, and 3) the Regulatory section, which contains documents describing the compliance strategy for each acceptance requirement for each feed type, and how this strategy will be implemented

  13. A survey of decontamination processes applicable to DOE nuclear facilities

    Energy Technology Data Exchange (ETDEWEB)

    Chen, L.; Chamberlain, D.B.; Conner, C.; Vandegrift, G.F.

    1997-05-01

    The objective of this survey was to select an appropriate technology for in situ decontamination of equipment interiors as part of the decommissioning of U.S. Department of Energy nuclear facilities. This selection depends on knowledge of existing chemical decontamination methods. This report provides an up-to-date review of chemical decontamination methods. According to available information, aqueous systems are probably the most universally used method for decontaminating and cleaning metal surfaces. We have subdivided the technologies, on the basis of the types of chemical solvents, into acid, alkaline permanganate, highly oxidizing, peroxide, and miscellaneous systems. Two miscellaneous chemical decontamination methods (electrochemical processes and foam and gel systems) are also described. A concise technical description of various processes is given, and the report also outlines technical considerations in the choice of technologies, including decontamination effectiveness, waste handing, fields of application, and the advantages and limitations in application. On the basis of this survey, six processes were identified for further evaluation. 144 refs., 2 tabs.

  14. A survey of decontamination processes applicable to DOE nuclear facilities

    International Nuclear Information System (INIS)

    Chen, L.; Chamberlain, D.B.; Conner, C.; Vandegrift, G.F.

    1997-05-01

    The objective of this survey was to select an appropriate technology for in situ decontamination of equipment interiors as part of the decommissioning of U.S. Department of Energy nuclear facilities. This selection depends on knowledge of existing chemical decontamination methods. This report provides an up-to-date review of chemical decontamination methods. According to available information, aqueous systems are probably the most universally used method for decontaminating and cleaning metal surfaces. We have subdivided the technologies, on the basis of the types of chemical solvents, into acid, alkaline permanganate, highly oxidizing, peroxide, and miscellaneous systems. Two miscellaneous chemical decontamination methods (electrochemical processes and foam and gel systems) are also described. A concise technical description of various processes is given, and the report also outlines technical considerations in the choice of technologies, including decontamination effectiveness, waste handing, fields of application, and the advantages and limitations in application. On the basis of this survey, six processes were identified for further evaluation. 144 refs., 2 tabs

  15. Beryllium-rich intermediate phases in beryllium alloys

    International Nuclear Information System (INIS)

    Raynor, G.V.

    1977-01-01

    The results of a survey of the factors affecting the formation of phases of stoichiometry MBe 5 , M 2 Be 17 , MBe 12 and MBe 13 are presented. Using published information it is shown that the structures adopted at the higher Be:M ratios involve different characteristics from those adopted at lower Be:M ratios. In the ThMn 12 and NaZn 13 structures adopted in the former case dsub(M-Be) > (rsub(M) + rsub(Be)), (where the atomic radii refer to coordination number 12) and the dsub(Be-Be) distances are contracted. In the CaCu 5 structure adopted at composition MBe 5 , dsub(M-Be) < (rsub(M) + rsub(Be)) and interactions between unlike atoms are significant. The structural characteristics, occurrence, and the stabilities of these phases, and of the others mentioned above, are discussed in terms of atomic radius ratios, the position of the M component in the periodic table, and the value of the univalent ionic radius of the M component, taken as a measure of the extension in space of the hard incompressible ionic core. Though the compound-forming characteristics of beryllium are largely dictated by its small atomic diameter, other factors such as the nature of the bonding which can be exerted by the partner atoms are also significant. In particular, the proportions of the volumes of the atoms which are occupied by the hard incompressible ionic cores assume importance. (author)

  16. Thermomechanical testing of beryllium for the JET/ISX-B beryllium limiter experiment

    International Nuclear Information System (INIS)

    Watson, R.D.; Smith, M.F.; Whitley, J.B.; McDonald, J.M.

    1984-01-01

    Materials testing of S-65-B beryllium has been conducted in support of the beryllium limiter experiment on the ISX-B tokamak. The S-65-B grade of hot-pressed beryllium was chosen over S-200-E because of its superior strength and ductility at elevated temperatures. The testing has included measurement of tensile and yield strength, ductility, Young's Modulus, thermal conductivity, and specific heat from 50 0 C to 700 0 C. Thermal fatigue testing of a 2.5 cm beryllium cube was conducted using an electron beam to apply a heat flux of 2.5 kw/cm 2 for 0.3 second pulses for 1500 cycles. Results from the tests are compared to elastic-plastic finite element stress calculations. The testing indicates that the ISX-B beryllium limiter should survive the tokamak environment without serious structural failure, although some surface cracking is expected to occur. (author)

  17. Design of remote handled process assemblies for the process facility modifications project

    International Nuclear Information System (INIS)

    Smets, J.L.; Ajifu, D.A.

    1987-01-01

    The modular design philosophy for the process facility modification project utilizes an integrated design of components to facilitate operations and maintenance of nuclear fuel reprocessing equipment in a hot cell environment. The utilization of a matrix of remoteable base frames combines with process equipment designed as remote assemblies and sub-assemblies has simplified the overall design. Modularity will allow future flexibility while providing advantages for construction and maintenance in the initial installation

  18. Participative Facility Planning for Obstetrical and Neonatal Care Processes: Beginning of Life Process

    Directory of Open Access Journals (Sweden)

    Jori Reijula

    2016-01-01

    Full Text Available Introduction. Old hospitals may promote inefficient patient care processes and safety. A new, functionally planned hospital presents a chance to create an environment that supports streamlined, patient-centered healthcare processes and adapts to users’ needs. This study depicts the phases of a facility planning project for pregnant women and newborn care processes (beginning of life process at Turku University Hospital. Materials and Methods. Project design reports and meeting documents were utilized to assess the beginning of life process as well as the work processes of the Women’s and Children’s Hospital. Results. The main elements of the facility design (FD project included rigorous preparation for the FD phase, functional planning throughout the FD process, and setting key values: (1 family-centered care, (2 Lean thinking and Lean tools as the framework for the FD process, (3 safety, and (4 cooperation. Conclusions. A well-prepared FD project with sufficient insight into functional planning, Lean thinking, and user-centricity seemed to facilitate the actual FD process. Although challenges occurred, the key values were not forgone and were successfully incorporated into the new hospital building.

  19. Recent results on high thermal energy load testing of beryllium for ITER first wall application

    Science.gov (United States)

    Kupriyanov, I. B.; Roedig, M.; Nikolaev, G. N.; Kurbatova, L. A.; Linke, J.; Gervash, A. A.; Giniyatulin, R. N.; Podkovyrov, V. L.; Muzichenko, A. D.; Khimchenko, L.

    2011-12-01

    In this paper, progress in the high heat flux (HHF) qualification testing of TGP-56FW beryllium grade for ITER first wall applications is presented. Two actively cooled Be/CuCrZr brazing mock-ups were tested under complex thermal loading conditions in the electron beam facility JUDITH-1 (step 1: vertical displacement event test at 40 MJ m-2, 0.3 s, 1 shot; step 2: disruption tests at 3 MJ m-2, 1 shot, Δt=5 ms; step 3: repetitive fatigue test at 80 MW m-2, 1000 shots, Δt=25 ms). After testing, metallographic investigations on the microstructure and crack morphology were carried out. The results of these studies of Be tiles are reported and discussed. The overall results of TGP-56FW grade qualification testing have demonstrated the reliable performance capability of TGP-56FW for application as the armor of the ITER first wall. In addition, the results of first experiments with TGP-56FW and S-65C beryllium grades in the QSPA-Be plasma gun facility are also reported. In these experiments, beryllium tiles (80×80×10 mm3) were tested in a hydrogen plasma stream (5 cm in diameter) with pulse duration 0.5 ms and heat loads of 0.5-2 MJ m-2. Experiments were performed at room temperature. The evolution of the surface microstructure and mass loss of beryllium exposed to up to 100 shots is presented.

  20. Recent results on high thermal energy load testing of beryllium for ITER first wall application

    International Nuclear Information System (INIS)

    Kupriyanov, I B; Nikolaev, G N; Kurbatova, L A; Roedig, M; Linke, J; Gervash, A A; Giniyatulin, R N; Podkovyrov, V L; Muzichenko, A D; Khimchenko, L

    2011-01-01

    In this paper, progress in the high heat flux (HHF) qualification testing of TGP-56FW beryllium grade for ITER first wall applications is presented. Two actively cooled Be/CuCrZr brazing mock-ups were tested under complex thermal loading conditions in the electron beam facility JUDITH-1 (step 1: vertical displacement event test at 40 MJ m - 2, 0.3 s, 1 shot; step 2: disruption tests at 3 MJ m - 2, 1 shot, Δt=5 ms; step 3: repetitive fatigue test at 80 MW m - 2, 1000 shots, Δt=25 ms). After testing, metallographic investigations on the microstructure and crack morphology were carried out. The results of these studies of Be tiles are reported and discussed. The overall results of TGP-56FW grade qualification testing have demonstrated the reliable performance capability of TGP-56FW for application as the armor of the ITER first wall. In addition, the results of first experiments with TGP-56FW and S-65C beryllium grades in the QSPA-Be plasma gun facility are also reported. In these experiments, beryllium tiles (80×80×10 mm 3 ) were tested in a hydrogen plasma stream (5 cm in diameter) with pulse duration 0.5 ms and heat loads of 0.5-2 MJ m - 2. Experiments were performed at room temperature. The evolution of the surface microstructure and mass loss of beryllium exposed to up to 100 shots is presented.

  1. Modern integrated environmental monitoring and processing systems for nuclear facilities

    International Nuclear Information System (INIS)

    Oprea, I.

    2000-01-01

    presentation by using on-line dynamic evolution of the events, environment information, evacuation optimization, image and voice processing. These modern systems are proposed for environmental monitoring around nuclear facilities, as open interactive systems supporting the operator in the global overview of the environment and the status of the situation updating the remote GIS data base, assuring man-computer interaction and a good information flow for emergency knowledge exchange, improving the protection of the population and decision makers efforts. The local monitoring systems could be integrated into national or international environmental monitoring systems, achieving desired interoperability between government, civilian and army in disaster preparedness efforts

  2. Preliminary design for the Waste Receiving And Processing Facility Module 1: Volume 3, Outline specifications

    International Nuclear Information System (INIS)

    1992-03-01

    This report presents specifications related to the buildings and equipment of the wrap facility. The facility will retrieve, process, and certify transuranic, mixed, and low-level radioactive wastes for disposal

  3. 75 FR 80734 - Chronic Beryllium Disease Prevention Program

    Science.gov (United States)

    2010-12-23

    ... are used in nuclear weapons as nuclear reactor moderators or reflectors and as nuclear reactor fuel...), grinding, and machine tooling of parts. Inhalation of beryllium particles may cause chronic beryllium...

  4. Analysis of a beryllium-copper diffusion joint after HHF test

    International Nuclear Information System (INIS)

    Guiniatouline, R.N.; Mazul, I.V.; Gorodetsky, A.E.; Zalavutdinov, R.Kh.; Rybakov, S.Yu.; Savenko, V.I.

    1996-01-01

    The development of beryllium-copper joints which can withstand relevant ITER divertor conditions is one of the important tasks at the present time. One of the main problems associated with these joints is the intermetallic layers. The strength and life of these joints significantly depend on the width and contents of the intermetallic layers. The objective of this work is to study the diffusion joint of TGP-56 beryllium to OFHC copper after thermal response and thermocyclic tests with a beryllium-copper mock-up. The HHF test was performed on the e-beam facility (EBTS, SNLA). The following methods were used for analysis: roentgenographic analysis, X-ray spectrum analysis and fracture analysis. During the investigation the following studies were undertaken: the analysis of the diffusion boundary layer, which was obtained at the cross-section of one of the tiles, the analysis of the debonded surfaces of several beryllium tiles and corresponding copper parts and the analysis of the upper surface of one of the tiles after HHF tests. The joint roentgenographic and element analyses revealed the following phases in the diffusion zone: Cu 2 Be (∝170 μm), CuBe (∝30 μm), CuBe 2 (∝1 μm) and a solid solution of copper in beryllium. The phases Cu 2 Be, CuBe and the solid solution of copper in beryllium were detected by the quantitative microanalysis and the phases CuBe, CuBe 2 and CuBe, by the roentgenographic analysis. The fracture (origin) is located in the central part of the tiles. This crack was caused by residual stresses and thermal fatigue testing. This analysis provides important data on the joint quality and may be used for all types of joints used for ITER applications. (orig.)

  5. Experiments on tritium behavior in beryllium, (1)

    International Nuclear Information System (INIS)

    Kawamura, Hiroshi; Ishizuka, Etsuo; Matsumoto, Mikio; Inada, Seiji; Sezaki, Katsuji; Saito, Minoru; Kato, Mineo.

    1989-06-01

    In JMTR, it was observed that the tritium concentration of the primary coolant increases with the reactor operation at 50 MW. As one of the tritium generation sources, we paid attention to a neutron reflector made of beryllium because the tritium generation rate in the beryllium is bigger than other components in the reactor core. On the other hand, the irradiation test of blanket materials (i.e. tritium breeding materials and neutron multipling materials) are planned for development of the fusion reactor in JMTR and the beryllium will be also irradiated as a neutron multiplier with tritium breeding materials. Therefore, as the irradiated specimens, we used a hot-pressed beryllium disk fabricated by the same method as the neutron reflector or the neutron multiplier and conducted the irradiation tests in JMTR. The purpose of these tests are to clarify the tritium behavior in the hot-pressed beryllium. In this paper, from a viewpoint of the fabrication of capsules for neutron irradiation, the specifications of the irradiated specimens and capsules are summarized. Additionally, the results on the puncture test of the container of the irradiation specimens are described. (author)

  6. Studies of neutron methods for process control and criticality surveillance of fissile material processing facilities

    International Nuclear Information System (INIS)

    Zoltowski, T.

    1988-01-01

    The development of radiochemical processes for fissile material processing and spent fuel handling need new control procedures enabling an improvement of plant throughput. This is strictly related to the implementation of continuous criticality control policy and developing reliable methods for monitoring the reactivity of radiochemical plant operations in presence of the process perturbations. Neutron methods seem to be applicable for fissile material control in some technological facilities. The measurement of epithermal neutron source multiplication with heuristic evaluation of measured data enables surveillance of anomalous reactivity enhancement leading to unsafe states. 80 refs., 47 figs., 33 tabs. (author)

  7. Defense Waste Processing Facility Canister Closure Weld Current Validation Testing

    Energy Technology Data Exchange (ETDEWEB)

    Korinko, P. S. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Maxwell, D. N. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2018-01-29

    Two closure welds on filled Defense Waste Processing Facility (DWPF) canisters failed to be within the acceptance criteria in the DWPF operating procedure SW4-15.80-2.3 (1). In one case, the weld heat setting was inadvertently provided to the canister at the value used for test welds (i.e., 72%) and this oversight produced a weld at a current of nominally 210 kA compared to the operating procedure range (i.e., 82%) of 240 kA to 263 kA. The second weld appeared to experience an instrumentation and data acquisition upset. The current for this weld was reported as 191 kA. Review of the data from the Data Acquisition System (DAS) indicated that three of the four current legs were reading the expected values, approximately 62 kA each, and the fourth leg read zero current. Since there is no feasible way by further examination of the process data to ascertain if this weld was actually welded at either the target current or the lower current, a test plan was executed to provide assurance that these Nonconforming Welds (NCWs) meet the requirements for strength and leak tightness. Acceptance of the welds is based on evaluation of Test Nozzle Welds (TNW) made specifically for comparison. The TNW were nondestructively and destructively evaluated for plug height, heat tint, ultrasonic testing (UT) for bond length and ultrasonic volumetric examination for weld defects, burst pressure, fractography, and metallography. The testing was conducted in agreement with a Task Technical and Quality Assurance Plan (TTQAP) (2) and applicable procedures.

  8. Occupational and non-occupational allergic contact dermatitis from beryllium

    Energy Technology Data Exchange (ETDEWEB)

    Vilaplana, J; Romaguera, C; Grimalt, F [Allergy Department, Dermatological Service Hospital Clinico, Barcelona (Spain)

    1992-01-01

    There are various references to sensitization to beryllium in the literature. Since introducing a patch testing series for patients with suspected sensitization to metals, we have found 3 cases of sensitization to beryllium. Of these 3 cases, we regard the first 2 as having relevant sensitization. Beryllium chloride (1% pet.) was positive in 3 patients and negative in 150 controls. (au).

  9. Occupational and non-occupational allergic contact dermatitis from beryllium

    International Nuclear Information System (INIS)

    Vilaplana, J.; Romaguera, C.; Grimalt, F.

    1992-01-01

    There are various references to sensitization to beryllium in the literature. Since introducing a patch testing series for patients with suspected sensitization to metals, we have found 3 cases of sensitization to beryllium. Of these 3 cases, we regard the first 2 as having relevant sensitization. Beryllium chloride (1% pet.) was positive in 3 patients and negative in 150 controls. (au)

  10. Beryllium phonon spectrum from cold neutron measurements

    International Nuclear Information System (INIS)

    Bulat, I.A.

    1979-01-01

    The inelastic coherent scattering of neutrons with the initial energy E 0 =4.65 MeV on the spectrometer according to the time of flight is studied in polycrystalline beryllium. The measurements are made for the scattering angles THETA=15, 30, 45, 60, 75 and 90 deg at 293 K. The phonon spectrum of beryllium, i-e. g(w) is reestablished from the experimental data. The data obtained are compared with the data of model calculations. It is pointed out that the phonon spectrum of beryllium has a bit excessive state density in the energy range from 10 to 30 MeV. It is caused by the insufficient statistical accuracy of the experiment at low energy transfer

  11. Spectrographic determination of impurities in beryllium oxide

    International Nuclear Information System (INIS)

    Paula Reino, L.C. de; Lordello, A.R.; Pereira, A.S.A.

