WorldWideScience

Sample records for part iii fuels

  1. 40 CFR Appendix III to Part 600 - Sample Fuel Economy Label Calculation

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 29 2010-07-01 2010-07-01 false Sample Fuel Economy Label Calculation...) ENERGY POLICY FUEL ECONOMY AND CARBON-RELATED EXHAUST EMISSIONS OF MOTOR VEHICLES Pt. 600, App. III Appendix III to Part 600—Sample Fuel Economy Label Calculation Suppose that a manufacturer called Mizer...

  2. Market Analysis and Consumer Impacts Source Document. Part III. Consumer Behavior and Attitudes Toward Fuel Efficient Vehicles

    Science.gov (United States)

    1980-12-01

    This source document on motor vehicle market analysis and consumer impacts consists of three parts. Part III consists of studies and reviews on: consumer awareness of fuel efficiency issues; consumer acceptance of fuel efficient vehicles; car size ch...

  3. Performance limits of coated particle fuel. Part III. Fission product migration in HTR fuel

    Energy Technology Data Exchange (ETDEWEB)

    Nabielek, H.; Hick, H.; Wagner-Loffler, M.; Voice, E. H.

    1974-06-15

    A general introduction and literature survey to the physics and mathematics of fission product migration in HTR fuel is given as well as a review of available experimental results and their evaluation in terms of models and materials data.

  4. Workshop 96. Part III

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-12-31

    Part III of the proceedings contain 155 contributions in various fields of science and technology including nuclear engineering, environmental science, and biomedical engineering. Out of these, 10 were selected to be inputted in INIS. (P.A.).

  5. Workshop 96. Part III

    International Nuclear Information System (INIS)

    1995-12-01

    Part III of the proceedings contain 155 contributions in various fields of science and technology including nuclear engineering, environmental science, and biomedical engineering. Out of these, 10 were selected to be inputted in INIS. (P.A.)

  6. Thorium utilisation in a small long-life HTR. Part III: Composite-rod fuel blocks

    Energy Technology Data Exchange (ETDEWEB)

    Verrue, Jacques, E-mail: jacques.verrue@polytechnique.org [Delft University of Technology, Reactor Institute Delft, Mekelweg 15, 2629 JB Delft (Netherlands); École Polytechnique (Member of ParisTech), 91128 Palaiseau Cedex (France); Ding, Ming, E-mail: dingm2005@gmail.com [Delft University of Technology, Reactor Institute Delft, Mekelweg 15, 2629 JB Delft (Netherlands); Harbin Engineering University, Nantong Street 145, 150001 Harbin (China); Kloosterman, Jan Leen, E-mail: j.l.kloosterman@tudelft.nl [Delft University of Technology, Reactor Institute Delft, Mekelweg 15, 2629 JB Delft (Netherlands)

    2014-02-15

    Highlights: • Composite-rod fuel blocks are proposed for a small block-type HTR. • An axial separation of fuel compacts is the most important feature. • Three patterns are presented to analyse the effects of the spatial distribution. • The spatial distribution has a large influence on the neutron spectrum. • Composite-rod fuel blocks reach a reactivity swing less than 4%. - Abstract: The U-Battery is a small long-life high temperature gas-cooled reactor (HTR) with power of 20 MWth. In order to increase its lifetime and diminish its reactivity swing, the concept of composite-rod fuel blocks with uranium and thorium was investigated. Composite-rod fuel blocks feature a specific axial separation between UO{sub 2} and ThO{sub 2} compacts in fuel rods. The design parameters, investigated by SCALE 6, include the number and spatial distribution of fuel compacts within the rods, the enrichment of uranium, the radii of fuel kernels and fuel compacts, and the packing fractions of uranium and thorium TRISO particles. The analysis shows that a lower moderation ratio and a larger inventory of heavy metals results in a lower reactivity swing. The optimal atomic carbon-to-heavy metal ratio depends on the mass fraction of U-235 and is commonly in the 160–200 range. The spatial distribution of the fuel compacts within the fuel rods has a large influence on the energy spectrum in each fuel compact and thus on the beginning-of-life reactivity and the reactivity swing. At end-of-life, the differences caused by the spatial distribution of the fuel compacts are smaller due to the fissions of U-233 in the ThO{sub 2} fuel compacts. This phenomenon enables to design fuel blocks with a very low reactivity swing, down to less than 4% in a 10-year lifetime. Among three types of thorium fuelled U-Battery blocks, the composite-rod fuel block achieves the highest end-of-life reactivity and the lowest reactivity swing.

  7. A novel concept of QUADRISO particles Part III: applications to the plutonium-thorium fuel cycle

    International Nuclear Information System (INIS)

    Talamo, A.

    2009-01-01

    In the present study, a plutonium-thorium fuel cycle is investigated including the 233 U production and utilization. A prismatic thermal High Temperature Gas Reactor (HTGR) and the novel concept of quadruple isotropic (QUADRISO) coated particles, designed at the Argonne National Laboratory, have been used for the study. In absorbing QUADRISO particles, a burnable poison layer surrounds the central fuel kernel to flatten the reactivity curve as a function of time. At the beginning of life, the fuel in the QUADRISO particles is hidden from neutrons, since they get absorbed in the burnable poison before they reach the fuel kernel. Only when the burnable poison depletes, neutrons start streaming into the fuel kernel inducing fission reactions and compensating the fuel depletion of ordinary TRISO particles. In fertile QUADRISO particles, the absorber layer is replaced by natural thorium with the purpose of flattening the excess of reactivity by the thorium resonances and producing 233 U. The above configuration has been compared with a configuration where fissile (neptunium-plutonium oxide from Light Water Reactors irradiated fuel) and fertile (natural thorium oxide) fuels are homogeneously mixed in the kernel of ordinary TRISO particles. For the 233 U utilization, the core has been equipped with europium oxide absorbing QUADRISO particles.

  8. Regulations for safe transport of spent fuels from nuclear power plants in CMEA member countries. Part III

    International Nuclear Information System (INIS)

    Zizka, B.

    1978-11-01

    The regulations for safe transport of spent fuel from nuclear power plants in the CMEA member countries consist of general provisions, technical requirements for spent fuel transport, transport conditions, procedures for submitting reports on transport, regulations for transport and protection of radioactive material to be transported, procedures for customs clearance, technical and organizational measures for the prevention of hypothetical accidents and the elimination of their consequences. The bodies responsible for spent fuel transport in the CMEA member countries are listed. (J.B.)

  9. Development of metal fuel and study of construction materials (I-IV), Part III; Razvoj metalnog goriva i ispitivanje konstrukcionih materijala (I-VI deo); III deo

    Energy Technology Data Exchange (ETDEWEB)

    Mihajlovic, A [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1965-11-15

    This volume includes the following reports: radiation damage of metal uranium; influence of burnup rate on the stability of metal uranium fuel, influence of precipitation and desorption of inert gas on the density change of samples.

  10. Fuel cells (part 2)

    International Nuclear Information System (INIS)

    Campanari, S.; Macchi, E.

    1999-01-01

    The article, following and completing the issues dealt with in part 1 (CH4 Energia Metano, 1/99, p. 7), describe the operating characteristic and construction features of molten carbonate and solid oxide fuel cells (MCFC and SOFC). For the latter type, construction cost are evaluated by various authors and research institutes. The article ends by presenting some tables showing the classification and the main characteristics of various fuel cells, and well as the effect of some gases on the behaviour of some of them [it

  11. Standards in neurosonology. Part III

    Directory of Open Access Journals (Sweden)

    Joanna Wojczal

    2016-06-01

    Full Text Available The paper presents standards related to ultrasound imaging of the cerebral vasculature and structures. The aim of this paper is to standardize both the performance and description of ultrasound imaging of the extracranial and intracranial cerebral arteries as well as a study of a specific brain structure, i.e. substantia nigra hyperechogenicity. The following aspects are included in the description of standards for each ultrasonographic method: equipment requirements, patient preparation, study technique and documentation as well as the required elements of ultrasound description. Practical criteria for the diagnosis of certain pathologies in accordance with the latest literature were also presented. Furthermore, additional comments were included in some of the sections. Part I discusses standards for the performance, documentation and description of different ultrasound methods (Duplex, Doppler. Part II and III are devoted to standards for specific clinical situations (vasospasm, monitoring after the acute stage of stroke, detection of a right-to-left shunts, confirmation of the arrest of the cerebral circulation, an assessment of the functional efficiency of circle of Willis, an assessment of the cerebrovascular vasomotor reserve as well as the measurement of substantia nigra hyperechogenicity.

  12. Presentation of safety after closure of the repository for spent nuclear fuel. Main report of the project SR-Site. Part III

    International Nuclear Information System (INIS)

    2011-01-01

    The purpose of the safety assessment SR-Site is to investigate whether a safe repository for spent nuclear fuel by KBS-3 type can be constructed at Forsmark in Oesthammar in Sweden. The location of the Forsmark has been selected based on results of several surveys from surface conditions at depth in Forsmark and in Laxemar in Oskarshamn. The choice of location is not justified in SR-Site Report, but in other attachments to SKB's permit applications. SR-Site Report is an important part of SKB's permit applications to construct and operate a repository for spent nuclear fuel at Forsmark in Oesthammar. The purpose of the report in the applications is to show that a repository at Forsmark is safe after closure

  13. Fuel cycle math - part one

    International Nuclear Information System (INIS)

    Anon.

    1992-01-01

    This article is Part One of a two-part article that reviews some of the numbers associated with the nuclear fuel cycle. The contents of Part One include: composition of the element uranium, considering atomic mass and weight-percent of the isotopes; uranium in the ground, including ore grades; mining, with dilution factors and recovery rates; ore sorting, including concentration factors; and uranium recovery. No financial information is presented in either Part One or Part Two

  14. Drilling miniature holes, Part III

    Energy Technology Data Exchange (ETDEWEB)

    Gillespie, L.K.

    1978-07-01

    Miniature components for precision electromechanical mechanisms such as switches, timers, and actuators typically require a number of small holes. Because of the precision required, the workpiece materials, and the geometry of the parts, most of these holes must be produced by conventional drilling techniques. The use of such techniques is tedious and often requires considerable trial and error to prevent drill breakage, minimize hole mislocation and variations in hole diameter. This study of eight commercial drill designs revealed that printed circuit board drills produced better locational and size repeatability than did other drills when centerdrilling was not used. Boring holes 1 mm in dia, or less, as a general rule did not improve hole location in brass or stainless steel. Hole locations of patterns of 0.66-mm holes can be maintained within 25.4-..mu..m diametral positional tolerance if setup misalignments can be eliminated. Size tolerances of +- 3.8 ..mu..m can be maintained under some conditions when drilling flat plates. While these levels of precision are possible with existing off-the-shelf drills, they may not be practical in many cases.

  15. AUTOMOTIVE DIESEL MAINTENANCE 1. UNIT XIII, I--MAINTAINING THE FUEL SYSTEM (PART III), CUMMINS DIESEL ENGINES, II--RADIATOR SHUTTER SYSTEM.

    Science.gov (United States)

    Human Engineering Inst., Cleveland, OH.

    THIS MODULE OF A 30-MODULE COURSE IS DESIGNED TO DEVELOP AN UNDERSTANDING OF THE CONSTRUCTION, OPERATION, AND MAINTENANCE OF THE DIESEL ENGINE FUEL AND RADIATOR SHUTTER SYSTEMS. TOPICS ARE (1) MORE ABOUT THE CUMMINS FUEL SYSTEM, (2) CALIBRATING THE PT FUEL PUMP, (3) CALIBRATING THE FUEL INJECTORS, (4) UNDERSTANDING THE SHUTTER SYSTEM, (5) THE…

  16. Neuroscience in Nazi Europe Part III

    DEFF Research Database (Denmark)

    Zeidman, Lawrence A; Kondziella, Daniel

    2012-01-01

    In Part I, neuroscience collaborators with the Nazis were discussed, and in Part II, neuroscience resistors were discussed. In Part III, we discuss the tragedy regarding european neuroscientists who became victims of the Nazi onslaught on “non-Aryan” doctors. Some of these unfortunate...... of neuroscience, we pay homage and do not allow humanity to forget, lest this dark period in history ever repeat itself....

  17. Fuel cycle math - part two

    International Nuclear Information System (INIS)

    Anon.

    1992-01-01

    This article is Part 2 of a two part series on simple mathematics associated with the nuclear fuel cycle. While not addressing any of the financial aspects of the fuel cycle, this article does discuss the following: conversion between English and metric systems; uranium content expressed in equivalent forms, such as U3O8, and the method of determining these equivalencies; the uranium conversion process, considering different input and output compounds; and the enrichment process, including feed, tails, and product assays, as well as SWU and feed requirements

  18. Presentation of safety after closure of the repository for spent nuclear fuel. Main report of the project SR-Site. Part III; Redovisning av saekerhet efter foerslutning av slutfoervaret foer anvaent kaernbraensle. Huvudrapport fraan projekt SR-Site. Del III

    Energy Technology Data Exchange (ETDEWEB)

    2011-07-01

    The purpose of the safety assessment SR-Site is to investigate whether a safe repository for spent nuclear fuel by KBS-3 type can be constructed at Forsmark in Oesthammar in Sweden. The location of the Forsmark has been selected based on results of several surveys from surface conditions at depth in Forsmark and in Laxemar in Oskarshamn. The choice of location is not justified in SR-Site Report, but in other attachments to SKB's permit applications. SR-Site Report is an important part of SKB's permit applications to construct and operate a repository for spent nuclear fuel at Forsmark in Oesthammar. The purpose of the report in the applications is to show that a repository at Forsmark is safe after closure

  19. AUTOMOTIVE DIESEL MAINTENANCE 1. UNIT XXIV, I--MAINTAINING THE FUEL SYSTEM PART III--CATERPILLAR DIESEL ENGINE, II--UNDERSTANDING THE VOLTAGE REGULATOR/ALTERNATOR.

    Science.gov (United States)

    Minnesota State Dept. of Education, St. Paul. Div. of Vocational and Technical Education.

    THIS MODULE OF A 30-MODULE COURSE IS DESIGNED TO DEVELOP AN UNDERSTANDING OF THE OPERATION AND MAINTENANCE OF THE DIESEL ENGINE FUEL AND BATTERY CHARGING SYSTEM. TOPICS ARE (1) INJECTION TIMING CONTROLS, (2) GOVERNOR, (3) FUEL SYSTEM MAINTENANCE TIPS, (4) THE CHARGING SYSTEM, (5) REGULATING THE GENERATOR/ALTERNATOR, AND (6) CHARGING SYSTEM SERVICE…

  20. Science based integrated approach to advanced nuclear fuel development - integrated multi-scale multi-physics hierarchical modeling and simulation framework Part III: cladding

    International Nuclear Information System (INIS)

    Tome, Carlos N.; Caro, J.A.; Lebensohn, R.A.; Unal, Cetin; Arsenlis, A.; Marian, J.; Pasamehmetoglu, K.

    2010-01-01

    Advancing the performance of Light Water Reactors, Advanced Nuclear Fuel Cycles, and Advanced Reactors, such as the Next Generation Nuclear Power Plants, requires enhancing our fundamental understanding of fuel and materials behavior under irradiation. The capability to accurately model the nuclear fuel systems to develop predictive tools is critical. Not only are fabrication and performance models needed to understand specific aspects of the nuclear fuel, fully coupled fuel simulation codes are required to achieve licensing of specific nuclear fuel designs for operation. The backbone of these codes, models, and simulations is a fundamental understanding and predictive capability for simulating the phase and microstructural behavior of the nuclear fuel system materials and matrices. In this paper we review the current status of the advanced modeling and simulation of nuclear reactor cladding, with emphasis on what is available and what is to be developed in each scale of the project, how we propose to pass information from one scale to the next, and what experimental information is required for benchmarking and advancing the modeling at each scale level.

  1. Comparison with experiment of COMETHE III-L fuel rod behaviour predictions

    International Nuclear Information System (INIS)

    Vliet, J. van; Billaux, M.

    1983-01-01

    A comparison is presented between experimental results and COMETHE III-L fuel rod behaviour predictions. The first part of the paper focuses on mechanical aspects, with as main experiments, AECL X-264 and Studsvik Interramp. The second part presents the results of a wide FGR benchmarking campaign, with a reference to previous COMETHE versions. It appears that the variance between experiment and calculation has decreased by a factor four when the III-J version was improved into the III-L version. As conclusion, some COMETHE III-L calculations are presented in order to illustrate its capability of predicting fuel rod performance limits. (author)

  2. Fuel Behaviour Simulations in Fumex III CRP at NRI

    International Nuclear Information System (INIS)

    Valach, M.; Klouzal, J.; Dostal, M.; Zymak, J.

    2013-01-01

    NRI Rez plc took part in the previous coordinated research projects focused on fuel behaviour modelling held by the IAEA - FUMEX-I and FUMEX-II. These were very helpful for the development and validation of various codes used in the Nuclear Research Institute Rez (NRI) for the evaluation of the fuel rod thermomechanical behaviour. Based on the considerations of our needs related to the modeling for Czech NPPs we have performed basic parametric calculations of two LOCA cases (IFA-650.1 and IFA-650.2) and detailed evaluation WWER related cases Kola MIR ramp rods. The AREVA ''Idealized case'' and 16x16 LTA cases were also calculated because of the high burnup reached. Report summarises simulated cases in the frame of FUMEX III Project at the NRI Rez plc. (author)

  3. Eleventh annual meeting, Bologna, Italy, 17-20 April 1978. Summary report. Part III

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1978-09-01

    The Summary Report - Part III of the Eleventh Annual Meeting of the IAEA International Working Group on Fast Reactors - contains the discussions on the commercialization LMFBRs according to national plans, mostly related to technology of fuel fabrication, PHENIX fuel pins testing, heterogeneous cores, in service inspection of fuel elements, regulations and licensing, and related OECD activities. Most of the discussions were related to the existing reactors: BR-10, BN-600, BN-350, BN-1600, RAPSODIE and PHENIX.

  4. Eleventh annual meeting, Bologna, Italy, 17-20 April 1978. Summary report. Part III

    International Nuclear Information System (INIS)

    1978-09-01

    The Summary Report - Part III of the Eleventh Annual Meeting of the IAEA International Working Group on Fast Reactors - contains the discussions on the commercialization LMFBRs according to national plans, mostly related to technology of fuel fabrication, PHENIX fuel pins testing, heterogeneous cores, in service inspection of fuel elements, regulations and licensing, and related OECD activities. Most of the discussions were related to the existing reactors: BR-10, BN-600, BN-350, BN-1600, RAPSODIE and PHENIX

  5. Tenth annual meeting, Vienna, Austria, 29 March - 1 April 1977. Summary report. Part III

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1977-11-01

    The Summary Report - Part III of the Tenth Annual Meeting of the IAEA International Working Group on Fast Reactors - contains the discussions on the commercial development of FBRs according to national plans, mostly related to technology problems of containment design, fuel fabrication, fuel failures, sodium pressure, fuel-sodium interaction, computer codes needed for licensing. Most of the discussions were related to the existing reactors: BN-600, BN-350, BN-1600, BOR-60, RAPSODIE, PHENIX.

  6. Tenth annual meeting, Vienna, Austria, 29 March - 1 April 1977. Summary report. Part III

    International Nuclear Information System (INIS)

    1977-11-01

    The Summary Report - Part III of the Tenth Annual Meeting of the IAEA International Working Group on Fast Reactors - contains the discussions on the commercial development of FBRs according to national plans, mostly related to technology problems of containment design, fuel fabrication, fuel failures, sodium pressure, fuel-sodium interaction, computer codes needed for licensing. Most of the discussions were related to the existing reactors: BN-600, BN-350, BN-1600, BOR-60, RAPSODIE, PHENIX

  7. Part 5. Fuel cycle options

    International Nuclear Information System (INIS)

    Lineberry, M.J.; McFarlane, H.F.; Amundson, P.I.; Goin, R.W.; Webster, D.S.

    1980-01-01

    The results of the FBR fuel cycle study that supported US contributions to the INFCE are presented. Fuel cycle technology is reviewed from both generic and historical standpoints. Technology requirements are developed within the framework of three deployment scenarios: the reference international, the secured area, and the integral cycle. Reprocessing, fabrication, waste handling, transportation, and safeguards are discussed for each deployment scenario. Fuel cycle modifications designed to increase proliferation defenses are described and assessed for effectiveness and technology feasibility. The present status of fuel cycle technology is reviewed and key issues that require resolution are identified

  8. Barnwell Nuclear Fuels Plant applicability study. Volume III. Appendices

    International Nuclear Information System (INIS)

    1978-03-01

    Volume III suppliees supporting information to assist Congress in making a decision on the optimum utilization of the Barnwell Nuclear Fuels Plant. Included are applicable fuel cycle policies; properties of reference fuels; description and evaluation of alternative operational (flue cycle) modes; description and evaluation of safeguards systems and techniques; description and evaluation of spiking technology; waste and waste solidification evaluation; and Department of Energy programs relating to nonproliferation

  9. Introduction to Part III: Application of LCA in Practice

    DEFF Research Database (Denmark)

    Rosenbaum, Ralph K.

    2018-01-01

    While Part II of this book presents the theoretical foundation and methodology of LCA, Part III is dedicated to a comprehensive discussion of how this methodology has been adapted and applied in practice. The chapters of Part III provide an easily readable and accessible introduction to different...

  10. COMETHE III-M for transient fuel rod behaviour prediction

    International Nuclear Information System (INIS)

    Billaux, M.; Vliet, J. van

    1983-01-01

    The COMETHE III-M version is being developed in order to provide fuel rod behaviour prediction capability both in steady-state and in transient situations. It also allows to estimate the fuel rod enthalpy evolution versus time or burnup which may be important in core-related safety studies. This paper describes the transient heat transfer models, including transient heat conduction inside the fuel rod, and a subchannel model providing transient flow as well as enthalpy calculation capability. Transient fission gas release is also modelled on basis of the change rate of oxide temperature. The models are illustrated by a few calculation examples. (author)

  11. Cubby : Multiscreen Desktop VR Part III

    NARCIS (Netherlands)

    Djajadiningrat, J.P.; Gribnau, M.W.

    2000-01-01

    In this month's final episode of our 'Cubby: Multiscreen Desktop VR' trilogy we explain how you read the InputSprocket driver from part II, how you use it as input for the cameras from part I and how you calibrate the input device so that it leads to the correct head position.

  12. Warship Radar Signatures (Ship Survivability Part III-A)

    NARCIS (Netherlands)

    Galle, L.F.; Heemskerk, H.J.M.; Ewijk, L.J. van

    2000-01-01

    Radar Cross Section (RCS) management is of paramount importance for a warships's survivability. In this first part of the paper (Part III-A), the operational benefits of low RCS will be explained. Basic RCS theory, measurement and simulation techniques will be addressed. The RCS of representative

  13. SIMMER-III applications to fuel-coolant interactions

    Energy Technology Data Exchange (ETDEWEB)

    Morita, K.; Kondo, Sa.; Tobita, Y.; Brear, D.J. [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1998-01-01

    The main purpose of the SIMMER-III code is to provide a numerical simulation of complex multiphase, multicomponent flow problems essential to investigate core disruptive accidents in liquid-metal fast reactors (LMFRs). However, the code is designed to be sufficiently flexible to be applied to a variety of multiphase flows, in addition to LMFR safety issues. In the present study, some typical experiments relating to fuel-coolant interactions (FCIs) have been analyzed by SIMMER-III to demonstrate that the code is applicable to such complex and highly transient multiphase flow situations. It is shown that SIMMER-III can reproduce the premixing phase both in water and sodium systems as well as the propagation of steam explosion. It is thus demonstrated the code is basically capable of simulating integral multiphase thermal-hydraulic problems included in FCI experiments. (author)

  14. Theoretical analysis of nuclear reactors (Phase III), I-V, Part IV, Influence of isotopic composition of nuclear fuel on the reactivity with constant flux; Razrada metoda teorijske analize nuklearnih reaktora (III faza) I-IV, IV Deo, Uticaj promene izotopnog sastava goriva na reaktivnost uz konstantan fluks

    Energy Technology Data Exchange (ETDEWEB)

    Pop-Jordanov, J [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1963-01-15

    Part one of this report presents a series of differential equations describing the nuclear fuel depletion during reactor operation. This series of differential equations is extended to describe the fission products. This part includes equations for effective multiplication factor k{sub eff} and reactivity {rho} as a function of irradiation {tau}. Part two includes results obtained on the analog computer PACE 231 R, and related to Calder Hall type reactor. Part three covers detailed preparation of the series of equations for solution by using the analog computer. Part four includes the list of references related to this task.

  15. Some Aspects of Facial Nerve Paralysis. Part III. Complications ...

    African Journals Online (AJOL)

    Some Aspects of Facial Nerve Paralysis. Part III. Complications, Prognosis and management. ... It should be possible to set a definite prognosis within 2 weeks after the onset of facial paralysis, and in many cases even sooner. In the prognosis of facial paralysis the aetiological and time factors involved, the completeness of ...

  16. 10 CFR Appendix II to Part 504 - Fuel Price Computation

    Science.gov (United States)

    2010-01-01

    ... DEPARTMENT OF ENERGY (CONTINUED) ALTERNATE FUELS EXISTING POWERPLANTS Pt. 504, App. II Appendix II to Part... effects of future real price increases for each fuel. The delivered price of an alternate fuel used to calculate delivered fuel expenses must reflect the petitioner's delivered price of the alternate fuel and...

  17. Study on the thermal-hydraulic stability of high burn up STEP III fuel in Japan

    International Nuclear Information System (INIS)

    Ishikawa, M.; Kitamura, H.; Toba, A.; Omoto, A.