    1986-03-01

    A method for the spectrographic determination of Al, B, Cd, Co, Cu, Cr, Fe, Mg, NaNi, Si and Zn in nuclear grade beryllium oxide has been developed. The determination of Co, Al, Na and Zn is besed upon a carrier distillation technique. Better results were obtained with 2% Ga 2 O 3 as carrier in beryllium oxide. For the elements B, Cd, Cu, Fe, Cr, Mg, Ni and Si the sample is loaded in a Scribner-Mullin shallow cup electrode, covered with graphite powder and excited in DC arc. The relative standard deviation values for different elements are in the range of 10 to 20%. The method fulfills requirements of precision and sensitivity for specification analysis of nuclear grade beryllium oxide.(Author) [pt

  12. Inherent structure features of beryllium and their influence on the performance polycrystalline metal under different conditions

    Energy Technology Data Exchange (ETDEWEB)

    Khomutov, A.M.; Mikhailov, V.S.; Pronin, V.N.; Pakhomov, Ya.D. [State Scientific Center of Russian Federation `A.A. Bochvar All-Russia Research Inst. of Inorganic Materials (VNIINM)`, Moscow (Russian Federation)

    1998-01-01

    The anisotropy of physical properties of beryllium single crystals resulting from covalent bonds in crystal lattice leads to significant residual thermal microstresses (RTM) in the polycrystalline metal. It is demonstrated experimentally that there is a simple linear dependence between the magnitude of RTM and the ultimate tensile strength. The factors controlling RTM are analysed and in the framework of powder metallurgy process the technological methods of producing beryllium with the needed properties are recommended. Primarily it is necessary to control the quantity and extent of dispersity of intergranular oxide inclusions and mean grain size in combination with the high degree of macro- and microhomogenity of the structure. The requirements to beryllium microstructure for different operating conditions including neutron fluxes and transient temperature fields are formulated. In the framework of the concept under development one can explain formerly not fully understandable effects, which are characteristic of polycrystalline beryllium such as unexpected Petch-Stro curve, the role of twinning etc., and predict new ones. In particular, it can be possible to expect the growth of ductility of high strength beryllium grades as neutron irradiated. (author)

  13. Solid state bonding of beryllium-copper for an ITER first wall application

    Energy Technology Data Exchange (ETDEWEB)

    Odegard, B.C. Jr.; Cadden, C.H. [Sandia National Labs., Livermore, CA (United States)

    1998-01-01

    Several different joint assemblies were evaluated in support of a manufacturing technology for diffusion bonding a beryllium armor tile to a copper alloy heat sink for fusion reactor applications. Because beryllium reacts with all but a few elements to form intermetallic compounds, this study considered several different surface treatments as a means of both inhibiting these reactions and promoting a good diffusion bond between the two substrates. All diffusion bonded assemblies used aluminum or an aluminum-beryllium composite (AlBeMet-150) as the interfacial material in contact with beryllium. In most cases, explosive bonding was utilized as a technique for joining the copper alloy heat sink to an aluminum or AlBeMet-150 substrate, which was subsequently diffusion bonded to an aluminum coated beryllium tile. In this approach, a 250 {mu}m thick titanium foil was used as a diffusion barrier between the copper and aluminum to prevent the formation of Cu-Al intermetallic phases. In all cases, a hot isostatic pressing (HIP) furnace was used in conjunction with canned assemblies in order to minimize oxidation and apply sufficient pressure on the assembly for excellent metal-to-metal contact and subsequent bonding. Several different processing schedules were evaluated during the course of this study; bonded assemblies were produced that failed outside the bond area indicating a 100% joint efficiency. (author)

  14. Solid state bonding of beryllium-copper for an ITER first wall application

    International Nuclear Information System (INIS)

    Odegard, B.C. Jr.; Cadden, C.H.

    1998-02-01

    Several different joint assemblies were evaluated in support of a manufacturing technology for diffusion bonding a beryllium armor tile to a copper alloy heat sink for fusion reactor applications. Because beryllium reacts with all but a few elements to form intermetallic compounds, this study considered several different surface treatments as a means of both inhibiting these reactions and promoting a good diffusion bond between the two substrates. A diffusion bonded assemblies used aluminum or an aluminum-beryllium composite (AlBeMet-150) as the interfacial material in contact with beryllium. In most cases, explosive bonding was utilized as a technique for joining the copper alloy heat sink to an aluminum or AlBeMet-150 substrate, which was subsequently diffusion bonded to an aluminum coated beryllium tile. In this approach, a 250 microm thick titanium foil was used as a diffusion barrier between the copper and aluminum to prevent the formation of Cu-Al intermetallic phases. In all cases, a hot isostatic pressing (HIP) furnace was used in conjunction with canned assemblies in order to minimize oxidation and apply sufficient pressure on the assembly for excellent metal-to-metal contact and subsequent bonding. Several different processing schedules were evaluated during the course of this study; bonded assemblies were produced that failed outside the bond area indicating a 100% joint efficiency

  15. Leachability characteristics of beryllium in redmud waste and its stabilization in cement

    International Nuclear Information System (INIS)

    Saradhi, I.V.; Mahadevan, T.N.; Krishnamoorthy, T.M.

    1999-01-01

    More than 70% of the beryl ore processed by the Beryllium Metal Plant at the BARC Vashi Complex ends up as redmud waste. The presence of significant quantities (0.4 to 0.8%) of beryllium in the redmud qualifies it as hazardous requiring safe handling, storage and disposal. The waste also contains 0.09% of water soluble fluoride. The various standard protocol of procedures were employed to estimate the leachability of beryllium from redmud for both short term and long term periods. Nearly 50% of beryllium present in redmud is leachable in water. We have tried the stabilization of redmud using portland cement. The proportion of redmud to cement was in the ratio of 1:1, 1:2 and 1:4. The blocks were cast, cured and used in the leachability experiments using standard protocols as above. The results of the TCLP test gave the levels of beryllium well below the standard limits in the TCLP extract of cement stabilized waste indicating the suitability of stabilization of redmud with cement whereas that of raw waste (redmud) are much higher than the prescribed limits. The total leach percent of beryllium in 1:2 block is 0.05% over period of 164 days whereas 1:1 and 1:4 gave a leach percent of 0.26 and 0.15% respectively. The DLT results indicate, diffusion controlled release of beryllium from the cement stabilized redmud blocks. The effective diffusion coefficient of beryllium obtained from the modelling study is 10 orders of magnitude less than the molecular diffusion coefficient of beryllium indicating the effectiveness of cement stabilization. From the detailed experiments performed, it is felt that 1:2 proportion of redmud and cement will be the best suited option for stabilization of redmud waste. The 1:1 proportion of redmud to cement mixture which could not be cast into compact cement blocks also exhibited very low leachability characteristics similar to 1:2 and 1:4 and can be be favourably considered for stabilization in case of space constraints at storage sites. The

  16. Evaluation of physical facilities and processing operations of major ...

    African Journals Online (AJOL)

    ... of these abattoirs were evaluated based on their presence and functional status. ... of safe and wholesome meat and meat products for human consumption. Keywords: Abattoir, Butcher, Meat, Physical facilities, Public health, Standard ...

  17. Evaluation of physical facilities and processing operations of major ...

    African Journals Online (AJOL)

    ADEYEYE

    abattoirs were as a result of failure to enforce the use of standard facilities in carrying out abattoir operations and general maintenance ... incinerator, chemical treatment and disposal. Sub- .... Veterinary laboratory .... sustainable food security.

  18. Sustainable Acquisition Process Improvement for Naval Facilities Engineering Command

    National Research Council Canada - National Science Library

    Sanders, Erin

    2003-01-01

    .... To meet the new requirements, laws must be implemented through effective policy. For over 6 years, the Navy has been acquiring sustainably designed facilities and has recently set sustainable development policy guidelines...

  19. Risk assessment of mitigated domino scenarios in process facilities

    International Nuclear Information System (INIS)

    Landucci, Gabriele; Necci, Amos; Antonioni, Giacomo; Argenti, Francesca; Cozzani, Valerio

    2017-01-01

    The propagation of accidents among process units may lead to severe cascading events or domino effects with catastrophic consequences. Prevention, mitigation and management of domino scenarios is of utmost importance and may be achieved in industrial facilities through the adoption of multiple safety layers. The present study was aimed at developing an innovative methodology to address the quantitative risk assessment (QRA) of domino scenarios accounting for the presence and role of safety barriers. Based on the expected performance of safety barriers, a dedicated event tree analysis allowed the identification and the assessment of the frequencies of the different end-point events deriving from unmitigated and partially mitigated domino chains. Specific criteria were introduced in consequence analysis to consider the mitigation effects of end-point scenarios deriving from safety barriers. Individual and societal risk indexes were calculated accounting for safety barriers and the mitigated scenarios that may result from their actions. The application of the methodology to case-studies of industrial interest proved the importance of introducing a specific systematic and quantitative analysis of safety barrier performance when addressing escalation leading to domino effect. - Highlights: • A methodology was developed to account for safety barrier performance in escalation prevention. • The methodology allows quantitative assessment accounting for safety barrier performance. • A detailed analysis of transient mitigated scenarios is allowed by the developed procedure. • The procedure allows accounting for safety barrier performance in QRA of domino scenarios. • An important reduction in the risk due to domino scenarios is evidenced when considering safety barriers.

  20. Energy dependences of absorption in beryllium windows and argon gas

    International Nuclear Information System (INIS)

    Chantler, C.T.; Staudenmann, J-P.

    1994-01-01

    In part of an ongoing work on x-ray form factors, new absorption coefficients are being evaluated for all elements, across the energy range from below 100 eV to above 100 keV. These new coefficients are applied herein to typical problems in synchrotron radiation stations, namely the use of beryllium windows and argon gas detectors. Results are compared with those of other authors. The electron-ion pair production process in ionization chambers is discussed, and the effects of 3d-element impurities are indicated. 15 refs., 6 figs

  1. A role for cell adhesion in beryllium-mediated lung disease

    Energy Technology Data Exchange (ETDEWEB)

    Hong-geller, Elizabeth [Los Alamos National Laboratory

    2008-01-01

    Chronic beryllium disease (CBD) is a debilitating lung disorder in which exposure to the lightweight metal beryllium (Be) causes the accumulation of beryllium-specific CD4+ T cells in the lung and formation of noncaseating pulmonary granulomas. Treatment for CBD patients who exhibit progressive pulmonary decline is limited to systemic corticosteroids, which suppress the severe host inflammatory response. Studies in the past several years have begun to highlight cell-cell adhesion interactions in the development of Be hypersensitivity and CBD. In particular, the high binding affinity between intercellular adhesion molecule 1 (I-CAM1) on lung epithelial cells and the {beta}{sub 2} integrin LFA-1 on migrating lymphocytes and macrophages regulates the concerted rolling of immune cells to sites of inflammation in the lung. In this review, we discuss the evidence that implicates cell adhesion processes in onset of Be disease and the potential of cell adhesion as an intervention point for development of novel therapies.

  2. Experimental and thermodynamic studies of beryllium replacement materials for CANDU brazed joints

    Energy Technology Data Exchange (ETDEWEB)

    Potter, K.N.; Ferrier, G.A.; Corcoran, E.C., E-mail: Kieran.Potter@rmc.ca [Royal Military College of Canada, Kingston, ON (Canada)

    2015-07-01

    Currently, appendages are joined to CANDU fuel elements via a brazing process, which uses beryllium as the filler material. A potential reduction in the occupational limit on airborne beryllium particulates has motivated research into alternative brazing materials. To this end, the Canadian nuclear industry has funded an initiative to identify and evaluate the suitability of several candidate materials. This work describes contributions toward the assessment of alternative brazing materials from the Royal Military College of Canada. Thermodynamic modelling was performed to predict the aqueous behaviour of each candidate material in CANDU coolant conditions characteristic of reactor shutdown, and experiments are underway to support modelling predictions. These results will assist in selecting a suitable replacement material for beryllium. (author)

  3. Experimental and thermodynamic assessment of beryllium-replacement materials for CANDU brazed joints

    Energy Technology Data Exchange (ETDEWEB)

    Potter, K.N.; Ferrier, G.A.; Corcoran, E.C., E-mail: Kieran.Potter@rmc.ca [Royal Military College of Canada, Kingston ON, (Canada); Dimayuga, F.C. [Canadian Nuclear Laboratories, Chalk River, ON (Canada)

    2015-07-01

    Currently, appendages are joined to CANDU fuel elements via a brazing process, with beryllium as the filler material. A potential reduction in the occupational limit on airborne beryllium particulates has motivated research into alternative brazing materials. To this end, the Canadian nuclear industry has funded an initiative to identify and evaluate the suitability of several candidate brazing materials. This work describes contributions toward the assessment of alternative brazing materials from the Royal Military College of Canada (RMCC). An impact testing method was developed to evaluate the mechanical strength of candidate braze joints.Thermodynamic modelling was performed to predict the aqueous behaviour of each candidate material in CANDU coolant conditions characteristic of reactor shutdown, and corrosion experiments are underway to support modelling predictions.The results of these activities will assist in selecting a suitable replacement material for beryllium. (author)

  4. Influence of purity level on the mechanical properties of hot isostatically pressed beryllium

    International Nuclear Information System (INIS)

    Odegard, B.C. Jr.

    1979-09-01

    The procurement of a quantity of ultra-pure beryllium powder combined with special handling from powder to billet form resulted in the fabrication of high purity beryllium. The mechanical properties of these billets were contrasted to those of commercial grade billets to determine the influence of impurities and powder processing. The tensile test results show that the strength values are primarily dependent on the grain size in a behavior predictable by the Hall-Petch relationship. Only a fraction of the strength differential can be attributed to metallic impurities in solution. The grain size is controlled by the powder size distribution. The ductility is dominated by both grain size and oxide content. The fine grained, low oxide billets exhibited the highest ductilities. There is evidence to suggest that oxide distribution has a large influence on the ductility. The fracture toughness was highest for the high purity beryllium billets

  5. Preliminary results for explosion bonding of beryllium to copper

    International Nuclear Information System (INIS)

    Butler, D.J.; Dombrowski, D.E.

    1995-01-01

    This program was undertaken to determine if explosive bonding is a viable technique for joining beryllium to copper substrates. The effort was a cursory attempt at trying to solve some of the problems associated with explosive bonding beryllium and should not be considered a comprehensive research effort. There are two issues that this program addressed. Can beryllium be explosive bonded to copper substrates and can the bonding take place without shattering the beryllium? Thirteen different explosive bonding iterations were completed using various thicknesses of beryllium that were manufactured with three different techniques

  6. Overview of NORM and activities by a NORM licensed permanent decontamination and waste processing facility

    Energy Technology Data Exchange (ETDEWEB)

    Mirro, G.A. [Growth Resources, Inc., Lafayette, LA (United States)

    1997-02-01

    This paper presents an overview of issues related to handling NORM materials, and provides a description of a facility designed for the processing of NORM contaminated equipment. With regard to handling NORM materials the author discusses sources of NORM, problems, regulations and disposal options, potential hazards, safety equipment, and issues related to personnel protection. For the facility, the author discusses: description of the permanent facility; the operations of the facility; the license it has for handling specific radioactive material; operating and safety procedures; decontamination facilities on site; NORM waste processing capabilities; and offsite NORM services which are available.

  7. Methodology for Determining Increases in Radionuclide Inventories for the Effluent Treatment Facility Process

    International Nuclear Information System (INIS)

    Blanchard, A.

    1998-01-01

    A study is currently underway to determine if the Effluent Treatment Facility can be downgraded from a Hazard Category 3 facility to a Radiological Facility per DOE STD-1027-92. This technical report provides a methodology to determine and monitor increases in the radionuclide inventories of the ETF process columns. It also provides guidelines to ensure that other potential increases to the ETF radionuclide inventory are evaluated as required to ensure that the ETF remains a Radiological Facility

  8. Characterization of a Fe inclusion in beryllium-matrix using auger electron spectroscopy

    International Nuclear Information System (INIS)

    Arkusk, R.; Moreno, D.; Simca, F.; Yeheskel, O.; Utzmoni, U.

    1991-04-01

    The auger electron spectroscopy techniques was employed to investigate the nature of an inclusion that had been revealed by radiography in a beryllium body produced by the hot isostatic press technique. The investigation's are that the inclusion is composed of several different iron-beryllium intermetallic compounds (BeFe 3 , BeFe 5 , Be 7 Fe). The conclusion drawn is that iron metal impurity was imbedded in the Be powder and that interdiffusion under the process's conditions gave rise to the enlarged inclusion. (author)

  9. Ageing Management of Beryllium and Graphite Blocks in Research Reactor MARIA

    Energy Technology Data Exchange (ETDEWEB)

    Golab, A. [National Centre for Nuclear Research, Warsaw (Poland)

    2013-07-01

    In the paper the phenomenon of beryllium moderator poisoning by thermal neutron absorption and the method and results of this phenomenon control is presented. Also the phenomenon of graphite blocks damage due to fast neutrons accumulation and the methods and results of this process supervising is described. These methods refer especially to: visual inspection of their state and radiography of graphite blocks. Special attention is paid to permanent estimate of fast neutron fluency accumulated in blocks and methods of their shuffling in the reactor core. The shuffling makes possible to increase the lifetime of beryllium and graphite blocks and decrease the cost of reactor operation.

  10. Results of investigations of non-specific humoral-type response in workers exposed to beryllium

    International Nuclear Information System (INIS)

    Cianciara, M.; Loch, T.

    1989-12-01

    In the group of 30 workers exposed to beryllium a statistically significant increase in the level of immunoglobulins of class A was found in comparison with the control group (p<0.01). This may be an early indicator of a process taking place in the respiratory system. Results of measurement of immunoglobulins, as non-specific indicators, are of limited value in the evaluation of the influence of the occupational factor; at the same time they point to the usefulness of testing for the specific indicators of humoral reactivity in persons whose work involves contact with beryllium or its compounds. (author)

  11. Statistical process control support during Defense Waste Processing Facility chemical runs

    International Nuclear Information System (INIS)

    Brown, K.G.

    1994-01-01

    The Product Composition Control System (PCCS) has been developed to ensure that the wasteforms produced by the Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS) will satisfy the regulatory and processing criteria that will be imposed. The PCCS provides rigorous, statistically-defensible management of a noisy, multivariate system subject to multiple constraints. The system has been successfully tested and has been used to control the production of the first two melter feed batches during DWPF Chemical Runs. These operations will demonstrate the viability of the DWPF process. This paper provides a brief discussion of the technical foundation for the statistical process control algorithms incorporated into PCCS, and describes the results obtained and lessons learned from DWPF Cold Chemical Run operations. The DWPF will immobilize approximately 130 million liters of high-level nuclear waste currently stored at the Site in 51 carbon steel tanks. Waste handling operations separate this waste into highly radioactive sludge and precipitate streams and less radioactive water soluble salts. (In a separate facility, soluble salts are disposed of as low-level waste in a mixture of cement slag, and flyash.) In DWPF, the precipitate steam (Precipitate Hydrolysis Aqueous or PHA) is blended with the insoluble sludge and ground glass frit to produce melter feed slurry which is continuously fed to the DWPF melter. The melter produces a molten borosilicate glass which is poured into stainless steel canisters for cooling and, ultimately, shipment to and storage in a geologic repository

  12. Facilities for the production and processing of radioisotopes

    International Nuclear Information System (INIS)

    Fourie, P.J.

    1980-01-01

    Radioisotopes which are used in South Africa are produced in the nuclear reactor SAFARI 1 of the AEB and the CSIR cyclotron in Pretoria or are being imported from various overseas manufactures. The safe and efficient production and use of radioisotopes is possible when being handled by sufficiently trained personnel using special designed equipment and facilities. The Isotope Production Centre is situated next to the reactor and waste treatment buildings. New production facilities shielded with lead and equipped with remote handling equipment are being erected and will be commissioned early during 1980 [af

  13. Method of beryllium implantation in germanium substrate

    International Nuclear Information System (INIS)

    Kagawa, S.; Baba, Y.; Kaneda, T.; Shirai, T.

    1983-01-01

    A semiconductor device is disclosed, as well as a method for manufacturing it in which ions of beryllium are implanted into a germanium substrate to form a layer containing p-type impurity material. There after the substrate is heated at a temperature in the range of 400 0 C. to 700 0 C. to diffuse the beryllium ions into the substrate so that the concentration of beryllium at the surface of the impurity layer is in the order of 10 17 cm- 3 or more. In one embodiment, a p-type channel stopper is formed locally in a p-type germanium substrate and an n-type active layer is formed in a region surrounded by, and isolated from, the channel stopper region. In another embodiment, a relatively shallow p-type active layer is formed at one part of an n-type germanium substrate and p-type guard ring regions are formed surrounding, and partly overlapping said p-type active layer. In a further embodiment, a p-type island region is formed at one part of an n-type germanium substrate, and an n-type region is formed within said p-type region. In these embodiments, the p-type channel stopper region, p-type guard ring regions and the p-type island region are all formed by implanting ions of beryllium into the germanium substrate

  14. ESR investigations of gamma irradiated beryllium ceramics

    International Nuclear Information System (INIS)

    Ryabikin, Yu.A.; Polyakov, A.I.; Petukhov, Yu.V.; Bitenbaev, M.I.; Zashkvara, O.V.

    2000-01-01

    In this report the result of ESR- investigation of kinetics of radiation paramagnetic defects accumulated in beryllium ceramics under gamma irradiation are presented. The data on quantum yield and destruction rate constants of these defects under ionizing irradiation are obtained. (orig.)

  15. ESR investigations of gamma irradiated beryllium ceramics

    Energy Technology Data Exchange (ETDEWEB)

    Ryabikin, Yu A; Polyakov, A I; Petukhov, Yu V; Bitenbaev, M I; Zashkvara, O V [Physical-Technical Inst., Almaty (Kazakhstan)

    2000-04-01

    In this report the result of ESR- investigation of kinetics of radiation paramagnetic defects accumulated in beryllium ceramics under gamma irradiation are presented. The data on quantum yield and destruction rate constants of these defects under ionizing irradiation are obtained. (orig.)

  16. Beryllium, a material of great promise

    International Nuclear Information System (INIS)

    Le Fauconnier, J.P.; Nomine, A.M.

    1992-01-01

    Beryllium is one of the lightest metals. It also owns an outstanding combination of physical, mechanical and nuclear properties which gives it a favorable position, compared to more usual materials, in various fields of applications. Constant technological advancements in the elaboration and working up have induced a significant improvement of its ductility and a reduction of the production costs. (Author). 12 refs., 7 figs

  17. Thermophysical properties of solid and liquid beryllium

    International Nuclear Information System (INIS)

    Boivineau, M.; Arles, L.; Vermeulen, J.M.; Thevenin, Th.

    1993-01-01

    A submillisecond resistive heating technique under high pressure (0.12 GPa) has been used to measure selected thermophysical properties of both solid and liquid beryllium. Data have been obtained between room temperature and 2900 K. Results on enthalpy, volume expansion, electrical resistivity, and sound velocity measurements are presented

  18. Critical parameters controlling irradiation swelling in beryllium

    International Nuclear Information System (INIS)

    Dubinko, V.I.

    1995-01-01

    Radiation effects in beryllium can hardly be explained within a framework of the conventional theory based on the bias concept due to elastic interaction difference (EID) between vacancies and self-interstitial atoms (SIAs) since beryllium belongs to hexagonal close-packed metals where diffusion has been shown to be anisotropic. Diffusional anisotropy difference (DAD) between point defects changes the cavity bias for their absorption and leads to dependence of the dislocation bias on the distribution of dislocations over crystallographic directions. On the other hand, the elastic interaction between point defects and cavities gives rise to the size and gas pressure dependencies of the cavity bias, resulting in new critical quantities for bubble-void transition effects at low temperature irradiation. In the present paper, we develop the concept of the critical parameters controlling irradiation swelling with account of both DAD and EID, and take care of thermal effects as well since they are of major importance for beryllium which has an anomalously low self-diffusion activation energy. Experimental data on beryllium swelling are analyzed on the basis of the present theory. (orig.)

  19. Potential exposures and risks from beryllium-containing products.

    Science.gov (United States)

    Willis, Henry H; Florig, H Keith

    2002-10-01

    Beryllium is the strongest of the lightweight metals. Used primarily in military applications prior to the end of the Cold War, beryllium is finding new applications in many commercial products, including computers, telecommunication equipment, and consumer and automotive electronics. The use of beryllium in nondefense consumer applications is of concern because beryllium is toxic. Inhalation of beryllium dust or vapor causes a chronic lung disease in some individuals at concentrations as low as 0.01 microg/m3 in air. As beryllium enters wider commerce, it is prudent to ask what risks this might present to the general public and to workers downstream of the beryllium materials industry. We address this question by evaluating the potential for beryllium exposure from the manufacturing, use, recycle, and disposal of beryllium-containing products. Combining a market study with a qualitative exposure analysis, we determine which beryllium applications and life cycle phases have the largest exposure potential. Our analysis suggests that use and maintenance of the most common types of beryllium-containing products do not result in any obvious exposures of concern, and that maintenance activities result in greater exposures than product use. Product disposal has potential to present significant individual risks, but uncertainties concerning current and future routes of product disposal make it difficult to be definitive. Overall, additional exposure and dose-response data are needed to evaluate both the health significance of many exposure scenarios, and the adequacy of existing regulations to protect workers and the public. Although public exposures to beryllium and public awareness and concern regarding beryllium risks are currently low, beryllium risks have psychometric qualities that may lead to rapidly heightened public concern.

  20. Semi-solid metal forming of beryllium-reinforced aluminum alloys

    International Nuclear Information System (INIS)

    Haws, W.; Lane, L.; Marder, J.; Nicholas, N.

    1995-01-01

    A Powder Metallurgy (PM) based, Semi-Solid Metal (SSM) forming process has been developed to produce low cost near-net shapes of beryllium-reinforced aluminum alloys. Beryllium acts as a reinforcing additive to the aluminum, in which there is nearly no mutual solid solubility. The modulus of elasticity of the alloy dramatically increases, while the density and thermal expansion coefficient decrease with increasing beryllium content. The material is suitable for complex thermal management and vibration resistance applications, as well as for airborne components which are density and stiffness sensitive. The forming process involves heating a blank of the material to a temperature at which the aluminum is semi-solid and the beryllium is solid. The semi-solid blank is then injected without turbulence into a permanent mold. High quality, near net shape components can be produced which are functionally superior to those produced by other permanent mold processes. Dimensional accuracy is equivalent to or better than that obtained in high pressure die casting. Cost effectiveness is the primary advantage of this technique compared to other forming processes. The advantages and limitations of the process are described. Physical and mechanical property data are presented, as well as directions for future investigation

  1. 303-K Storage Facility closure plan

    International Nuclear Information System (INIS)

    1993-01-01

    Recyclable scrap uranium with zircaloy-2 and copper silicon alloy, uranium-titanium alloy, beryllium/zircaloy-2 alloy, and zircaloy-2 chips and fines were secured in concrete billets (7.5-gallon containers) in the 303-K Storage Facility, located in the 300 Area. The beryllium/zircaloy-2 alloy and zircaloy-2 chips and fines are designated as mixed waste with the characteristic of ignitability. The concretion process reduced the ignitability of the fines and chips for safe storage and shipment. This process has been discontinued and the 303-K Storage Facility is now undergoing closure as defined in the Resource Conservation and Recovery Act (RCRA) of 1976 and the Washington Administrative Code (WAC) Dangerous Waste Regulations, WAC 173-303-040. This closure plan presents a description of the 303-K Storage Facility, the history of materials and waste managed, and the procedures that will be followed to close the 303-K Storage Facility. The 303-K Storage Facility is located within the 300-FF-3 (source) and 300-FF-5 (groundwater) operable units, as designated in the Hanford Federal Facility Agreement and Consent Order (Tri-Party Agreement) (Ecology et al. 1992). Contamination in the operable units 300-FF-3 and 300-FF-5 is scheduled to be addressed through the Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) of 1980 remedial action process. Therefore, all soil remedial action at the 304 Facility will be conducted as part of the CERCLA remedial action of operable units 300-FF-3 and 300-FF-5

  2. Detection of beryllium treatment of natural sapphires by NRA

    Energy Technology Data Exchange (ETDEWEB)

    Gutierrez, P.C., E-mail: carolina.gutierrez@uam.e [Centro de Micro-Analisis de Materiales (CMAM), Universidad Autonoma de Madrid, Campus de Cantoblanco, 28049 Madrid (Spain); Ynsa, M.-D.; Climent-Font, A. [Centro de Micro-Analisis de Materiales (CMAM), Universidad Autonoma de Madrid, Campus de Cantoblanco, 28049 Madrid (Spain); Dpto. Fisica Aplicada C-12, Universidad Autonoma de Madrid, Campus de Cantoblanco, 28049 Madrid (Spain); Calligaro, T. [Centre de Recherche et de Restauration des musees de France C2RMF, CNRS-UMR171, 14 quai Francois Mitterrand, 75001 Paris (France)

    2010-06-15

    Since the 1990's, artificial treatment of natural sapphires (Al{sub 2}O{sub 3} crystals coloured by impurities) by diffusion of beryllium at high temperature has become a growing practice. This process permits to enhance the colour of these gemstones, and thus to increase their value. Detection of such a treatment - diffusion of tens of {mu}g/g of beryllium in Al{sub 2}O{sub 3} crystals - is usually achieved using high sensitivity techniques like laser-ablation inductively coupled plasma mass spectrometry (LA-ICP/MS) or laser-induced breakdown spectrometry (LIBS) which are unfortunately micro-destructive (leaving 50-100-{mu}m diameter craters on the gems). The simple and non-destructive alternative method proposed in this work is based on the nuclear reaction {sup 9}Be({alpha}, n{gamma}){sup 12}C with an external helium ion beam impinging on the gem directly placed in air. The 4439 keV prompt {gamma}-ray tagging Be atoms are detected with a high efficiency bismuth germanate scintillator. Beam dose is monitored using the 2235 keV prompt {gamma}-ray produced during irradiation by the aluminium of the sapphire matrix through the {sup 27}Al({alpha}, p{gamma}){sup 30}Si nuclear reaction. The method is tested on a series of Be-treated sapphires previously analyzed by LA-ICP/MS to determine the optimal conditions to obtain a peak to background appropriate to reach the required {mu}g/g sensitivity. Using a 2.8-MeV external He beam and a beam dose of 200 {mu}C, beryllium concentrations from 5 to 16 {mu}g/g have been measured in the samples, with a detection limit of 1 {mu}g/g.

  3. 10 CFR 95.17 - Processing facility clearance.

    Science.gov (United States)

    2010-01-01

    ... that the facility is not under foreign ownership, control, or influence to such a degree that a determination could not be made. An NRC finding of foreign ownership, control, or influence is based on factors concerning the foreign intelligence threat, risk of unauthorized technology transfer, type and sensitivity of...

  4. Development of cables for nuclear fuel processing facilities

    International Nuclear Information System (INIS)

    Kamimura, Seiji; Seki, Ikuo; Yamamoto, Yasuaki; Matsuyama, Shigeki; Endo, Shigeru; Yagi, Toshiaki; Kawakami, Waichiro.

    1988-01-01

    Accompanying the development of nuclear power stations, the expansion and repletion of the facilities for nuclear fuel cycle such as fuel reprocessing facilities and waste treatment facilities are requested. In these facilities, there is the environment which is exposed to very high level radiation, and in this case, the cables withstanding 10 MGy radiation dose are required. As the cables meeting such requirement, the new cables having excellent flexibility and radiation resistance were developed. In this paper, the points of material development and the characteristics of cables are reported. Considering the radiation resistance and others, ethylene propylene rubber was selected as the base polymer of the insulator, and polyethylene chlorosulfonate was selected as the sheath material. In order to give excellent radiation resistance, as the anti-rad, energy transfer type aromatic oil that absorbs and dissipates radiation energy and radical trap type anti-oxidant of amine group that catches and stabilizes the radicals generated in the polymer were added. The bromine group burning retarding agent having excellent radiation resistance was applied. In this way, the cables withstanding high radiation dose up to 10 MGy were able to be developed. (K.I.)

  5. Dosimetry. Standard practice for dosimetry in gamma irradiation facilities for food and non-food processing

    International Nuclear Information System (INIS)

    2008-01-01

    This Ghana Standard outlines the installation qualification program for an irradiator and the dosimetry procedures to be followed during operational qualification, performance qualification and routine processing in facilities that process food and non-food with gamma rays. This is to ensure that the product has been treated with predetermined range of absorbed dose. It is not intended for use in X-ray and electron beam facilities and therefore dosimetry systems in such facilities are not covered

  6. Experimental results of beryllium exposed to intense high energy proton beam pulses

    Energy Technology Data Exchange (ETDEWEB)

    Ammigan, K. [Fermilab; Hartsell, B. [Fermilab; Hurh, P. [Fermilab; Zwaska, R. [Fermilab; Butcher, M. [CERN; Guinchard, M. [CERN; Calviani, M. [CERN; Losito, R. [CERN; Roberts, S. [Culham Lab; Kuksenko, V. [Oxford U.; Atherton, A. [Rutherford; Caretta, O. [Rutherford; Davenne, T. [Rutherford; Densham, C. [Rutherford; Fitton, M. [Rutherford; Loveridge, J. [Rutherford; O' Dell, J. [Rutherford

    2017-02-10

    Beryllium is extensively used in various accelerator beam lines and target facilities as a material for beam windows, and to a lesser extent, as secondary particle production targets. With increasing beam intensities of future accelerator facilities, it is critical to understand the response of beryllium under extreme conditions to reliably operate these components as well as avoid compromising particle production efficiency by limiting beam parameters. As a result, an exploratory experiment at CERN’s HiRadMat facility was carried out to take advantage of the test facility’s tunable high intensity proton beam to probe and investigate the damage mechanisms of several beryllium grades. The test matrix consisted of multiple arrays of thin discs of varying thicknesses as well as cylinders, each exposed to increasing beam intensities. This paper outlines the experimental measurements, as well as findings from Post-Irradiation-Examination (PIE) work where different imaging techniques were used to analyze and compare surface evolution and microstructural response of the test matrix specimens.

  7. Deep layer-resolved core-level shifts in the beryllium surface

    DEFF Research Database (Denmark)

    Aldén, Magnus; Skriver, Hans Lomholt; Johansson, Börje

    1993-01-01

    Core-level energy shifts for the beryllium surface region are calculated by means of a Green’s function technique within the tight-binding linear muffin-tin orbitals method. Both initial- and final-state effects in the core-ionization process are fully accounted for. Anomalously large energy shifts...

  8. Process for decontamination of surfaces in an facility of natural uranium hexafluoride production (UF6)

    International Nuclear Information System (INIS)

    Almeida, Claudio C. de; Silva, Teresinha M.; Rodrigues, Demerval L.; Carneiro, Janete C.G.G.

    2017-01-01

    The experience acquired in the actions taken during the decontamination process of an IPEN-CNEN / SP Nuclear and Energy Research Institute facility, for the purpose of making the site unrestricted, is reported. The steps of this operation involved: planning, training of facility operators, workplace analysis and radiometric measurements. The facility had several types of equipment from the natural uranium hexafluoride (UF 6 ) production tower and other facility materials. Rules for the transportation of radioactive materials were established, both inside and outside the facility and release of materials and installation

  9. The defense waste processing facility: the final processing step for defense high-level waste disposal

    International Nuclear Information System (INIS)

    Cowan, S.P.; Sprecher, W.M.; Walton, R.D.