    2004-01-01

    Japanese BWR utilities have performed a joint study of the Thermal Hydraulic Stability of High Burn up STEP III Fuel. In this study, the parametric dependency of thermal hydraulic stability threshold was obtained. It was confirmed through experiments that the STEP III Fuel has sufficient stability characteristics. (author)

  18. Solar neutrino oscillation parameters after SNO Phase-III and SAGE Part-III

    International Nuclear Information System (INIS)

    Yang Ping; Liu Qiuyu

    2009-01-01

    We analyse the recently published results from solar neutrino experiments SNO Phase-III and SAGE Part-III and show their constraints on solar neutrino oscillation parameters, especially for the mixing angle θ 12 . Through a global analysis using all existing data from SK, SNO, Ga and Cl radiochemical experiments and long base line reactor experiment KamLAND , we obtain the parameters Δm 12 2 =7.684 -0.208 +0.212 x 10 -5 eV 2 , tan 2 θ 12 =0.440 -0.057 +0.059 . We also find that the discrepancy between the KamLAND and solar neutrino results can be reduced by choosing a small non-zero value for the mixing angle θ 13 . (authors)

  19. Hydrogen Fuel Cells: Part of the Solution

    Science.gov (United States)

    Busby, Joe R.; Altork, Linh Nguyen

    2010-01-01

    With the decreasing availability of oil and the perpetual dependence on foreign-controlled resources, many people around the world are beginning to insist on alternative fuel sources. Hydrogen fuel cell technology is one answer to this demand. Although modern fuel cell technology has existed for over a century, the technology is only now becoming…

  20. 12 CFR Appendix III to Part 27 - Fair Housing Lending Inquiry/Application Log Sheet

    Science.gov (United States)

    2010-01-01

    ... 12 Banks and Banking 1 2010-01-01 2010-01-01 false Fair Housing Lending Inquiry/Application Log Sheet III Appendix III to Part 27 Banks and Banking COMPTROLLER OF THE CURRENCY, DEPARTMENT OF THE TREASURY FAIR HOUSING HOME LOAN DATA SYSTEM Pt. 27, App. III Appendix III to Part 27—Fair Housing Lending...

  1. FEMAXI-III: a computer code for the analysis of thermal and mechanical behavior of fuel rods

    International Nuclear Information System (INIS)

    Nakajima, Tetsuo; Ichikawa, Michio; Iwano, Yoshihiko; Ito, Kenichi; Saito, Hiroaki; Kashima, Koichi; Kinoshita, Motoyasu; Okubo, Tadatsune.

    1985-12-01

    FEMAXI-III is a computer code to predict the thermal and mechanical behavior of a light water fuel rod during its irradiation life. It can analyze the integral behavior of a whole fuel rod throughout its life, as well as the localized behavior of a small part of fuel rod. The localized mechanical behavior such as the cladding ridge deformation is analyzed by the two-dimensional axisymmetric finite element method. FEMAXI-III calculates, in particular, the temperature distribution, the radial deformation, the fission gas release, and the inner gas pressure as a function of irradiation time and axial position, and the stresses and strains in the fuel and cladding at a small part of fuel rod as a function of irradiation time. For this purpose, Elasto-plasticity, creep, thermal expansion, fuel cracking and crack healing, relocation, densification, swelling, hot pressing, heat generation distribution, fission gas release, and fuel-cladding mechanical interaction are modelled and their interconnected effects are considered in the code. Efforts have been made to improve the accuracy and stability of finite element solution and to minimize the computer memory and running time. This report describes the outline of the code and the basic models involved, and also includes the application of the code and its input manual. (author)

  2. Alternate-Fueled Combustor-Sector Performance: Part A: Combustor Performance Part B: Combustor Emissions

    Science.gov (United States)

    Shouse, D. T.; Neuroth, C.; Henricks, R. C.; Lynch, A.; Frayne, C.; Stutrud, J. S.; Corporan, E.; Hankins, T.

    2010-01-01

    Alternate aviation fuels for military or commercial use are required to satisfy MIL-DTL-83133F(2008) or ASTM D 7566 (2010) standards, respectively, and are classified as drop-in fuel replacements. To satisfy legacy issues, blends to 50% alternate fuel with petroleum fuels are certified individually on the basis of feedstock. Adherence to alternate fuels and fuel blends requires smart fueling systems or advanced fuel-flexible systems, including combustors and engines without significant sacrifice in performance or emissions requirements. This paper provides preliminary performance (Part A) and emissions and particulates (Part B) combustor sector data for synthetic-parafinic-kerosene- (SPK-) type fuel and blends with JP-8+100 relative to JP-8+100 as baseline fueling.

  3. Development of failed fuel detection system for PWR (III)

    International Nuclear Information System (INIS)

    Hwang, Churl Kew; Kang, Hee Dong; Jeong, Seung Ho; Cho, Byung Sub; Yoon, Byeong Joo; Yoon, Jae Seong

    1987-12-01

    Ultrasonic transducers satisfying the conditions for failed fuel rod detection for failed fuel rod detection have been designed and built. And performance tests for them have been carried out. Ultrasonic signal processing units, a manipulator guiding the ultrasonic probe through the fuel assembly lanes and its control units have been constructed. The performance of the system has been verified experimentally to be successful in failed fuel rod detection. (Author)

  4. Final Report. Fumex-III. Improvement of Models Used for Fuel Behaviour Simulation

    International Nuclear Information System (INIS)

    Kulacsy, Katalin

    2013-01-01

    The FUMEX-III coordinated research programme organised by the IAEA was the first FUMEX exercise in which AEKI (Hungarian Academy of Sciences KFKI Atomic Energy Research Institute) took part with the partial support of Paks NPP. The aim of the participation was to test the code FUROM developed at AEKI against not only measurements but also other fuel behaviour simulation codes, to share and discuss modelling experience and issues, and to establish acquaintance with fuel modellers in other countries. Among the numerous cases proposed for the programme, AEKI chose to simulate normal operation up to high burn-up and ramp tests, with special interest in VVER rods and PWR rods with annular pellets. The US PWR 16x16, the SPC RE GINNA, the Kola3-MIR, the IFA-519.9 cases and the AREVA idealised rod were thus selected. The present Final Report gives a short description of the FUROM models relevant to the selected cases, presents the results for the 5 cases and summarises the conclusions of the FUMEX-III programme. The input parameters used for the simulations can be found in the Appendix at the end of the Report. Observations concerning the IFPE datasets are collected for each dataset in their respective Sections for possible use in the IFPE database. (author)

  5. Compatibility analysis of DUPIC fuel (part 3) - radiation physics analysis

    International Nuclear Information System (INIS)

    Kim, Chun Soo; Bae, Dae Seok; Kim, Kyung Su; Park, Byung Yun; Koh, Young Kown

    2000-04-01

    As a part of the compatibility analysis of DUPIC fuel in CANDU reactors, the radiation physics calculations have been performed for the CANDU primary shielding system, thermal shield, radiation damage, transportation cask and storage. At first, the primary shield system was assessed for the DUPIC fuel core, which has shown that the dose rates and heat deposition rates through the primary shield of the DUPIC fuel core are not much different from those of natural uranium core because the power levels on the core periphery are similar for both cores. Secondly, the radiation effects on the critical components and the themal shields were assessed when the DUPIC fuel is loaded in CANDU reactors. Compared with the displacement per atom (DPA) of the critical component for natural uranium core, that for the DUPIC fuel core was increased by -30% for the innermost groove and the weld points and by -10% for the corner of the calandria subshells and annular plates in the calandria, respectivdely. Finally, the feasibility study of the DUPIC fuel handling was performed, which has shown that all handling and inspection of the DUPIC fuel bundles be done remotely and behind a shielding wall. For the transportation of the DUPIC fuel, the preliminary study has shown that there shold be no technical problem th design a transportation cask for the fresh and spent DUPIC fuel bundles. For the storage of the fresh and spent DUPIC fuels, there is no the criticality safety problem unless the fuel bundle geometry is destroyed

  6. Compatibility analysis of DUPIC fuel (part 3) - radiation physics analysis

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Chun Soo; Bae, Dae Seok; Kim, Kyung Su; Park, Byung Yun; Koh, Young Kown

    2000-04-01

    As a part of the compatibility analysis of DUPIC fuel in CANDU reactors, the radiation physics calculations have been performed for the CANDU primary shielding system, thermal shield, radiation damage, transportation cask and storage. At first, the primary shield system was assessed for the DUPIC fuel core, which has shown that the dose rates and heat deposition rates through the primary shield of the DUPIC fuel core are not much different from those of natural uranium core because the power levels on the core periphery are similar for both cores. Secondly, the radiation effects on the critical components and the themal shields were assessed when the DUPIC fuel is loaded in CANDU reactors. Compared with the displacement per atom (DPA) of the critical component for natural uranium core, that for the DUPIC fuel core was increased by -30% for the innermost groove and the weld points and by -10% for the corner of the calandria subshells and annular plates in the calandria, respectivdely. Finally, the feasibility study of the DUPIC fuel handling was performed, which has shown that all handling and inspection of the DUPIC fuel bundles be done remotely and behind a shielding wall. For the transportation of the DUPIC fuel, the preliminary study has shown that there shold be no technical problem th design a transportation cask for the fresh and spent DUPIC fuel bundles. For the storage of the fresh and spent DUPIC fuels, there is no the criticality safety problem unless the fuel bundle geometry is destroyed.

  7. Thallium (III) salts utilization in organic synthesis. Part II

    International Nuclear Information System (INIS)

    Ferraz, H.M.C.

    1989-01-01

    The utilizations of thallium (III) salts in organic synthesis with carbonylic and acitylenic substrates are presented. The reactions of carbonylic substra3ts with kitones and the oxidation reactions of acetylenic substrates are shown. Others reactions including thallium (III) salts and non aromatic unsatured substracts, as cleasage of ethers and epoxide using thallium trinitrate, hydrazones treatments with thallium triacetates, etc, are also mentioned. (C.G.C.) [pt

  8. Applications for fueling of Forsmark-3 and Oskarshamn III

    International Nuclear Information System (INIS)

    1984-01-01

    The method for handling and final disposal of spent nuclear fuel outlined in the report KBS-3 has been found acceptable in relation to conventional and radiation safety. The main institutions agree on this, IAEA and other international institutions deem the method over-safe. The KBS-3 study has left some problems unsolved. Further research projects have been identified. (Aa)

  9. Tanzania 1895-1920 : Part III: 1914-1920s

    NARCIS (Netherlands)

    Dietz, A.J.

    2016-01-01

    An earlier version of this African Postal Heritage Paper was published as African Studies Centre Leiden Working Paper 119 / 2015: "A postal history of the First World War in Africa and its aftermath - German colonies; III Deutsch Ostafrika / German East Africa", written by Ton Dietz.

  10. Inteligencia Artificial y Neurología. (III Parte

    Directory of Open Access Journals (Sweden)

    Mario Camacho Pinto

    1987-04-01

    Full Text Available

    De acuerdo con mi anuncio esta III Parte estaría constituida por los mecanismos cerebrales susceptibles de extrapolación tal como fueron enumerados por mí: control de input-output para realizar conductas, y de inteligencia y aprendizaje, de los cuales por razón de espacio sólo se publica la mitad en esta edición de Medicina. Se trata de una presentación esquemática, auncuando ahora encuentro quizás más atractivo el enfoque de J’urgen Ruech expuesto en el Capítulo Comunicación y Psiquiatría de la obra extensa de Freedman (1 así: Input = percepción; análisis de datos = reconocimiento; procesamiento de datos = pensamiento; almacenamiento de datos = memoria; output = expresión y acción. A mi modo de ver se completaría este encuadre funcional con el tópico aprendizaje, proceso contiguo al de la memoria. Antes de entrar en materia hago unas consideraciones preliminares. En la primera me refiero a otro enfoque del concepto de LA. no incluido anteriormente. Se trata de Schank Roger y Hunter Larry (2 para quienes las indagaciones a que conduce el trasegar acerca de lA son las más atrevidas de nuestra existencia: ¿cuál es la naturaleza de la mente, qué pasa cuando estamos pensando, sintiendo, viendo o entendiendo? ¿Es posible comprender cómo trabaja nuestra mente realmente? Preguntas milenarias en cuyas respuestas no se ha registrado progreso. La lA ofrece una nueva herramienta para avanzar en este sentido: el computador.

    Las teorías sobre la mente han consistido en procesos descriptivos. Y los planteamientos iníciales hechos sobre lA por los investigadores han sido enfocados hacia lo que ellos mismos consideraron como manifestaciones de alta inteligencia: problemas matemáticos, ajedrez, rompecabezas complejos, etc.; gran cantidad de energía fue dedicada y se encontraron técnicas computacionales exitosas. Pero se comprendió que las técnicas desarrolladas no eran las mismas que emplea el cerebro, por lo cual se

  11. Irradiated uranium reprocessing, Final report - I-IV, Part III

    International Nuclear Information System (INIS)

    Gal, I.

    1961-12-01

    This third part of the final report include the following: Annex 5 - device for opening the cover; Annex 6 - inner part of the device for sampling of the radioactive solution; Annex 7 - outer part of the device for sampling of the radioactive solution; Annex 8 - pneumatic taps [sr

  12. Review of oxidation rates of DOE spent nuclear fuel : Part 1 : nuclear fuel

    International Nuclear Information System (INIS)

    Hilton, B.A.

    2000-01-01

    The long-term performance of Department of Energy (DOE) spent nuclear fuel (SNF) in a mined geologic disposal system depends highly on fuel oxidation and subsequent radionuclide release. The oxidation rates of nuclear fuels are reviewed in this two-volume report to provide a baseline for comparison with release rate data and technical rationale for predicting general corrosion behavior of DOE SNF. The oxidation rates of nuclear fuels in the DOE SNF inventory were organized according to metallic, Part 1, and non-metallic, Part 2, spent nuclear fuels. This Part 1 of the report reviews the oxidation behavior of three fuel types prototypic of metallic fuel in the DOE SNF inventory: uranium metal, uranium alloys and aluminum-based dispersion fuels. The oxidation rates of these fuels were evaluated in oxygen, water vapor, and water. The water data were limited to pure water corrosion as this represents baseline corrosion kinetics. Since the oxidation processes and kinetics discussed in this report are limited to pure water, they are not directly applicable to corrosion rates of SNF in water chemistry that is significantly different (such as may occur in the repository). Linear kinetics adequately described the oxidation rates of metallic fuels in long-term corrosion. Temperature dependent oxidation rates were determined by linear regression analysis of the literature data. As expected the reaction rates of metallic fuels dramatically increase with temperature. The uranium metal and metal alloys have stronger temperature dependence than the aluminum dispersion fuels. The uranium metal/water reaction exhibited the highest oxidation rate of the metallic fuel types and environments that were reviewed. Consequently, the corrosion properties of all DOE SNF may be conservatively modeled as uranium metal, which is representative of spent N-Reactor fuel. The reaction rate in anoxic, saturated water vapor was essentially the same as the water reaction rate. The long-term intrinsic

  13. A user input manual for single fuel rod behaviour analysis code FEMAXI-III

    International Nuclear Information System (INIS)

    Saito, Hiroaki; Yanagisawa, Kazuaki; Fujita, Misao.

    1983-03-01

    Principal objectives of Safety related research in connection with lighr water reactor fuel rods under normal operating condition are mainly addressed 1) to assess fuel integrity under steady state condition and 2) to generate initial condition under hypothetical accident. These assessments have to be relied principally upon steady state fuel behaviour computing code that is able to calculate fuel conditions to tbe occurred in a various manner. To achieve these objectives, efforts have been made to develope analytical computer code that calculates in-reactor fuel rod behaviour in best estimate manner. The computer code developed for the prediction of the long-term burnup response of single fuel rod under light water reactor condition is the third in a series of code versions:FEMAMI-III. The code calculates temperature, rod internal gas pressure, fission gas release and pellet-cladding interaction related rod deformation as a function of time-dependent fuel rod power and coolant boundary conditions. This document serves as a user input manual for the code FEMAMI-III which has opened to the public in year of 1982. A general description of the code input and output are included together with typical examples of input data. A detailed description of structures, analytical submodels and solution schemes in the code shall be given in the separate document to be published. (author)

  14. Application of the BISON Fuel Performance Code of the FUMEX-III Coordinated Research Project

    International Nuclear Information System (INIS)

    Williamson, R.L.; Novascone, S.R.

    2013-01-01

    Since 1981, the International Atomic Energy Agency (IAEA) has sponsored a series of Coordinated Research Projects (CRP) in the area of nuclear fuel modeling. These projects have typically lasted 3-5 years and have had broad international participation. The objectives of the projects have been to assess the maturity and predictive capability of fuel performance codes, support interaction and information exchange between countries with code development and application needs, build a database of well- defined experiments suitable for code validation, transfer a mature fuel modeling code to developing countries, and provide guidelines for code quality assurance and code application to fuel licensing. The fourth and latest of these projects, known as FUMEX-III1 (FUel Modeling at EXtended Burnup- III), began in 2008 and ended in December of 2011. FUMEX-III was the first of this series of fuel modeling CRP's in which the INL participated. Participants met at the beginning of the project to discuss and select a set of experiments ('priority cases') for consideration during the project. These priority cases were of broad interest to the participants and included reasonably well-documented and reliable data. A meeting was held midway through the project for participants to present and discuss progress on modeling the priority cases. A final meeting was held at close of the project to present and discuss final results and provide input for a final report. Also in 2008, the INL initiated development of a new multidimensional (2D and 3D) multiphysics nuclear fuel performance code called BISON, with code development progressing steadily during the three-year FUMEX-III project. Interactions with international fuel modeling researchers via FUMEX-III played a significant role in the BISON evolution, particularly influencing the selection of material and behavioral models which are now included in the code. The FUMEX-III cases are generally integral fuel rod experiments occurring

  15. Innovative nuclear fuels and applications. Part 1: limits of today's fuels and concepts for innovative fuels. Part 2: materials properties, irradiation performance and gaps in our knowledge

    International Nuclear Information System (INIS)

    Matzke, H.

    2000-01-01

    Part I of this contribution on innovative nuclear fuels gives a summary of current developments and problems of today's fuels, i.e. enriched UO 2 and UO 2 with a few % of PUO 2 (MOX fuel) or Gd 2 O 3 (as burnable neutron poison). The problems and property changes caused by high burnups (e.g. degradation of the thermal conductivity, polygonization or formation of the rim-structure) are discussed. Subsequently, the concepts for new fuels to burn excess Pu and to achieve an effective transmutation of the minor actinides Np, Am and Cm are treated. The criteria for the choice of suitable fuels and different fuel types (high Pu-content fuels, nitrides, U-free fuels, inert matrix supported fuels, cercers, cermets, etc.) are discussed. Part II of this contribution on innovative nuclear fuels deals with the properties of relevance of the different materials suggested to be used in innovative fuels which range from pure actinide fuel such as PuN and AmO 2 to spinel MgAl 2 O 4 and zircon ZrSiO 4 for inert matrix-based fuels, etc. The available knowledge on materials research aspects is summarized with emphasis on the physics of radiation damage. It is shown that significant gaps in the present knowledge exist, e.g. for the minor actinide compounds, and suggestions are made to fill these gaps in order to achieve a sufficient data base to design and operate suitable innovative fuels in a near future. (author)

  16. 29 CFR Appendix III to Part 1918 - The Mechanics of Conventional Cargo Gear (Non-mandatory)

    Science.gov (United States)

    2010-07-01

    ... 29 Labor 7 2010-07-01 2010-07-01 false The Mechanics of Conventional Cargo Gear (Non-mandatory.... 1918, App. III Appendix III to Part 1918—The Mechanics of Conventional Cargo Gear (Non-mandatory) Note: This appendix is non-mandatory and provides an explanation of the mechanics in the correct spotting of...

  17. Fuels and fire in land-management planning. Part 1. Forest-fuel classification.

    Science.gov (United States)

    Wayne G. Maxwell; Franklin R. Ward

    1981-01-01

    This report describes a way to collect and classify the total fuel complex within a planning area. The information can be used as input for appraising and rating probable fire behavior and calculating expected costs and losses from various land uses and management alternatives, reported separately as Part 2 and Part 3 of this series. This total package can be used...

  18. Eighth annual meeting, Vienna, Austria, 15-18 April 1975. Summary report. Part III

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1976-01-01

    The Summary Report of the Eighth Annual Meeting of the International Working Group on Fast Reactors contains the minutes of the meeting (Part 1); papers which review the national programmes in the field of LMFBRs (Part 2) and the discussions on the review of national programmes (Part 3). The agenda of the meeting involved design, construction, operating experiences of demonstration fast power reactors, reprocessing of spent fuel from LMFBRs, reliability of decay heat removal systems, fuel failure mechanisms, safety of LMFBRs.

  19. Eighth annual meeting, Vienna, Austria, 15-18 April 1975. Summary report. Part III

    International Nuclear Information System (INIS)

    1976-01-01

    The Summary Report of the Eighth Annual Meeting of the International Working Group on Fast Reactors contains the minutes of the meeting (Part 1); papers which review the national programmes in the field of LMFBRs (Part 2) and the discussions on the review of national programmes (Part 3). The agenda of the meeting involved design, construction, operating experiences of demonstration fast power reactors, reprocessing of spent fuel from LMFBRs, reliability of decay heat removal systems, fuel failure mechanisms, safety of LMFBRs

  20. Antipsychotics and Sexual Dysfunction: Sexual Dysfunction - Part III

    Directory of Open Access Journals (Sweden)

    Anil Kumar Mysore Nagaraj

    2009-11-01

    Full Text Available Satisfying sexual experience is an essential part of a healthy and enjoyable life for most people. Antipsychotic drugs are among the various factors that affect optimal sexual functioning. Both conventional and novel antipsychotics are associated with significant sexual side effects. This review has presented various studies comparing different antipsychotic drugs. Dopamine antagonism, increased serum prolactin, serotonergic, adrenergic and cholinergic mechanisms are all proposed to be the mechanisms for sexual dysfunction. Drug treatment for this has not given satisfactory long-term results. Knowledge of the receptor pharmacology of an individual antipsychotic will help to determine whether it is more or less likely to cause sexual side effects and its management.

  1. COMETHE III J a computer code for predicting mechanical and thermal behaviour of a fuel pin

    International Nuclear Information System (INIS)

    Verbeek, P.; Hoppe, N.

    1976-01-01

    The design of fuel pins for power reactors requires a realistic evaluation of their thermal and mechanical performances throughout their irradiation life. This evaluation involves the knowledge of a number of parameters, very intricate and interconnected, for example, the temperature, the restructuring and the swelling rates of the fuel pellets, the dimensions, the stresses and the strains in the clad, the composition and the properties of gases, the inner gas pressure etc. This complex problem can only be properly handled by a computer programme which analyses the fuel pin thermal and mechanical behaviour at successive steps of its irradiation life. This report presents an overall description of the COMETHE III-J computer programme, designed to calculate the integral performance of oxide fuel pins with cylindrical metallic cladding irradiated in thermal or fast flux. (author)

  2. Radiobiology in clinical radiation therapy - Part III: Normal tissue damage

    International Nuclear Information System (INIS)

    Travis, Elizabeth L.

    1996-01-01

    Objective: This is the third part of a course designed for residents in radiation oncology preparing for their boards. This part of the course will focus on the mechanisms underlying damage in normal tissues. Although conventional wisdom long held that killing and depletion of a critical cell(s) in a tissue was responsible for the later expression of damage, histopathologic changes in normal tissue can now be explained and better understood in terms of the new molecular biology. The concept that depletion of a single cell type is responsible for the observed histopathologic changes in normal tissues has been replaced by the hypothesis that damage results from the interaction of many different cell systems, including epithelial, endothelial, macrophages and fibroblasts, via the production of specific autocrine, paracrine and endocrine growth factors. A portion of this course will discuss the clinical and experimental data on the production and interaction of those cytokines and cell systems considered to be critical to tissue damage. It had long been suggested that interindividual differences in radiation-induced normal tissue damage was genetically regulated, at least in part. Both clinical and experimental data supported this hypothesis but it is the recent advances in human and mouse molecular genetics which have provided the tools to dissect out the genetic component of normal tissue damage. These data will be presented and related to the potential to develop genetic markers to identify sensitive individuals. The impact on clinical outcome of the ability to identify prospectively sensitive patients will be discussed. Clinically it is well-accepted that the volume of tissue irradiated is a critical factor in determining tissue damage. A profusion of mathematical models for estimating dose-volume relationships in a number of organs have been published recently despite the fact that little data are available to support these models. This course will review the

  3. 18 CFR 410.1 - Basin regulations-Water Code and Administrative Manual-Part III Water Quality Regulations.

    Science.gov (United States)

    2010-04-01

    ... Code and Administrative Manual-Part III Water Quality Regulations. 410.1 Section 410.1 Conservation of... CODE AND ADMINISTRATIVE MANUAL-PART III WATER QUALITY REGULATIONS § 410.1 Basin regulations—Water Code and Administrative Manual—Part III Water Quality Regulations. (a) The Water Code of the Delaware River...

  4. Fossil fuels: Kyoto initiatives and opportunities. Part 1

    International Nuclear Information System (INIS)

    Pinelli, G.; Zerlia, T.

    2008-01-01

    GHG emission in the upstream step of fossil fuel chains could give an environmental as well as economic opportunity for traditional sectors. This study deepens the matter showing an increasing number of initiative over the last few years taken both the involved sectors and by various stake holders (public and private subjects) within the Kyoto flexible mechanism (CDM and JI) or linked to voluntary national or at a global level actions. The above undertakings give evidence for an increased interest and an actual activity dealing with GHG reduction whose results play an evident and positive role for the environment too. Part 1. of this study deals with fossil fuel actions within the Kyoto protocol mechanism. Part 2. will show international and national voluntary initiative [it

  5. International Working Group on Past Reactors Thirteenth Annual Meeting. Summary Report. Part III

    International Nuclear Information System (INIS)

    1981-04-01

    The Thirteenth Annual Meeting of the IAEA International Working Group on Fast Reactors was held at the IAEA Headquarters, Vienna, Austria from 9 to 11 April 1980. The Summary Report (Part I) contains the Minutes of the Meeting. The Summary Report (Part II) contains the papers which review the national programme in the field of LMFBRs and other presentations at the Meeting. The Summary Report (Part III) contains the discussions on the review of the national programmes

  6. Nuclear fuel for light water reactors. Part 2 and conclusion

    International Nuclear Information System (INIS)

    1983-01-01

    The article gives brief descriptions of a new cycle for nuclear fuel in the core and, in particular, fuel replacement, stock pool management for irradiated fuel elements, transport containers for irradiated nuclear fuels, treatment of low activity waste, the Climax system for long-term stocking of irradiated fuel, and transport of irradiated fuel over the Nevada Test Site. (A.E.W.)

  7. Criticality and shielding calculations for containers in dry of spent fuel of TRIGA Mark III reactor of ININ

    International Nuclear Information System (INIS)

    Barranco R, F.