    1983-01-01

    The policy of the U.S. Department of Energy is to pursue an aggressive and credible waste management program that advocates final disposal of government generated (defense) high-level nuclear wastes in a manner consistent with environmental, health, and safety responsibilities and requirements. The Defense Waste Processing Facility (DWPF) is an essential component of the Department's program. It is the first project undertaken in the United States to immobilize government generated high-level nuclear wastes for geologic disposal. The DWPF will be built at the Department's Savannah River Plant near Aiken, South Carolina. When construction is complete in 1989, the DWPF will begin processing the high-level waste at the Savannah River Plant into a borosilicate glass form, a highly insoluble and non-dispersable product, in easily handled canisters. The immobilized waste will be stored on site followed by transportation to and disposal in a Federal repository. The focus of this paper is on the DWPF. The paper discusses issues which justify the project, summarizes its technical attributes, analyzes relevant environmental and insitutional factors, describes the management approach followed in transforming technical and other concepts into concrete and steel, and concludes with observations about the future role of the facility

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-08-01

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

  11. Mixed U/Pu oxide fuel fabrication facility co-processed feed, pelletized fuel

    International Nuclear Information System (INIS)

    1978-09-01

    Two conceptual MOX fuel fabrication facilities are discussed in this study. The first facility in the main body of the report is for the fabrication of LWR uranium dioxide - plutonium dioxide (MOX) fuel using co-processed feed. The second facility in the addendum is for the fabrication of co-processed MOX fuel spiked with 60 Co. Both facilities produce pellet fuel. The spiked facility uses the same basic fabrication process as the conventional MOX plant but the fuel feed incorporates a high energy gamma emitter as a safeguard measure against diversion; additional shielding is added to protect personnel from radiation exposure, all operations are automated and remote, and normal maintenance is performed remotely. The report describes the fuel fabrication process and plant layout including scrap and waste processing; and maintenance, ventilation and safety measures

  12. Hardware Development Process for Human Research Facility Applications

    Science.gov (United States)

    Bauer, Liz

    2000-01-01

    The simple goal of the Human Research Facility (HRF) is to conduct human research experiments on the International Space Station (ISS) astronauts during long-duration missions. This is accomplished by providing integration and operation of the necessary hardware and software capabilities. A typical hardware development flow consists of five stages: functional inputs and requirements definition, market research, design life cycle through hardware delivery, crew training, and mission support. The purpose of this presentation is to guide the audience through the early hardware development process: requirement definition through selecting a development path. Specific HRF equipment is used to illustrate the hardware development paths. The source of hardware requirements is the science community and HRF program. The HRF Science Working Group, consisting of SCientists from various medical disciplines, defined a basic set of equipment with functional requirements. This established the performance requirements of the hardware. HRF program requirements focus on making the hardware safe and operational in a space environment. This includes structural, thermal, human factors, and material requirements. Science and HRF program requirements are defined in a hardware requirements document which includes verification methods. Once the hardware is fabricated, requirements are verified by inspection, test, analysis, or demonstration. All data is compiled and reviewed to certify the hardware for flight. Obviously, the basis for all hardware development activities is requirement definition. Full and complete requirement definition is ideal prior to initiating the hardware development. However, this is generally not the case, but the hardware team typically has functional inputs as a guide. The first step is for engineers to conduct market research based on the functional inputs provided by scientists. CommerCially available products are evaluated against the science requirements as

  13. Effect of copper on crack propagation in beryllium single crystals

    International Nuclear Information System (INIS)

    Aldinger, F.; Wilhelm, M.

    The effect of copper additives on the fracture energy and the development of cracks parallel to the basal plane was studied in zone-refined single crystalline beryllium. At 77 0 K the cleavage planes are very smooth, so the crack propagation energy, which is independent of copper content (less than 2 at. percent Cu) in the range of measurement accuracy, is only a little higher than the surface energy of the basal plane. At room temperature, due to intense plastic processes taking place in front of the crack tip, the fracture energy is an order of magnitude higher than at low temperatures. The effect of copper on the plastic processes can be divided into two regions. In region I (less than 1.2 at. percent Cu), in which the crack propagation energy increases sharply with increasing copper content, crack propagation is controlled by prism slips. The decrease in crack propagation energy in region II (greater than 1.2 at. percent Cu) can be attributed to a reduction of beryllium twinning energy with increasing copper content. (auth)

  14. Processing of tetraphenylborate precipitates in the Savannah River Site Defense Waste Processing Facility

    International Nuclear Information System (INIS)

    Eibling, R.E.

    1990-01-01

    The Savannah River Site has generated 77 million gallons of high level radioactive waste since the early 1950's. By 1987, evaporation had reduced the concentration of the waste inventory to 35 million gallons. Currently, the wastes reside in large underground tanks as a soluble fraction stored, crystallized salts, and an insoluble fraction, sludge, which consists of hydrated transition metal oxides. The bulk of the radionuclides, 67 percent, are in the sludge while the crystallized salts and supernate are composed of the nitrates, nitrites, sulfates and hydroxides of sodium, potassium, and cesium. The principal radionuclide in the soluble waste is 137 Cs with traces of 90 Sr. The transformation of the high level wastes into a borosilicate glass suitable for permanent disposal is the goal of the Defense Waste Processing Facility (DWPF). To minimize the volume of glass produced, the soluble fraction of the waste is treated with sodium tetraphenylborate and sodium titanate in the waste tanks to precipitate the radioactive cesium ion and absorb the radioactive strontium ion. The precipitate is washed in the waste tanks and is then pumped to the DWPF. The precipitate, as received, is incompatible with the vitrification process because of the high aromatic carbon content and requires further chemical treatment. Within the DWPF, the precipitate is processed in the Salt Processing Cell to remove the aromatic carbon as benzene. The precipitate hydrolysis process hydrolyzes the tetraphenylborate anion to produce borate anion and benzene. The benzene is removed by distillation, decontaminated and transferred out of the DWPF for disposal

  15. The Defense Waste Processing Facility: an innovative process for high-level waste immobilization

    International Nuclear Information System (INIS)

    Cowan, S.P.

    1985-01-01

    The Defense Waste Processing Facility (DWPF), under construction at the Department of Energy's Savannah River Plant (SRP), will process defense high-level radioactive waste so that it can be disposed of safely. The DWPF will immobilize the high activity fraction of the waste in borosilicate glass cast in stainless steel canisters which can be handled, stored, transported and disposed of in a geologic repository. The low-activity fraction of the waste, which represents about 90% of the high-level waste HLW volume, will be decontaminated and disposed of on the SRP site. After decontamination the canister will be welded shut by an upset resistance welding technique. In this process a slightly oversized plug is pressed into the canister opening. At the same time a large current is passed through the canister and plug. The higher resistance of the canister/plug interface causes the heat which welds the plug in place. This process provides a high quality, reliable weld by a process easily operated remotely

  16. Guidelines for operator competence - Optimising facility management processes; Leitfaden Betreiberkompetenz. Schritt fuer Schritt Facility Management Prozesse optimieren

    Energy Technology Data Exchange (ETDEWEB)

    Moser, R

    2005-06-15

    This brochure issued by IFMA (International Facility Management Association) Switzerland and the Swiss Federal Office of Energy (SFOE) presents interactive guidelines for energy management in the area of facility management. These guidelines are based on the results of a project carried out by the International Energy Agency's Annex 40 'Operator competence'. The guidelines provide a step-by-step guide from initial analysis through to successful project completion and answer many questions that may crop up during the process. The focus is placed on energy aspects. Tools and 14 sample process descriptions are provided along with practical examples. Theoretical aspects are also presented and discussed, including models for operator roles and the processes involved. Also, change, risk and knowledge management are examined. Notes and information on possibilities for further education are presented.

  17. Guidelines for operator competence - Optimising facility management processes; Leitfaden Betreiberkompetenz. Schritt fuer Schritt Facility Management Prozesse optimieren

    Energy Technology Data Exchange (ETDEWEB)

    Moser, R.

    2005-06-15

    This brochure issued by IFMA (International Facility Management Association) Switzerland and the Swiss Federal Office of Energy (SFOE) presents interactive guidelines for energy management in the area of facility management. These guidelines are based on the results of a project carried out by the International Energy Agency's Annex 40 'Operator competence'. The guidelines provide a step-by-step guide from initial analysis through to successful project completion and answer many questions that may crop up during the process. The focus is placed on energy aspects. Tools and 14 sample process descriptions are provided along with practical examples. Theoretical aspects are also presented and discussed, including models for operator roles and the processes involved. Also, change, risk and knowledge management are examined. Notes and information on possibilities for further education are presented.

  18. Design and construction of the defense waste processing facility project at the Savannah River Plant

    International Nuclear Information System (INIS)

    Baxter, R.G.

    1986-01-01

    The Du Pont Company is building for the Department of Energy a facility to vitrify high-level radioactive waste at the Savannah River Plant (SRP) near Aiken, South Carolina. The Defense Waste Processing Facility (DWPF) will solidify existing and future radioactive wastes by immobilizing the waste in Processing Facility (DWPF) will solidify existing and future radioactives wastes by immobilizing the waste in borosilicate glass contained in stainless steel canisters. The canisters will be sealed, decontaminated and stored, prior to emplacement in a federal repository. At the present time, engineering and design is 90% complete, construction is 25% complete, and radioactive processing in the $870 million facility is expected to begin by late 1989. This paper describes the SRP waste characteristics, the DWPF processing, building and equipment features, and construction progress of the facility

  19. Economic comparison of centralizing or decentralizing processing facilities for defense transuranic waste

    International Nuclear Information System (INIS)

    Brown, C.M.

    1980-07-01

    This study is part of a set of analyses under direction of the Transuranic Waste Management Program designed to provide comprehensive, systematic methodology and support necessary to better understand options for national long-term management of transuranic (TRU) waste. The report summarizes activities to evaluate the economics of possible alternatives in locating facilities to process DOE-managed transuranic waste. The options considered are: (1) Facilities located at all major DOE TRU waste generating sites. (2) Two or three regional facilities. (3) Central processing facility at only one DOE site. The study concludes that processing at only one facility is the lowest cost option, followed, in order of cost, by regional then individual site processing

  20. Secondary process for securing emergency cooling in nuclear facilities

    International Nuclear Information System (INIS)

    Bachl, H.

    1975-01-01

    An auxiliary process for securing the emergency cooling of nuclear power plants is described which is characterized in that a two-material heat power auxiliary process is connected at the cold end of the cooling circuit to a main heat power process to obtain mechanical energy from thermal, which in normal operation works as a cold-absorption process, but with failure of the main process changes to a heat power process with full evaporation and subsequent superheating of the two-materials mixture. (RW/LH) [de

  1. Alaska Synthetic Aperture Radar (SAR) Facility science data processing architecture

    Science.gov (United States)

    Hilland, Jeffrey E.; Bicknell, Thomas; Miller, Carol L.

    1991-01-01

    The paper describes the architecture of the Alaska SAR Facility (ASF) at Fairbanks, being developed to generate science data products for supporting research in sea ice motion, ice classification, sea-ice-ocean interaction, glacier behavior, ocean waves, and hydrological and geological study areas. Special attention is given to the individual substructures of the ASF: the Receiving Ground Station (RGS), the SAR Processor System, and the Interactive Image Analysis System. The SAR data will be linked to the RGS by the ESA ERS-1 and ERS-2, the Japanese ERS-1, and the Canadian Radarsat.

  2. Sustainable Acquisition Process Improvement for Naval Facilities Engineering Command

    National Research Council Canada - National Science Library

    Sanders, Erin

    2003-01-01

    .... While this guidance is intended to create an open and competitive process to achieve lowest cost or best value, conflicts among traditional acquisition processes and new law requirements are emerging...

  3. Toxicokinetics of beryllium following inhalation of beryllium oxide by Beagle dogs. III

    International Nuclear Information System (INIS)

    Finch, G.L.; Haley, P.J.; Hoover, M.D.; Mewhinney, J.A.; Bice, D.E.; Eidson, A.F.

    1988-01-01

    Young adult Beagle dogs inhaled radiolabeled beryllium oxide aerosols ( 7 BeO) prepared at either 500 deg. or 1000 deg. C to achieve one of two initial lung burdens (ILBs) of BeO. After exposure, animals were monitored by whole body counting for 7 Be, and excreta, clinical, and radiographic data were collected. One group of dogs was assigned for serial sacrifice for quantitation of beryllium clearance from lung, translocation to other organs, and histopathologic analysis of lung and lymph nodes. A second group of dogs was subjected to periodic bronchopulmonary lavage for analysis of lymphocyte responsiveness to beryllium. These latter dogs were subsequently re-exposed to the high ILB level of BeO prepared t 500 deg. C. ILBs following the second exposure were higher than that after the first exposure (74 vs. 42 μg BeO/kg, respectively). Except for one dog that exhibited enhanced beryllium retention after the second exposure, patterns of whole body clearance were similar to those observed after the initial exposures to the 500 deg. C-BeO. Analysis of lymphocyte responsiveness to beryllium in the second group of dogs is continuing. (author)

  4. Toxicokinetics of beryllium following inhalation of beryllium oxide by Beagle dogs. III

    Energy Technology Data Exchange (ETDEWEB)

    Finch, G L; Haley, P J; Hoover, M D; Mewhinney, J A; Bice, D E; Eidson, A F

    1988-12-01

    Young adult Beagle dogs inhaled radiolabeled beryllium oxide aerosols ({sup 7}BeO) prepared at either 500 deg. or 1000 deg. C to achieve one of two initial lung burdens (ILBs) of BeO. After exposure, animals were monitored by whole body counting for {sup 7}Be, and excreta, clinical, and radiographic data were collected. One group of dogs was assigned for serial sacrifice for quantitation of beryllium clearance from lung, translocation to other organs, and histopathologic analysis of lung and lymph nodes. A second group of dogs was subjected to periodic bronchopulmonary lavage for analysis of lymphocyte responsiveness to beryllium. These latter dogs were subsequently re-exposed to the high ILB level of BeO prepared t 500 deg. C. ILBs following the second exposure were higher than that after the first exposure (74 vs. 42 {mu}g BeO/kg, respectively). Except for one dog that exhibited enhanced beryllium retention after the second exposure, patterns of whole body clearance were similar to those observed after the initial exposures to the 500 deg. C-BeO. Analysis of lymphocyte responsiveness to beryllium in the second group of dogs is continuing. (author)

  5. Defense Waste Processing Facility Nitric- Glycolic Flowsheet Chemical Process Cell Chemistry: Part 2

    Energy Technology Data Exchange (ETDEWEB)

    Zamecnik, J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Edwards, T. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-06-06

    The conversions of nitrite to nitrate, the destruction of glycolate, and the conversion of glycolate to formate and oxalate were modeled for the Nitric-Glycolic flowsheet using data from Chemical Process Cell (CPC) simulant runs conducted by Savannah River National Laboratory (SRNL) from 2011 to 2016. The goal of this work was to develop empirical correlation models to predict these values from measureable variables from the chemical process so that these quantities could be predicted a-priori from the sludge or simulant composition and measurable processing variables. The need for these predictions arises from the need to predict the REDuction/OXidation (REDOX) state of the glass from the Defense Waste Processing Facility (DWPF) melter. This report summarizes the work on these correlations based on the aforementioned data. Previous work on these correlations was documented in a technical report covering data from 2011-2015. This current report supersedes this previous report. Further refinement of the models as additional data are collected is recommended.

  6. Status of material development for lifetime expansion of beryllium reflector

    Energy Technology Data Exchange (ETDEWEB)

    Dorn, C [Materion Brush Beryllium and Composites, California (United States); Tsuchiya, Kunihiko; Kawamura, Hiroshi [Japan Atomic Energy Agency, Oarai Research and Development Center, Oarai, Ibaraki (Japan); Hatano, Y [Univ. of Toyama, Toyama (Japan); Chakrov, P [INP-KNNC, Almaty (Kazakhstan); Kodama, M [Nippon Nuclear Fuel Development Co., Ltd., Oarai, Ibaraki (Japan)

    2012-03-15

    Beryllium has been used as the reflector element material in the reactor, specifically S-200F structural grade beryllium manufactured by Materion Brush Beryllium and Composites (former, Brush Wellman Inc.). As a part of the reactor upgrade, the Japan Atomic Energy Agency (JAEA) also has carried out the cooperation experiments to extend the operating lifetime of the beryllium reflector elements. It will first be necessary to determine which of the material's physical, mechanical and chemical properties will be the most influential on that choice. The irradiation testing plans to evaluate the various beryllium grades are also briefly considered and prepared. In this paper, material selection, irradiation test plan and PEI development for lifetime expansion of beryllium are described for material testing reactors. (author)

  7. Mechanisms of hydrogen retention in metallic beryllium and beryllium oxide and properties of ion-induced beryllium nitride; Rueckhaltemechanismen fuer Wasserstoff in metallischem Beryllium und Berylliumoxid sowie Eigenschaften von ioneninduziertem Berylliumnitrid

    Energy Technology Data Exchange (ETDEWEB)

    Oberkofler, Martin

    2011-09-22

    In the framework of this thesis laboratory experiments on atomically clean beryllium surfaces were performed. They aim at a basic understanding of the mechanisms occurring upon interaction of a fusion plasma with a beryllium first wall. The retention and the temperature dependent release of implanted deuterium ions are investigated. An atomistic description is developed through simulations and through the comparison with calculations based on density functional theory. The results of these investigations are compared to the behaviour of hydrogen upon implantation into thermally grown beryllium oxide layers. Furthermore, beryllium nitride is produced by implantation of nitrogen into metallic beryllium and its properties are investigated. The results are interpreted with regard to the use of beryllium in a fusion reactor. (orig.)

  8. Study on neutron irradiation behavior of beryllium as neutron multiplier

    Energy Technology Data Exchange (ETDEWEB)

    Ishitsuka, Etsuo [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

    1998-03-01

    More than 300 tons beryllium is expected to be used as a neutron multiplier in ITER, and study on the neutron irradiation behavior of beryllium as the neutron multiplier with Japan Materials Testing Reactor (JMTR) were performed to get the engineering data for fusion blanket design. This study started as the study on the tritium behavior in beryllium neutron reflector in order to make clear the generation mechanism on tritium of JMTR primary coolant since 1985. These experiences were handed over to beryllium studies for fusion study, and overall studies such as production technology of beryllium pebbles, irradiation behavior evaluation and reprocessing technology have been started since 1990. In this presentation, study on the neutron irradiation behavior of beryllium as the neutron multiplier with JMTR was reviewed from the point of tritium release, thermal properties, mechanical properties and reprocessing technology. (author)

  9. Use of data processing tools in decommissioning nuclear facilities

    International Nuclear Information System (INIS)

    Petrasch, P.; Lukacs, G.

    1995-01-01

    With the present level of electronic data processing technology, no project of the scale of nuclear reactor decommissioning could be carried out without the use of data processing systems. On the contrary, a reactor decommissioning project requires essential support not only for the technical but also the economic side through the use of proper data processing programs, and not only general applications in the area of personal computers such as MS-EXCEL or MS Project, but also special data processing systems designed for the reactor decommissioning tasks. Various data processing supports are required depending upon the progress of a reactor decommissioning project. (orig./DG) [de

  10. Measuring device for bending of beryllium reflector

    International Nuclear Information System (INIS)

    Nishida, Seiri; Sakamoto, Naoki.

    1994-01-01

    The device of the present invention can measure bending of a beryllium reflector formed in a reactor core of a nuclear reactor by a relatively easy operation. Namely, a sensor portion comprises a long-support that can be inserted to a fuel element-insertion hole disposed in the reactor and a plurality of distance sensors disposed in a longitudinal direction of the support. A supersonic wave sensor which is advantageous in the heat resistance, the size and the accuracy and can conduct measurement in water relatively easily is used as the distance sensors. However, other sensors, instead of the sensor described above, may also be used. The plurality of distance sensors detect the bending amount of the beryllium reflector in the longitudinal direction by such an easy operation of inserting such a sensor portion to the fuel element-insertion hole upon exchange of fuel elements. (I.S.)

  11. Facility design philosophy: Tank Waste Remediation System Process support and infrastructure definition

    International Nuclear Information System (INIS)

    Leach, C.E.; Galbraith, J.D.; Grant, P.R.; Francuz, D.J.; Schroeder, P.J.

    1995-11-01

    This report documents the current facility design philosophy for the Tank Waste Remediation System (TWRS) process support and infrastructure definition. The Tank Waste Remediation System Facility Configuration Study (FCS) initially documented the identification and definition of support functions and infrastructure essential to the TWRS processing mission. Since the issuance of the FCS, the Westinghouse Hanford Company (WHC) has proceeded to develop information and requirements essential for the technical definition of the TWRS treatment processing programs

  12. Interaction of hydrogen and its isotopes with irradiated beryllium

    International Nuclear Information System (INIS)

    Tazhibaeva, I.L.; Shestakov, V.P.; Klepikov, A.Kh.; Pomanenko, O.G.; Chikhraj, E.V.; Kenzhin, E.A.; Zverev, V.V.; Kolbanenkov, A.N.

    2000-01-01

    In the article the results of experiments on hydrogen and its isotopes accumulation and gas-release from irradiated beryllium are presented. The irradiation was conducted at different media and temperatures in the RA and IVG.1M reactors. The measurements were carried out by thermal desorption method. Hydrogen release from beryllium samples saturated at different conditions were calculated. Dependence of hydrogen confinement character in beryllium from grain orientation in the sample, temperature and irradiation rate was revealed

  13. Premelting hcp to bcc Transition in Beryllium

    Science.gov (United States)

    Lu, Y.; Sun, T.; Zhang, Ping; Zhang, P.; Zhang, D.-B.; Wentzcovitch, R. M.

    2017-04-01

    Beryllium (Be) is an important material with wide applications ranging from aerospace components to x-ray equipment. Yet a precise understanding of its phase diagram remains elusive. We have investigated the phase stability of Be using a recently developed hybrid free energy computation method that accounts for anharmonic effects by invoking phonon quasiparticles. We find that the hcp → bcc transition occurs near the melting curve at 0 materials.

  14. Analysis of surface contaminants on beryllium windows

    International Nuclear Information System (INIS)

    Gmur, N.F.

    1986-12-01

    It is known that various crystalline and liquid compounds form on the downstream surfaces of beryllium windows exposed to air. It is also known that the integrity of such windows may be compromised resulting in leaks through the window. The purpose of this report is to document the occurrences described as they pertain to the NSLS and to analyze, where possible, the various substances formed

  15. Preparation and properties of beryllium diphosphide

    International Nuclear Information System (INIS)

    Brice, J.F.; Gerardin, R.; Zanne, M.; Gleitzer, C.; Aubry, J.

    1975-01-01

    The compound BeP 2 can be obtained by direct action of phosphor on beryllium metal at 800-1000 0 C, and by removal of arsenic with phosphor in the diarsenide BeAs 2 . BeP 2 is a non hygroscopic brown-red powder. The X rays diffraction provide evidence for a quadratic cell with a = 7.08 A and c = 15.06 A. The atomic stacking is diamond type

  16. Beryllium window flange for synchrotron radiation X-ray beamline fabricated by hot isostatic press method

    International Nuclear Information System (INIS)

    Asaoka, Seiji; Maezawa, Hideki; Nishida, Kiyotoshi; Sakamoto, Naoki.

    1995-01-01

    The synchrotron radiation experimental facilities in National Laboratory for High Energy Physics are the experimental facilities for joint utilization, that possess the positron storage ring of 2.5 GeV exclusively used for synchrotron radiation. Synchrotron radiation is led through a mainstay beam channel to the laboratory, and in the beam line of X-ray, it is used for experiment through the taking-out window made of beryllium. At this time, the function of the taking-out window is to shut off between the ultrahigh vacuum in the mainstay beam channel and the atmosphere, and to cut the low energy component of synchrotron radiation spectra. The experiment using X-ray is carried out mostly in the atmosphere. The design of the efficient cooling water channel which is compatible with the flange construction is important under the high thermal load of synchrotron radiation. The beryllium window flange for synchrotron radiation X-ray was made by HIP method, and the ultrahigh vacuum test, the high pressure water flow test and the actual machine test were carried out by heat cycle. The properties required for the window material, the requirement of the construction, the new development of HIP method, and the experiments for evaluating the manufactured beryllium window are described. (K.I.)

  17. Comprehensive Measurement of Neutron Yield Produced by 62 MeV Protons on Beryllium Target

    International Nuclear Information System (INIS)

    Osipenko, M.; Ripani, M.; Ricco, G.; Alba, R.; Schillaci, M.; Cosentino, L.; Del Zoppo, A.; Di Pietro, A.; Figuera, P.; Finocchiaro, P.; Maiolino, C.; Santonocito, D.; Scuderi, V.; Barbagallo, M.; Colonna, N.; Boccaccio, P.; Esposito, J.; Celentano, A.; Viberti, C.M.; Kostyukov, A.

    2013-06-01

    A low-power prototype of neutron amplifier, based on a 70 MeV, high current proton cyclotron being installed at LNL for the SPES RIB facility, was recently proposed within INFN-E project. This prototype uses a thick Beryllium converter to produce a fast neutron spectrum feeding a sub-critical reactor core. To complete the design of such facility the new measurement of neutron yield from a thick Beryllium target was performed at LNS. This measurement used liquid scintillator detectors to identify produced neutrons by Pulse Shape Discrimination and Time of Flight technique to measure neutron energy in the range 0.5-62 MeV. To extend the covered neutron energy range 3 He detector was used to measure neutrons below 0.5 MeV. The obtained yields were normalized to the charge deposited by the proton beam on the metallic Beryllium target. These techniques allowed to achieve a wide angular coverage from 0 to 150 degrees and to explore almost complete neutron energy interval. (authors)

  18. Training manual for process operation and management of radioactive waste treatment facility

    Energy Technology Data Exchange (ETDEWEB)

    Shon, J. S.; Kim, K. J.; Ahn, S. J. [and others

    2004-12-01

    Radioactive Waste Treatment Facility (RWTF) has been operating for safe and effective treatment of radioactive wastes generated in the Korea Atomic Energy Research Institute (KAERI). In RWTF, there are evaporation, bituminization and solar evaporation processes for liquid waste, solid waste treatment process and laundry process. As other radioactive waste treatment facilities in foreign countries, the emergency situation such as fire and overflow of liquid waste can be taken place during the operation and result in the spread of contamination of radioactivity. So, easy and definite operating procedure is necessary for the safe operation of the facility. This manual can be available as easy and concise training materials for new employees and workers dispatched from service agency. Especially, in case of emergency urgently occurred during operation, everyone working in the facility can quickly stop the facility following this procedure.

  19. Training manual for process operation and management of radioactive waste treatment facility

    International Nuclear Information System (INIS)

    Shon, J. S.; Kim, K. J.; Ahn, S. J.

    2004-12-01

    Radioactive Waste Treatment Facility (RWTF) has been operating for safe and effective treatment of radioactive wastes generated in the Korea Atomic Energy Research Institute (KAERI). In RWTF, there are evaporation, bituminization and solar evaporation processes for liquid waste, solid waste treatment process and laundry process. As other radioactive waste treatment facilities in foreign countries, the emergency situation such as fire and overflow of liquid waste can be taken place during the operation and result in the spread of contamination of radioactivity. So, easy and definite operating procedure is necessary for the safe operation of the facility. This manual can be available as easy and concise training materials for new employees and workers dispatched from service agency. Especially, in case of emergency urgently occurred during operation, everyone working in the facility can quickly stop the facility following this procedure

  20. Thin-Film Materials Synthesis and Processing Facility

    Data.gov (United States)

    Federal Laboratory Consortium — FUNCTION: Provides a wide capability for deposition and processing of thin films, including sputter and ion-beam deposition, thermal evaporation, electro-deposition,...

  1. Permeation behavior of deuterium implanted into beryllium

    Energy Technology Data Exchange (ETDEWEB)

    Nakamura, Hirofumi; Hayashi, Takumi; O' hira, Shigeru; Nishi, Masataka [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2001-09-01

    Study on Implantation Driven Permeation (IDP) behavior of deuterium through pure beryllium was investigated as a part of the research to predict the tritium permeation through the first wall components ITER (International Thermonuclear Experimental Reactor). The permeation experiments were carried out with two beryllium specimens, one was an unannealed specimen and the other was that annealed at 1173 K. The permeation flux was measured as a function of specimen temperature and incident ion flux. Surface analysis of specimen was also carried out after the permeation experiment. Permeation was observed only with the annealed specimen and no significant permeation was observed with unannealed specimen under the present experimental condition (maximum temperature: 685 K, detection limit: 1x10{sup 13} D atoms/m{sup 2}s). It could be attributed that the intrinsic lattice defects, which act as diffusion preventing site, decreased with the specimen annealing. Based on the result of steady and transient permeation behavior and surface analysis, it was estimated that the deuterium permeation implanted into annealed beryllium was controlled by surface recombination due to the oxide layer on the surface of the permeated side. (author)

  2. Behavior of beryllium pebbles under irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Dalle-Donne, M.; Scaffidi-Argentina, F. [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany). Inst. fuer Neutronenphysik und Reactortechnik; Baldwin, D.L.; Gelles, D.S.; Greenwood, L.R.; Kawamura, H.; Oliver, B.M.

    1998-01-01

    Beryllium pebbles are being considered in fusion reactor blanket designs as neutron multiplier. An example is the European `Helium Cooled Pebble Bed Blanket.` Several forms of beryllium pebbles are commercially available but little is known about these forms in response to fast neutron irradiation. Commercially available beryllium pebbles have been irradiated to approximately 1.3 x 10{sup 22} n/cm{sup 2} (E>1 MeV) at 390degC. Pebbles 1-mm in diameter manufactured by Brush Wellman, USA and by Nippon Gaishi Company, Japan, and 3-mm pebbles manufactured by Brush Wellman were included. All were irradiated in the below-core area of the Experimental Breeder Reactor-II in Idaho Falls, USA, in molybdenum alloy capsules containing helium. Post-irradiation results are presented on density change measurements, tritium release by assay, stepped-temperature anneal, and thermal ramp desorption tests, and helium release by assay and stepped-temperature anneal measurements, for Be pebbles from two manufacturing methods, and with two specimen diameters. The experimental results on density change and tritium and helium release are compared with the predictions of the code ANFIBE. (author)

  3. Dynamic behaviour of S200F beryllium

    International Nuclear Information System (INIS)

    Montoya, Dominique

    1991-01-01

    Compression tests have been made on a large scale of strain, strain rate (up to 2000 s -1 ) and temperature (between 20 C and 300 C). From these experiences, we have calculated a constitutive model for beryllium S200F, which can be used by computer codes. Its formulation is not far from Steinberg, Cochran and Guinan's. But in our case, the influences of temperature and strain rate appear clearly within the expression. To validate our equation, we have used it in a computer code. Its extrapolation for higher strain rates is in good agreement with experiments such as Taylor impact tests or plate impact tests (strain rates greater than 10 4 s -1 ). With micrography, we could settle a link between the main strain mode within the material, and the variation of one parameter of the model. Beside the constitutive model, we have shown that shock loaded beryllium behaves in two different ways. If the strain rate is lower than 5.10 6 s -1 , then it is proportional to the squared shock pressure. Beyond, it is a linear function of shock pressure to the power of four. By a spall study on beryllium, we have confirmed that it is excessively fragile. Its fracture is sudden, at a strength near 1 GPa. (author) [fr

  4. Quantitative method of determining beryllium or a compound thereof in a sample

    Science.gov (United States)

    McCleskey, T. Mark; Ehler, Deborah S.; John, Kevin D.; Burrell, Anthony K.; Collis, Gavin E.; Minogue, Edel M.; Warner, Benjamin P.

    2010-08-24

    A method of determining beryllium or a beryllium compound thereof in a sample, includes providing a sample suspected of comprising beryllium or a compound thereof, extracting beryllium or a compound thereof from the sample by dissolving in a solution, adding a fluorescent indicator to the solution to thereby bind any beryllium or a compound thereof to the fluorescent indicator, and determining the presence or amount of any beryllium or a compound thereof in the sample by measuring fluorescence.

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

    International Nuclear Information System (INIS)

    Charak, I; Pedersen, D.R.; Forrester, R.J.; Phipps, R.D.

    1993-01-01

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

  6. The Cryogenic Properties of Several Aluminum-Beryllium Alloys and a Beryllium Oxide Material

    Science.gov (United States)

    Gamwell, Wayne R.; McGill, Preston B.

    2003-01-01

    Performance related mechanical properties for two aluminum-beryllium (Al-Be) alloys and one beryllium-oxide (BeO) material were developed at cryogenic temperatures. Basic mechanical properties (Le., ultimate tensile strength, yield strength, percent elongation, and elastic modulus were obtained for the aluminum-beryllium alloy, AlBeMetl62 at cryogenic [-195.5"C (-320 F) and -252.8"C (-423"F)I temperatures. Basic mechanical properties for the Be0 material were obtained at cyrogenic [- 252.8"C (-423"F)] temperatures. Fracture properties were obtained for the investment cast alloy Beralcast 363 at cryogenic [-252.8"C (-423"F)] temperatures. The AlBeMetl62 material was extruded, the Be0 material was hot isostatic pressing (HIP) consolidated, and the Beralcast 363 material was investment cast.

  7. Safety and environmental process for the design and construction of the National Ignition Facility

    Energy Technology Data Exchange (ETDEWEB)

    Brereton, S.J., LLNL

    1998-05-27

    The National Ignition Facility (NIF) is a U.S. Department of Energy (DOE) laser fusion experimental facility currently under construction at the Lawrence Livermore National Laboratory (LLNL). This paper describes the safety and environmental processes followed by NIF during the design and construction activities.

  8. Formation of cellular structure in beryllium at plastic working

    International Nuclear Information System (INIS)

    Papirov, I.I.; Nikolaenko, A.A.; Shokurov, V.S.; Pikalov, A.I.

    2013-01-01

    Conditions of cellular structure formation are investigated at various kinds of deformation and heat treatment of beryllium ingots. It is shown that the cellular structure plays the important role in formation of complex of physical mechanical properties of beryllium. Influence of impurity, various conditions of deformation (temperature, squeezing degree) and heat treatments on substructure, texture and mechanical properties of metal is investigated. Optimum conditions of rolling and heat treatments of beryllium are defined. The way of sign-variable cyclic deformation of beryllium ingots is offered for reception quasi-isotropic fine-grained metal. Physical-mechanical properties of ultra fine-grained metal are studied

  9. Practice for dosimetry in electron and bremsstrahlung irradiation facilities for food processing. 2. ed.

    International Nuclear Information System (INIS)

    2002-01-01

    This practice describes dosimetric procedures to be followed in facility characterization, process qualification, and routine processing for electron beam and bremsstrahlung irradiation facilities for food processing to ensure that product receives an acceptable range of absorbed doses. Other procedures related to facility characterization, process qualification, and routine product processing that may influence and be used to monitor absorbed dose in the product are also discussed. Information about effective or regulatory dose limits for food products is not within the scope of this practice (see ASTM Guides F 1355 and F 1356). The electron energy range covered in this practice is from 0.3 MeV to 10 MeV. Such electrons can be generated in continuous or pulse modes. The maximum electron energy of bremsstrahlung facilities covered in this practice is 10 MeV. A photon beam can be generated by inserting a bremsstrahlung converter in the electron beam path (See ISO/ASTM Practice 51608

  10. Age hardening in rapidly solidified and hot isostatically pressed beryllium-aluminum-silver alloys

    International Nuclear Information System (INIS)

    Carter, D.H.; McGeorge, A.C.; Jacobson, L.A.; Stanek, P.W.