    2015-01-01

    In this thesis criticality and shielding calculations to evaluate the design of a container of dry storage of spent nuclear fuel generated in research reactors were made. The design of such container was originally proposed by Argentina and Brazil, and the Instituto Nacional de Investigaciones Nucleares (ININ) of Mexico. Additionally, it is proposed to modify the design of this container to store spent fuel 120 that are currently in the pool of TRIGA Mark III reactor, the Nuclear Center of Mexico and calculations and analyzes are made to verify that the settlement of these fuel elements is subcritical limits and dose rates to workers and the general public are not exceeded. These calculations are part of the design criteria for security protection systems in dry storage system (Dss for its acronym in English) proposed by the Nuclear Regulatory Commission (NRC) of the United States. To carry out these calculations simulation codes of Monte Carlo particle transport as MCNPX and MCNP5 were used. The initial design (design 1) 78 intended to store spent fuel with a maximum of 115. The ININ has 120 fuel elements and spent 3 control rods (currently stored in the reactor pool). This leads to the construction of two containers of the original design, but for economic reasons was decided to modify (design 2) to store in a single container. Criticality calculations are performed to 78, 115 and fresh fuel elements 124 within the container, to the two arrangements described in Chapter 4, modeling the three-dimensional geometry assuming normal operating conditions and accident. These calculations are focused to demonstrate that the container will remain subcritical, that is, that the effective multiplication factor is less than 1, in particular not greater than 0.95 (as per specified by the NRC). Spent fuel 78 and 124 within the container, both gamma radiation to neutron shielding calculations for only two cases were simulated. First actinides and fission products generated

  8. 16 CFR Appendix A to Part 306 - Summary of Labeling Requirements for Biodiesel Fuels

    Science.gov (United States)

    2010-01-01

    ... Biodiesel Fuels A Appendix A to Part 306 Commercial Practices FEDERAL TRADE COMMISSION REGULATIONS UNDER... Part 306—Summary of Labeling Requirements for Biodiesel Fuels (Part 1 of 2) Fuel type Blends of 5 percent or less Blends of more than 5 but not more than 20 percent Header Text Color Biodiesel No label...

  9. ATWS: a reappraisal. Part III. Frequency of anticipated transients. Interim report

    International Nuclear Information System (INIS)

    Leverenz, F.L. Jr.; Koren, J.M.; Erdmann, R.C.; Lellouche, G.S.

    1978-07-01

    The document is Part III of the Institute study of the ATWS question. The frequencies of the various events which have led to a reactor scram are documented from the nuclear power plant records. Some of these events, in the absence of scram, could lead to undesirable system response and are the ''transients of significance'' which comprise the anticipated transients of the ATWS question

  10. A cold demonstration of fuel consolidation. Part 1

    International Nuclear Information System (INIS)

    Matheson, J.E.

    1989-01-01

    Spent fuel consolidation is an option for increasing spent fuel storage capacities being considered by many utilities. The process of consolidating fuel involves separating the fuel rods from the structural frame which holds them in a square array. The rods are then repackaged into a tightly packed bundle which occupies about half the cross-sectional area of fuel assembly. Thus approximately twice as much fuel can be stored in the underwater racks at a spent fuel storage pool. There have been several demonstrations of fuel consolidation to date. The focus of this paper is the development and subsequent demonstration program of a shear/compactor

  11. 49 CFR Appendix D to Part 238 - Requirements for External Fuel Tanks on Tier I Locomotives

    Science.gov (United States)

    2010-10-01

    ... 49 Transportation 4 2010-10-01 2010-10-01 false Requirements for External Fuel Tanks on Tier I..., App. D Appendix D to Part 238—Requirements for External Fuel Tanks on Tier I Locomotives The... properties of the locomotive fuel tank to reduce the risk of fuel spillage to acceptable levels under...

  12. 40 CFR Appendix II to Part 600 - Sample Fuel Economy Calculations

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 29 2010-07-01 2010-07-01 false Sample Fuel Economy Calculations II... FUEL ECONOMY AND CARBON-RELATED EXHAUST EMISSIONS OF MOTOR VEHICLES Pt. 600, App. II Appendix II to Part 600—Sample Fuel Economy Calculations (a) This sample fuel economy calculation is applicable to...

  13. Gas Generation from K East Basin Sludges and Irradiated Metallic Uranium Fuel Particles Series III Testing

    International Nuclear Information System (INIS)

    Schmidt, Andrew J.; Delegard, Calvin H.; Bryan, Samuel A.; Elmore, Monte R.; Sell, Rachel L.; Silvers, Kurt L.; Gano, Susan R.; Thornton, Brenda M.

    2003-01-01

    The path forward for managing of Hanford K Basin sludge calls for it to be packaged, shipped, and stored at T Plant until final processing at a future date. An important consideration for the design and cost of retrieval, transportation, and storage systems is the potential for heat and gas generation through oxidation reactions between uranium metal and water. This report, the third in a series (Series III), describes work performed at the Pacific Northwest National Laboratory (PNNL) to assess corrosion and gas generation from irradiated metallic uranium particles (fuel particles) with and without K Basin sludge addition. The testing described in this report consisted of 12 tests. In 10 of the tests, 4.3 to 26.4 g of fuel particles of selected size distribution were placed into 60- or 800-ml reaction vessels with 0 to 100 g settled sludge. In another test, a single 3.72-g fuel fragment (i.e., 7150-mm particle) was placed in a 60 ml reaction vessel with no added sludge. The twelfth test contained only sludge. The fuel particles were prepared by crushing archived coupons (samples) from an irradiated metallic uranium fuel element. After loading the sludge materials (whether fuel particles, mixtures of fuel particles and sludge, or sludge-only) into reaction vessels, the solids were covered with an excess of K Basin water, the vessels closed and connected to a gas measurement manifold, and the vessels back-flushed with inert neon cover gas. The vessels were then heated to a constant temperature. The gas pressures and temperatures were monitored continuously from the times the vessels were purged. Gas samples were collected at various times during the tests, and the samples analyzed by mass spectrometry. Data on the reaction rates of uranium metal fuel particles with water as a function of temperature and particle size were generated. The data were compared with published studies on metallic uranium corrosion kinetics. The effects of an intimate overlying sludge layer

  14. Automotive fuels survey. Part 4. Innovations or illusions

    International Nuclear Information System (INIS)

    Troelstra, W.P.; Van Walwijk, M.; Bueckmann, M.

    1999-01-01

    Volumes 1 to 3 of the IEA/AFIS Automotive Fuels Survey, address the most well-known automotive fuels and fuel production routes. Less well-known fuels and energy sources that are not used in combustion engines, e.g. electricity, were excluded from these volumes. In this report fuel routes and fuels that have not been addressed in the first volumes will be analysed. In this report, each chapter starts with a short description of the fuel(route) and its status of development (e.g. if the idea has been abandoned or if the fuel is already sold at a fuel station). Then the different aspects of that fuel are described as far as the information is available. This is limited to information that can not be found in volumes one and two of the Automotive Fuels Survey. For example: for the diesel-water mixtures, the production of diesel is not be described. If comparisons are made, they are made either relative to an already described fuel(route) that is related (e.g. biogas will be compared with natural gas) or relative to diesel and gasoline as was done in volume 1 and 2 of the Automotive Fuels Survey. For some of the fuels, the relation with a fuel already covered in volume one and two is very strong. For these fuels more information can be found in the chapters on the related fuel in the other volumes of the Automotive Fuels Survey. The following fuels are covered in this report: biodiesel from used oil and fat, biodiesel and biogasoline from algae, diesel from hydrothermal upgrading, biogas, hythane, Fischer-Tropsch diesel, diesel-water blends, higher ethers, and electricity. 74 refs

  15. FEMAXI-III, a computer code for fuel rod performance analysis

    International Nuclear Information System (INIS)

    Ito, K.; Iwano, Y.; Ichikawa, M.; Okubo, T.

    1983-01-01

    This paper presents a method of fuel rod thermal-mechanical performance analysis used in the FEMAXI-III code. The code incorporates the models describing thermal-mechanical processes such as pellet-cladding thermal expansion, pellet irradiation swelling, densification, relocation and fission gas release as they affect pellet-cladding gap thermal conductance. The code performs the thermal behavior analysis of a full-length fuel rod within the framework of one-dimensional multi-zone modeling. The mechanical effects including ridge deformation is rigorously analyzed by applying the axisymmetric finite element method. The finite element geometrical model is confined to a half-pellet-height region with the assumption that pellet-pellet interaction is symmetrical. The 8-node quadratic isoparametric ring elements are adopted for obtaining accurate finite element solutions. The Newton-Raphson iteration with an implicit algorithm is applied to perform the analysis of non-linear material behaviors accurately and stably. The pellet-cladding interaction mechanism is exactly treated using the nodal continuity conditions. The code is applicable to the thermal-mechanical analysis of water reactor fuel rods experiencing variable power histories. (orig.)

  16. Visual inspection system and sipping design for spent fuel at TRIGA MARK III reactor of Mexico

    International Nuclear Information System (INIS)

    Delfin, A.; Mazon, R.

    2002-01-01

    In the framework of the Technical Cooperation Regional Project for Latin America RLA/4/018 for the biennium 2001-2002, one of the activities identified is the characterization of spent fuel. Of these activities an important one is not doubt the physical condition of spent fuel because an appropriate identification of the fuel status will prevent problems of fuel leaks, corrosion problems etc. As part of the activities of the project was decided that countries no having visual inspection and sipping systems should be very desirable to have them as a result of this project. The Triga reactor of Mexico does not have both of them, therefore, it was decided the need of having both system. The paper describe first the way we designed and constructed a remote Visual Inspection System and example of how is operated. Along the experience and problems we have had with the system. Also we will present the design of the Sipping system were two option were considered. First to take a sample of water after a convenient period of time passing through a circuit to a multichannel analyzer and to identify leakage by way of measuring Caesium-137. Second, exists the possibility that the Stainless Steel sleeve of the fuel has only very small failures, so it is going to be very difficult to have leakages unless the fuel is hot. Therefore we are evaluating the possibility of using heaters to increase the temperature of the fuel and succeed on detecting leakages. The results - we hope - will be ready to be presented at the meeting. (author)

  17. Japanese contributions to IAEA INTOR workshop, phase two A, part 2, chapter III: impurity control (engineering)

    International Nuclear Information System (INIS)

    Seki, Masahiro; Miki, Nobuharu; Shibutani, Yoji; Fujimura, Kaoru; Adachi, Jun-ichi; Sato, Kosuke; Fujii, Masaharu; Yamazaki, Seiichiro; Itoh, Shin-ichi.

    1985-07-01

    This report corresponds to the second half of Chapter III of Japanese contribution report to IAEA INTOR Workshop, Phase Two A, Part 2. Data base assessment are made on candidate materials for the divertor, limiter, and the first wall. Engineering trade-off studies are made for the high-recycling and low temperature conditions. The studies include material considerations, configuration, thermohydraulic and stress analysis, disruption, lifetime analysis, and tritium permeation. (author)

  18. 40 CFR Appendix III to Part 266 - Tier II Emission Rate Screening Limits for Free Chlorine and Hydrogen Chloride

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 26 2010-07-01 2010-07-01 false Tier II Emission Rate Screening Limits for Free Chlorine and Hydrogen Chloride III Appendix III to Part 266 Protection of Environment... to Part 266—Tier II Emission Rate Screening Limits for Free Chlorine and Hydrogen Chloride Terrain...

  19. Neutron spectra in two beam ports of a TRIGA Mark III reactor with HEU fuel

    International Nuclear Information System (INIS)

    Vega C, H. R.; Hernandez D, V. M.; Paredes G, L.; Aguilar, F.

    2012-10-01

    Before to change the HEU for Leu fuel of the ININ's TRIGA Mark III nuclear reactor the neutron spectra were measured in two beam ports using 5 and 10 W. Measurements were carried out in a tangential and a radial beam port using a Bonner sphere spectrometer. It was found that neutron spectra are different in the beam ports, in radial beam port the amplitude of thermal and fast neutrons are approximately the same while, in the tangential beam port thermal neutron peak is dominant. In the radial beam port the fluence-to-ambient dose equivalent factors are 131±11 and 124±10 p Sv-cm 2 for 5 and 10 W respectively while in the tangential beam port the fluence-to-ambient dose equivalent factor is 55±4 p Sv-cm 2 for 10 W. (Author)

  20. Hanford spent nuclear fuel project recommended path forward, volume III: Alternatives and path forward evaluation supporting documentation

    International Nuclear Information System (INIS)

    Fulton, J.C.

    1994-10-01

    Volume I of the Hanford Spent Nuclear Fuel Project - Recommended Path Forward constitutes an aggressive series of projects to construct and operate systems and facilities to safely retrieve, package, transport, process, and store K Basins fuel and sludge. Volume II provided a comparative evaluation of four Alternatives for the Path Forward and an evaluation for the Recommended Path Forward. Although Volume II contained extensive appendices, six supporting documents have been compiled in Volume III to provide additional background for Volume II

  1. Optimization of binary breeder reactor IV - Conception of mixed fuel at central part of the core

    International Nuclear Information System (INIS)

    Dias, A.F.; Ishiguro, Y.

    1986-04-01

    Neutronic characteristics of some LMFBRs are analized for a fueling mode that is different from those reported previously. In an inner part of the core both 233 U/ 232 Th and Pu/U assemblies are placed while the outer zone is fueled with Pu/U assemblies. Both oxide metal fuels and 232 Th and 238 U blankets are considered. (Author) [pt

  2. 10 CFR Appendix to Part 474 - Sample Petroleum-Equivalent Fuel Economy Calculations

    Science.gov (United States)

    2010-01-01

    ... 10 Energy 3 2010-01-01 2010-01-01 false Sample Petroleum-Equivalent Fuel Economy Calculations..., DEVELOPMENT, AND DEMONSTRATION PROGRAM; PETROLEUM-EQUIVALENT FUEL ECONOMY CALCULATION Pt. 474, App. Appendix to Part 474—Sample Petroleum-Equivalent Fuel Economy Calculations Example 1: An electric vehicle is...

  3. Part 6. Internationalization and collocation of FBR fuel cycle facilities

    International Nuclear Information System (INIS)

    Stevenson, M.G.; Abramson, P.B.; LeSage, L.G.

    1980-01-01

    This report examines some of the non-proliferation, technical, and institutional aspects of internationalization and/or collocation of major facilities of the Fast Breeder Reactor (FBR) fuel cycle. The national incentives and disincentives for establishment of FBR Fuel Cycle Centers are enumerated. The technical, legal, and administrative considerations in determining the feasibility of FBR Fuel Cycle Centers are addressed by making comparisons with Light Water Reactor (LWR) centers which have been studied in detail by the IAEA and UNSRC

  4. FEMAXI-III. An axisymmetric finite element computer code for the analysis of fuel rod performance

    International Nuclear Information System (INIS)

    Ichikawa, M.; Nakajima, T.; Okubo, T.; Iwano, Y.; Ito, K.; Kashima, K.; Saito, H.

    1980-01-01

    For the analysis of local deformation of fuel rods, which is closely related to PCI failure in LWR, FEMAXI-III has been developed as an improved version based on the essential models of FEMAXI-II, MIPAC, and FEAST codes. The major features of FEMAXI-III are as follows: Elasto-plasticity, creep, pellet cracking, relocation, densification, hot pressing, swelling, fission gas release, and their interrelated effects are considered. Contact conditions between pellet and cladding are exactly treated, where sliding or sticking is defined by iterations. Special emphasis is placed on creep and pellet cracking. In the former, an implicit algorithm is applied to improve numerical stability. In the latter, the pellet is assumed to be non-tension material. The recovery of pellet stiffness under compression is related to initial relocation. Quadratic isoparametric elements are used. The skyline method is applied to solve linear stiffness equation to reduce required core memories. The basic performance of the code has been proven to be satisfactory. (author)

  5. Building human resources capability in health care: a global analysis of best practice--Part III.

    Science.gov (United States)

    Zairi, M

    1998-01-01

    This is the last part of a series of three papers which discussed very comprehensively best practice applications in human resource management by drawing special inferences to the healthcare context. It emerged from parts I and II that high performing organisations plan and intend to build sustainable capability through a systematic consideration of the human element as the key asset and through a continuous process of training, developing, empowering and engaging people in all aspects of organisational excellence. Part III brings this debate to a close by demonstrating what brings about organisational excellence and proposes a road map for effective human resource development and management, based on world class standards. Healthcare human resource professionals can now rise to the challenge and plan ahead for building organisational capability and sustainable performance.

  6. Neuroscience in Nazi Europe Part III: victims of the Third Reich.

    Science.gov (United States)

    Zeidman, Lawrence A; Kondziella, Daniel

    2012-11-01

    In Part I, neuroscience collaborators with the Nazis were discussed, and in Part II, neuroscience resistors were discussed. In Part III, we discuss the tragedy regarding european neuroscientists who became victims of the Nazi onslaught on “non-Aryan” doctors. Some of these unfortunate neuroscientists survived Nazi concentration camps, but most were murdered. We discuss the circumstances and environment which stripped these neuroscientists of their profession, then of their personal rights and freedom, and then of their lives. We include a background analysis of anti-Semitism and Nazism in their various countries, then discuss in depth seven exemplary neuroscientist Holocaust victims; including Germans Ludwig Pick, Arthur Simons, and Raphael Weichbrodt, Austrians Alexander Spitzer and Viktor Frankl, and Poles Lucja Frey and Wladyslaw Sterling. by recognizing and remembering these victims of neuroscience, we pay homage and do not allow humanity to forget, lest this dark period in history ever repeat itself.

  7. Social class, political power, and the state: their implications in medicine--part III.

    Science.gov (United States)

    Navarro, V

    1977-01-01

    This is the third part of an article on the distribution of power and the nature of the state in Western industrialized societies and their implications in medicine. Parts I and II were published in the preceding issue of this Journal. Part I presented a critique of contemporary theories of the Western system of power; discussed the countervailing pluralist and power of elite theories, as well as those of bureaucratic and professional control; and concluded with an examination of the Marxist theories of economic determinism, structural determinism, and corporate statism. Part II presented a Marxist theory of the role, nature, and characteristics of state intervention. Part III focuses on the mode of that intervention and the reasons for its growth, with an added analysis of the attributes of state intervention in the health sector, and of the dialectical relationship between its growth and the current fiscal crisis of the state. In all three parts, the focus is on Western European countries and on North America, with many examples and categories from the area of medicine.

  8. Compatibility analysis of DUPIC fuel (part5) - DUPIC fuel cycle economics analysis

    International Nuclear Information System (INIS)

    Ko, Won Il; Choi, Hang Bok; Yang, Myung Seung

    2000-08-01

    This study examines the economics of the DUPIC fuel cycle using unit costs of fuel cycle components estimated based on conceptual designs. The fuel cycle cost (FCC) was calculated by a deterministic method in which reference values of fuel cycle components are used. The FCC was then analyzed by a Monte Carlo simulation to get the uncertainty of the FCC associated with the unit costs of the fuel cycle components. From the deterministic analysis on the one-batch equilibrium fuel cycle model, the DUPIC FCC was estimated to be 6.55-6.72 mills/kWh for proposed DUPIC fuel options, which is a little smaller than that of the once-through FCC by 0.04-0.28 mills/kWh. Considering the uncertainty (0.45-0.51 mills/kWh) of the FCC estimated by the Monte Carlo simulation method, the cost difference between the DUPIC and once-through fuel cycle is negligible. On the other hand, the material balance calculation has shown that the DUPIC fuel cycle can save natural uranium resources by -20% and reduce the spent fuel arising by -65%, compared with the once-through fuel cycle. In conclusion, the DUPIC fuel cycle possesses a strong advantage over the once-through fuel cycle from the viewpoint of the environmental effect

  9. Compatibility analysis of DUPIC fuel (part5) - DUPIC fuel cycle economics analysis

    Energy Technology Data Exchange (ETDEWEB)

    Ko, Won Il; Choi, Hang Bok; Yang, Myung Seung

    2000-08-01

    This study examines the economics of the DUPIC fuel cycle using unit costs of fuel cycle components estimated based on conceptual designs. The fuel cycle cost (FCC) was calculated by a deterministic method in which reference values of fuel cycle components are used. The FCC was then analyzed by a Monte Carlo simulation to get the uncertainty of the FCC associated with the unit costs of the fuel cycle components. From the deterministic analysis on the one-batch equilibrium fuel cycle model, the DUPIC FCC was estimated to be 6.55-6.72 mills/kWh for proposed DUPIC fuel options, which is a little smaller than that of the once-through FCC by 0.04-0.28 mills/kWh. Considering the uncertainty (0.45-0.51 mills/kWh) of the FCC estimated by the Monte Carlo simulation method, the cost difference between the DUPIC and once-through fuel cycle is negligible. On the other hand, the material balance calculation has shown that the DUPIC fuel cycle can save natural uranium resources by -20% and reduce the spent fuel arising by -65%, compared with the once-through fuel cycle. In conclusion, the DUPIC fuel cycle possesses a strong advantage over the once-through fuel cycle from the viewpoint of the environmental effect.

  10. 10 CFR Appendix III to Part 960 - Application of the System and Technical Guidelines During the Siting Process

    Science.gov (United States)

    2010-01-01

    ... 960—Application of the System and Technical Guidelines During the Siting Process 1. This appendix... 10 Energy 4 2010-01-01 2010-01-01 false Application of the System and Technical Guidelines During the Siting Process III Appendix III to Part 960 Energy DEPARTMENT OF ENERGY GENERAL GUIDELINES FOR THE...

  11. AUTOMOTIVE DIESEL MAINTENANCE 1. UNIT III, MAINTAINING THE FUEL SYSTEM--DETROIT DIESEL ENGINE.

    Science.gov (United States)

    Human Engineering Inst., Cleveland, OH.

    THIS MODULE OF A 30-MODULE COURSE IS DESIGNED TO DEVELOP AN UNDERSTANDING OF THE OPERATION AND MAINTENANCE OF THE DIESEL ENGINE FUEL SYSTEM. TOPICS ARE (1) PURPOSE OF THE FUEL SYSTEM, (2) TRACING THE FUEL FLOW, (3) MINOR COMPONENTS OF THE FUEL SYSTEM, (4) MAINTENANCE TIPS, (5) CONSTRUCTION AND FUNCTION OF THE FUEL INJECTORS, AND (6)…

  12. Calculation of DND-signals in case of fuel pin failures in KNK II with the computer code FICTION III

    International Nuclear Information System (INIS)

    Schmuck, I.

    1990-11-01

    In KNK II two delayed neutron detectors are installed for quick detection of fuel subassembly cladding failures. They record the release of the precursors of the emitters of delayed neutrons into the sodium. The computer code FICTION III calculates the expected delayed neutron signals for certain fuel pin failures, where the user has to set the boundary conditions interactively. In view of FICTION II the advancement of FICTION III consists of the following items: application of the data sets of 105 isotopes, distinction of thermal and fast neutron induced fission, partitioning of the sodium flow into two circuits, consideration of the specific fission rates in 10 fuel pin sections, elaboration of the user's interaction possibilities for input/ output. The capability of FICTION III is shown by means of two applications (UNi-test pin on position 100 and the third KNK fuel subassembly cladding failure). Object of further evaluations will be among other things the analysis of increased delayed neutron signals in regard to the fault location and dimension

  13. Nondestructive nuclear measurement in the fuel cycle. Part 1

    International Nuclear Information System (INIS)

    Lyoussi, A.

    2005-01-01

    Nondestructive measurement techniques are today widely used in practically all steps of the fuel cycle. This article is devoted to the presentation of the control and characterization needs and to the main passive nondestructive nuclear methods used: 1 - nondestructive nuclear measurement, needs and motivation: nuclear fuel cycle, nondestructive nuclear measurements (passive and active methods), comments; 2 - main passive nondestructive nuclear measurement methods: gamma spectroscopy (principle, detectors, electronic systems, data acquisition and signal processing, domains of application, main limitations), passive neutronic measurements (needs and motivations, neutron detectors, total neutronic counting, neutronic coincidences counting, neutronic multiplicities counting, comments). (J.S.)

  14. Spent Fuel Performance Assessment and Research. Final Report of a Coordinated Research Project on Spent Fuel Performance Assessment and Research (SPAR-III) 2009–2014

    International Nuclear Information System (INIS)

    2015-10-01

    At the beginning of 2014, there were 437 nuclear power reactors in operation and 72 reactors under construction. To date, around 370 500 t (HM) (tonnes of heavy metal) of spent fuel have been discharged from reactors, and approximately 253 700 t (HM) are stored at various storage facilities. Although wet storage at reactor sites still dominates, the amount of spent fuel being transferred to dry storage technologies has increased significantly since 2005. For example, around 28% of the total fuel inventory in the United States of America is now in dry storage. Although the licensing for the construction of geological disposal facilities is under way in Finland, France and Sweden, the first facility is not expected to be available until 2025 and for most States with major nuclear programmes not for several decades afterwards. Spent fuel is currently accumulating at around 7000 t (HM) per year worldwide. The net result is that the duration of spent fuel storage has increased beyond what was originally foreseen. In order to demonstrate the safety of both spent fuel and the storage system, a good understanding of the processes that might cause deterioration is required. To address this, the IAEA continued the Coordinated Research Project (CRP) on Spent Fuel Performance Assessment and Research (SPAR-III) in 2009 to evaluate fuel and materials performance under wet and dry storage and to assess the impact of interim storage on associated spent fuel management activities (such as handling and transport). This has been achieved through: evaluating surveillance and monitoring programmes of spent fuel and storage facilities; collecting and exchanging relevant experience of spent fuel storage and the impact on associated spent fuel management activities; facilitating the transfer of knowledge by documenting the technical basis for spent fuel storage; creating synergy among research projects of the participating Member States; and developing the capability to assess the impact

  15. The European carbon balance. Part 1: fossil fuel emissions

    NARCIS (Netherlands)

    Ciais, P.; Paris, J.D.; Marland, G.; Peylin, P.; Piao, S.L.; levin, I.; Pregger, T.; Scholz, Y.; Friedrich, R.; Rivier, L.; Houweling, S.; Schulze, E.D.

    2010-01-01

    We analyzed the magnitude, the trends and the uncertainties of fossil-fuel CO2 emissions in the European Union 25 member states (hereafter EU-25), based on emission inventories from energy-use statistics. The stability of emissions during the past decade at EU-25 scale masks decreasing trends in

  16. Criticality and shielding calculations for containers in dry of spent fuel of TRIGA Mark III reactor of ININ; Calculos de criticidad y blindaje para contenedores en seco de combustible gastado del reactor Triga Mark III del ININ

    Energy Technology Data Exchange (ETDEWEB)

    Barranco R, F.

    2015-07-01

    In this thesis criticality and shielding calculations to evaluate the design of a container of dry storage of spent nuclear fuel generated in research reactors were made. The design of such container was originally proposed by Argentina and Brazil, and the Instituto Nacional de Investigaciones Nucleares (ININ) of Mexico. Additionally, it is proposed to modify the design of this container to store spent fuel 120 that are currently in the pool of TRIGA Mark III reactor, the Nuclear Center of Mexico and calculations and analyzes are made to verify that the settlement of these fuel elements is subcritical limits and dose rates to workers and the general public are not exceeded. These calculations are part of the design criteria for security protection systems in dry storage system (Dss for its acronym in English) proposed by the Nuclear Regulatory Commission (NRC) of the United States. To carry out these calculations simulation codes of Monte Carlo particle transport as MCNPX and MCNP5 were used. The initial design (design 1) 78 intended to store spent fuel with a maximum of 115. The ININ has 120 fuel elements and spent 3 control rods (currently stored in the reactor pool). This leads to the construction of two containers of the original design, but for economic reasons was decided to modify (design 2) to store in a single container. Criticality calculations are performed to 78, 115 and fresh fuel elements 124 within the container, to the two arrangements described in Chapter 4, modeling the three-dimensional geometry assuming normal operating conditions and accident. These calculations are focused to demonstrate that the container will remain subcritical, that is, that the effective multiplication factor is less than 1, in particular not greater than 0.95 (as per specified by the NRC). Spent fuel 78 and 124 within the container, both gamma radiation to neutron shielding calculations for only two cases were simulated. First actinides and fission products generated

  17. Startup of Torrey Pines Mark III and Puerto Rico Nuclear Center reactors with TRIGA-FLIP fuel

    Energy Technology Data Exchange (ETDEWEB)

    Chesworth, R. H. [Gulf E and ES, San Diego, CA (United States)

    1972-07-01

    This paper discusses the characteristics of TRIGA FLIP cores in two different geometries: the normal TRIGA single-rod geometry as typified by the installation in the Torrey Pines Mark III reactor; and the four-rod cluster geometry as typified by the conversion core installed in the Puerto Rico Nuclear Center reactor at Mayaguez. In both reactors the fuel is 8-1/2 wt % uranium, 70% enriched in U-235. The hydrogen to zirconium atom ratio is 1.5 to 1.65 and the cladding material is stainless steel. The basic neutronic characteristics of the fuel in both reactor installations are briefly discussed.

  18. 40 CFR Appendix Viii to Part 600 - Fuel Economy Label Formats

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 29 2010-07-01 2010-07-01 false Fuel Economy Label Formats VIII... POLICY FUEL ECONOMY AND CARBON-RELATED EXHAUST EMISSIONS OF MOTOR VEHICLES Pt. 600, App. VIII Appendix VIII to Part 600—Fuel Economy Label Formats EC01MY92.117 EC01MY92.118 EC01MY92.119 EC01MY92.120...

  19. Theoretical analysis of nuclear reactors (Phase I), I-V, Part IV, Nuclear fuel depletion

    International Nuclear Information System (INIS)

    Pop-Jordanov, J.

    1962-07-01

    Nuclear fuel depletion is analyzed in order to estimate the qualitative and quantitative fuel property changes during irradiation and the influence of changes on the reactivity during long-term reactor operation. The changes of fuel properties are described by changes of neutron absorption and fission cross sections. Part one of this report covers the economic significance of fuel burnup and the review of fuel isotopic changes during depletion. Pat two contains the analysis of the U 235 chain, analytical expressions for the concentrations of U 235 , U 236 and Np 237 as a function of burnup. Part three contains the analysis of neutron spectrum influence on the Westcott method for calculating the cross sections. Part four contains the calculation method applied on Calder Hall type reactor. The results were obtained by applying ZUSE-22 R digital computer

  20. Fuels for homogeneous charge compression ignition (HCCI) engines. Automotive fuels survey. Part 6

    Energy Technology Data Exchange (ETDEWEB)

    Van Walwijk, M.

    2001-01-01

    Homogeneous charge compression ignition (HCCI) is a third mode of operation for internal combustion engines, beside spark ignition and conventional compression ignition. This report concentrates on the requirements that HCCI operation puts on fuels for these engines. For readers with limited time available, this summary describes the main findings. Policy makers that need some more background information may turn directly to chapter 7, 'Fuels for HCCI engines'. The rest of this report can be considered as a reference guide for more detailed information. The driving force to investigate HCCI engines is the potential of low emissions and simultaneously high energy efficiency. HCCI is gaining attention the last few years. However, HCCI engines are still in the research phase. After many experiments with prototype engines, people have now started working on computer simulations of the combustion process, to obtain a fundamental understanding of HCCI combustion and to steer future engine developments. In HCCI engines, an air/fuel mixture is prepared before it enters the combustion chamber. The homogeneous mixture is in the combustion chamber compressed to auto-ignition. Unlike in conventional engines, combustion starts at many different locations simultaneously and the speed of combustion is very high, so there is no flame front. Lean air/fuel mixtures (excess air) are used to control combustion speed. Because of the excess air, combustion temperature is relatively low, resulting in low NOx emissions. When the fuel is vaporised to a truly homogeneous mixture, complete combustion results in low particulate emissions. The most important advantages of HCCI engines are: - Emissions of NOx and particulates are very low. - Energy efficiency is high. It is comparable to diesel engines. - Many different fuels (one at a time) can be used in the HCCI concept. There are also some hurdles to overcome: - Controlling combustion is difficult, it complicates engine design

  1. 49 CFR 536.10 - Treatment of dual-fuel and alternative fuel vehicles-consistency with 49 CFR part 538.

    Science.gov (United States)

    2010-10-01

    ... vehicles-consistency with 49 CFR part 538. 536.10 Section 536.10 Transportation Other Regulations Relating... vehicles—consistency with 49 CFR part 538. (a) Statutory alternative fuel and dual-fuel vehicle fuel... economy in a particular compliance category by more than the limits set forth in 49 U.S.C. 32906(a), the...

  2. Safety analysis of RA reactor operation, I-III, Part III - Environmental effect of the maximum credible accident

    International Nuclear Information System (INIS)

    Raisic, N.

    1963-02-01

    Maximum credible accident at the RA reactor would consider release of fission products into the environment. This would result from fuel elements failure or meltdown due to loss of coolant. The analysis presented in this report assumes that the reactor was operating at nominal power at the moment of maximum possible accident. The report includes calculations of fission products activity at the moment of accident, total activity release during the accident, concentration of radioactive material in the air in the reactor neighbourhood, and the analysis of accident environmental effects

  3. Thermal sensation and comfort models for non-uniform and transient environments: Part III: whole-body sensation and comfort

    OpenAIRE

    Zhang, Hui; Arens, Edward; Huizenga, Charlie; Han, Taeyoung

    2009-01-01

    A three-part series presents the development of models for predicting the local thermal sensation (Part I) and local thermal comfort (Part II) of different parts of the human body, and also the whole-body sensation and comfort (Part III) that result from combinations of local sensation and comfort. The models apply to sedentary activities in a range of environments: uniform and non-uniform, stable and transient. They are based on diverse findings from the literature and from body-part-specifi...

  4. Nondestructive nuclear measurement in the fuel cycle. Part 2

    International Nuclear Information System (INIS)

    Lyoussi, A.

    2005-01-01

    Nondestructive measurement techniques are today widely used in practically all steps of the fuel cycle. This article is devoted to the presentation of the control and characterization needs and to the main active nondestructive nuclear methods used: 1 - main active nondestructive nuclear measurement methods: active neutronic measurement (needs and motivations, physical principle, measurement of delayed neutrons following a continuous irradiation, measurement of prompt neutrons (differential die-away technique - DDT), measurement of prompt and delayed neutrons (Sphincs method), neutronic method coupled to gamma spectroscopy), measurement by induced photo-fissions (needs and motivations, physical principle); 2 - conclusion. (J.S.)

  5. A structural modification of the two dimensional fuel behaviour analysis code FEMAXI-III with high-speed vectorized operation

    International Nuclear Information System (INIS)

    Yanagisawa, Kazuaki; Ishiguro, Misako; Yamazaki, Takashi; Tokunaga, Yasuo.

    1985-02-01

    Though the two-dimensional fuel behaviour analysis code FEMAXI-III has been developed by JAERI in form of optimized scalar computer code, the call for more efficient code usage generally arized from the recent trends like high burn-up and load follow operation asks the code into further modification stage. A principal aim of the modification is to transform the already implemented scalar type subroutines into vectorized forms to make the programme structure efficiently run on high-speed vector computers. The effort of such structural modification has been finished on a fair way to success. The benchmarking two tests subsequently performed to examine the effect of the modification led us the following concluding remarks: (1) In the first benchmark test, comparatively high-burned three fuel rods that have been irradiated in HBWR, BWR, and PWR condition are prepared. With respect to all cases, a net computing time consumed in the vectorized FEMAXI is approximately 50 % less than that consumed in the original one. (2) In the second benchmark test, a total of 26 PWR fuel rods that have been irradiated in the burn-up ranges of 13-30 MWd/kgU and subsequently power ramped in R2 reactor, Sweden is prepared. In this case the code is purposed to be used for making an envelop of PCI-failure threshold through 26 times code runs. Before coming to the same conclusion, the vectorized FEMAXI-III consumed a net computing time 18 min., while the original FEMAXI-III consumed a computing time 36 min. respectively. (3) The effects obtained from such structural modification are found to be significantly attributed to saving a net computing time in a mechanical calculation in the vectorized FEMAXI-III code. (author)

  6. Physiotherapy and low back pain - part iii: outcomes research utilising the biosychosocial model: psychosocial outcomes

    Directory of Open Access Journals (Sweden)

    L. D. Bardin

    2003-02-01

    has evolved that necessitates the use of a biopsychosocial model, focusing on illness rather than disease and incorporating the biological, psychological and social aspects that are important to understand and to study LBP in its chronic form. Traditional outcome measures that measure elements within the biological component are limited to assess the spectrum of impacts caused by chronic low back pain (CLBP and the validity, reliability and sensitivity of some of these measures has been questioned.Few physiologic tests of spine function are clinically meaningful to patients, objective physical findings can be absent, and in CLBP disability and activity intolerance are often disproportional to the original injury. Biological outcomes should be complemented by outcomes of the psychosocial aspects of back pain that measure the considerable functional and emotional impact on the quality of life of patients experiencing low back dysfunction. Outcomes research is an analysis of clinical practice as it actually occurs and can  make a valuable contribution to understanding the multidimensional impact of LBP. Psychosocial aspects of the biopsychosocial model for outcomes research are discussed in part III: functional status/disability, psychological impairment, patient satisfaction, health related quality of life

  7. A Structural Molar Volume Model for Oxide Melts Part III: Fe Oxide-Containing Melts

    Science.gov (United States)

    Thibodeau, Eric; Gheribi, Aimen E.; Jung, In-Ho

    2016-04-01

    As part III of this series, the model is extended to iron oxide-containing melts. All available experimental data in the FeO-Fe2O3-Na2O-K2O-MgO-CaO-MnO-Al2O3-SiO2 system were critically evaluated based on the experimental condition. The variations of FeO and Fe2O3 in the melts were taken into account by using FactSage to calculate the Fe2+/Fe3+ distribution. The molar volume model with unary and binary model parameters can be used to predict the molar volume of the molten oxide of the Li2O-Na2O-K2O-MgO-CaO-MnO-PbO-FeO-Fe2O3-Al2O3-SiO2 system in the entire range of compositions, temperatures, and oxygen partial pressures from Fe saturation to 1 atm pressure.

  8. Technical realisation of the VISA-3 project, Parts I-II, Part I; Tehnicka realizacija projekta VISA-3, I-II deo, I Deo

    Energy Technology Data Exchange (ETDEWEB)

    Pavicevic, M; Smokovic, Z [Institute of Nuclear Sciences Boris Kidric, Odeljenje za reaktorsku eksperimentalnu tehniku, Vinca, Beograd (Serbia and Montenegro)

    1966-11-15

    This task is related to irradiation of reactor materials (steel, Al, MgO, Al{sub 2}O{sub 3}, ets.) at higher temperatures (200-500 deg C) in the fast neutron flux. These conditions would be more realistic to real reactor conditions than the conditions achieved within VISA-2 project. The experimental space will be the same as in VISA-2 project, i.e. refurbished reactor channels and within the fuel elements. The irradiation capsule will be leak tight with thermal isolation layer and supplied with electric heater to enable temperature variation.

  9. Prototypical spent nuclear nuclear fuel rod consolidation equipment, Phase 2: Final design report: Volume 2, Appendices: Part 1

    International Nuclear Information System (INIS)

    Ciez, A.P.

    1987-01-01

    The purpose of this specification is to establish functional and design requirements for the Prototypical Spent Nuclear Fuel Rod Consolidation System. The Department of Energy-Idaho Operations Office (DOE-ID) is responsible for the implementation of the Prototypic Dry Rod Consolidation Demonstration Project. This program is to develop and demonstrate a fully qualified, licensable, cost-effective, dry spent fuel rod consolidation system by July 1989. The work is divided into four phases as follows: Phase I--Preliminary Design, Phase II--Final Design Option, Phase III--Fabrication and System Checkout Option, and Phase IV--Installation and Hot Demonstration Option. This specification is part of the Phase II effort. The objectives of this specification are to provide functional and design requirements for the Prototypical Spent Nuclear Fuel Rod Consolidation equipment; establish specific tool and subsystem requirements such that the integrated and overall system requirements are satisfied; and establish positioning, envelope and size interface control requirements for each tool or subsystem such that the individual components will interface properly with the overall system design

  10. Acuity and case management: a healthy dose of outcomes, part III.

    Science.gov (United States)

    Huber, Diane L; Craig, Kathy

    2007-01-01

    This is the third of a 3-part series presenting 2 effective applications--acuity and dosage--that describe how the business case for case management (CM) can be made. In Part I, dosage and acuity concepts were explained as client need-severity, CM intervention-intensity, and CM activity-dose prescribed by amount, frequency, duration, and breadth of activities. Concepts were presented that related the practice of CM to the use of evidence-based practice (EBP), knowledge, and methods and the development of instruments that measure and score pivotal CM actions. Part I also featured a specific exemplar, the CM Acuity Tool, and described how to use acuity to identify and score the complexity of a CM case. Part II further explained dosage and 2 acuity instruments, the Acuity Tool and AccuDiff. Part III presents linkage to EBP and practical applications. The information contained in the 3-part series applies to all CM practice settings and contains ideas and recommendations useful to CM generalists, specialists, supervisors, and business and outcomes managers. The Acuity Tools Project was developed from frontline CM practice in one large, national telephonic CM company. Dosage: A literature search failed to find research into dosage of a behavioral intervention. The Huber-Hall model was developed and tested in a longitudinal study of CM models in substance abuse treatment and reported in the literature. Acuity: A structured literature search and needs assessment launched the development of the suite of acuity tools. A gap analysis identified that an instrument to assign and measure case acuity specific to CM activities was needed. Clinical experts, quality specialists, and business analysts (n = 7) monitored the development and testing of the tools, acuity concepts, scores, differentials, and their operating principles and evaluated the validity of the acuity tools' content related to CM activities. During the pilot phase of development, interrater reliability testing of

  11. Investigation of the Stress Intensity Limits of ASME Section III Div.5 for Structure Design Criteria of SFR Fuel Assembly

    Energy Technology Data Exchange (ETDEWEB)

    Choo, Jin-Yup; Kim, Hyung-Kyu; Cheon, Jin-Sik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    These affect the mechanical design of the fuel assembly components. And thus, appropriate structural design criteria should also be chosen to incorporate the specific design conditions of the SFR fuel assemblies. Among them, the temperature is one of the most crucial conditions to be concerned because the sodium coolant temperature is normally more than 500ºC which is much higher than that of the LWR (< 350ºC). This implies that a thermal creep should be significantly considered in the SFR fuel assembly mechanical design. In addition to the high temperature condition, an irradiation swelling is also an important behavior that the SFR fuel assembly material should accommodate. To incorporate the temperature and irradiation impacts, the material of the fuel assembly components is presently determined to be made of HT-9, the ferriticmartensitic steel. In this paper, the ASME Sec. III Div. 5 (referred to as ‘Div. 5’ hereinafter), which was developed for a ‘high temperature reactor’, is considered as one of the structural design criteria for the mechanical design of SFR fuel assemblies. In this paper, the stress intensity limits, S{sub m} and S{sub t} of HT-9 were built for the structural criteria of an SFR fuel assembly. S{sub m} is obtained from the ultimate strength. As for S{sub t}, it is more complicated because of its dependency of time duration in addition to temperature. Following the definition of S{sub mt}, the method in the ASME Sec. III Div. 1, Subsec. NH was consulted. We found that the Sm is adopted as S{sub mt} under the temperature about 470ºC which is relatively low temperature range and over 470ºC with relatively short time duration as 1000 hours. And the S{sub t} is adopted as Smt at over 470ºC and long time duration over 34800 hours, and over 520ºC and 10{sup 4} hours too. And at over 570ºC and 1000 hours, and at over 630ºC and 100 hours, S{sub t} is also adopted for S{sub mt}.

  12. Design and construction of the SIPPING for fuels of the TRIGA Mark III reactor

    International Nuclear Information System (INIS)

    Castaneda J, G.; Delfin L, A.; Alvarado P, R.; Mazon R, R.; Ortega V, B.

    2003-01-01

    The sipping technique, it has been used by several possessors of nuclear research reactors in its irradiated nuclear fuels, likewise in some fuel storage sites, with the objective of to determine the quantity of radioactivity that the fuel liberates in the means in that it is. The irradiated fuel in storage of some nuclear research reactors, its can have cracks that cross the cladding of the same one, generating the liberation of fission products that its need to determine to maintain safety measures appropriate as much as the fuel as of the facilities where they are. It doesn't exist until now, some method published for the non destructive sipping test technique. Based on that described, the Reactor Department of the National Institute of Nuclear Research, it has designed and built an inspection system of irradiated fuel that it will allow the detection of gassy fission products in site, and solids by means of the measurement of the activity of the Cs-137 contained in water samples. (Author)

  13. Alternate-Fueled Combustor-Sector Performance. Parts A and B; (A) Combustor Performance; (B) Combustor Emissions

    Science.gov (United States)

    Shouse, D. T.; Hendricks, R. C.; Lynch, A.; Frayne, C. W.; Stutrud, J. S.; Corporan, E.; Hankins, T.

    2012-01-01

    Alternate aviation fuels for military or commercial use are required to satisfy MIL-DTL-83133F(2008) or ASTM D 7566 (2010) standards, respectively, and are classified as "drop-in" fuel replacements. To satisfy legacy issues, blends to 50% alternate fuel with petroleum fuels are certified individually on the basis of processing and assumed to be feedstock agnostic. Adherence to alternate fuels and fuel blends requires "smart fueling systems" or advanced fuel-flexible systems, including combustors and engines, without significant sacrifice in performance or emissions requirements. This paper provides preliminary performance (Part A) and emissions and particulates (Part B) combustor sector data. The data are for nominal inlet conditions at 225 psia and 800 F (1.551 MPa and 700 K), for synthetic-paraffinic-kerosene- (SPK-) type (Fisher-Tropsch (FT)) fuel and blends with JP-8+100 relative to JP-8+100 as baseline fueling. Assessments are made of the change in combustor efficiency, wall temperatures, emissions, and luminosity with SPK of 0%, 50%, and 100% fueling composition at 3% combustor pressure drop. The performance results (Part A) indicate no quantifiable differences in combustor efficiency, a general trend to lower liner and higher core flow temperatures with increased FT fuel blends. In general, emissions data (Part B) show little differences, but with percent increase in FT-SPK-type fueling, particulate emissions and wall temperatures are less than with baseline JP-8. High-speed photography illustrates both luminosity and combustor dynamic flame characteristics.

  14. Alternate-Fueled Combustor-Sector Performance—Part A: Combustor Performance and Part B: Combustor Emissions

    OpenAIRE

    Shouse, D. T.; Neuroth, C.; Hendricks, R. C.; Lynch, A.; Frayne, C. W.; Stutrud, J. S.; Corporan, E.; Hankins, Capt. T.

    2012-01-01

    Alternate aviation fuels for military or commercial use are required to satisfy MIL-DTL-83133F or ASTM D 7566 standards, respectively, and are classified as “drop-in’’ fuel replacements. To satisfy legacy issues, blends to 50% alternate fuel with petroleum fuels are acceptable. Adherence to alternate fuels and fuel blends requires “smart fueling systems’’ or advanced fuel-flexible systems, including combustors and engines, without significant sacrifice in performance or emissions requirements...

  15. Post-Irradiation Examination Test of the Parts of X-Gen Nuclear Fuel Assembly

    International Nuclear Information System (INIS)

    Ahn, S. B.; Ryu, W. S.; Choo, Y. S.

    2008-08-01

    The mechanical properties of the parts of a nuclear fuel assembly are degraded during the operation of the reactor, through the mechanism of irradiation damage. The properties changes of the parts of the fuel assembly should be quantitatively estimated to ensure the safety of the fuel assembly and rod during the operation. The test techniques developed in this report are used to produce the irradiation data of the grid 1x1 cell spring, the grid 1x1 cell, the spring on one face of the 1x1 cell, the inner/outer strip of the grid and the welded part. The specimens were irradiated in the CT test hole of HANARO of a 30 MW thermal output at 300 deg. C during about 100 days From the spring test of mid grid 1x1 cell and grid plate, the irradiation effects can be examined. The irradiation effects on the irradiation growth also were occurred. The buckling load of mid grid 1x1 cell does not change with a neutron irradiation. From the tensile tests, the strengths increased but the elongations decreased due to an irradiation. The tensile test and microstructure examination of the spot and fillet welded parts are performed for the evaluation of an irradiation effects. Through these tests of components, the essential data on the fuel assembly design could be obtained. These results will be used to update the irradiation behavior databases, to improve the performance of fuel assembly, and to predict the service life of the fuel assembly in a reactor

  16. BEHAVE: fire behavior prediction and fuel modeling system-BURN Subsystem, part 1

    Science.gov (United States)

    Patricia L. Andrews

    1986-01-01

    Describes BURN Subsystem, Part 1, the operational fire behavior prediction subsystem of the BEHAVE fire behavior prediction and fuel modeling system. The manual covers operation of the computer program, assumptions of the mathematical models used in the calculations, and application of the predictions.

  17. Providing for energy efficiency in homes and small buildings. Part III. Determining which practices are most effective and installing materials

    Energy Technology Data Exchange (ETDEWEB)

    None

    1980-06-01

    The training program is designed to educate students and individuals in the importance of conserving energy and to provide for developing skills needed in the application of energy-saving techniques that result in energy-efficient buildings. A teacher guide and student workbook are available to supplement the basic manual. Subjects covered in Part III are: determining which practices are most efficient and economical; installing energy-saving materials; and improving efficiency of equipment.

  18. Technical Reports (Part I). End of Project Report, 1968-1971, Volume III.

    Science.gov (United States)

    Western Nevada Regional Education Center, Lovelock.

    The pamphlets included in this volume are technical reports prepared as outgrowths of the Student Information Systems of the Western Nevada Regional Education Center (WN-REC) funded by a Title III (Elementary and Secondary Education Act) grant. These reports describe methods of interpreting the printouts from the Student Information System;…

  19. Fabrication and testing of ceramic UO2 fuel - I-III. Part I

    International Nuclear Information System (INIS)

    Novakovic, M.

    1961-12-01

    The task described consists of the following: fabrication of UO 2 with different granulation from uranyl nitrate by ammonia diuranate; determination of size and shape distributions of metal and ceramic powders; fabrication of sintered pressed samples UO 2 ; investigating the properties of sintered uranium dioxide dependent on the fabrication process; producing a vibrator for compacting UO 2 powder. This volume includes reports on the first two tasks

  20. IMPROVEMENT OF PERFORMANCE OF DUAL FUEL ENGINE OPERATED AT PART LOAD

    Directory of Open Access Journals (Sweden)

    N. Kapilan

    2010-12-01

    Full Text Available Rising petroleum prices, an increasing threat to the environment from exhaust emissions, global warming and the threat of supply instabilities has led to the choice of inedible Mahua oil (MO as one of the main alternative fuels to diesel oil in India. In the present work, MO was converted into biodiesel by transesterification using methanol and sodium hydroxide. The cost of Mahua oil biodiesel (MOB is higher than diesel. Hence liquefied petroleum gas (LPG, which is one of the cheapest gaseous fuels available in India, was fumigated along with the air to reduce the operating cost and to reduce emissions. The dual fuel engine resulted in lower efficiency and higher emissions at part load. Hence in the present work, the injection time was varied and the performance of the dual fuel engine was studied. From the engine tests, it is observed that an advanced injection time results in higher efficiency and lower emissions. Hence, advancing the injection timing is one of the ways of increasing the efficiency of LPG+MOB dual fuel engine operated at part load.

  1. Material control in nuclear fuel fabrication facilities. Part I. Fuel descriptions and fabrication processes, P.O. 1236909 Final report

    International Nuclear Information System (INIS)

    Borgonovi, G.M.; McCartin, T.J.; Miller, C.L.

    1978-12-01

    The report presents information on foreign nuclear fuel fabrication facilities. Fuel descriptions and fuel fabrication information for three basic reactor types are presented: The information presented for LWRs assumes that Pu--U Mixed Oxide Fuel (MOX) will be used as fuel

  2. Operation and maintenance of the RA reactor in 1964, I-II, Part I

    International Nuclear Information System (INIS)

    Pavicevic, M.

    1964-12-01

    During 1964, the Reactor as operated about 20 days each months at nominal power of 6.5 MW, 5 days at lower power levels and 5 days were used for maintenance. Total production was 27930 MWh which is 11.7% higher than the planned value. Fuel exchange was done 3 times during this period, 98 spent fuel channels were exchanged. In addition to routine maintenance of reactor components and instruments a series of analyses of heavy water and helium were done. Special attention was devoted to corrosion analyses of the reactor materials because of the heavy water system was refurbished decontaminated in 1963. Utilization of the experimental space in the reactor was better that previously. 546 samples were irradiated till the end of November, of which 443 for users from the Institute. Specific irradiations in the fast neutron flux were done in six VISA-2 channels in the core

  3. Theoretical analysis of nuclear reactors (Phase I), I-V, Part III, Reactor poisoning

    International Nuclear Information System (INIS)

    Pop-Jordanov, J.