    1995-01-01

    Three different alloys of beryllium, aluminum and silver were processed to powder by centrifugal atomization in a helium atmosphere. Alloy compositions were, by weight, 50% Be, 47.5% Al, 2.5% Ag, 50% Be, 47% Al, 3% Ag, and 50% Be, 46% Al, 4% Ag. Due to the low solubility of both aluminum and silver in beryllium, the silver was concentrated in the aluminum phase, which appeared to separate from the beryllium in the liquid phase. A fine, continuous composite beryllium-aluminum microstructure was formed, which did not significantly change after hot isostatically pressing at 550 C for one hour at 30,000 psi argon pressure. Samples of HIP material were solution treated at 550 C for one hour, followed by a water quench. Aging temperatures were 150, 175, 200 and 225 C for times ranging from one half hour to 65 hours. Hardness measurements were made using a diamond pyramid indenter with a load of 1 kg. Results indicate that peak hardness was reached in 36--40 hours at 175 C and 12--16 hours at 200 C aging temperature, relatively independent of alloy composition

  11. Corrosion of beryllium oxide; Corrosion de l'oxyde de beryllium

    Energy Technology Data Exchange (ETDEWEB)

    Elston, J; Caillat, R [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1958-07-01

    Data are reported on the volatilization rate of beryllium oxide in moist air depending on temperature and water vapour concentration. They are concerned with powder samples or sintered shapes of various densities. For sintered samples, the volatilization rate is very low under the following conditions: - temperature: 1300 deg. C, - water vapour concentration in moist air: 25 g/m{sup 3}, - flow rate: 12 I/hour corresponding to a speed of 40 m/hour on the surface of the sample. For calcinated powders (1300 deg. C), grain growth has been observed under a stream of moist air at 1100 deg. C. For instance, grain size changes from 0,5 to at least 2 microns after 500 hours of exposure at this temperature. Furthermore, results data are reported on corrosion of sintered beryllium oxide in pressurized water. At 250 deg. C, under a pressure of 40 kg/cm{sup 2} water is very slightly corrosive; however, internal strains are revealed. Finally, some features on the corrosion in liquid sodium are exposed. (author)Fren. [French] La volatilisation de l'oxyde de beryllium dans l'air humide est etudiee en fonction de la temperature pour differentes teneurs de vapeur d'eau. Les essais decrits portent sur de l'oxyde de beryllium en poudre ou sur des echantillons d'oxyde de beryllium fritte de differentes densites. Avec un debit d'air de 12 I/h contenant 25 g de vapeur par m{sup 3} correspondant a une vitesse de 40 m/h sur la surface de l'echantillon, la volatilisation des frittes a 1300 deg. C reste tres faible. Sur de la poudre d'oxyde de beryllium calcinee initialement a 1300 deg. C, on observe un grossissement de la taille des grains sous l'influence de l'air humide a 1100 deg. C. Par exemple, elle passe de 0,5 a au moins 2 microns apres 500 heures d'exposition a cette temperature. On donne d'autre part les resultats d'une etude de la corrosion de frittes d'oxyde de beryllium par l'eau, en autoclave. A 250 deg. C, sous une pression de 40 kg/cm{sup 2}, l'action de l'eau reste tres

  12. Status of the beryllium replacement project

    Energy Technology Data Exchange (ETDEWEB)

    Dimayuga, I. [Atomic Energy of Canada, Chalk River, Ontario (Canada); Corcoran, E. [Royal Military College of Canada, Kingston, Ontario (Canada); Daniels, T. [Ontario Power Generation, Pickering, Ontario (Canada); Harmsen, J. [Cameco Fuel Manufacturing Inc., Port Hope, Ontario (Canada); Lu, E. [Bruce Power, Tiverton, Ontario (Canada); Onderwater, T.; Palleck, S. [General Electric- Hitachi, Wilmington, North Carolina (United States); Pant, A. [Cameco Fuel Manufacturing Inc., Port Hope, Ontario (Canada)

    2013-07-01

    Currently, beryllium (Be) is used as the filler metal for brazing appendages on the sheaths of CANDU® fuel elements. Because of its toxicity, occupational exposure limits for Be are being reduced to very low levels, resulting in significant challenges to CANDU® fuel fabricators. The CANDU® Owners Group (COG) initiated a test program to identify a filler material to replace Be and confirm that the brazed joints meet the established technical requirements for CANDU® fuel. Together with eliminating health risks associated with the use of Be, the industry needs to be assured that continuation of fuel supply remains unaffected and that fuel fabrication processes continue to comply with health and safety standards. A literature survey of studies on brazing and joining of Zircaloy identified potential filler materials that can meet or exceed existing design requirements of the brazed joint, including the required mechanical, microstructural, corrosion resistance, and irradiation properties equivalent to those obtained with Be as braze material. Candidate materials were evaluated against several criteria, including manufacturability, melting point, wettability, mechanical properties, corrosion resistance, effect on neutron economy, potential activation products, and interaction with fuel channels and other related disciplines. This exercise resulted in a list of promising candidate materials that were recommended for the first phase of testing. These materials include stainless steel (304 or 316), Al-Si, Ni-P, and Zr-Mn alloys. To allow a CANDU® utility have sufficient confidence in considering implementation of a different braze filler material, a Be Replacement Test Program, involving out-reactor and in-reactor tests, is being undertaken as a collaborative endeavour by the Canadian nuclear industry. The out-reactor tests consist of: a constructability assessment to determine the material’s suitability with current fuel manufacturing methods; evaluation of

  13. Seismic qualification program plan for continued operation at DOE-SRS nuclear material processing facilities

    International Nuclear Information System (INIS)

    Talukdar, B.K.; Kennedy, W.N.

    1991-01-01

    The Savannah River Facilities for the most part were constructed and maintained to standards that were developed by Du Pont and are not rigorously in compliance with the current General Design Criteria (GDC); DOE Order 6430.IA requirements. In addition, many of the facilities were built more than 30 years ago, well before DOE standards for design were issued. The Westinghouse Savannah River Company (WSRC) his developed a program to address the evaluation of the Nuclear Material Processing (NMP) facilities to GDC requirements. The program includes a facility base-line review, assessment of areas that are not in compliance with the GDC requirements, planned corrective actions or exemptions to address the requirements, and a safety assessment. The authors from their direct involvement with the Program, describe the program plan for seismic qualification including other natural phenomena hazards,for existing NMP facility structures to continue operation Professionals involved in similar effort at other DOE facilities may find the program useful

  14. Advanced Process Control Application and Optimization in Industrial Facilities

    Directory of Open Access Journals (Sweden)

    Howes S.

    2015-01-01

    Full Text Available This paper describes application of the new method and tool for system identification and PID tuning/advanced process control (APC optimization using the new 3G (geometric, gradient, gravity optimization method. It helps to design and implement control schemes directly inside the distributed control system (DCS or programmable logic controller (PLC. Also, the algorithm helps to identify process dynamics in closed-loop mode, optimizes controller parameters, and helps to develop adaptive control and model-based control (MBC. Application of the new 3G algorithm for designing and implementing APC schemes is presented. Optimization of primary and advanced control schemes stabilizes the process and allows the plant to run closer to process, equipment and economic constraints. This increases production rates, minimizes operating costs and improves product quality.

  15. Defense Waste Processing Facility Simulant Chemical Processing Cell Studies for Sludge Batch 9

    International Nuclear Information System (INIS)

    Smith, Tara E.; Newell, J. David; Woodham, Wesley H.

    2016-01-01

    The Savannah River National Laboratory (SRNL) received a technical task request from Defense Waste Processing Facility (DWPF) and Saltstone Engineering to perform simulant tests to support the qualification of Sludge Batch 9 (SB9) and to develop the flowsheet for SB9 in the DWPF. These efforts pertained to the DWPF Chemical Process Cell (CPC). CPC experiments were performed using SB9 simulant (SB9A) to qualify SB9 for sludge-only and coupled processing using the nitric-formic flowsheet in the DWPF. Two simulant batches were prepared, one representing SB8 Tank 40H and another representing SB9 Tank 51H. The simulant used for SB9 qualification testing was prepared by blending the SB8 Tank 40H and SB9 Tank 51H simulants. The blended simulant is referred to as SB9A. Eleven CPC experiments were run with an acid stoichiometry ranging between 105% and 145% of the Koopman minimum acid equation (KMA), which is equivalent to 109.7% and 151.5% of the Hsu minimum acid factor. Three runs were performed in the 1L laboratory scale setup, whereas the remainder were in the 4L laboratory scale setup. Sludge Receipt and Adjustment Tank (SRAT) and Slurry Mix Evaporator (SME) cycles were performed on nine of the eleven. The other two were SRAT cycles only. One coupled flowsheet and one extended run were performed for SRAT and SME processing. Samples of the condensate, sludge, and off-gas were taken to monitor the chemistry of the CPC experiments.

  16. Defense Waste Processing Facility Simulant Chemical Processing Cell Studies for Sludge Batch 9

    Energy Technology Data Exchange (ETDEWEB)

    Smith, Tara E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Newell, J. David [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Woodham, Wesley H. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2016-08-10

    The Savannah River National Laboratory (SRNL) received a technical task request from Defense Waste Processing Facility (DWPF) and Saltstone Engineering to perform simulant tests to support the qualification of Sludge Batch 9 (SB9) and to develop the flowsheet for SB9 in the DWPF. These efforts pertained to the DWPF Chemical Process Cell (CPC). CPC experiments were performed using SB9 simulant (SB9A) to qualify SB9 for sludge-only and coupled processing using the nitric-formic flowsheet in the DWPF. Two simulant batches were prepared, one representing SB8 Tank 40H and another representing SB9 Tank 51H. The simulant used for SB9 qualification testing was prepared by blending the SB8 Tank 40H and SB9 Tank 51H simulants. The blended simulant is referred to as SB9A. Eleven CPC experiments were run with an acid stoichiometry ranging between 105% and 145% of the Koopman minimum acid equation (KMA), which is equivalent to 109.7% and 151.5% of the Hsu minimum acid factor. Three runs were performed in the 1L laboratory scale setup, whereas the remainder were in the 4L laboratory scale setup. Sludge Receipt and Adjustment Tank (SRAT) and Slurry Mix Evaporator (SME) cycles were performed on nine of the eleven. The other two were SRAT cycles only. One coupled flowsheet and one extended run were performed for SRAT and SME processing. Samples of the condensate, sludge, and off-gas were taken to monitor the chemistry of the CPC experiments.

  17. Safety evaluation report of hot cell facilities for demonstration of advanced spent fuel conditioning process

    International Nuclear Information System (INIS)

    You, Gil Sung; Choung, W. M.; Ku, J. H.; Cho, I. J.; Kook, D. H.; Park, S. W.; Bek, S. Y.; Lee, E. P.

    2004-10-01

    The advanced spent fuel conditioning process(ACP) proposed to reduce the overall volume of the PWR spent fuel and improve safety and economy of the long-term storage of spent fuel. In the next phase(2004∼2006), the hot test will be carried out for verification of the ACP in a laboratory scale. For the hot test, the hot cell facilities of α- type and auxiliary facilities are required essentially for safe handling of high radioactive materials. As the hot cell facilities for demonstration of the ACP, a existing hot cell of β- type will be refurbished to minimize construction expenditures of hot cell facility. Up to now, the detail design of hot cell facilities and process were completed, and the safety analysis was performed to substantiate secure of conservative safety. The design data were submitted for licensing which was necessary for construction and operation of hot cell facilities. The safety investigation of KINS on hot cell facilities was completed, and the license for construction and operation of hot cell facilities was acquired already from MOST. In this report, the safety analysis report submitted to KINS was summarized. And also, the questionnaires issued from KINS and answers of KAERI in process of safety investigation were described in detail

  18. Characterization of decontamination and decommissioning wastes expected from the major processing facilities in the 200 Areas

    International Nuclear Information System (INIS)

    Amato, L.C.; Franklin, J.D.; Hyre, R.A.; Lowy, R.M.; Millar, J.S.; Pottmeyer, J.A.; Duncan, D.R.

    1994-08-01

    This study was intended to characterize and estimate the amounts of equipment and other materials that are candidates for removal and subsequent processing in a solid waste facility when the major processing and handling facilities in the 200 Areas of the Hanford Site are decontaminated and decommissioned. The facilities in this study were selected based on processing history and on the magnitude of the estimated decommissioning cost cited in the Surplus Facilities Program Plan; Fiscal Year 1993 (Winship and Hughes 1992). The facilities chosen for this study include B Plant (221-B), T Plant (221-T), U Plant (221-U), the Uranium Trioxide (UO 3 ) Plant (224-U and 224-UA), the Reduction Oxidation (REDOX) or S Plant (202-S), the Plutonium Concentration Facility for B Plant (224-B), and the Concentration Facility for the Plutonium Finishing Plant (PFP) and REDOX (233-S). This information is required to support planning activities for current and future solid waste treatment, storage, and disposal operations and facilities

  19. Characterization of decontamination and decommissioning wastes expected from the major processing facilities in the 200 Areas

    Energy Technology Data Exchange (ETDEWEB)

    Amato, L.C.; Franklin, J.D.; Hyre, R.A.; Lowy, R.M.; Millar, J.S.; Pottmeyer, J.A. [Los Alamos Technical Associates, Kennewick, WA (United States); Duncan, D.R. [Westinghouse Hanford Co., Richland, WA (United States)

    1994-08-01

    This study was intended to characterize and estimate the amounts of equipment and other materials that are candidates for removal and subsequent processing in a solid waste facility when the major processing and handling facilities in the 200 Areas of the Hanford Site are decontaminated and decommissioned. The facilities in this study were selected based on processing history and on the magnitude of the estimated decommissioning cost cited in the Surplus Facilities Program Plan; Fiscal Year 1993 (Winship and Hughes 1992). The facilities chosen for this study include B Plant (221-B), T Plant (221-T), U Plant (221-U), the Uranium Trioxide (UO{sub 3}) Plant (224-U and 224-UA), the Reduction Oxidation (REDOX) or S Plant (202-S), the Plutonium Concentration Facility for B Plant (224-B), and the Concentration Facility for the Plutonium Finishing Plant (PFP) and REDOX (233-S). This information is required to support planning activities for current and future solid waste treatment, storage, and disposal operations and facilities.

  20. Zero-Release Mixed Waste Process Facility Design and Testing

    International Nuclear Information System (INIS)

    Richard D. Boardman; John A. Deldebbio; Robert J. Kirkham; Martin K. Clemens; Robert Geosits; Ping Wan

    2004-01-01

    A zero-release off-gas cleaning system for mixed-waste thermal treatment processes has been evaluated through experimental scoping tests and process modeling. The principles can possibly be adapted to a fluidized-bed calcination or stream reforming process, a waste melter, a rotary kiln process, and possibly other waste treatment thermal processes. The basic concept of a zero-release off-gas cleaning system is to recycle the bulk of the off-gas stream to the thermal treatment process. A slip stream is taken off the off-gas recycle to separate and purge benign constituents that may build up in the gas, such as water vapor, argon, nitrogen, and CO2. Contaminants are separated from the slip stream and returned to the thermal unit for eventual destruction or incorporation into the waste immobilization media. In the current study, a standard packed-bed scrubber, followed by gas separation membranes, is proposed for removal of contaminants from the off-gas recycle slipstream. The scrub solution is continuously regenerated by cooling and precipitating sulfate, nitrate, and other salts that reach a solubility limit in the scrub solution. Mercury is also separated by the scrubber. A miscible chemical oxidizing agent was shown to effectively oxidize mercury and also NO, thus increasing their removal efficiency. The current study indicates that the proposed process is a viable option for reducing off-gas emissions. Consideration of the proposed closed-system off-gas cleaning loop is warranted when emissions limits are stringent, or when a reduction in the total gas emissions volume is desired. Although the current closed-loop appears to be technically feasible, economical considerations must be also be evaluated on a case-by-case basis

  1. The plastic deformation of copper-beryllium alloy

    International Nuclear Information System (INIS)

    Gadalla, A.A.

    1980-01-01

    Copper-2 at. % Be alloy specimens containing coherent and/or incoherent spherical beryllium precipitates have been tested at different temperatures. The precipitation process and dislocation substructure were examined by electron microscopy. The yielding process is consistent with Orwan mechanism, and the work-hardening of the alloy is parabolic in nature for smaller particles but changes to three-stage hardening for larger particles. The extent of stage I deformation is temperature dependent, and the rate of work-hardening is quite steep and may be described by either the Ashby or the Hirsch parabolic models. There is a noticeable softening during this stage which may be attributed to shearing of particles during deformation. The dislocation substructure shows a uniform distribution of fine dislocations as well as propagation of cracks across grains. The precipitates are a mixture of semi-coherent and incoherent particles. (author)

  2. Criticality Safety Evaluation Report for the Cold Vacuum Drying (CVD) Facility's Process Water Handling System

    International Nuclear Information System (INIS)

    KESSLER, S.F.

    2000-01-01

    This report addresses the criticality concerns associated with process water handling in the Cold Vacuum Drying Facility. The controls and limitations on equipment design and operations to control potential criticality occurrences are identified

  3. Criticality safety evaluation report for the cold vacuum drying facility's process water handling system

    International Nuclear Information System (INIS)

    NELSON, J.V.

    1999-01-01

    This report addresses the criticality concerns associated with process water handling in the Cold Vacuum Drying Facility. The controls and limitations on equipment design and operations to control potential criticality occurrences are identified

  4. NO/sub x/ removal facility: MON process

    Energy Technology Data Exchange (ETDEWEB)

    Fukuda, Y

    1974-05-01

    A newly developed process for nitrogen oxides removal is described. The MON process, named for Mitsubishi Kizoku, Okabe of Tohoku Univ., and Nippon Kagaku, uses potassium permanganate as an oxidizing agent. Potassium permanganate in alkaline solution converts nitric oxide into nitrate and nitrogen dioxide into nitric acid. The resulting MnO/sub 2/ is easily filtered and recovered as material for the manufacturing of KMnO/sub 4/. Contrary to the conventional methods, the NO/sub x/ conversion rate increases with increasing temperature. Test results at a pilot plant showed that NO/sub x/ was reduced from 570 ppM (nitric oxide 520 ppM) to 27 ppM (mostly NO) at 97 to 98 percent conversion. Another advantage of the process is that other acidic gases such as sulfur dioxide are also removed.