    1962-07-01

    Method was developed for calculation of Xe 135 static effect and kinetic effects of Xe 135 and Sm 149 with separate treatment of iodine effect and influence of reactor poisoning during power increase. Mentioned effects are treated first for uranium fuel and then the basic formulae were generalized for a mixture of fissile material. The annex contains a table with data needed for calculations and the Xe 13 5 microscopic capture cross section dependent on temperature [sr

  4. Fracture toughness of A533B Part III - variability of A533B fracture toughness as determined from Charpy data

    International Nuclear Information System (INIS)

    Druce, S.G.; Eyre, B.L.

    1978-08-01

    This is the final part of a series of three reports examining the upper shelf fracture toughness of A533B Class 1 pressure vessel steel. Part I (AERE R 8968) critically reviews the current elasto plastic fracture mechanics methodologies employed to characterise toughness following extensive yielding and Part II (AERE R 8969) examines several sources of fracture mechanics data pertinent to A533B Class 1 in the longitudinal (RW) orientation. Part III is a review of the effects of (i) position and orientation within the plate (ii) welding processes and post weld heat treatment and (iii) neutron irradiation as measured by Charpy impact testing. It is concluded that the upper shelf factor energy is dependent on orientation and position and can be reduced by welding, extended post weld heat treatments and neutron irradiation. Neutron irradiation effects are known to be strongly dependent on composition and metallurgical conditions, but an explanation for the variability following extended post weld treatments has yet to be resolved. (author)

  5. Thermal sensation and comfort models for non-uniform and transient environments, part III: Whole-body sensation and comfort

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Hui; Arens, Edward; Huizenga, Charlie [Center for the Built Environment, UC Berkeley (United States); Han, Taeyoung [General Motors Company (United States)

    2010-02-15

    A three-part series presents the development of models for predicting the local thermal sensation (Part I) and local thermal comfort (Part II) of different parts of the human body, and also the whole-body sensation and comfort (Part III) that result from combinations of local sensation and comfort. The models apply to sedentary activities in a range of environments: uniform and non-uniform, stable and transient. They are based on diverse findings from the literature and from body-part-specific human subject tests in a climate chamber. They were validated against a test of automobile passengers. The series is intended to present the models' rationale, structure, and coefficients, so that others can test them and develop them further as additional empirical data becomes available. A) The whole-body (overall) sensation model has two forms, depending on whether all of the body's segments have sensations effectively in the same direction (e.g warm or cool), or whether some segments have sensations opposite to those of the rest of the body. For each, individual body parts have different weights for warm versus cool sensations, and strong local sensations dominate the overall sensation. If all sensations are near neutral, the overall sensation is close to the average of all body sensations. B) The overall comfort model also has two forms. Under stable conditions, people evaluate their overall comfort by a complaint-driven process, meaning that when two body parts are strongly uncomfortable, no matter how comfortable the other body parts might be, the overall comfort will be near the discomfort level of the two most uncomfortable parts. When the environmental conditions are transient, or people have control over their environments, overall comfort is better than that of the two most uncomfortable body parts. This can be accounted for by adding the most comfortable vote to the two most uncomfortable ones. (author)

  6. Economic Analysis of Symbiotic Light Water Reactor/Fast Burner Reactor Fuel Cycles Proposed as Part of the U.S. Advanced Fuel Cycle Initiative (AFCI)

    International Nuclear Information System (INIS)

    Williams, Kent Alan; Shropshire, David E.

    2009-01-01

    A spreadsheet-based 'static equilibrium' economic analysis was performed for three nuclear fuel cycle scenarios, each designed for 100 GWe-years of electrical generation annually: (1) a 'once-through' fuel cycle based on 100% LWRs fueled by standard UO2 fuel assemblies with all used fuel destined for geologic repository emplacement, (2) a 'single-tier recycle' scenario involving multiple fast burner reactors (37% of generation) accepting actinides (Pu,Np,Am,Cm) from the reprocessing of used fuel from the uranium-fueled LWR fleet (63% of generation), and (3) a 'two-tier' 'thermal+fast' recycle scenario where co-extracted U,Pu from the reprocessing of used fuel from the uranium-fueled part of the LWR fleet (66% of generation) is recycled once as full-core LWR MOX fuel (8% of generation), with the LWR MOX used fuel being reprocessed and all actinide products from both UO2 and MOX used fuel reprocessing being introduced into the closed fast burner reactor (26% of generation) fuel cycle. The latter two 'closed' fuel cycles, which involve symbiotic use of both thermal and fast reactors, have the advantages of lower natural uranium requirements per kilowatt-hour generated and less geologic repository space per kilowatt-hour as compared to the 'once-through' cycle. The overall fuel cycle cost in terms of $ per megawatt-hr of generation, however, for the closed cycles is 15% (single tier) to 29% (two-tier) higher than for the once-through cycle, based on 'expected values' from an uncertainty analysis using triangular distributions for the unit costs for each required step of the fuel cycle. (The fuel cycle cost does not include the levelized reactor life cycle costs.) Since fuel cycle costs are a relatively small percentage (10 to 20%) of the overall busbar cost (LUEC or 'levelized unit electricity cost') of nuclear power generation, this fuel cycle cost increase should not have a highly deleterious effect on the competitiveness of nuclear power. If the reactor life cycle

  7. Direct disposal of spent fuel. Simulation of shaft transport. Phase III. Final report

    International Nuclear Information System (INIS)

    Filbert, W.; Heda, M.; Khamis, M.; Neydek, J.; Niehues, N.; Rissel, J.; Schrimpf, C.; Weber, W.; Fuchs, D.; Gerlach, A.; Langebrake, F.; Sindern, W.; Gasch, A.; Leicht, R.; Schwab, B.; Hecke, R. van; Kipka, P.; Simmich, K.; Weber, H.

    1994-01-01

    The aim of these demonstration tests was to verify the technical feasibility of a shaft hoisting equipment with a payload of 85 t as well as the safe transport of POLLUX-casks. In phase III of the project the components and the test stand were built, their proper functioning and reliability were tested to demonstrate that they are state-of-the-art. The following additional investigations were carried out: - Tests to fix operational disturbances and simulation tests for the new components to demonstrate their licensibility -Selection and lifetime tests of ropes and investigation of the present state-of-the-art of rope slighting under the conditions of the conceptual design of the shaft hoisting facility - Execution of a probabilistic safety analysis e.g. the determination of the release of radioactive material (result: probability [de

  8. Characterization of Delayed-Particle Emission Signatures for Pyroprocessing. Part 1: ABTR Fuel Assembly.

    Energy Technology Data Exchange (ETDEWEB)

    Durkee, Jr., Joe W. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-06-19

    A three-part study is conducted using the MCNP6 Monte Carlo radiation-transport code to calculate delayed-neutron (DN) and delayed-gamma (DG) emission signatures for nondestructive assay (NDA) metal-fuel pyroprocessing. In Part 1, MCNP6 is used to produce irradiation-induced used nuclear fuel (UNF) isotopic inventories for an Argonne National Laboratory (ANL) Advanced Burner Test Reactor (ABTR) preconceptual design fuel assembly (FA) model. The initial fuel inventory consists of uranium mixed with light-water-reactor transuranic (TRU) waste and 10 wt% zirconium (U-LWR-SFTRU-10%Zr). To facilitate understanding, parametric evaluation is done using models for 3% and 5% initial 235U a% enrichments, burnups of 5, 10, 15, 20, 30, …, 120 GWd/MTIHM, and 3-, 5-, 10-, 20-, and 30- year cooling times. Detailed delayed-particle radioisotope source terms for the irradiate FA are created using BAMF-DRT and SOURCES3A. Using simulation tallies, DG activity ratios (DGARs) are developed for 134Cs/137Cs 134Cs/154Eu, and 154Eu/137Cs markers as a function of (1) burnup and (2) actinide mass, including elemental uranium, neptunium, plutonium, americium, and curium. Spectral-integrated DN emission is also tallied. The study reveals a rich assortment of DGAR behavior as a function of DGAR type, enrichment, burnup, and cooling time. Similarly, DN emission plots show variation as a function of burnup and of actinide mass. Sensitivity of DGAR and DN signatures to initial 235U enrichment, burnup, and cooling time is evident. Comparisons of the ABTR radiation signatures and radiation signatures previously reported for a generic Westinghouse oxide-fuel assembly indicate that there are pronounced differences in the ABTR and Westinghouse oxide-fuel DN and DG signatures. These differences are largely attributable to the initial TRU inventory in the ABTR fuel. The actinide and nonactinide inventories for the

  9. Irradiation of Parts of the X-Gen Nuclear Fuel Assembly made by KNF in HANARO

    International Nuclear Information System (INIS)

    Choo, K. N.; Cho, M. S.; Shin, Y. T.; Kim, B. G.; Lee, S. H.; Eom, K. B.

    2008-01-01

    An instrumented capsule has been developed at HANARO (High flux Advanced Neutron Application ReactOr) for the neutron irradiation tests of materials. The capsule system has been actively utilized for the various material irradiation tests requested by users from research institutes, universities, and the industries. As a preliminary test, some specimens made of the parts of a nuclear fuel assembly were inserted in the 05M-07U instrumented capsule and successfully irradiated at HANARO. Based on the results and experience, a new irradiation capsule of 07M-13N was designed, fabricated, and irradiated at HANARO for the evaluation of the neutron irradiation properties of the parts of the X-Gen nuclear fuel assembly made by KNF (Korea Nuclear Fuel). Specimens such as bucking and spring test specimens of spacer grid, microstructure and tensile test specimens of welded parts, tensile, irradiation growth and spring test specimens made of HANA tube, Zirlo, Zircaloy-4 and Inconel-718 were placed in the capsule. The capsule was loaded into the CT test hole of HANARO of a 30MW thermal output and the specimens were irradiated at 295 - 460 .deg. C up to a fast neutron fluence of 1.2x10 21 (n/cm 2 ) (E>1.0MeV)

  10. PIO I-II tendencies. Part 2. Improving the pilot modeling

    Directory of Open Access Journals (Sweden)

    Ioan URSU

    2011-03-01

    Full Text Available The study is conceived in two parts and aims to get some contributions to the problem ofPIO aircraft susceptibility analysis. Part I, previously published in this journal, highlighted the mainsteps of deriving a complex model of human pilot. The current Part II of the paper considers a properprocedure of the human pilot mathematical model synthesis in order to analyze PIO II typesusceptibility of a VTOL-type aircraft, related to the presence of position and rate-limited actuator.The mathematical tools are those of semi global stability theory developed in recent works.

  11. Active Control of Low-Speed Fan Tonal Noise Using Actuators Mounted in Stator Vanes: Part III Results

    Science.gov (United States)

    Sutliff, Daniel L.; Remington, Paul J.; Walker, Bruce E.

    2003-01-01

    A test program to demonstrate simplification of Active Noise Control (ANC) systems relative to standard techniques was performed on the NASA Glenn Active Noise Control Fan from May through September 2001. The target mode was the m = 2 circumferential mode generated by the rotor-stator interaction at 2BPF. Seven radials (combined inlet and exhaust) were present at this condition. Several different error-sensing strategies were implemented. Integration of the error-sensors with passive treatment was investigated. These were: (i) an in-duct linear axial array, (ii) an induct steering array, (iii) a pylon-mounted array, and (iv) a near-field boom array. The effect of incorporating passive treatment was investigated as well as reducing the actuator count. These simplified systems were compared to a fully ANC specified system. Modal data acquired using the Rotating Rake are presented for a range of corrected fan rpm. Simplified control has been demonstrated to be possible but requires a well-known and dominant mode signature. The documented results here in are part III of a three-part series of reports with the same base title. Part I and II document the control system and error-sensing design and implementation.

  12. EFSUMB Guidelines on Interventional Ultrasound (INVUS), Part III - Abdominal Treatment Procedures (Long Version)

    DEFF Research Database (Denmark)

    Dietrich, Christoph F; Lorentzen, T.; Appelbaum, L.

    2016-01-01

    The third part of the European Federation of Societies for Ultrasound in Medicine and Biology (EFSUMB) Guidelines on Interventional Ultrasound (INVUS) assesses the evidence for ultrasound-guided and assisted interventions in abdominal treatment procedures. Recommendations for clinical practice ar...

  13. EFSUMB Guidelines on Interventional Ultrasound (INVUS), Part III - Abdominal Treatment Procedures (Short Version)

    DEFF Research Database (Denmark)

    Dietrich, Christoph F; Lorentzen, T.; Appelbaum, L.

    2016-01-01

    The third part of the European Federation of Societies for Ultrasound in Medicine and Biology (EFSUMB) Guidelines on Interventional Ultrasound assesses the evidence for ultrasound-guided and assisted interventions in abdominal treatment procedures. Recommendations for clinical practice are presen...

  14. RA reactor safety analysis I-III, Part III - Environmental effect of the maximum credible accident; Analiza sigurnosti rada Reaktora RA I-III, III deo - Posledica maksimalno moguceg akcidenta na okolinu reaktora

    Energy Technology Data Exchange (ETDEWEB)

    Raisic, N [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1963-02-15

    The objective of the maximum credible accident analysis was to determine the integral radiation doses in the vicinity of the reactor and in the environment. In case of RA reactor the maximum credible accident, meaning release of the fission products, would be caused by fuel elements meltdown. This analysis includes the following calculation results: activity of the fission products, volatility of the fission products, concentration of radioactive materials in the air, analysis of the accident environmental effects.

  15. Antibacterianos de acción sistémica: Parte III. Sulfonamidas y tetraciclinas

    Directory of Open Access Journals (Sweden)

    Manuel Cué Brugueras

    1999-04-01

    Full Text Available Resumen: Se presenta la tercera parte de una revisión bibliográfica sobre los antibacterianos de elección, en la cual se abordan los grupos sulfonamidas y tetraciclinas; además, se incluye un cuadro resumen con los antibióticos tratados en las tres partes, así como sus vías de administración y nombres comerciales que se utilizan en Cuba. Se hacen algunas consideraciones sobre la manera de enfrentar la gran variedad de antibióticos y el costo de la antibióticoterapia por parte de los países en vías de desarrollo, tomando como referencia algunas recomendaciones hechas por la OPS/OMSSummary: The third part of a bibliographic review on the elective antibacterials in which the groups of sulfonamides and tetracyclines are approached is presented. It is also included a summary picture with the antibiotics dealt with in the three parts, the routes of administration, and the trade names used in Cuba. Some considerations are made on the way to face the wide range of antibiotics and the cost of antibiotic therapy in the developing countries, taking into account some recommendations made by the PAHO/WHO

  16. Transport properties of gaseous ions over a wide energy range. Part III

    International Nuclear Information System (INIS)

    Ellis, H.W.; Thackston, M.G.; McDaniel, E.W.; Mason, E.A.

    1984-01-01

    This paper updates and extends in scope our two previous papers entitled ''Transport Properties of Gaseous Ions over a Wide Energy Range.'' The references to the earlier publications (referred to as ''Part I'' and ''Part II'') are I, H. W. Ellis, R. Y. Pai, E. W. McDonald, E. A. Mason, and L. A. Viehland, ATOMIC DATA AND NUCLEAR DATA TABLES 17, 177--210 (19876); and II, H. W. Ellis, E. W. McDaniel, D. L. Albritton, L. A. Veihland, S. L. Lin, and E. A. Mason, ATOMIC DATA AND NUCLEAR DATA TABLES 22, 179--217 (1978). Parts I and II contained compilations of experimental data on ionic mobilities and diffusion coefficients (both longitudinal and transverse) for ions in neutral gase (almost exclusively at room temperature) in an externally applied electric field

  17. Theoretical analysis of nuclear reactors (Phase III), I-V, Part III, Reactor poisoning; Razrada metoda teorijske analize nuklearnih reaktora (III faza) I-IV, III Deo, Zatrovanje reaktora

    Energy Technology Data Exchange (ETDEWEB)

    Pop-Jordanov, J [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1963-01-15

    Report on calculation of poisoning in experimental and power reactor includes four parts. Part one describes the influence of poisoning on the physical parameters of a reactor. part two includes transformation of differential equations for iodine and xenon. It was needed for easier solution of of differential equation using the analog computer. This calculation was done for RA reactor operating at 5 MW power. The RA reactor was used an example of calculation by the proposed method. Part four shows the application of the method for calculating the Calder Hall power reactor.

  18. Tensile Test of Welding Joint Parts for a Plate-type Fuel Assembly

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, K. H.; Kim, J. Y.; Kim, H. J.; Yim, J. S. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-05-15

    The tensile tests were performed using an INSTRON 4505 (universal tensile) testing machine. These welding joints are composed of two parts for the soundness of the fuel assembly; one is the side plate with a fixing bar and the other is a side plate with an end fitting. These two joint parts are fabricated by TIG welding method. The tensile tests of the welding joints of a plate-type FA are executed by a tensile test. The fixture configurations for the specimen are very important to obtain the strict test results. The maximum strength has an approximately linear correlation with the unit bonding length of the welding joints. In spite of these results, the maximum strengths of the welding joints are satisfied according to the minimum requirement. These tensile tests of the joint parts for a plate-type fuel assembly (FA) have to be executed to evaluate the structural strength. For the tensile test, the joint parts of a FA used in the test are made of aluminum alloy (Al6061-T6)

  19. Tensile Test of Welding Joint Parts for a Plate-type Fuel Assembly

    International Nuclear Information System (INIS)

    Yoon, K. H.; Kim, J. Y.; Kim, H. J.; Yim, J. S.

    2013-01-01

    The tensile tests were performed using an INSTRON 4505 (universal tensile) testing machine. These welding joints are composed of two parts for the soundness of the fuel assembly; one is the side plate with a fixing bar and the other is a side plate with an end fitting. These two joint parts are fabricated by TIG welding method. The tensile tests of the welding joints of a plate-type FA are executed by a tensile test. The fixture configurations for the specimen are very important to obtain the strict test results. The maximum strength has an approximately linear correlation with the unit bonding length of the welding joints. In spite of these results, the maximum strengths of the welding joints are satisfied according to the minimum requirement. These tensile tests of the joint parts for a plate-type fuel assembly (FA) have to be executed to evaluate the structural strength. For the tensile test, the joint parts of a FA used in the test are made of aluminum alloy (Al6061-T6)

  20. The energy balance experiment EBEX-2000. Part III: Behaviour and quality of the radiation measurements

    NARCIS (Netherlands)

    Kohsiek, W.; Liebethal, C.; Foken, T.; Vogt, R.; Oncley, S.P.; Bernhofer, C.; Debruin, H.A.R.

    2007-01-01

    An important part of the Energy Balance Experiment (EBEX-2000) was the measurement of the net radiation and its components. Since the terrain, an irrigated cotton field, could not be considered homogeneous, radiation measurements were made at nine sites using a variety of radiation instruments,

  1. Broadcasting Stations of the World; Part III. Frequency Modulation Broadcasting Stations.

    Science.gov (United States)

    Foreign Broadcast Information Service, Washington, DC.

    This third part of "Broadcasting Stations of the World", which lists all reported radio broadcasting and television stations, with the exception of those in the United States which broadcast on domestic channels, covers frequency modulation broadcasting stations. It contains two sections: one indexed alphabetically by country and city, and the…

  2. CBIOS Science Sessions - 2016 - Part I and III National Symposium on Nanoscience and Biomedical Nanotechnology - Proceedings

    Directory of Open Access Journals (Sweden)

    L. Monteiro Rodrigues, et al.

    2016-05-01

    Full Text Available CBiOS Science Sessions - 2016 – Part 1 New methods to explore efficacy and safety of natural origin products; Stefânia Duz Delsin Effectiveness of Hypopressive Exercises in women with pelvic floor dysfunctions; Beatriz Navarro Brazález Indoor air quality in baby rooms: a study about VOC levels; Raquel Rodrigues dos Santos, Ana Sofia Fernandes e Liliana Mendes A medicinal chemistry approach for the development of novel anti-tumor agentes; Maria M. M. Santos Isolation, modelling and phytosome forms of antibacterial and anti-proliferative compounds from Plectranthus spp; Diogo Matias Intellectual Property – Patenting Propriedade Intelectual – Patenteamento Rui Gomes Biomarkers in wastewater; Álvaro Lopes A Contribution for a Better Comprehension of Donkey Dentistry: the Importance of Dental Care; João Brandão Rodrigues Characterization of Lusitano’s Pure Blood Pressure Centers using two pressure plates; Pequito M.; Gomes-Costa M.; Prazeres J.; Bragança M.; Roupa I.; Fonseca R.G.; Abrantes J. Application of photoplethysmography to monitor heart rate in dogs; Rui Assunção, Henrique Silva, João Requicha, Luis Lobo, Luis Monteiro Rodrigues Looking into the oscillatory properties of the laser Doppler flowmetry signal in human microcirculation; Henrique Silva, Hugo Ferreira, Marie-Ange Renault, Alain-Pierre Gadeau, Julia Buján, LM Rodrigues III Symposium of Nanoscience and Biomedical Nanotechnology – Proceedings April 15/04/2016 Lisboa - Universidade Lusófona Honor Commitee /Comissão de Honra Magnífico Reitor da Universidade Lusófona, Mário Moutinho Presidente do Conselho de Administração da Universidade Lusófona, Manuel de Almeida Damásio Sr. Bastonário da Ordem dos Engenheiros, Carlos Matias Ramos Sr. Bastonário da Ordem dos Médicos, José Silva Vice-presidente do Conselho de Enfermagem, Maria José Costa Dias Presidente da Associação Nacional de Farmácias, Paulo Cleto Duarte Presidente da Sociedade Portuguesa de Ci

  3. Safety analysis of RA reactor operation, I-III, Part III - Environmental effect of the maximum credible accident; Analiza sigurnosti rada reaktora RA - I-III, III deo - Posledica maksimalno moguceg akcidenta na okolinu reaktora

    Energy Technology Data Exchange (ETDEWEB)

    Raisic, N [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1963-02-15

    Maximum credible accident at the RA reactor would consider release of fission products into the environment. This would result from fuel elements failure or meltdown due to loss of coolant. The analysis presented in this report assumes that the reactor was operating at nominal power at the moment of maximum possible accident. The report includes calculations of fission products activity at the moment of accident, total activity release during the accident, concentration of radioactive material in the air in the reactor neighbourhood, and the analysis of accident environmental effects.

  4. Phase Formation and Transformations in Transmutation Fuel Materials for the LIFE Engine Part I - Path Forward

    International Nuclear Information System (INIS)

    Turchi, P.E.; Kaufman, L.; Fluss, M.J.

    2008-01-01

    The current specifications of the LLNL fusion-fission hybrid proposal, namely LIFE, impose severe constraints on materials, and in particular on the nuclear fissile or fertile nuclear fuel and its immediate environment. This constitutes the focus of the present report with special emphasis on phase formation and phase transformations of the transmutation fuel and their consequences on particle and pebble thermal, chemical and mechanical integrities. We first review the work that has been done in recent years to improve materials properties under the Gen-IV project, and with in particular applications to HTGR and MSR, and also under GNEP and AFCI in the USA. Our goal is to assess the nuclear fuel options that currently exist together with their issues. Among the options, it is worth mentioning TRISO, IMF, and molten salts. The later option will not be discussed in details since an entire report is dedicated to it. Then, in a second part, with the specific LIFE specifications in mind, the various fuel options with their most critical issues are revisited with a path forward for each of them in terms of research, both experimental and theoretical. Since LIFE is applicable to very high burn-up of various fuels, distinctions will be made depending on the mission, i.e., energy production or incineration. Finally a few conclusions are drawn in terms of the specific needs for integrated materials modeling and the in depth knowledge on time-evolution thermochemistry that controls and drastically affects the performance of the nuclear materials and their immediate environment. Although LIFE demands materials that very likely have not yet been fully optimized, the challenge are not insurmountable and a well concerted experimental-modeling effort should lead to dramatic advances that should well serve other fission programs such as Gen-IV, GNEP, AFCI as well as the international fusion program, ITER

  5. Phase Formation and Transformations in Transmutation Fuel Materials for the LIFE Engine Part I - Path Forward

    Energy Technology Data Exchange (ETDEWEB)

    Turchi, P E; Kaufman, L; Fluss, M J

    2008-11-10

    The current specifications of the LLNL fusion-fission hybrid proposal, namely LIFE, impose severe constraints on materials, and in particular on the nuclear fissile or fertile nuclear fuel and its immediate environment. This constitutes the focus of the present report with special emphasis on phase formation and phase transformations of the transmutation fuel and their consequences on particle and pebble thermal, chemical and mechanical integrities. We first review the work that has been done in recent years to improve materials properties under the Gen-IV project, and with in particular applications to HTGR and MSR, and also under GNEP and AFCI in the USA. Our goal is to assess the nuclear fuel options that currently exist together with their issues. Among the options, it is worth mentioning TRISO, IMF, and molten salts. The later option will not be discussed in details since an entire report is dedicated to it. Then, in a second part, with the specific LIFE specifications in mind, the various fuel options with their most critical issues are revisited with a path forward for each of them in terms of research, both experimental and theoretical. Since LIFE is applicable to very high burn-up of various fuels, distinctions will be made depending on the mission, i.e., energy production or incineration. Finally a few conclusions are drawn in terms of the specific needs for integrated materials modeling and the in depth knowledge on time-evolution thermochemistry that controls and drastically affects the performance of the nuclear materials and their immediate environment. Although LIFE demands materials that very likely have not yet been fully optimized, the challenge are not insurmountable and a well concerted experimental-modeling effort should lead to dramatic advances that should well serve other fission programs such as Gen-IV, GNEP, AFCI as well as the international fusion program, ITER.

  6. LHC Beam Dump Design Study - Part III : Off-normal operating conditions

    CERN Document Server

    Bruno, L; Ross, M; Sala, P

    2000-01-01

    The LHC beam dump design study has been preliminarily substantiated by energy deposition simulations (Part I) and heat transfer analyses (Part II). The present report is devoted to the abnormal operating conditions induced by a malfunction of the beam diluters. A general approach to the analysis of off-normal operation is presented, which is derived from standard design norms adopted in the nuclear industry. Attention is focused mainly on the carbon core, which is longitudinally split into segments of different density in order to better distribute the deposited energy. The maximum energy density it absorbs decreases by at least 33%, compared to a uniform standard density carbon core. This structure may sustain any partial sweep failure without major damage, up to the ultimate beam intensity and energy. To minimise the risks inherent in a fully unswept beam, a sacrificial graphite mandrel will be placed on the core axis, surrounded by a thick high strength carbon-carbon composite tube. With this arrangement, ...

  7. Normal and sonographic anatomy of selected peripheral nerves. Part III: Peripheral nerves of the lower limb.

    Science.gov (United States)

    Kowalska, Berta; Sudoł-Szopińska, Iwona

    2012-06-01

    The ultrasonographic examination is currently increasingly used in imaging peripheral nerves, serving to supplement the physical examination, electromyography and magnetic resonance imaging. As in the case of other USG imaging studies, the examination of peripheral nerves is non-invasive and well-tolerated by patients. The typical ultrasonographic picture of peripheral nerves as well as the examination technique have been discussed in part I of this article series, following the example of the median nerve. Part II of the series presented the normal anatomy and the technique for examining the peripheral nerves of the upper limb. This part of the article series focuses on the anatomy and technique for examining twelve normal peripheral nerves of the lower extremity: the iliohypogastric and ilioinguinal nerves, the lateral cutaneous nerve of the thigh, the pudendal, sciatic, tibial, sural, medial plantar, lateral plantar, common peroneal, deep peroneal and superficial peroneal nerves. It includes diagrams showing the proper positioning of the sonographic probe, plus USG images of the successively discussed nerves and their surrounding structures. The ultrasonographic appearance of the peripheral nerves in the lower limb is identical to the nerves in the upper limb. However, when imaging the lower extremity, convex probes are more often utilized, to capture deeply-seated nerves. The examination technique, similarly to that used in visualizing the nerves of upper extremity, consists of locating the nerve at a characteristic anatomic reference point and tracking it using the "elevator technique". All 3 parts of the article series should serve as an introduction to a discussion of peripheral nerve pathologies, which will be presented in subsequent issues of the "Journal of Ultrasonography".

  8. Burnout in boiling heat transfer. Part III. High-quality forced-convection systems

    International Nuclear Information System (INIS)

    Bergles, A.E.

    1979-01-01

    This is the final part of a review of burnout during boiling heat transfer. The status of burnout in high-quality forced-convection systems is reviewed, and recent developments are summarized in detail. A general guide to the considerable literature is given. Parametric effects and correlations for water in circular and noncircular ducts are presented. Other topics discussed include transients, steam-generator applications, correlations for other fluids, fouling, and augmentation

  9. PIO I-II tendencies case study. Part 1. Mathematical modeling

    Directory of Open Access Journals (Sweden)

    Adrian TOADER

    2010-03-01

    Full Text Available In the paper, a study is performed from the perspective of giving a method to reduce the conservatism of the well known PIO (Pilot-Induced Oscillation criteria in predicting the susceptibility of an aircraft to this very harmful phenomenon. There are three interacting components of a PIO – the pilot, the vehicle, and the trigger (in fact, the hazard. The study, conceived in two parts, aims to underline the importance of human pilot model involved in analysis. In this first part, it is shown, following classical sources, how the LQG theory of control and estimation is used to obtain a complex model of human pilot. The approach is based on the argument, experimentally proved, that the human behaves “optimally” in some sense, subject to his inherent psychophysical limitations. The validation of such model is accomplished based on the experimental model of a VTOL-type aircraft. Then, the procedure of inserting typical saturation nonlinearities in the open loop transfer function is presented. A second part of the paper will illustrate PIO tendencies evaluation by means of a grapho-analytic method.

  10. Building Worlds and Learning Astronomy on Facebook Part III: Testing, Launch, and Evaluation

    Science.gov (United States)

    Harold, J.; Hines, D.; Vidugiris, E.; Goldman, K. H.

    2015-11-01

    James Harold (SSI), Dean Hines (STScI/SSI) and a team at the National Center for Interactive Learning at the Space Science Institute are developing Starchitect, an end-to-end stellar and planetary evolution game for the Facebook platform. Supported by NSF and NASA, and based in part on a prototype presented at ASP several years ago, Starchitect uses the “sporadic play” model of games such as Farmville, where players might only take actions a few times a day, but may continue playing for months. This paper is an update to a presentation at last year's ASP conference.

  11. National Energy Board 1992-93 estimates part III expenditure plan

    International Nuclear Information System (INIS)

    1992-01-01

    This Expenditure Plan is designed to be used as a reference document. As such, it contains several levels of detail to respond to the various needs of its audience. This Plan is divided into two sections. Section 1 presents an overview of the Program including a description, information on its background, objectives and planning perspective as well as performance information that forms the basis for the resources requested. Section 2 provides further information on costs and resources as well as special analyses that the reader may require to understand the Program more fully. Section 1 is preceded by details of Spending Authorities from Part 2 of the Estimates and Volume 2 of the Public Accounts. This is to provide continuity with other Estimates documents and to help in assessing the Program's financial performance over the past year. The format and organization of the document should be regarded as transitional. For the most part the document is organized to be consistent with the program structure approved for the National Energy Board. 16 figs

  12. National Energy Board 1992-93 estimates part III expenditure plan

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1993-12-31

    This Expenditure Plan is designed to be used as a reference document. As such, it contains several levels of detail to respond to the various needs of its audience. This Plan is divided into two sections. Section 1 presents an overview of the Program including a description, information on its background, objectives and planning perspective as well as performance information that forms the basis for the resources requested. Section 2 provides further information on costs and resources as well as special analyses that the reader may require to understand the Program more fully. Section 1 is preceded by details of Spending Authorities from Part 2 of the Estimates and Volume 2 of the Public Accounts. This is to provide continuity with other Estimates documents and to help in assessing the Program`s financial performance over the past year. The format and organization of the document should be regarded as transitional. For the most part the document is organized to be consistent with the program structure approved for the National Energy Board. 16 figs.

  13. Investigation of the Stress Intensity Limits of ASME Section III Div.5 for Structure Design Criteria of SFR Fuel Assembly

    International Nuclear Information System (INIS)

    Choo, Jin-Yup; Kim, Hyung-Kyu; Cheon, Jin-Sik

    2017-01-01

    In this paper, the stress intensity limits, Sm and St of HT-9 were built for the structural criteria of an SFR fuel assembly. Sm is obtained from the ultimate strength. As for St, it is more complicated because of its dependency of time duration in addition to temperature. Following the definition of Smt, the method in the ASME Sec. III Div. 1, Subsec. NH was consulted. We found that the Sm is adopted as Smt under the temperature about 470 .deg. C which is relatively low temperature range and over 470 .deg. C with relatively short time duration as 1000 hours. And the St is adopted as Smt at over 470 .deg. C and long time duration over 34800 hours, and over 520 .deg. C and 104 hours too. And at over 570 .deg. C and 1000 hours, and at over 630 .deg. C and 100 hours, St is also adopted for Smt. To use the present result as design criteria, a stringent examination needs to be carried out, because those are calculated from the formulae of HT-9 without an experimental validation. Therefore, an experimental work on the mechanical properties of HT-9 will be necessary.

  14. Final disposal of spent fuels and high activity waste: status and trends in the world. Part 2

    International Nuclear Information System (INIS)

    Herscovich de Pahissa, Marta

    2008-01-01

    The proper management of spent fuel arising from nuclear power production is a key issue for the sustainable development of nuclear energy. Some countries have adopted reprocessing of spent fuel and part of them has continued to develop and improve closed fuel cycle technologies; some other countries have adopted a direct final disposal. The objective in this article is to provide an update on the latest development in the world related with the geological disposal of spent nuclear fuel and high level wastes. (author) [es

  15. Two-loop renormalization in the standard model, part III. Renormalization equations and their solutions

    International Nuclear Information System (INIS)

    Actis, S.; Passarino, G.

    2006-12-01

    In part I and II of this series of papers all elements have been introduced to extend, to two loops, the set of renormalization procedures which are needed in describing the properties of a spontaneously broken gauge theory. In this paper, the final step is undertaken and finite renormalization is discussed. Two-loop renormalization equations are introduced and their solutions discussed within the context of the minimal standard model of fundamental interactions. These equations relate renormalized Lagrangian parameters (couplings and masses) to some input parameter set containing physical (pseudo-)observables. Complex poles for unstable gauge and Higgs bosons are used and a consistent setup is constructed for extending the predictivity of the theory from the Lep1 Z-boson scale (or the Lep2 WW scale) to regions of interest for LHC and ILC physics. (orig.)

  16. Two-loop renormalization in the standard model, part III. Renormalization equations and their solutions

    Energy Technology Data Exchange (ETDEWEB)

    Actis, S. [Deutsches Elektronen-Synchrotron (DESY), Zeuthen (Germany); Passarino, G. [Torino Univ. (Italy). Dipt. di Fisica Teorica; INFN, Sezione di Torino (Italy)

    2006-12-15

    In part I and II of this series of papers all elements have been introduced to extend, to two loops, the set of renormalization procedures which are needed in describing the properties of a spontaneously broken gauge theory. In this paper, the final step is undertaken and finite renormalization is discussed. Two-loop renormalization equations are introduced and their solutions discussed within the context of the minimal standard model of fundamental interactions. These equations relate renormalized Lagrangian parameters (couplings and masses) to some input parameter set containing physical (pseudo-)observables. Complex poles for unstable gauge and Higgs bosons are used and a consistent setup is constructed for extending the predictivity of the theory from the Lep1 Z-boson scale (or the Lep2 WW scale) to regions of interest for LHC and ILC physics. (orig.)

  17. Study of irradiated bone: Part III. /sup 99m/Tc pyrophosphate autoradiographic changes

    International Nuclear Information System (INIS)

    King, M.A.; Corriveau, O.; Casarett, G.W.; Weber, D.A.

    1978-01-01

    The macroautoradiographic and microautoradiographic localization of /sup 99m/Tc-pyrophosphate (/sup 99m/TcPPi) was studied in x-irradiated bone of rabbits up to one year post-irradiation. In cortical bone, /sup 99m/TcPPi was concentrated on bone surfaces near vasculature. Both forming and resorbing bone surfaces were comparably labeled at 2 hrs post-injection. Uptake on the surface of sites of haversian bone remodeling was observed to be at least part of the increased /sup 99m/TcPPi observed in irradiated bone in camera images. In irradiated trabecular bone 12 months following irradiation, a patchy decrease in /sup 99m/TcPPi uptake was correlated with localized decreases in vasculature

  18. Atomic Energy Control Board 1991-92 estimates part III expenditure plan

    International Nuclear Information System (INIS)

    1991-01-01

    This Expenditure Plan is designed to be used as a reference document. As such, it contains several levels of detail to respond to the various needs of its audience. This Plan is divided into two sections. Section 1 presents an overview of the Program including a description, information on its background, objectives and planning perspective as well as performance information that forms the basis for the resources requested. Section 2 provides further information on costs and resources as well as special analyses that the reader may require to understand the Program more fully. Section 1 is preceded by details of Spending Authorities from Part 2 of the Estimates and Volume 2 of the Public Accounts. This is to provide continuity with other Estimates documents and to help in assessing the Program's financial performance over the past year. 22 figs

  19. Atomic Energy Control Board 1991-92 estimates part III expenditure plan

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1992-12-31

    This Expenditure Plan is designed to be used as a reference document. As such, it contains several levels of detail to respond to the various needs of its audience. This Plan is divided into two sections. Section 1 presents an overview of the Program including a description, information on its background, objectives and planning perspective as well as performance information that forms the basis for the resources requested. Section 2 provides further information on costs and resources as well as special analyses that the reader may require to understand the Program more fully. Section 1 is preceded by details of Spending Authorities from Part 2 of the Estimates and Volume 2 of the Public Accounts. This is to provide continuity with other Estimates documents and to help in assessing the Program`s financial performance over the past year. 22 figs.

  20. Atomic Energy Control Board 1992-93 estimates part III expenditure plan

    International Nuclear Information System (INIS)

    1992-01-01

    This Expenditure Plan is designed to be used as a reference document. As such, it contains several levels of detail to respond to the various needs of its audience. This Plan is divided into two sections. Section 1 presents an overview of the Program including a description, information on its background, objectives and planning perspective as well as performance information that forms the basis for the resources requested. Section 2 provides further information on costs and resources as well as special analyses that the reader may require to understand the Program more fully. Section 1 is preceded by details of Spending Authorities from Part 2 of the Estimates and Volume 2 of the Public Accounts. This is to provide continuity with other Estimates documents and to help in assessing the Program's financial performance over the past year. 7 refs., 21 figs

  1. Atomic Energy Control Board 1992-93 estimates part III expenditure plan

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1993-12-31

    This Expenditure Plan is designed to be used as a reference document. As such, it contains several levels of detail to respond to the various needs of its audience. This Plan is divided into two sections. Section 1 presents an overview of the Program including a description, information on its background, objectives and planning perspective as well as performance information that forms the basis for the resources requested. Section 2 provides further information on costs and resources as well as special analyses that the reader may require to understand the Program more fully. Section 1 is preceded by details of Spending Authorities from Part 2 of the Estimates and Volume 2 of the Public Accounts. This is to provide continuity with other Estimates documents and to help in assessing the Program`s financial performance over the past year. 7 refs., 21 figs.

  2. Notes on the Birds of Central Oaxaca, Part III: Hirundinidae to Fringillidae

    Directory of Open Access Journals (Sweden)

    John M. Forcey

    2015-06-01

    Full Text Available Notas sobre las aves de Oaxaca central, parte II: Hirundinidae a Fringillidae Se reportan nuevos datos que amplían y clarifican nuestro conocimiento del estatus y distribución de 112 especies de aves en la región central del Estado de Oaxaca. Las observaciones se realizaron abarcando partes de los distritos de Centro, Etla, Ixtlan, Tlacolula, y Zaachila dentro de 35 km alrededor de la Ciudad de Oaxaca. El reporte se basa en observaciones tomadas durante 752 días, comprendidos entre diciembre 1996 y mayo 2002. Los principales hábitats del área son de pino-encino (incluyendo zonas pequeñas de pino-encino-oyamel y pino-encino mezclado con pastizales, matorral de encino, matorral subtropical, vegetación riparia, y vegetación secundaria, campos agrícolas y otros (incluyendo áreas urbanas, jardines, y parques. Las siguientes especies se reportan por primera vez en la zona: Progne sinaloae (registro nuevo en el estado de Oaxaca, Thryothorus felix, Hylocichla mustelina, Vermivora pinus, Vermivora chrysoptera, Dendroica pensylvanica, Dendroica magnolia, Dendroica fusca, Dendroica graciae, Oporornis philadelphia, Wilsonia canadensis, y Spiza americana.Además, las siguientes nueve especies se han reportado solamente en los Conteos Navideños o por registros únicos: Tachycineta bicolor, Dumetella carolinensis, Vermivora peregrina, Dendroica dominica, Dendroica discolor, y Piranga erythrocephala (en temporada de reproducción. Se reportan datos de la reproducción de 43 especies, 18 de los cuales no se habían registrado en estado reproductivo antes en esta zona. De estos 43, 39 se pueden agrupar como reproduciéndose en los meses de abril a julio.

  3. Potential sites for a spent unreprocessed fuel facility (SURFF), southwesten part of the Nevada Test Site

    International Nuclear Information System (INIS)

    Hoover, D.L.; Eckel, E.B.; Ohl, J.P.

    1978-01-01

    In the absence of specific criteria, the topography, geomorphology, and geology of Jackass Flats and vicinity in the southwestern part of the Nevada Test Site are evaluated by arbitrary guidelines for a Spent Unreprocessed Fuel Facility. The guidelines include requirements for surface slopes of less than 5%, 61 m of alluvium beneath the site, an area free of active erosion or deposition, lack of faults, a minimum area of 5 km 2 , no potential for flooding, and as many logistical support facilities as possible. The geology of the Jackass Flats area is similar to the rest of the Nevada Test Site in topographic relief (305-1,200 m), stratigraphy (complexly folded and faulted Paleozoic sediments overlain by Tertiary ash-flow tuffs and lavas overlain in turn by younger alluvium), and structure (Paleozoic thrust faults and folds, strike-slip faults, proximity to volcanic centers, and Basin and Range normal faults). Of the stratigraphic units at the potential Spent Unreprocessed Fuel Facility site in Jackass Flats, only the thickness and stability of the alluvium are of immediate importance. Basin and Range faults and a possible extension of the Mine Mountain fault need further investigation. The combination of a slope map and a simplified geologic and physiographic map into one map shows several potential sites for a Spent Unreprocessed Fuel Facility in Jackass Flats. The potential areas have slopes of less than 5% and contain only desert pavement or segmented pavement--the two physiographic categories having the greatest geomorphic and hydraulic stability. Before further work can be done, specific criteria for a Spent Unreprocessed Fuel Facility site must be defined. Following criteria definition, potential sites will require detailed topographic and geologic studies, subsurface investigations (including geophysical methods, trenching, and perhaps shallow drilling for faults in alluvium), detailed surface hydrologic studies, and possibly subsurface hydrologic studies

  4. Circuit modeling of the electrical impedance: part III. Disuse following bone fracture

    International Nuclear Information System (INIS)

    Shiffman, C A

    2013-01-01

    Multifrequency measurements of the electrical impedance of muscle have been extended to the study of disuse following bone fracture, and analyzed using the five-element circuit model used earlier in the study of the effects of disease. Eighteen subjects recovering from simple fractures on upper or lower limbs were examined (ten males, eight females, aged 18–66). Muscles on uninjured contralateral limbs were used as comparison standards, and results are presented in terms of the ratios p(injured)/p(uninjured), where p stands for the circuit parameter r 1 , r 2 , r 3 , 1/c 1 or 1/c 2 . These are strikingly similar to the diseased-to-healthy ratios for patients with neuromuscular disease, reported in part I of this series. In particular, r 1 is virtually unaffected and the ratios for r 2 , r 3 , 1/c 1 and 1/c 2 can be as large as in serious disease. Furthermore, the same pattern of relationships between the parameters is found, suggesting that there is a common underlying mechanism for the impedance changes. Atrophy and fibrosis are examined as candidates for that mechanism, but it is argued that their effects are too small to explain the observed changes. Fundamental considerations aside, the sensitivity, reproducibility and technical simplicity of the technique recommend its use for in-flight assessments of muscles during orbital or interplanetary missions. (paper)

  5. The Systemic Theory of Living Systems and Relevance to CAM: the Theory (Part III

    Directory of Open Access Journals (Sweden)

    José A. Olalde Rangel

    2005-01-01

    Full Text Available Western medical science lacks a solid philosophical and theoretical approach to disease cognition and therapeutics. My first two articles provided a framework for a humane medicine based on Modern Biophysics. Its precepts encompass modern therapeutics and CAM. Modern Biophysics and its concepts are presently missing in medicine, whether orthodox or CAM, albeit they probably provide the long sought explanation that bridges the abyss between East and West. Key points that differentiate Systemic from other systems' approaches are ‘Intelligence’, ‘Energy’ and the objective ‘to survive’. The General System Theory (GST took a forward step by proposing a departure from the mechanistic biological concept—of analyzing parts and processes in isolation—and brought us towards an organismic model. GST examines the system's components and results of their interaction. However, GST still does not go far enough. GST assumes ‘Self-Organization’ as a spontaneous phenomenon, ignoring a causative entity or central controller to all systems: Intelligence. It also neglects ‘Survive’ as the directional motivation common to any living system, and scarcely assigns ‘Energy’ its true inherent value. These three parameters, Intelligence, Energy and Survive, are vital variables to be considered, in our human quest, if we are to achieve a unified theory of life.

  6. Short-stack modeling of degradation in solid oxide fuel cells. Part I. Contact degradation

    Energy Technology Data Exchange (ETDEWEB)

    Gazzarri, J.I. [Department of Mechanical Engineering, University of British Columbia, 2054-6250 Applied Science Lane, Vancouver, BC V6T 1Z4 (Canada); Kesler, O. [Department of Mechanical and Industrial Engineering, University of Toronto, 5 King' s College Road, Toronto, ON M5S 3G8 (Canada)

    2008-01-21

    As the first part of a two paper series, we present a two-dimensional impedance model of a working solid oxide fuel cell (SOFC) to study the effect of contact degradation on the impedance spectrum for the purpose of non-invasive diagnosis. The two dimensional modeled geometry includes the ribbed interconnect, and is adequate to represent co- and counter-flow configurations. Simulated degradation modes include: cathode delamination, interconnect oxidation, and interconnect-cathode detachment. The simulations show differences in the way each degradation mode impacts the impedance spectrum shape, suggesting that identification is possible. In Part II, we present a sensitivity analysis of the results to input parameter variability that reveals strengths and limitations of the method, as well as describing possible interactions between input parameters and concurrent degradation modes. (author)

  7. Short-stack modeling of degradation in solid oxide fuel cells. Part I. Contact degradation

    Science.gov (United States)

    Gazzarri, J. I.; Kesler, O.

    As the first part of a two paper series, we present a two-dimensional impedance model of a working solid oxide fuel cell (SOFC) to study the effect of contact degradation on the impedance spectrum for the purpose of non-invasive diagnosis. The two dimensional modeled geometry includes the ribbed interconnect, and is adequate to represent co- and counter-flow configurations. Simulated degradation modes include: cathode delamination, interconnect oxidation, and interconnect-cathode detachment. The simulations show differences in the way each degradation mode impacts the impedance spectrum shape, suggesting that identification is possible. In Part II, we present a sensitivity analysis of the results to input parameter variability that reveals strengths and limitations of the method, as well as describing possible interactions between input parameters and concurrent degradation modes.

  8. Design and construction of the SIPPING for fuels of the TRIGA Mark III reactor; Diseno y construccion del SIPPING para combustibles del reactor TRIGA Mark III

    Energy Technology Data Exchange (ETDEWEB)

    Castaneda J, G.; Delfin L, A.; Alvarado P, R.; Mazon R, R.; Ortega V, B. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)]. e-mail: adl@nuclear.inin.mx

    2003-07-01

    The sipping technique, it has been used by several possessors of nuclear research reactors in its irradiated nuclear fuels, likewise in some fuel storage sites, with the objective of to determine the quantity of radioactivity that the fuel liberates in the means in that it is. The irradiated fuel in storage of some nuclear research reactors, its can have cracks that cross the cladding of the same one, generating the liberation of fission products that its need to determine to maintain safety measures appropriate as much as the fuel as of the facilities where they are. It doesn't exist until now, some method published for the non destructive sipping test technique. Based on that described, the Reactor Department of the National Institute of Nuclear Research, it has designed and built an inspection system of irradiated fuel that it will allow the detection of gassy fission products in site, and solids by means of the measurement of the activity of the Cs-137 contained in water samples. (Author)

  9. Storage device for a long nuclear reactor fuel element and/or a long nuclear reactor fuel element part

    International Nuclear Information System (INIS)

    Vogt, M.; Schoenwitz, H.P.; Dassbach, W.

    1986-01-01

    The storage device can be erected in a dry storage room for new fuel elements and also in a storage pond for irradiated fuel elements. It consists of shells, which are arranged vertically and which have a lid. A suspension for the fuel element is provided on the underside of the lid, which acts as a support against squashing or bending in case of vertical forces acting (earthquake). (DG) [de

  10. Nuclear and radiological safety in the substitution process of the fuel HEU to LEU 30/20 in the Reactor TRIGA Mark III of the ININ

    International Nuclear Information System (INIS)

    Hernandez G, J.

    2012-10-01

    Inside the safety initiative in the international ambit, with the purpose of reducing the risks associated with the use of high enrichment nuclear fuels (HEU) for different proposes to the peaceful uses of the nuclear energy, Mexico contributes by means of the substitution of the high enrichment fuel HEU for low enrichment fuel LEU 30/20 in the TRIGA Mark III Reactor, belonging to Instituto Nacional de Investigaciones Nucleares (ININ). The conversion process was carried out by means of the following activities: analysis of the proposed core, reception and inspection of the fuel LEU 30/20, the discharge of the fuels of the mixed reactor core, shipment of the fuels HEU fresh and irradiated to the origin country, reload activities with the fuels LEU 30/20 and parameters measurement of the core operation. In order to maintaining the personnel's integrity and infrastructure associated to the Reactor, during the whole process the measurements of nuclear and radiological safety were controlled to detail, in execution with the license requirements of the installation. This work describes the covering activities and radiological inspections more relevant, as well as the measurements of radiological control implemented with base in the estimate of the equivalent dose of the substitution process. (Author)

  11. Dispersion of a Passive Scalar Fluctuating Plume in a Turbulent Boundary Layer. Part III: Stochastic Modelling

    Science.gov (United States)

    Marro, Massimo; Salizzoni, Pietro; Soulhac, Lionel; Cassiani, Massimo

    2018-01-01

    We analyze the reliability of the Lagrangian stochastic micromixing method in predicting higher-order statistics of the passive scalar concentration induced by an elevated source (of varying diameter) placed in a turbulent boundary layer. To that purpose we analyze two different modelling approaches by testing their results against the wind-tunnel measurements discussed in Part I (Nironi et al., Boundary-Layer Meteorology, 2015, Vol. 156, 415-446). The first is a probability density function (PDF) micromixing model that simulates the effects of the molecular diffusivity on the concentration fluctuations by taking into account the background particles. The second is a new model, named VPΓ, conceived in order to minimize the computational costs. This is based on the volumetric particle approach providing estimates of the first two concentration moments with no need for the simulation of the background particles. In this second approach, higher-order moments are computed based on the estimates of these two moments and under the assumption that the concentration PDF is a Gamma distribution. The comparisons concern the spatial distribution of the first four moments of the concentration and the evolution of the PDF along the plume centreline. The novelty of this work is twofold: (i) we perform a systematic comparison of the results of micro-mixing Lagrangian models against experiments providing profiles of the first four moments of the concentration within an inhomogeneous and anisotropic turbulent flow, and (ii) we show the reliability of the VPΓ model as an operational tool for the prediction of the PDF of the concentration.