  5. Quality Assurance Program description, Defense Waste Processing Facility (DWPF)

    International Nuclear Information System (INIS)

    Maslar, S.R.

    1992-01-01

    This document describes the Westinghouse Savannah River Company's (WSRC) Quality Assurance Program for Defense Waste Processing at the Savannah River Site (SRS). WSRC is the operating contractor for the US Department of Energy (DOE) at the SRS. The following objectives are achieved through developing and implementing the Quality Assurance Program: (1) Ensure that the attainment of quality (in accomplishing defense high-level waste processing objectives at the SRS) is at a level commensurate with the government's responsibility for protecting public health and safety, the environment, the public investment, and for efficiently and effectively using national resources. (2) Ensure that high-level waste from qualification and production activities conform to requirements defined by OCRWM. These activities include production processes, equipment, and services; and products that are planned, designed, procured, fabricated, installed, tested, operated, maintained, modified, or produced

  6. Plutonium production story at the Hanford site: processes and facilities history

    Energy Technology Data Exchange (ETDEWEB)

    Gerber, M.S., Westinghouse Hanford

    1996-06-20

    This document tells the history of the actual plutonium production process at the Hanford Site. It contains five major sections: Fuel Fabrication Processes, Irradiation of Nuclear Fuel, Spent Fuel Handling, Radiochemical Reprocessing of Irradiated Fuel, and Plutonium Finishing Operations. Within each section the story of the earliest operations is told, along with changes over time until the end of operations. Chemical and physical processes are described, along with the facilities where these processes were carried out. This document is a processes and facilities history. It does not deal with the waste products of plutonium production.

  7. Economics of gamma processing in cobalt-60 irradiation facilities

    International Nuclear Information System (INIS)

    Gay, H. G.; Kotler, J. G.

    1985-01-01

    Gamma processing by cobalt-60 is well established. However, since irradiation of food is relatively new from the commercial point of view, it is important to assess costs of gamma irradiation in the context of food processing. Five different types of AECL-RCC irradiation equipment are examined in terms of their throughputs, and capital and operating costs. Using these figures, costs of irradiation of nine types of food products are presented. In general, these represent about 2-10% of the wholesale cost of these products

  8. Computer program for source distribution process in radiation facility

    International Nuclear Information System (INIS)

    Al-Kassiri, H.; Abdul Ghani, B.

    2007-08-01

    Computer simulation for dose distribution using Visual Basic has been done according to the arrangement and activities of Co-60 sources. This program provides dose distribution in treated products depending on the product density and desired dose. The program is useful for optimization of sources distribution during loading process. there is good agreement between calculated data for the program and experimental data.(Author)

  9. New facility for processing and storage of radioactive and toxic chemical waste

    International Nuclear Information System (INIS)

    Gallagher, F.E. III

    1976-01-01

    A new facility for the processing and storage of radioactive and toxic chemical waste is described. The facility is located in the science and engineering complex of the Santa Barbara campus of the University of California, near the Pacific Ocean. It is designed to provide a safe and secure processing and storage area for hazardous wastes, while meeting the high aesthetic standards and ecological requirements of campus and community regulatory boards. The ventilation system and fire prevention features will be described in detail. During the design phase, a small laboratory was added to provide an area for the radiation protection and industrial hygiene programs. Operational experience with this new facility is discussed

  10. Waste management, waste resource facilities and waste conversion processes

    International Nuclear Information System (INIS)

    Demirbas, Ayhan

    2011-01-01

    In this study, waste management concept, waste management system, biomass and bio-waste resources, waste classification, and waste management methods have been reviewed. Waste management is the collection, transport, processing, recycling or disposal, and monitoring of waste materials. A typical waste management system comprises collection, transportation, pre-treatment, processing, and final abatement of residues. The waste management system consists of the whole set of activities related to handling, treating, disposing or recycling the waste materials. General classification of wastes is difficult. Some of the most common sources of wastes are as follows: domestic wastes, commercial wastes, ashes, animal wastes, biomedical wastes, construction wastes, industrial solid wastes, sewer, biodegradable wastes, non-biodegradable wastes, and hazardous wastes.

  11. Case cluster of pneumoconiosis at a coal slag processing facility.

    Science.gov (United States)

    Fagan, Kathleen M; Cropsey, Erin B; Armstrong, Jenna L

    2015-05-01

    During an inspection by the Occupational Safety and Health Administration (OSHA) of a small coal slag processing plant with 12 current workers, four cases of pneumoconiosis were identified among former workers. The OSHA investigation consisted of industrial hygiene sampling, a review of medical records, and case interviews. Some personal sampling measurements exceeded the OSHA Permissible Exposure Limit (PEL) for total dust exposures of 15 mg/m(3), and the measured respirable silica exposure of 0.043 mg/m(3), although below OSHA's current PEL for respirable dust containing silica, was above the American Conference of Governmental Industrial Hygienists' Threshold Limit Value (TLV). Chest x-rays for all four workers identified small opacities consistent with pneumoconiosis. This is the first known report of lung disease in workers processing coal slag and raises concerns for workers exposed to coal slag dust. © 2015 Wiley Periodicals, Inc.

  12. Processing device for discharged water from radioactive material handling facility

    International Nuclear Information System (INIS)

    Kono, Takao; Kono, Hiroyuki; Yasui, Katsuaki; Kataiki, Koichi.

    1995-01-01

    The device of the present invention comprises a mechanical floating material-removing means for removing floating materials in discharged water, an ultrafiltration device for separating processed water discharged from the removing means by membranes, a reverse osmotic filtration device for separating the permeated water and a condensing means for evaporating condensed water. Since processed water after mechanically removing floating materials is supplied to the ultrafiltration device, the load applied on the filtering membrane is reduced, to simplify the operation control as a total. In addition, since the amount of resultant condensed water is reduced, and the devolumed condensed water is condensed and dried, the condensing device is made compact and the amount of resultant wastes is reduced. (T.M.)

  13. Defense waste processing facility at Savannah River Plant. Instrument and power jumpers

    International Nuclear Information System (INIS)

    Heckendorm, F.M. II.

    1983-06-01

    The Defense Waste Processing Facility (DWPF) for waste vitrification at the Savannah River Plant is in the final design stage. Development of equipment interconnecting devices or jumpers for use within the remotely operated processing canyon is now complete. These devices provide for the specialized instrument and electrical requirements of the DWPF process for low-voltage, high-frequency, and high-power interconnections

  14. Hanford Central Waste Complex: Waste Receiving and Processing Facility dangerous waste permit application

    International Nuclear Information System (INIS)

    1991-10-01

    The Hanford Central Waste Complex is an existing and planned series of treatment, and/or disposal (TSD) unites that will centralize the management of solid waste operations at a single location on the Hanford Facility. The Complex includes two units: the WRAP Facility and the Radioactive Mixed Wastes Storage Facility (RMW Storage Facility). This Part B permit application addresses the WRAP Facility. The Facility will be a treatment and storage unit that will provide the capability to examine, sample, characterize, treat, repackage, store, and certify radioactive and/or mixed waste. Waste treated and stored will include both radioactive and/or mixed waste received from onsite and offsite sources. Certification will be designed to ensure and demonstrate compliance with waste acceptance criteria set forth by onsite disposal units and/or offsite facilities that subsequently are to receive waste from the WRAP Facility. This permit application discusses the following: facility description and general provisions; waste characterization; process information; groundwater monitoring; procedures to prevent hazards; contingency plant; personnel training; exposure information report; waste minimization plan; closure and postclosure requirements; reporting and recordkeeping; other relevant laws; certification

  15. Lessons learned from the Siting Process of an Interim Storage Facility in Spain - 12024

    Energy Technology Data Exchange (ETDEWEB)

    Lamolla, Meritxell Martell [MERIENCE Strategic Thinking, 08734 Olerdola, Barcelona (Spain)

    2012-07-01

    On 29 December 2009, the Spanish government launched a site selection process to host a centralised interim storage facility for spent fuel and high-level radioactive waste. It was an unprecedented call for voluntarism among Spanish municipalities to site a controversial facility. Two nuclear municipalities, amongst a total of thirteen municipalities from five different regions, presented their candidatures to host the facility in their territories. For two years the government did not make a decision. Only in November 30, 2011, the new government elected on 20 November 2011 officially selected a non-nuclear municipality, Villar de Canas, for hosting this facility. This paper focuses on analysing the factors facilitating and hindering the siting of controversial facilities, in particular the interim storage facility in Spain. It demonstrates that involving all stakeholders in the decision-making process should not be underestimated. In the case of Spain, all regional governments where there were candidate municipalities willing to host the centralised interim storage facility, publicly opposed to the siting of the facility. (author)

  16. Transient response and radiation dose estimates for breaches to a spent fuel processing facility

    Energy Technology Data Exchange (ETDEWEB)

    Solbrig, Charles W., E-mail: soltechco@aol.com; Pope, Chad; Andrus, Jason

    2014-08-15

    Highlights: • We model doses received from a nuclear fuel facility from boundary leaks due to an earthquake. • The supplemental exhaust system (SES) starts after breach causing air to be sucked into the cell. • Exposed metal fuel burns increasing pressure and release of radioactive contamination. • Facility releases are small and much less than the limits showing costly refits are unnecessary. • The method presented can be used in other nuclear fuel processing facilities. - Abstract: This paper describes the analysis of the design basis accident for Idaho National Laboratory Fuel Conditioning Facility (FCF). The facility is used to process spent metallic nuclear fuel. This analysis involves a model of the transient behavior of the FCF inert atmosphere hot cell following an earthquake initiated breach of pipes passing through the cell boundary. Such breaches allow the introduction of air and subsequent burning of pyrophoric metals. The model predicts the pressure, temperature, volumetric releases, cell heat transfer, metal fuel combustion, heat generation rates, radiological releases and other quantities. The results show that releases from the cell are minimal and satisfactory for safety. This analysis method should be useful in other facilities that have potential for damage from an earthquake and could eliminate the need to back fit facilities with earthquake proof boundaries or lessen the cost of new facilities.

  17. Transient response and radiation dose estimates for breaches to a spent fuel processing facility

    International Nuclear Information System (INIS)

    Solbrig, Charles W.; Pope, Chad; Andrus, Jason

    2014-01-01

    Highlights: • We model doses received from a nuclear fuel facility from boundary leaks due to an earthquake. • The supplemental exhaust system (SES) starts after breach causing air to be sucked into the cell. • Exposed metal fuel burns increasing pressure and release of radioactive contamination. • Facility releases are small and much less than the limits showing costly refits are unnecessary. • The method presented can be used in other nuclear fuel processing facilities. - Abstract: This paper describes the analysis of the design basis accident for Idaho National Laboratory Fuel Conditioning Facility (FCF). The facility is used to process spent metallic nuclear fuel. This analysis involves a model of the transient behavior of the FCF inert atmosphere hot cell following an earthquake initiated breach of pipes passing through the cell boundary. Such breaches allow the introduction of air and subsequent burning of pyrophoric metals. The model predicts the pressure, temperature, volumetric releases, cell heat transfer, metal fuel combustion, heat generation rates, radiological releases and other quantities. The results show that releases from the cell are minimal and satisfactory for safety. This analysis method should be useful in other facilities that have potential for damage from an earthquake and could eliminate the need to back fit facilities with earthquake proof boundaries or lessen the cost of new facilities

  18. Deuterium/hydrogen isotope exchange on beryllium and beryllium nitride; Deuterium/Wasserstoff-Isotopenaustausch an Beryllium und Berylliumnitrid

    Energy Technology Data Exchange (ETDEWEB)

    Dollase, Petra; Eichler, Michael; Koeppen, Martin; Dittmar, Timo; Linsmeier, Christian [Forschungszentrum Juelich GmbH, Institut fuer Energie- und Klimaforschung - Plasmaphysik (Germany)

    2016-07-01

    In the fusion experiments JET and ITER, the first wall is made up of beryllium. The use of nitrogen is discussed for radiative cooling in the divertor. This can react with the surface of the first wall to form beryllium nitride (Be{sub 3}N{sub 2}). The hydrogen isotopes deuterium and tritium, which react in the fusion reaction to helium and a neutron, are used as fuel. Since the magnetic confinement of the plasma is not perfect, deuterium and tritium ions are also found on the beryllium wall and can accumulate there. This should be avoided due to the radioactivity of tritium. Therefore the isotope exchange with deuterium is investigated to regenerate the first wall. We investigate the isotopic exchange of deuterium and protium in order to have not to work with radioactive tritium. The ion bombardment is simulated with an ion source. With voltages up to a maximum of 5 kV, deuterium and protic hydrogen ions are implanted in polycrystalline Be and Be{sub 3}N{sub 2}. The samples are then analyzed in situ using X-ray photoelectron spectroscopy (XPS) and thermal desorption spectroscopy (TDS). Subsequently, samples prepared under the same conditions are characterized ex-situ by means of nuclear reaction analysis (NRA). [German] In den Fusionsexperimenten JET und ITER besteht die erste Wand im Hauptraum aus Beryllium (Be). Zur Strahlungskuehlung im Divertor wird der Einsatz von Stickstoff diskutiert. Dieser kann mit der Oberflaeche der ersten Wand zu Berylliumnitrid (Be{sub 3}N{sub 2}) reagieren. Als Brennstoff werden die Wasserstoffisotope Deuterium und Tritium eingesetzt, die in der Fusionsreaktion zu Helium und einem Neutron reagieren. Da der magnetische Einschluss des Plasmas nicht perfekt ist, treffen auch Deuterium- und Tritiumionen auf die Berylliumwand auf und koennen sich dort anreichern. Das soll aufgrund der Radioaktivitaet von Tritium unbedingt vermieden werden. Daher wird zur Regenerierung der ersten Wand der Isotopenaustausch mit Deuterium untersucht. Wir

  19. Interfacial properties of HIP joint between beryllium and reduced activation ferritic/martensitic steel

    International Nuclear Information System (INIS)

    Hirose, T.; Ogiwara, H.; Enoeda, M.; Akiba, M.

    2007-01-01

    Full text of publication follows: ITER test blanket module is the most important components to validate energy production and fuel breeding process for future demonstration reactor. Reduced activation ferritic / martensitic steel is recognized as a promising structural material for breeding blanket systems. And Beryllium must be used as plasma facing materials for ITER in vessel components. In this work, interfacial properties of beryllium/reduced activation ferritic/martensitic steel (RAF/Ms) joint were investigated for a first wall of ITER test blanket module (TBM). The starting materials were ITER grade Beryllium, S65C and a Japanese RAF/M, F82H. The joint was produced by solid state hot isostatic pressing (HIP) method. Chromium layer with the thickness of 1 μm and 10 μm were formed by plasma vapor deposition on the beryllium surface as a diffusion barrier. The HIP was carried out at 1023 K and 1233 K which are determined by standard normalizing and tempering temperature of F82H. The joint made at 1233 K was followed by tempering at 1033 K. The bonding interface was characterized by electron probe microanalysis (EPMA). The bonding strength was also investigated by isometric four point bending tests at ambient temperature. EPMA showed chromium layer effectively worked as a diffusion barrier at 1023 K. However, the beryllium rich layer was formed in F82H after HIP at 1233 K followed by tempering. Bending tests revealed that thin chromium layer and low temperature HIP is preferable. The high temperature HIP introduce brittle BeFe inter metallic compounds along bonding interface. On the other hand, joint with thick chromium layer suffer from brittleness of chromium itself. (authors)

  20. Interfacial properties of HIP joint between beryllium and reduced activation ferritic/martensitic steel

    Energy Technology Data Exchange (ETDEWEB)

    Hirose, T. [Blanket Engineering Group, Japan Atomic Energy Agency, Naka, Ibaraki (Japan); Ogiwara, H. [Japan Atomic Energy Agency, Tokai-mura, Naga-gun, Ibaraki-ken (Japan); Enoeda, M. [Naka Fusion Research Establishment, J.A.E.R.I., Japan Atomic Energy Research Institute, Naka-gun, Ibaraki-ken (Japan); Akiba, M. [Naka Fusion Institute, Japan Atomic Energy Agency, Naka, Ibaraki (Japan)

    2007-07-01

    Full text of publication follows: ITER test blanket module is the most important components to validate energy production and fuel breeding process for future demonstration reactor. Reduced activation ferritic / martensitic steel is recognized as a promising structural material for breeding blanket systems. And Beryllium must be used as plasma facing materials for ITER in vessel components. In this work, interfacial properties of beryllium/reduced activation ferritic/martensitic steel (RAF/Ms) joint were investigated for a first wall of ITER test blanket module (TBM). The starting materials were ITER grade Beryllium, S65C and a Japanese RAF/M, F82H. The joint was produced by solid state hot isostatic pressing (HIP) method. Chromium layer with the thickness of 1 {mu}m and 10 {mu}m were formed by plasma vapor deposition on the beryllium surface as a diffusion barrier. The HIP was carried out at 1023 K and 1233 K which are determined by standard normalizing and tempering temperature of F82H. The joint made at 1233 K was followed by tempering at 1033 K. The bonding interface was characterized by electron probe microanalysis (EPMA). The bonding strength was also investigated by isometric four point bending tests at ambient temperature. EPMA showed chromium layer effectively worked as a diffusion barrier at 1023 K. However, the beryllium rich layer was formed in F82H after HIP at 1233 K followed by tempering. Bending tests revealed that thin chromium layer and low temperature HIP is preferable. The high temperature HIP introduce brittle BeFe inter metallic compounds along bonding interface. On the other hand, joint with thick chromium layer suffer from brittleness of chromium itself. (authors)

  1. New treatment facility for low level process effluents at the Savannah River site

    International Nuclear Information System (INIS)

    Ebra, M.A.; Bibler, J.P.; Johnston, B.S.; Kilpatrick, L.L.; Poy, F.L.; Wallace, R.M.