  12. Dispersion of a Passive Scalar Fluctuating Plume in a Turbulent Boundary Layer. Part III: Stochastic Modelling

    Science.gov (United States)

    Marro, Massimo; Salizzoni, Pietro; Soulhac, Lionel; Cassiani, Massimo

    2018-06-01

    We analyze the reliability of the Lagrangian stochastic micromixing method in predicting higher-order statistics of the passive scalar concentration induced by an elevated source (of varying diameter) placed in a turbulent boundary layer. To that purpose we analyze two different modelling approaches by testing their results against the wind-tunnel measurements discussed in Part I (Nironi et al., Boundary-Layer Meteorology, 2015, Vol. 156, 415-446). The first is a probability density function (PDF) micromixing model that simulates the effects of the molecular diffusivity on the concentration fluctuations by taking into account the background particles. The second is a new model, named VPΓ, conceived in order to minimize the computational costs. This is based on the volumetric particle approach providing estimates of the first two concentration moments with no need for the simulation of the background particles. In this second approach, higher-order moments are computed based on the estimates of these two moments and under the assumption that the concentration PDF is a Gamma distribution. The comparisons concern the spatial distribution of the first four moments of the concentration and the evolution of the PDF along the plume centreline. The novelty of this work is twofold: (i) we perform a systematic comparison of the results of micro-mixing Lagrangian models against experiments providing profiles of the first four moments of the concentration within an inhomogeneous and anisotropic turbulent flow, and (ii) we show the reliability of the VPΓ model as an operational tool for the prediction of the PDF of the concentration.

  13. A Study of Future Communications Concepts and Technologies for the National Airspace System-Part III

    Science.gov (United States)

    Ponchak, Denise S.; Apaza, Rafael D.; Wichgersm Joel M.; Haynes, Brian; Roy, Aloke

    2014-01-01

    The National Aeronautics and Space Administration (NASA) Glenn Research Center (GRC) is investigating current and anticipated wireless communications concepts and technologies that the National Airspace System (NAS) may need in the next 50 years. NASA has awarded three NASA Research Announcements (NAR) studies with the objective to determine the most promising candidate technologies for air-to-air and air-to-ground data exchange and analyze their suitability in a post-NextGen NAS environment. This paper will present progress made in the studies and describe the communications challenges and opportunities that have been identified as part of the study. NASA's NextGen Concepts and Technology Development (CTD) Project integrates solutions for a safe, efficient and high-capacity airspace system through joint research efforts and partnerships with other government agencies. The CTD Project is one of two within NASA's Airspace Systems Program and is managed by the NASA Ames Research Center. Research within the CTD Project is in support the 2011 NASA Strategic Plan Sub-Goal 4.1: Develop innovative solutions and advanced technologies, through a balanced research portfolio, to improve current and future air transportation. The focus of CTD is on developing capabilities in traffic flow management, dynamic airspace configuration, separation assurance, super density operations and airport surface operations. Important to its research is the development of human/automation information requirements and decisionmaking guidelines for human-human and human-machine airportal decision-making. Airborne separation, oceanic intrail climb/descent and interval management applications depend on location and intent information of surrounding aircraft. ADS-B has been proposed to provide the information exchange, but other candidates such as satellite-based receivers, broadband or airborne internet, and cellular communications are possible candidate's.

  14. Ultrasound assessment of selected peripheral nerve pathologies. Part III: Injuries and postoperative evaluation

    Directory of Open Access Journals (Sweden)

    Berta Kowalska

    2013-03-01

    Full Text Available The previous articles of the series devoted to ultrasound diagnostics of peripheral nerves concerned the most common nerve pathologies, i.e. entrapment neuropathies. The aim of the last part of the series is to present ultrasound possibilities in the postoperative control of the peripheral nerves as well as in the diagnostics of the second most common neuropathies of peripheral nerves, i.e. posttraumatic lesions. Early diagnostics of posttraumatic changes is of fundamental importance for the course of treatment and its long-term effects. It aids surgeons in making treatment decisions (whether surgical or conservative. When surgical treatment is necessary, the surgeon, based on US findings, is able to plan a given type of operative method. In certain cases, may even abandon the corrective or reconstructive surgery of the nerve trunk (when there are extensive defects of the nerve trunks and instead, proceed with muscle transfers. Medical literature proposes a range of divisions of the kinds of peripheral nerve injuries depending on, among others, the mechanism or degree of damage. However, the most important issue in the surgeon-diagnostician communication is a detailed description of stumps of the nerve trunks, their distance and location. In the postoperative period, ultrasound is used for monitoring the operative or conservative treatment effects including the determination of the causes of a persistent or recurrent neuropathy. It facilitates decision-making concerning a repeated surgical procedure or assuming a wait-and-see attitude. It is a difficult task for a diagnostician and it requires experience, close cooperation with a clinician and knowledge concerning surgical techniques. Apart from a static assessment, a dynamic assessment of possible adhesions constitutes a crucial element of postoperative examination. This feature distinguishes ultrasound scanning from other methods used in the diagnostics of peripheral neuropathies.

  15. Systems Analysis of Technologies for Energy Recovery from Waste. Part I. Gasification followed by Catalytic Combustion, PEM Fuel Cells and Solid Oxide Fuel Cells for Stationary Applications in Comparison with Incineration. Part - II. Catalytic combustion - Experimental part

    Energy Technology Data Exchange (ETDEWEB)

    Assefa, Getachew; Frostell, Bjoern [Royal Inst. of Technology, Stockholm (Sweden). Div. of Industrial Ecology; Jaeraas, Sven; Kusar, Henrik [Royal Inst. of Technology, Stockholm (Sweden). Div. of Chemical Technology

    2005-02-01

    This project is entitled 'Systems Analysis: Energy Recovery from waste, catalytic combustion in comparison with fuel cells and incineration'. Some of the technologies that are currently developed by researchers at the Royal Institute of Technology include catalytic combustion and fuel cells as downstream units in a gasification system. The aim of this project is to assess the energy turnover as well as the potential environmental impacts of biomass/waste-to-energy technologies. In second part of this project economic analyses of the technologies in general and catalytic combustion and fuel cell technologies in particular will be carried out. Four technology scenarios are studied: (1) Gasification followed by Low temperature fuel cells (Proton Exchange Membrane (PEM) fuel cells) (2) Gasification followed by high temperature fuel cells (Solid Oxide Fuel Cells (SOFC) (3) Gasification followed by catalytic combustion and (4) Incineration with energy recovery. The waste used as feedstock is an industrial waste containing parts of household waste, paper waste, wood residues and poly ethene. In the study compensatory district heating is produced by combustion of biofuel. The power used for running the processes in the scenarios will be supplied by the waste-to-energy technologies themselves while compensatory power is assumed to be produced from natural gas. The emissions from the system studied are classified and characterised using methodology from Life Cycle Assessment in to the following environmental impact categories: Global Warming Potential, Acidification Potential, Eutrophication Potential and finally Formation of Photochemical Oxidants. Looking at the result of the four technology chains in terms of the four impact categories with impact per GWh electricity produced as a unit of comparison and from the perspective of the rank each scenario has in all the four impact categories, SOFC appears to be the winner technology followed by PEM and CC as second

  16. Systems Analysis of Technologies for Energy Recovery from Waste. Part I. Gasification followed by Catalytic Combustion, PEM Fuel Cells and Solid Oxide Fuel Cells for Stationary Applications in Comparison with Incineration. Part - II. Catalytic combustion - Experimental part

    International Nuclear Information System (INIS)

    Assefa, Getachew; Frostell, Bjoern; Jaeraas, Sven; Kusar, Henrik

    2005-02-01

    This project is entitled 'Systems Analysis: Energy Recovery from waste, catalytic combustion in comparison with fuel cells and incineration'. Some of the technologies that are currently developed by researchers at the Royal Institute of Technology include catalytic combustion and fuel cells as downstream units in a gasification system. The aim of this project is to assess the energy turnover as well as the potential environmental impacts of biomass/waste-to-energy technologies. In second part of this project economic analyses of the technologies in general and catalytic combustion and fuel cell technologies in particular will be carried out. Four technology scenarios are studied: (1) Gasification followed by Low temperature fuel cells (Proton Exchange Membrane (PEM) fuel cells) (2) Gasification followed by high temperature fuel cells (Solid Oxide Fuel Cells (SOFC) (3) Gasification followed by catalytic combustion and (4) Incineration with energy recovery. The waste used as feedstock is an industrial waste containing parts of household waste, paper waste, wood residues and poly ethene. In the study compensatory district heating is produced by combustion of biofuel. The power used for running the processes in the scenarios will be supplied by the waste-to-energy technologies themselves while compensatory power is assumed to be produced from natural gas. The emissions from the system studied are classified and characterised using methodology from Life Cycle Assessment in to the following environmental impact categories: Global Warming Potential, Acidification Potential, Eutrophication Potential and finally Formation of Photochemical Oxidants. Looking at the result of the four technology chains in terms of the four impact categories with impact per GWh electricity produced as a unit of comparison and from the perspective of the rank each scenario has in all the four impact categories, SOFC appears to be the winner technology followed by PEM and CC as second and third

  17. The evaluation of failure stress and released amount of fission product gas of power ramped rod by fuel behaviour analysis code 'FEMAXI-III'

    International Nuclear Information System (INIS)

    Yanagisawa, Kazuaki; Fujita, Misao

    1984-01-01

    Pellet-Cladding Interaction(PCI) related in-pile failure of Zircaloy sheathed fuel rod is in general considered to be caused by combination of pellet-cladding mechanical interaction(PCMI) with fuel-cladding chemical interaction(FCCI). An understanding of a basic mechanism of PCI-related fuel failure is therefore necessary to get actual cladding hoop stress from mechanical interaction and released amounts of fission product(FP) gas of aggressive environmental agency from chemical interaction. This paper describes results of code analysis performed on fuel failure to cladding hoop stress and amounts of FP gas released under the condition associated with power ramping. Data from Halden(HBWR) and from Studsvik(R2) are used for code analysis. The fuel behaviour analysis code ''FEMAXI-III'' is used as an analytical tool. The followings are revealed from the study: (1) PCI-related fuel failure is dependent upon cladding hoop stress and released amounts of FP gas at power ramping. (2) Preliminary calculated threshold values of hoop stress and of released amounts of FP gas to PCI failure are respectively 330MPa, 10% under the Halden condition, 190MPa, 5% under the Inter ramp(BWR) condition, and 270MPa, 14% under the Over ramp(PWR) condition. The values of hoop stress calculated are almost in the similar range of those obtained from ex-reactor PCI simulated tests searched from references published. (3) The FEMAXI-III code verification is made in mechanical manner by using in-pile deformation data(diametral strain) obtained from power ramping test undertaken by JAERI. While, the code verification is made in thermal manner by using punctured FP gas data obtained from post irradiation examination performed on non-defected power ramped fuel rods. The calculations are resulted in good agreements to both, mechanical and thermal experimental data suggesting the validity of the code evaluation. (J.P.N.)

  18. PACE. A Program for Acquiring Competence in Entrepreneurship. Part III: Being an Entrepreneur. Unit F: Managing Human Resources. Research and Development Series No. 194 C-6.

    Science.gov (United States)

    Ohio State Univ., Columbus. National Center for Research in Vocational Education.

    This three-part curriculum for entrepreneurship education is primarily for postsecondary level, including four-level colleges and adult education, but it can be adapted for special groups or vocational teacher education. The emphasis of the eight instructional units in Part III is operating a business. Unit F focuses on proper management of human…

  19. PACE. A Program for Acquiring Competence in Entrepreneurship. Part III: Being an Entrepreneur. Unit A: Managing the Business. Research and Development Series No. 194 C-1.

    Science.gov (United States)

    Ohio State Univ., Columbus. National Center for Research in Vocational Education.

    This three-part curriculum for entrepreneurship education is primarily for postsecondary level, including four-year colleges and adult education, but it can be adapted for special groups or vocational teacher education. The emphasis of the eight instructional units in Part III is operating a business. Unit A focuses on the management process. It…

  20. PACE. A Program for Acquiring Competence in Entrepreneurship. Part III: Being an Entrepreneur. Unit B: Financial Management. Research and Development Series No. 194 C-2.

    Science.gov (United States)

    Ohio State Univ., Columbus. National Center for Research in Vocational Education.

    This three-part curriculum for entrepreneurship education is primarily for postsecondary level, including four-year colleges and adult education, but it can be adapted for special groups or vocational teacher education. The emphasis of the eight instructional units in Part III is operating a business. Unit B focuses on good financial management…

  1. PACE. A Program for Acquiring Competence in Entrepreneurship. Part III: Being an Entrepreneur. Unit H: Business Protection. Research and Development Series No. 194 C-8.

    Science.gov (United States)

    Ohio State Univ., Columbus. National Center for Research in Vocational Education.

    This three-part curriculum for entrepreneurship education is primarily for postsecondary level, including four-year colleges and adult education, but it can be adapted for special groups or vocational teacher education. The emphasis of the eight instructional units in part III is operating a business. Unit H focuses on business protection. It…

  2. PACE. A Program for Acquiring Competence in Entrepreneurship. Part III: Being an Entrepreneur. Unit G: Community Relations. Research and Development Series No. 194 C-7.

    Science.gov (United States)

    Ohio State Univ., Columbus. National Center for Research in Vocational Education.

    This three-part curriculum for entrepreneurship education is primarily for postsecondary level, including four-year colleges and adult education, but it can be adapted for special groups of vocational teacher education. The emphasis of the eight instructional units in Part III is operating a business. Unit G focuses on community relations. It…

  3. PACE. A Program for Acquiring Competence in Entrepreneurship. Part III: Being an Entrepreneur. Unit E: Successful Selling. Research and Development Series No. 194 C-5.

    Science.gov (United States)

    Ohio State Univ., Columbus. National Center for Research in Vocational Education.

    This three-part curriculum for entrepreneurship education is primarily for postsecondary level, including four-year colleges and adult education, but it can be adapted for special groups or vocational teacher education. The emphasis of the eight instructional units in Part III is operating a business. Unit E focuses on personal (face-to-face)…

  4. PACE. A Program for Acquiring Competence in Entrepreneurship. Part III: Being an Entrepreneur. Unit D: Marketing Management. Research and Development Series No. 194 C-4.

    Science.gov (United States)

    Ohio State Univ., Columbus. National Center for Research in Vocational Education.

    This three-part curriculum for entrepreneurship education is primarily for postsecondary level, including four-year colleges and adult education, but it can be adapted for special groups or vocational teacher education. The emphasis of the eight instructional units in Part III is operating a business. Unit D focuses on market management. It…

  5. Electronics and telecommunications in Poland, issues and perspectives: Part III. Innovativeness, applications, economy, development scenarios, politics

    Science.gov (United States)

    Modelski, Józef; Romaniuk, Ryszard

    2010-09-01

    important role of ET is combined with the existence in the society of an adequate infrastructure which recreates the full development cycle of high technology embracing: people, institutions, finances and logistics, in this also science, higher education, education, continuous training, dissemination and outreach, professional social environment, legal basis, political support and lobbying, innovation structures, applications, industry and economy. The digest of chosen development tendencies in ET was made here from the academic perspective, in a wider scale and on this background the national one, trying to situate this branch in the society, determine its changing role to build a new technical infrastructure of a society based on knowledge, a role of builder of many practical gadgets facilitating life, a role of a big future integrator of today's single bricks into certain more useful unity. This digest does not have a character of a systematic analysis of ET. It is a kind of an arbitrary utterance of the authors inside their field of competence. The aim of this paper is to take an active part in the discussion of the academic community in this country on the development strategy of ET, choice of priorities for cyclically rebuilding economy, in competitive environments. The review paper was initiated by the Committee of Electronics and Telecommunications of Polish Academy of Sciences and was published in Polish as introductory chapter of a dedicated expertise, printed in a book format. This version makes the included opinions available for a wider community.

  6. Irradiation Test in HANARO of the Parts of an X-Gen Nuclear Fuel Assembly

    Energy Technology Data Exchange (ETDEWEB)

    Choo, K. N.; Kim, B. G.; Kang, Y. H. (and others)

    2008-08-15

    An instrumented capsule of 07M-13N was designed, fabricated and irradiated for an evaluation of the neutron irradiation properties of the parts of an X-Gen nuclear fuel assembly for PWR requested by KNF. Some specimens requested by Westinghouse Co. and Hanyang university were also inserted. 389 KNF specimens such as bucking and spring test specimens of 1x1 cell spacer grid, tensile, microstructure and tensile of welded parts, irradiation growth, spring test specimens made of HANA tube, Zirlo, Zircaloy-4, Inconel-718 were placed in the capsule. The capsule was composed of 5 stages having many kinds of specimens and an independent electric heater at each stage. During the irradiation test, the temperature of the specimens and the thermal/fast neutron fluences were measured by 14 thermocouples and 7 sets of Ni-Ti-Fe (2 sets contain additional Nb-Ag) neutron fluence monitors installed in the capsule. The capsule was irradiated for 59.19days (4 cycles) in the CT test hole of HANARO of a 30MW thermal output at 300 {approx} 420 .deg. C(for KNF specimens) up to a fast neutron fluence of 1.27x10{sup 21}(n/cm{sup 2}) (E>1MeV). After an irradiation test, the main body of the capsule was cut off at the bottom of the protection tube with a cutting system and it was transported to the IMEF (Irradiated Materials Examination Facility). The irradiated specimens were tested to evaluate the irradiation performance of the parts of an X-Gen fuel assembly in the IMEF hot cell.

  7. Mass, energy and material balances of SRF production process. Part 3: solid recovered fuel produced from municipal solid waste.

    Science.gov (United States)

    Nasrullah, Muhammad; Vainikka, Pasi; Hannula, Janne; Hurme, Markku; Kärki, Janne

    2015-02-01

    This is the third and final part of the three-part article written to describe the mass, energy and material balances of the solid recovered fuel production process produced from various types of waste streams through mechanical treatment. This article focused the production of solid recovered fuel from municipal solid waste. The stream of municipal solid waste used here as an input waste material to produce solid recovered fuel is energy waste collected from households of municipality. This article presents the mass, energy and material balances of the solid recovered fuel production process. These balances are based on the proximate as well as the ultimate analysis and the composition determination of various streams of material produced in a solid recovered fuel production plant. All the process streams are sampled and treated according to CEN standard methods for solid recovered fuel. The results of the mass balance of the solid recovered fuel production process showed that 72% of the input waste material was recovered in the form of solid recovered fuel; 2.6% as ferrous metal, 0.4% as non-ferrous metal, 11% was sorted as rejects material, 12% as fine faction and 2% as heavy fraction. The energy balance of the solid recovered fuel production process showed that 86% of the total input energy content of input waste material was recovered in the form of solid recovered fuel. The remaining percentage (14%) of the input energy was split into the streams of reject material, fine fraction and heavy fraction. The material balances of this process showed that mass fraction of paper and cardboard, plastic (soft) and wood recovered in the solid recovered fuel stream was 88%, 85% and 90%, respectively, of their input mass. A high mass fraction of rubber material, plastic (PVC-plastic) and inert (stone/rock and glass particles) was found in the reject material stream. © The Author(s) 2014.

  8. Comparison of thermal and radical effects of EGR gases on combustion process in dual fuel engines at part loads

    International Nuclear Information System (INIS)

    Pirouzpanah, V.; Khoshbakhti Saray, R.; Sohrabi, A.; Niaei, A.

    2007-01-01

    Dual fuel engines at part load inevitably suffer from lower thermal efficiency and higher emission of carbon monoxide and unburned fuel. This work is conducted to investigate the combustion characteristics of a dual fuel (Diesel-gas) engine at part loads using a single zone combustion model with detailed chemical kinetics for combustion of natural gas fuel. In this home made software, the presence of the pilot fuel is considered as a heat source that is deriving form two superposed Wiebe's combustion functions to account for its contribution to ignition of the gaseous fuel and the rest of the total released energy. The chemical kinetics mechanism consists of 112 reactions with 34 species. This combustion model is able to establish the development of the combustion process with time and the associated important operating parameters, such as pressure, temperature, heat release rate (HRR) and species concentration. Therefore, this work is an attempt to investigate the combustion phenomenon at part load and using exhaust gas recirculation (EGR) to improve the above mentioned problems. Also, the results of this work show that each of the different cases of EGR (thermal, chemical and radical cases) has an important role on the combustion process in dual fuel engines at part loads. It is found that all the different cases of EGR have positive effects on the performance and emission parameters of dual fuel engines at part loads despite the negative effect of some diluent gases in the chemical case, which moderates too much the positive effects of the thermal and radical cases of EGR. Predicted values show good agreement with corresponding experimental values over the whole range of engine operating conditions. Implications will be discussed in detail

  9. AUTOMOTIVE DIESEL MAINTENANCE L. UNIT XII, PART I--MAINTAINING THE FUEL SYSTEM (PART II), CUMMINS DIESEL ENGINE, PART II--UNIT INSTALLATION (ENGINE).

    Science.gov (United States)

    Human Engineering Inst., Cleveland, OH.

    THIS MODULE OF A 30-MODULE COURSE IS DESIGNED TO DEVELOP AN UNDERSTANDING OF THE OPERATION AND MAINTENANCE OF THE DIESEL ENGINE FUEL SYSTEM AND THE PROCEDURES FOR DIESEL ENGINE INSTALLATION. TOPICS ARE FUEL FLOW CHARACTERISTICS, PTG FUEL PUMP, PREPARATION FOR INSTALLATION, AND INSTALLING ENGINE. THE MODULE CONSISTS OF A SELF-INSTRUCTIONAL BRANCH…

  10. 14 CFR Appendix M to Part 25 - Fuel Tank System Flammability Reduction Means

    Science.gov (United States)

    2010-01-01

    ... 14 Aeronautics and Space 1 2010-01-01 2010-01-01 false Fuel Tank System Flammability Reduction... 25—Fuel Tank System Flammability Reduction Means M25.1Fuel tank flammability exposure requirements. (a) The Fleet Average Flammability Exposure of each fuel tank, as determined in accordance with...

  11. Workshop 95. Part III

    International Nuclear Information System (INIS)

    1995-01-01

    Out of 140 short communications presented in the proceedings, 13 have been inputted into INIS. The topics covered include lifetime control in semiconductor devices by ion irradiation, single crystal scintillation detectors, environmental monitoring, diffusion and sorption of radionuclides in soils, accelerator driven reactors, radioactive waste disposal, digital reactor control systems and research reactors. (Z.S.)

  12. Workshop 95. Part III

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-12-31

    Out of 140 short communications presented in the proceedings, 13 have been inputted into INIS. The topics covered include lifetime control in semiconductor devices by ion irradiation, single crystal scintillation detectors, environmental monitoring, diffusion and sorption of radionuclides in soils, accelerator driven reactors, radioactive waste disposal, digital reactor control systems and research reactors. (Z.S.).

  13. The history, genotoxicity, and carcinogenicity of carbon-based fuels and their emissions. Part 2: solid fuels.

    Science.gov (United States)

    Claxton, Larry D

    2014-01-01

    The combustion of solid fuels (like wood, animal dung, and coal) usually involves elevated temperatures and altered pressures and genotoxicants (e.g., PAHs) are likely to form. These substances are carcinogenic in experimental animals, and epidemiological studies implicate these fuels (especially their emissions) as carcinogens in man. Globally, ∼50% of all households and ∼90% of all rural households use solid fuels for cooking or heating and these fuels often are burnt in simple stoves with very incomplete combustion. Exposed women and children often exhibit low birth weight, increased infant and perinatal mortality, head and neck cancer, and lung cancer although few studies have measured exposure directly. Today, households that cannot meet the expense of fuels like kerosene, liquefied petroleum gas, and electricity resort to collecting wood, agricultural residue, and animal dung to use as household fuels. In the more developed countries, solid fuels are often used for electric power generation providing more than half of the electricity generated in the United States. The world's coal reserves, which equal approximately one exagram, equal ∼1 trillion barrels of crude oil (comparable to all the world's known oil reserves) and could last for 600 years. Studies show that the PAHs that are identified in solid fuel emissions react with NO2 to form direct-acting mutagens. In summary, many of the measured genotoxicants found in both the indoor and electricity-generating combustors are the same; therefore, the severity of the health effects vary with exposure and with the health status of the exposed population. Copyright © 2014. Published by Elsevier B.V.

  14. The history, genotoxicity and carcinogenicity of carbon-based fuels and their emissions: part 4 - alternative fuels.

    Science.gov (United States)

    Claxton, Larry D

    2015-01-01

    Much progress has been made in reducing the pollutants emitted from various combustors (including diesel engines and power plants) by the use of alternative fuels; however, much more progress is needed. Not only must researchers improve fuels and combustors, but also there is a need to improve the toxicology testing and analytical chemistry methods associated with these complex mixtures. Emissions from many alternative carbonaceous fuels are mutagenic and carcinogenic. Depending on their source and derivation, alternative carbonaceous fuels before combustion may or may not be genotoxic; however, in order to know their genotoxicity, appropriate chemical analysis and/or bioassay must be performed. Newly developed fuels and combustors must be tested to determine if they provide a public health advantage over existing technologies - including what tradeoffs can be expected (e.g., decreasing levels of PAHs versus increasing levels of NOx and possibly nitroarenes in ambient air). Another need is to improve exposure estimations which presently are a weak link in doing risk analyses. Copyright © 2014 Elsevier B.V. All rights reserved.

  15. Operation and maintenance of the RA reactor in 1964, I-II, Part I; Pogon i odrzavanje reaktora RA u 1964. godini, I-II, I Deo

    Energy Technology Data Exchange (ETDEWEB)

    Pavicevic, M [Institute of Nuclear Sciences Boris Kidric, Reaktor RA, Vinca, Beograd (Serbia and Montenegro)

    1964-12-15

    During 1964, the Reactor as operated about 20 days each months at nominal power of 6.5 MW, 5 days at lower power levels and 5 days were used for maintenance. Total production was 27930 MWh which is 11.7% higher than the planned value. Fuel exchange was done 3 times during this period, 98 spent fuel channels were exchanged. In addition to routine maintenance of reactor components and instruments a series of analyses of heavy water and helium were done. Special attention was devoted to corrosion analyses of the reactor materials because of the heavy water system was refurbished decontaminated in 1963. Utilization of the experimental space in the reactor was better that previously. 546 samples were irradiated till the end of November, of which 443 for users from the Institute. Specific irradiations in the fast neutron flux were done in six VISA-2 channels in the core.