    1987-01-01

    A new facility, the F/H Effluent Treatment Facility (F/H ETF) is under construction at the Savannah River site. It will decontaminate process effluents containing low levels of radionuclides and hazardous chemicals prior to discharge to a surface stream. These effluents, which are currently discharged to seepage basins, originate in the chemical separations and high-level radioactive waste processing areas, known as F-Area and H-Area. The new facility will allow closure of the basins in order to meet the provisions of the Resource Conservation and Recovery Act by November 1988. A high degree of reliability is expected from this design as a result of extensive process development work that has been conducted at the Savannah River Laboratory. This work has included both bench scale testing of individual unit operations and pilot scale testing of an integrated facility, 150 to 285 L/min (40 to 75 gpm), that contains the major operations

  2. Beryllium-stimulated neopterin as a diagnostic adjunct in chronic beryllium disease.

    Science.gov (United States)

    Maier, Lisa A; Kittle, Lori A; Mroz, Margaret M; Newman, Lee S

    2003-06-01

    The diagnosis of chronic beryllium disease (CBD) relies on the beryllium lymphocyte proliferation test (BeLPT) to demonstrate a Be specific immune response. This test has improved early diagnosis, but cannot discriminate beryllium sensitization (BeS) from CBD. We previously found high neopterin levels in CBD patients' serum and questioned whether Be-stimulated neopterin production by peripheral blood cells in vitro might be useful in the diagnosis of CBD. CBD, BeS, Be exposed workers without disease (Be-exp) normal controls and sarcoidosis subjects were enrolled. Peripheral blood mononuclear cells (PBMN) were cultured in the presence and absence of beryllium sulfate. Neopterin levels were determined from cell supernatants by enzyme linked immunosorbent assay (ELISA). Clinical evaluation of CBD subjects included chest radiography, pulmonary function testing, exercise testing, and the BeLPT. CBD patients produced higher levels of neopterin in both unstimulated and Be-stimulated conditions compared to all other subjects (P workplace screening. Copyright 2003 Wiley-Liss, Inc.

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

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  4. Development of an integrated facility for processing transuranium solid wastes at the Savannah River Plant

    International Nuclear Information System (INIS)

    Boersma, M.D.; Hootman, H.E.; Permar, P.H.

    1978-01-01

    An integrated facility is being designed for processing solid wastes contaminated with long-lived alpha emitting (TRU) nuclides; this waste has been stored retrievably at the Savannah River Plant since 1965. The stored waste, having a volume of 10 4 m 3 and containing 3x10 5 Ci of transuranics, consists of both mixed combustible trash and failed and obsolete equipment primarily from transuranic production and associated laboratory operations. The facility for processing solid transuranic waste will consist of five processing modules: 1) unpackaging, sorting, and assaying; 2) treatment of combustibles by controlled air incineration; 3) size reduction of noncombustibles by plasma-arc cutting followed by decontamination by electropolishing; 4) fixation of the processed waste in cement; and 5) packaging for shipment to a federal repository. The facility is projected for construction in the mid-1980's. Pilot facilities, sized to manage currently generated wastes, will also demonstrate the key process steps of incineration of combustibles and size reduction/decontamination of noncombustibles; these facilities are projected for 1980-81. Development programs leading to these extensive new facilities are described

  5. Development of an integrated facility for processing TRU solid wastes at the Savannah River Plant

    International Nuclear Information System (INIS)

    Boersma, M.D.; Hootman, H.E.; Permar, P.H.

    1977-01-01

    An integrated facility is being designed for processing solid wastes contaminated with long-lived alpha emitting (TRU) nuclides; this waste has been stored retrievably at the Savannah River Plant since 1965. The stored waste, having a volume of 10 4 m 3 and containing 3 x 10 5 Ci of transuranics, consists of both mixed combustible trash and failed and obsolete equipment primarily from transuranic production and associated laboratory operations. The facility for processing solid transuranic waste will consist of five processing modules: (1) unpackaging, sorting, and assaying; (2) treatment of combustibles by controlled air incineration; (3) size reduction of noncombustibles by plasma-arc cutting followed by decontamination by electropolishing; (4) fixation of the processed waste in cement; and (5) packaging for shipment to a federal repository. The facility is projected for construction in the mid-1980's. Pilot facilities, sized to manage currently generated wastes, will also demonstrate the key process steps of incineration of combustibles and size reduction/decontamination of noncombustibles; these facilities are projected for 1980-81. Development programs leading to these extensive new facilities are described

  6. Implanted deuterium retention and release in carbon-coated beryllium

    International Nuclear Information System (INIS)

    Anderl, R.A.; Longhurst, G.R.; Pawelko, R.J.; Oates, M.A.

    1997-01-01

    Deuterium implantation experiments have been conducted on samples of clean and carbon-coated beryllium. These studies entailed preparation and characterization of beryllium samples coated with carbon thicknesses of 100, 500, and 1000 angstrom. Heat treatment of a beryllium sample coated with carbon to a thickness of approximately 100 angstrom revealed that exposure to a temperature of 400 degrees C under high vacuum conditions was sufficient to cause substantial diffusion of beryllium through the carbon layer, resulting in more beryllium than carbon at the surface. Comparable concentrations of carbon and beryllium were observed in the bulk of the coating layer. Higher than expected oxygen levels were observed throughout the coating layer as well. Samples were exposed to deuterium implantation followed by thermal desorption without exposure to air. Differences were observed in deuterium retention and postimplantation release behavior in the carbon-coated samples as compared with bare samples. For comparable implantation conditions (sample temperature of 400 degrees C and an incident deuterium flux of approximately 6 X 10 19 D/m 2 sec), the quantity of deuterium retained in the bare sample was less than that retained in the carbon-coated samples. Further, the release of the deuterium took place at lower temperatures for the bare beryllium surfaces than for carbon-coated beryllium samples. 4 refs., 8 figs., 1 tab

  7. Implanted Deuterium Retention and Release in Carbon-Coated Beryllium

    Science.gov (United States)

    Anderl, R. A.; Longhurst, G. R.; Pawelko, R. J.; Oates, M. A.

    1997-06-01

    Deuterium implantation experiments have been conducted on samples of clean and carbon-coated beryllium. These studies entailed preparation and characterization of beryllium samples coated with carbon thicknesses of 100, 500, and 1000 Å. Heat treatment of a beryllium sample coated with carbon to a thickness of approximately 100 Å revealed that exposure to a temperature of 400°C under high vacuum conditions was sufficient to cause substantial diffusion of beryllium through the carbon layer, resulting in more beryllium than carbon at the surface. Comparable concentrations of carbon and beryllium were observed in the bulk of the coating layer. Higher than expected oxygen levels were observed throughout the coating layer as well. Samples were exposed to deuterium implantation followed by thermal desorption without exposure to air. Differences were observed in deuterium retention and postimplantation release behavior in the carbon-coated samples as compared with bare samples. For comparable implantation conditions (sample temperature of 400°C and an incident deuterium flux of approximately 6 × 1019 D/m2-s), the quantity of deuterium retained in the bare sample was less than that retained in the carbon-coated samples. Further, the release of the deuterium took place at lower temperatures for the bare beryllium surfaces than for carbon-coated beryllium samples.

  8. 10 CFR 850.20 - Baseline beryllium inventory.

    Science.gov (United States)

    2010-01-01

    ... 10 Energy 4 2010-01-01 2010-01-01 false Baseline beryllium inventory. 850.20 Section 850.20 Energy... Baseline beryllium inventory. (a) The responsible employer must develop a baseline inventory of the... inventory, the responsible employer must: (1) Review current and historical records; (2) Interview workers...

  9. Atlas of hot isostatic beryllium powder pressing diagrams

    International Nuclear Information System (INIS)

    Stoev, P.I.; Papirov, I.I.; Tikhinskij, G.F.; Vasil'ev, A.A.

    1995-01-01

    Diagrams of hot isotopic pressing (HIP) of beryllium powder with different grain size in a wide range of pressing parameters are built by mathematical modeling methods. The HIP diagrams presented are divided into 3 groups: parametric dependencies D=f(P,T); technological HIP diagrams; compacting mechanisms. The created data bank permits to optimise beryllium powder HIP with changing parameters. 4 refs., 23 figs

  10. Reactivity effects due to beryllium poisoning of BR2

    International Nuclear Information System (INIS)

    Kalcheva, S.; Ponsard, B.; Koonen, E.

    2004-01-01

    This paper illustrates the impact of the poisoning of the beryllium reflector on reactivity variations of the Belgian MTR BR2 in SCK.CEN. Detailed calculations by MCNP-4C of reactivity effects caused by strong neutron absorbers 3 He and 6 Li during reactor operation history are presented. The importance of beryllium poisoning for the accuracy of reactivity predictions is discussed. (authors)

  11. Progress report of preliminary studies of beryllium toxicity

    Energy Technology Data Exchange (ETDEWEB)

    Hodge, H.C.

    1947-09-01

    This document was prepared in connection with a symposium of beryllium poisoning held at the Saranac Laboratories and describes progress made and a research program aimed at characterizing the toxicity of beryllium. Seven individual papers in this document are separately indexed and cataloged for the database.

  12. Joining of beryllium by braze welding technique: preliminary results

    Energy Technology Data Exchange (ETDEWEB)

    Banaim, P.; Abramov, E. [Ben-Gurion Univ. of the Negev, Beersheba (Israel); Zalkind, S.; Eden, S.

    1998-01-01

    Within the framework of some applications, there is a need to join beryllium parts to each other. Gas Tungsten Arc Braze Welds were made in beryllium using 0.3 mm commercially Aluminum (1100) shim preplaced at the joint. The welds exhibited a tendency to form microcracks in the Fusion Zone and Heat Affected Zone. All the microcracks were backfilled with Aluminum. (author)

  13. 324 Facility B-Cell quality process plan

    International Nuclear Information System (INIS)

    Carlson, J.L.

    1998-01-01

    B-Cell is currently being cleaned out (i.e., removal of equipment, fixtures and residual radioactive materials) and deactivated. TPA Milestone M-89-02 dictates that all mixed waste and equipment be removed from B-Cell by 5/31/99. The following sections describe the major activities that remain for completion of the TPA milestone. This includes: (1) Size Reduce Tank 119 and Miscellaneous Equipment. This activity is the restart of hotwork in B-Cell to size reduce the remainder of Tank 119 and other miscellaneous pieces of equipment into sizes that can be loaded into a grout container. This activity also includes the process of preparing the containers for shipment from the cell. The specific activities and procedures used are detailed in a table. (2) Load and Ship Low-Level Waste. This activity covers the process of taking a grouted LLW container from B-Cell and loading it into the cask in the REC airlock and Cask Handling Area (CHA) for shipment to the LLBG. The detailed activities and procedures for this part of cell cleanout are included in second table

  14. Development of Advanced Multizone Facilities for Microgravity Processing

    Science.gov (United States)

    1998-01-01

    NASA has been interested in experimental ground based study to investigate the fundamental processes involved in phase transformation processes during growth of metallic, nonmetallic and electronic materials. Solidification, vapor growth and solution growth techniques of growing crystals are of special interest because of the inherent importance of convection in the nutrient solution. Convection enhances the mass transport through the nutrient and results in faster growth rates. Availability of low gravity environment of space has provided scientists a new variable to control the extent of convection and thus isolate the diffusive phenomena for their better understanding. The thermal gradient at the liquid-solid interface is determined by the alloy characteristics, the hot zone temperature, cold zone temperature and the width of the insulating zone. The thermal profiles get established by the existing material and geometrical constraints of the experimental set up. The major effort under this research was devoted to designing a programmable furnace which can be used to obtain thermal profiles along the length of the sample as per the demands of the scientists. The furnace did not have active cooling of the zones. Only active heating and passive cooling were utilized.

  15. The LEU target development and conversion program for the MAPLE reactors and new processing facility

    International Nuclear Information System (INIS)

    Malkoske, G.R.

    2003-01-01

    The availability of isotope grade, Highly Enriched Uranium (HEU), from the United States for use in the manufacture of targets for molybdenum-99 production in AECL's NRU research reactor has been a key factor to enable MDS Nordion to develop a reliable, secure supply of medical isotopes for the international nuclear medicine community. The molybdenum extraction process from HEU targets is a proven and established method that has reliably produced medical isotopes for several decades. The HEU process provides predictable, consistent yields for our high-volume, molybdenum-99 production. Other medical isotopes such as I-131 and Xe-133, which play an important role in nuclear medicine applications, are also produced from irradiated HEU targets as a by-product of the molybdenum-99 process. To ensure a continued reliable and timely supply of medical isotopes, MDS Nordion is completing the commissioning of two MAPLE reactors and an associated isotope processing facility (the New Processing Facility). The new MAPLE facilities, which will be dedicated exclusively to medical isotope production, will provide an essential contribution to a secure, robust global healthcare system. Design and construction of these facilities has been based on a life cycle management philosophy for the isotope production process. This includes target irradiation, isotope extraction and waste management. The MAPLE reactors will operate with Low Enriched Uranium (LEU) fuel, a significant contribution to the objectives of the RERTR program. The design of the isotope production process in the MAPLE facilities is based on an established process - extraction of isotopes from HEU target material. This is a proven technology that has been demonstrated over more than three decades of operation. However, in support of the RERTR program and in compliance with U.S. legislation, MDS Nordion has undertaken a LEU Target Development and Conversion Program for the MAPLE facilities. This paper will provide an

  16. Characterization of transuranic solid wastes from a plutonium processing facility

    International Nuclear Information System (INIS)

    Mulkin, R.

    1975-06-01

    Transuranic-contaminated wastes generated in the processing areas of the Plutonium Chemistry and Metallurgy Group at the Los Alamos Scientific Laboratory (LASL) were studied in detail to identify their chemical and physical composition. Nondestructive Assay (NDA) equipment was developed to measure transuranic activity at the 10-nCi/g level in low-density residues typically found in room-generated waste. This information will supply the Waste Management Program with a more positive means of identifying concerns in waste storage and the challenge of optimizing the system of waste form, packaging, and environment of the storage area for 20-yr retrievable waste. A positive method of measuring transuranic activity in waste at the 10-nCi/g level will eliminate the need for administrative control in a sensitive area, and will provide the economic advantage of minimizing the volume of waste stored as retrievable waste. (U.S.)

  17. Fluid simulation of beryllium transport in the ITER gaseous divertor

    International Nuclear Information System (INIS)

    Knoll, D.A.; Campbell, R.B.; McHugh, P.R.

    1994-01-01

    The transport of either intrinsic or injected impurities will play a crucial role in the energy loss mechanisms in the ITER gaseous/cold plasma target divertor. Both 1-D and 2-D multi-charge state fluid codes are used to model the transport of beryllium in the ITER SOL. Our major conclusion is that in order to model the containment of impurities, the background flow field must be known in detail. Comparing 1-D and 2-D solutions, hydrogen flow reversal plays an important role in the entrainment process. Further, the flow of particles from the core plasma also has a strong impact on the resultant entrainment of the impurities in both 1-D and 2-D. It is imperative that those components of poloidal velocity due to E x B and diamagnetic drifts be included in the models. (orig.)

  18. Development of beryllium bonds for plasma-facing components

    International Nuclear Information System (INIS)

    Franconi, E.; Ceccotti, G.C.; Magnoli, L.

    1992-01-01

    This study concerns the techniques of bonding beryllium to both structural material (AISI 316 SS) and heat sink material (copper and DS-copper) plates, and the characterization of the bonding material obtained. Conventional bonding techniques for joining Be to SS and copper using brazing alloys were first investigated. The best result was obtained using a silver-copper eutetic alloy as a brazing alloy. However, the high-temperature capability of the materials prepared by this method is limited by the performance of brazing alloys at the operating temperature. To avoid this problem, we are developing a joining process known as solid-state reaction bonding that improves the capability at the operating temperature. (orig.)

  19. Modelling of radiation impact on ITER Beryllium wall

    Science.gov (United States)

    Landman, I. S.; Janeschitz, G.

    2009-04-01

    In the ITER H-Mode confinement regime, edge localized instabilities (ELMs) will perturb the discharge. Plasma lost after each ELM moves along magnetic field lines and impacts on divertor armour, causing plasma contamination by back propagating eroded carbon or tungsten. These impurities produce enhanced radiation flux distributed mainly over the beryllium main chamber wall. The simulation of the complicated processes involved are subject of the integrated tokamak code TOKES that is currently under development. This work describes the new TOKES model for radiation transport through confined plasma. Equations for level populations of the multi-fluid plasma species and the propagation of different kinds of radiation (resonance, recombination and bremsstrahlung photons) are implemented. First simulation results without account of resonance lines are presented.

  20. Modelling of radiation impact on ITER Beryllium wall

    Energy Technology Data Exchange (ETDEWEB)

    Landman, I.S. [Forschungszentrum Karlsruhe, IHM, FUSION, P.O. Box 3640, 76021 Karlsruhe (Germany)], E-mail: igor.landman@ihm.fzk.de; Janeschitz, G. [Forschungszentrum Karlsruhe, IHM, FUSION, P.O. Box 3640, 76021 Karlsruhe (Germany)

    2009-04-30

    In the ITER H-Mode confinement regime, edge localized instabilities (ELMs) will perturb the discharge. Plasma lost after each ELM moves along magnetic field lines and impacts on divertor armour, causing plasma contamination by back propagating eroded carbon or tungsten. These impurities produce enhanced radiation flux distributed mainly over the beryllium main chamber wall. The simulation of the complicated processes involved are subject of the integrated tokamak code TOKES that is currently under development. This work describes the new TOKES model for radiation transport through confined plasma. Equations for level populations of the multi-fluid plasma species and the propagation of different kinds of radiation (resonance, recombination and bremsstrahlung photons) are implemented. First simulation results without account of resonance lines are presented.