  16. Improvement of Computer Codes Used for Fuel Behaviour Simulation (FUMEX-III). Report of a Coordinated Research Project 2008-2012

    International Nuclear Information System (INIS)

    2013-03-01

    It is fundamental to the future of nuclear power that reactors can be run safely and economically to compete with other forms of power generation. As a consequence, it is essential to develop the understanding of fuel performance and to embody that knowledge in codes to provide best estimate predictions of fuel behaviour. This in turn leads to a better understanding of fuel performance, a reduction in operating margins, flexibility in fuel management and improved operating economics. The IAEA has therefore embarked on a series of programmes addressing different aspects of fuel behaviour modelling with the following objectives: - To assess the maturity and prediction capabilities of fuel performance codes, and to support interaction and information exchange between countries with code development and application needs (FUMEX series); - To build a database of well defined experiments suitable for code validation in association with the OECD Nuclear Energy Agency (OECD/NEA); - To transfer a mature fuel modelling code to developing countries, to support teams in these countries in their efforts to adapt the code to the requirements of particular reactors, and to provide guidance on applying the code to reactor operation and safety assessments; - To provide guidelines for code quality assurance, code licensing and code application to fuel licensing. This report describes the results of the coordinated research project on the ''Improvement of computer codes used for fuel behaviour simulation (FUMEX-III)''. This programme was initiated in 2008 and completed in 2012. It followed previous programmes on fuel modelling: D-COM 1982-1984, FUMEX 1993-1996 and FUMEX-II 2002-2006. The participants used a mixture of data derived from commercial and experimental irradiation histories, in particular data designed to investigate the mechanical interactions occurring in fuel during normal, transient and severe transient operation. All participants carried out calculations on priority

  17. Combustion studies of coal derived solid fuels by thermogravimetric analysis. III. Correlation between burnout temperature and carbon combustion efficiency

    Science.gov (United States)

    Rostam-Abadi, M.; DeBarr, J.A.; Chen, W.T.

    1990-01-01

    Burning profiles of 35-53 ??m size fractions of an Illinois coal and three partially devolatilized coals prepared from the original coal were obtained using a thermogravimetric analyzer. The burning profile burnout temperatures were higher for lower volatile fuels and correlated well with carbon combustion efficiencies of the fuels when burned in a laboratory-scale laminar flow reactor. Fuels with higher burnout temperatures had lower carbon combustion efficiencies under various time-temperature conditions in the laboratory-scale reactor. ?? 1990.

  18. Short stack modeling of degradation in solid oxide fuel cells. Part II. Sensitivity and interaction analysis

    Science.gov (United States)

    Gazzarri, J. I.; Kesler, O.

    In the first part of this two-paper series, we presented a numerical model of the impedance behaviour of a solid oxide fuel cell (SOFC) aimed at simulating the change in the impedance spectrum induced by contact degradation at the interconnect-electrode, and at the electrode-electrolyte interfaces. The purpose of that investigation was to develop a non-invasive diagnostic technique to identify degradation modes in situ. In the present paper, we appraise the predictive capabilities of the proposed method in terms of its robustness to uncertainties in the input parameters, many of which are very difficult to measure independently. We applied this technique to the degradation modes simulated in Part I, in addition to anode sulfur poisoning. Electrode delamination showed the highest robustness to input parameter variations, followed by interconnect oxidation and interconnect detachment. The most sensitive degradation mode was sulfur poisoning, due to strong parameter interactions. In addition, we simulate several simultaneous two-degradation-mode scenarios, assessing the method's capabilities and limitations for the prediction of electrochemical behaviour of SOFC's undergoing multiple simultaneous degradation modes.

  19. Short stack modeling of degradation in solid oxide fuel cells. Part II. Sensitivity and interaction analysis

    Energy Technology Data Exchange (ETDEWEB)

    Gazzarri, J.I. [Department of Mechanical Engineering, University of British Columbia, 2054-6250 Applied Science Lane, Vancouver, BC V6T 1Z4 (Canada); Kesler, O. [Department of Mechanical and Industrial Engineering, University of Toronto, 5 King' s College Road, Toronto, ON M5S 3G8 (Canada)

    2008-01-21

    In the first part of this two-paper series, we presented a numerical model of the impedance behaviour of a solid oxide fuel cell (SOFC) aimed at simulating the change in the impedance spectrum induced by contact degradation at the interconnect-electrode, and at the electrode-electrolyte interfaces. The purpose of that investigation was to develop a non-invasive diagnostic technique to identify degradation modes in situ. In the present paper, we appraise the predictive capabilities of the proposed method in terms of its robustness to uncertainties in the input parameters, many of which are very difficult to measure independently. We applied this technique to the degradation modes simulated in Part I, in addition to anode sulfur poisoning. Electrode delamination showed the highest robustness to input parameter variations, followed by interconnect oxidation and interconnect detachment. The most sensitive degradation mode was sulfur poisoning, due to strong parameter interactions. In addition, we simulate several simultaneous two-degradation-mode scenarios, assessing the method's capabilities and limitations for the prediction of electrochemical behaviour of SOFC's undergoing multiple simultaneous degradation modes. (author)

  20. 14 CFR Appendix N to Part 25 - Fuel Tank Flammability Exposure and Reliability Analysis

    Science.gov (United States)

    2010-01-01

    ..., Definitions). A non-flammable ullage is one where the fuel-air vapor is too lean or too rich to burn or is... Office for approval the fuel tank flammability analysis, including the airplane-specific parameters...

  1. Fuel Element Technical Manual

    Energy Technology Data Exchange (ETDEWEB)

    Burley, H.H. [ed.

    1956-08-01

    It is the purpose of the Fuel Element Technical Manual to Provide a single document describing the fabrication processes used in the manufacture of the fuel element as well as the technical bases for these processes. The manual will be instrumental in the indoctrination of personnel new to the field and will provide a single data reference for all personnel involved in the design or manufacture of the fuel element. The material contained in this manual was assembled by members of the Engineering Department and the Manufacturing Department at the Hanford Atomic Products Operation between the dates October, 1955 and June, 1956. Arrangement of the manual. The manual is divided into six parts: Part I--introduction; Part II--technical bases; Part III--process; Part IV--plant and equipment; Part V--process control and improvement; and VI--safety.

  2. Hybrid fuel cell/diesel generation total energy system, part 2

    Science.gov (United States)

    Blazek, C. F.

    1982-11-01

    Meeting the Goldstone Deep Space Communications Complex (DGSCC) electrical and thermal requirements with the existing system was compared with using fuel cells. Fuel cell technology selection was based on a 1985 time frame for installation. The most cost-effective fuel feedstock for fuel cell application was identified. Fuels considered included diesel oil, natural gas, methanol and coal. These fuel feedstocks were considered not only on the cost and efficiency of the fuel conversion process, but also on complexity and integration of the fuel processor on system operation and thermal energy availability. After a review of fuel processor technology, catalytic steam reformer technology was selected based on the ease of integration and the economics of hydrogen production. The phosphoric acid fuel cell was selected for application at the GDSCC due to its commercial readiness for near term application. Fuel cell systems were analyzed for both natural gas and methanol feedstock. The subsequent economic analysis indicated that a natural gas fueled system was the most cost effective of the cases analyzed.

  3. Elasto-dynamic analysis of a gear pump-Part III: Experimental validation procedure and model extension to helical gears

    Science.gov (United States)

    Mucchi, E.; Dalpiaz, G.

    2015-01-01

    This work concerns external gear pumps for automotive applications, which operate at high speed and low pressure. In previous works of the authors (Part I and II, [1,2]), a non-linear lumped-parameter kineto-elastodynamic model for the prediction of the dynamic behaviour of external gear pumps was presented. It takes into account the most important phenomena involved in the operation of this kind of machine. The two main sources of noise and vibration are considered: pressure pulsation and gear meshing. The model has been used in order to foresee the influence of working conditions and design modifications on vibration generation. The model's experimental validation is a difficult task. Thus, Part III proposes a novel methodology for the validation carried out by the comparison of simulations and experimental results concerning forces and moments: it deals with the external and inertial components acting on the gears, estimated by the model, and the reactions and inertial components on the pump casing and the test plate, obtained by measurements. The validation is carried out comparing the level of the time synchronous average in the time domain and the waterfall maps in the frequency domain, with particular attention to identify system resonances. The validation results are satisfactory globally, but discrepancies are still present. Moreover, the assessed model has been properly modified for the application to a new virtual pump prototype with helical gears in order to foresee gear accelerations and dynamic forces. Part IV is focused on improvements in the modelling and analysis of the phenomena bound to the pressure evolution around the gears in order to achieve results closer to the measured values. As a matter of fact, the simulation results have shown that a variable meshing stiffness has a notable contribution on the dynamic behaviour of the pump but this is not as important as the pressure phenomena. As a consequence, the original model was modified with the

  4. Material control in nuclear fuel fabrication facilities. Part II. Accountability, instrumentation and measurement techniques in fuel fabrication facilities

    International Nuclear Information System (INIS)

    Borgonovi, G.M.; McCartin, T.J.; McDaniel, T.; Miller, C.L.; Nguyen, T.

    1978-01-01

    This report describes the measurement techniques, the instrumentation, and the procedures used in accountability and control of nuclear materials, as they apply to fuel fabrication facilities. A general discussion is given of instrumentation and measurement techniques which are presently used being considered for fuel fabrication facilities. Those aspects which are most significant from the point of view of satisfying regulatory constraints have been emphasized. Sensors and measurement devices have been discussed, together with their interfacing into a computerized system designed to permit real-time data collection and analysis. Estimates of accuracy and precision of measurement techniques have been given, and, where applicable, estimates of associated costs have been presented. A general description of material control and accounting is also included. In this section, the general principles of nuclear material accounting have been reviewed first (closure of material balance). After a discussion of the most current techniques used to calculate the limit of error on inventory difference, a number of advanced statistical techniques are reviewed. The rest of the section deals with some regulatory aspects of data collection and analysis, for accountability purposes, and with the overall effectiveness of accountability in detecting diversion attempts in fuel fabrication facilities. A specific example of application of the accountability methods to a model fuel fabrication facility is given. The effect of random and systematic errors on the total material uncertainty has been discussed, together with the effect on uncertainty of the length of the accounting period

  5. Material control in nuclear fuel fabrication facilities. Part II. Accountability, instrumentation and measurement techniques in fuel fabrication facilities

    Energy Technology Data Exchange (ETDEWEB)

    Borgonovi, G.M.; McCartin, T.J.; McDaniel, T.; Miller, C.L.; Nguyen, T.

    1978-01-01

    This report describes the measurement techniques, the instrumentation, and the procedures used in accountability and control of nuclear materials, as they apply to fuel fabrication facilities. A general discussion is given of instrumentation and measurement techniques which are presently used being considered for fuel fabrication facilities. Those aspects which are most significant from the point of view of satisfying regulatory constraints have been emphasized. Sensors and measurement devices have been discussed, together with their interfacing into a computerized system designed to permit real-time data collection and analysis. Estimates of accuracy and precision of measurement techniques have been given, and, where applicable, estimates of associated costs have been presented. A general description of material control and accounting is also included. In this section, the general principles of nuclear material accounting have been reviewed first (closure of material balance). After a discussion of the most current techniques used to calculate the limit of error on inventory difference, a number of advanced statistical techniques are reviewed. The rest of the section deals with some regulatory aspects of data collection and analysis, for accountability purposes, and with the overall effectiveness of accountability in detecting diversion attempts in fuel fabrication facilities. A specific example of application of the accountability methods to a model fuel fabrication facility is given. The effect of random and systematic errors on the total material uncertainty has been discussed, together with the effect on uncertainty of the length of the accounting period.

  6. Operation and maintenance of the RA reactor in 1964, I-II, Part II; Pogon i odrzavanje reaktora RA u 1964. godini, I-II, II Deo

    Energy Technology Data Exchange (ETDEWEB)

    Pavicevic, M [Institute of Nuclear Sciences Boris Kidric, Reaktor RA, Vinca, Beograd (Serbia and Montenegro)

    1964-12-15

    This volume of the report contains the following 15 Annexes: Improvement of the fuel cycle economy (record No. 37009803 in INIS DB); Analysis of neutron flux increase in horizontal experimental channels of the RA reactor record No. 37005698 in INIS DB); Application of the critical system for determining the thermal neutron flux in a research reactor with central horizontal reflector ( record No. 37055005 in INIS DB); Determining the capacity of the RA reactor heat exchanger dependent on the coolant water temperature and flow; Operation of the RA reactor in forced regime; Analysis of the CEN-132 heavy water pumps failures at the RA reactor from decontamination till present; Modifications in the vacuum loop of the distillation system; Report on decontamination of the evaporator and cleaning of the condenser of the distillation system; Operation of reactor at nominal power with reduced D{sub 2}O circulation; Cooling of the RA reactor with reduced flow rate in the heavy water loop; Measurement of the heavy water level in the fuel channels of the RA reactor; Conclusions of the experts group of the RA reactor at the meeting held on November 2 and 3 1964; Conclusions of the experts group at the meeting held on November 23 1964; After heat and the cooling problem after RA reactor shut-down; Measurement of noise and vibrations on the Ra reactor heavy water system; Calculation and measurement of the uranium temperature during irradiation in the experimental channel in the reflector of the RA reactor; Temperature measurement of the reactor materials samples irradiated in the fuel channels of the RA reactor; Study of the modifications in the synchronous generators, heavy water pumps and condenser batteries of the RA reactor.

  7. Emission factors of air pollutants from CNG-gasoline bi-fuel vehicles: Part I. Black carbon.

    Science.gov (United States)

    Wang, Yang; Xing, Zhenyu; Xu, Hui; Du, Ke

    2016-12-01

    Compressed natural gas (CNG) is considered to be a "cleaner" fuel compared to other fossil fuels. Therefore, it is used as an alternative fuel in motor vehicles to reduce emissions of air pollutants in transportation. To quantify "how clean" burning CNG is compared to burning gasoline, quantification of pollutant emissions under the same driving conditions for motor vehicles with different fuels is needed. In this study, a fleet of bi-fuel vehicles was selected to measure the emissions of black carbon (BC), carbon monoxide (CO), hydrocarbon (HC) and nitrogen oxide (NO x ) for driving in CNG mode and gasoline mode respectively under the same set of constant speeds and accelerations. Comparison of emission factors (EFs) for the vehicles burning CNG and gasoline are discussed. This part of the paper series reports BC EFs for bi-fuel vehicles driving on the real road, which were measured using an in situ method. Our results show that burning CNG will lead to 54%-83% reduction in BC emissions per kilometer, depending on actual driving conditions. These comparisons show that CNG is a cleaner fuel than gasoline for motor vehicles in terms of BC emissions and provide a viable option for reducing BC emissions cause by transportation. Copyright © 2016 Elsevier B.V. All rights reserved.

  8. The Feasibility of Administering a Practical Clinical Examination in Podiatry at a College of Podiatric Medicine: Results of a Field Trial Under Simulated Part III Test Conditions.

    Science.gov (United States)

    And Others; Valletta, Michael

    1978-01-01

    The results of a practical clinical examination in podiatric medicine administered to fourth-year students are presented. The examination could become the prototype of a Part III practical clinical examination under the auspices of the National Board of Podiatry Examiners. Its feasibility is established and problems and issues are discussed.…

  9. Proceedings of the Annual Meeting of the Association for Education in Journalism and Mass Communication (83rd, Phoenix, Arizona, August 9-12, 2000). Miscellaneous, Part III.

    Science.gov (United States)

    Association for Education in Journalism and Mass Communication.

    The Miscellaneous, part III section of the proceedings contains the following 11 papers: "The Relationship between Health and Fitness Magazine Reading and Eating-Disordered Weight-Loss Methods among High School Girls" (Steven R. Thomsen, Michelle M. Weber, and Lora Beth Brown); "A Practical Exercise for Teaching Ethical Decision…

  10. Compatibility analysis of DUPIC fuel (Part II) - Reactor physics design and analysis

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Chang Joon; Choi, Hang Bok; Rhee, Bo Wook; Roh, Gyu Hong; Kim, Do Hun [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2000-03-01

    The compatibility analysis of the DUPIC fuel in a CANDU reactor has been assessed. This study includes the fuel composition adjustment, comparison of lattice properties, performance analysis of reactivity devices, determination of regional over-power (ROP) trip setpoint, and uncertainty estimation of core performance parameters. For the DUPIC fuel composition adjustment, three options have been proposed, which can produce uniform neutronic characteristics of the DUPIC fuel. The lattice analysis has shown that the characteristics of the DUPIC fuel is compatible with those of natural uranium fuel. The reactivity devices of the CANDU-6 reactor maintain their functional requirements even for the DUPIC fuel system. The ROP analysis has shown that the trip setpoint is not sacrificed for the DUPIC fuel system owing to the power shape that enhances more thermal margin. The uncertainty analysis of the core performance parameter has shown that the uncertainty associated with the fuel composition variation is reduced appreciably, which is primarily due to the fuel composition adjustment and secondly the on-power refueling feature and spatial control function of the CANDU reactor. The reactor physics calculation has also shown that it is feasible to use spent PWR fuel directly in CANDU reactors without deteriorating the CANDU-6 core physics design requirements. 29 refs., 67 figs., 60 tabs. (Author)

  11. Study of brushless fuel pump (improvement of pump and motor parts). 2nd Report. Blushless dendo fuel pump no kento. 2

    Energy Technology Data Exchange (ETDEWEB)

    Mine, K; Takada, S; Tatematsu, M; Takeuchi, H [Aisan Industry Co. Ltd., Aichi (Japan)

    1992-10-01

    A methanol use electrically driven fuel pump was developed as reported in the present report. Mixed fuel of gasoline with alcohol can be handled by a brushless fuel pump which was proposed and improved as reported. The flow rate performance was heightened to 25g/sec by heightening in output power of motor, while the high temperature performance was 17% heightened against the conventional ratio of lowering in flow rate by heightening in vapor jet capacity. Against the corrosiveness of methanol, an in-tank type was applied to the pump, and all its electrically conductive and other mechanical parts were made to be both anti-corrosive and anti-abrasive. It is structurally of a two-stage series turbine type of non-volume form. A sensor method was applied to the motor by confining the miniaturized control circuit of brushless motor in the motor so that the transistor is controlled against the heightening in temperature. The motor is a three-phase half-wave driving motor. Also developed was a fuel supply system which is useful for the mixed fuel covering a range of 100% methanol through 100% gasoline. The present pump is dimensionally interchangeable with the conventional gasoline use one. Its operational life is more than 10000 hours. 3 refs., 17 figs., 1 tab.

  12. Further analysis of extended storage of spent fuel. Final report of a co-ordinated research programme on the behaviour of spent fuel assemblies during extended storage (BEFAST-III) 1991-1996

    International Nuclear Information System (INIS)

    1997-05-01

    Considerable quantities of spent fuel continue to be produced and to accumulate in a number of countries. Although some new reprocessing facilities have been constructed, many countries are investigating the option of extended spent fuel storage prior to reprocessing or fuel disposal. Wet storage continues to predominate as an established technology. However, dry storage is becoming increasingly used with many countries considering dry storage for the longer term. This Technical Document is the final report of the IAEA Co-ordinated Research Programme on the Behaviour of Spent Fuel Assemblies During Extended Storage (BEFAST-III, 1991-1996). It contains analyses of wet and dry spent fuel storage technologies obtained from 16 organizations representing 13 countries (Canada, Finland, France, Germany, Hungary, the Republic of Korea, Japan, the Russian Federation, Slovakia, Spain, Sweden, the United Kingdom and the USA) which participated in the co-ordinated research programme as participants or observers. The report contains information presented during the three Research Co-ordination meetings and also data which were submitted by the participants in response to request by the Scientific Secretary. 48 refs, 4 tabs

  13. Oxidation of dibenzothiophene as a model substrate for the removal of organic sulphur from fossil fuels by iron(III ions generated from pyrite by Acidithiobacillus ferrooxidans

    Directory of Open Access Journals (Sweden)

    VLADIMIR P. BESKOSKI

    2007-06-01

    Full Text Available Within this paper a new idea for the removal of organically bonded sulphur from fossil fuels is discussed. Dibenzothiophene (DBT was used as a model compound of organicmolecules containing sulphur. This form of (biodesulphurization was performed by an indirect mechanism in which iron(III ions generated from pyrite by Acidithiobacillus ferrooxidans performed the abiotic oxidation. The obtained reaction products, dibenzothiopene sulfoxide and dibenzothiophene sulfone, are more soluble in water than the basic substrate and the obtained results confirmed the basic hypothesis and give the posibility of continuing the experiments related to application of this (biodesulphurization process.

  14. Fission products and nuclear fuel behaviour under severe accident conditions part 2: Fuel behaviour in the VERDON-1 sample

    Science.gov (United States)

    Geiger, E.; Le Gall, C.; Gallais-During, A.; Pontillon, Y.; Lamontagne, J.; Hanus, E.; Ducros, G.

    2017-11-01

    Within the framework of the International Source Term Programme (ISTP), the VERDON programme aims at quantifying the source term of radioactive materials in case of a hypothetical severe accident in a light water reactor (LWR). Tests were performed in a new experimental laboratory (VERDON) built in the LECA-STAR facility (CEA Cadarache). The VERDON-1 test was devoted to the study of a high burn-up UO2 fuel and FP releases at very high temperature (≈2873 K) in a reducing atmosphere. Post-test qualitative and quantitative characterisations of the VERDON-1 sample led to the proposal of a scenario explaining the phenomena occurring during the experimental sequence. Hence, the fuel and the cladding may have interacted which led to the melting of UO2-ZrO2 alloy. Although no relocation was observed during the test, it may have been imminent.

  15. Assesment of the energy quality of the synthesis gas produced from biomass derived fuels conversion: Part I: Liquid Fuels, Ethanol

    International Nuclear Information System (INIS)

    Arteaga Perez, Luis E; Casas, Yannay; Peralta, Luis M; Granda, Daikenel; Prieto, Julio O

    2011-01-01

    The use of biofuels plays an important role to increase the efficiency and energetic safety of the energy processes in the world. The main goal of the present research is to study from the thermodynamics and kinetics the effect of the operational variables on the thermo-conversion processes of biomass derived fuels focused on ethanol reforming. Several models are developed to assess the technological proposals. The minimization of Gibbs free energy is the criterion applied to evaluate the performance of the different alternatives considering the equilibrium constraints. All the models where validated on an experimental data base. The gas composition, HHV and the ratio H2/CO are used as measures for the process efficiency. The operational parameters are studied in a wide range (reactants molar ratio, temperature and oxygen/fuel ratio). (author)

  16. Stationary liquid fuel fast reactor SLFFR – Part I: Core design

    Energy Technology Data Exchange (ETDEWEB)

    Jing, T.; Yang, G.; Jung, Y.S.; Yang, W.S., E-mail: yang494@purdue.edu

    2016-12-15

    Highlights: • An innovative fast reactor concept SLFFR based on liquid metal fuel is proposed for TRU burning. • A compact core design of 1000 MWt SLFFR is developed to achieve a zero conversion ratio and passive safety. • The core size and the control requirement are significantly reduced compared to the conventional solid fuel reactor with same conversion ratio. - Abstract: For effective burning of hazardous transuranic (TRU) elements of used nuclear fuel, a transformational advanced reactor concept named the stationary liquid fuel fast reactor (SLFFR) has been proposed based on a stationary molten metallic fuel. A compact core design of a 1000 MWt SLFFR has been developed using TRU-Ce-Co fuel, Ta-10W fuel container, and sodium coolant. Conservative design approaches have been adopted to stay within the current material performance database. Detailed neutronics and thermal-fluidic analyses have been performed to evaluate the steady-state performance characteristics. The analysis results indicate that the SLFFR of a zero TRU conversion ratio is feasible while satisfying the conservatively imposed thermal design constraints. A theoretical maximum TRU consumption rate of 1.01 kg/day is achieved with uranium-free fuel. Compared to the solid fuel reactors with the same TRU conversion ratio, the core size and the reactivity control requirement are reduced significantly. The primary and secondary control systems provide sufficient shutdown margins, and the calculated reactivity feedback coefficients show that the prompt fuel expansion coefficient is sufficiently negative.

  17. Cost evaluation of a commercial-scale DUPIC fuel fabrication facility (Part I) -Summary

    Energy Technology Data Exchange (ETDEWEB)

    Ko, Won Il; Choi, Hang Bok; Yang, Myung Seung [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-08-01

    A conceptual design of a commercial scale DUPIC fuel fabrication facility was initiated to provide some insights into the costs associated with construction, operation, and decommissioning. The primary conclusion of this report is that it is feasible to design, license, construct, test, and operate a facility that will process 400 MTHE/yr of spent PWR fuel and reconfigure the fuel into CANDU fuel bundles at a reasonable unit cost of the fuel material. Although DUPIC fuel fabrication by vibropacking method is clearly cheaper than that of the pellet method, the feasibility of vibropac technology for DUPIC fuel fabrication and use of vibroac fuel in CANDU reactors may has to be studied in depth in order to use as an alternative to the conventional pellet fuel method. Especially, there are some questions on meeting the CANDU requirements in thermal and mechanical terms as well as density of fuel. Wherever possible, this report used representative costs of currently available technologies as the bases for cost estimation. It should also be noted that the conceptual design and cost information contained in this report was extracted from the public domain and general open literature. Later studies have to focus on other important areas of concern such as safety, security, safeguards, process optimization etc. 7 figs., 6 tabs. (Author)

  18. Water reactor fuel characterization. Part of a coordinated programme on non-destructive techniques for reactor fuel characterization

    International Nuclear Information System (INIS)

    Levai, F.

    1983-06-01

    The report describes an optical/mechanical system of examining nuclear reactor fuel bundles by tomographic imaging using high contrast X-ray film. This low cost system does not use expensive detectors or digital computers. The apparatus assembled from ordinary and available components consists of a 2π scanner, a back projector and filters. Although the system described and tested is based on transmission tomography, the report also discusses the extension of the concept to emission tomography

  19. Effect of secondary fuels and combustor temperature on mercury speciation in pulverized fuel co-combustion: part 1

    Energy Technology Data Exchange (ETDEWEB)

    Shishir P. Sable; Wiebren de Jong; Ruud Meij; Hartmut Spliethoff [Delft University Technology, Delft (Netherlands). Section Energy Technology, Department of Process and Energy

    2007-08-15

    The present work mainly involves bench scale studies to investigate partitioning of mercury in pulverized fuel co-combustion at 1000 and 1300{sup o}C. High volatile bituminous coal is used as a reference case and chicken manure, olive residue, and B quality (demolition) wood are used as secondary fuels with 10 and 20% thermal shares. The combustion experiments are carried out in an entrained flow reactor with a fuel input of 7-8 kWth. Elemental and total gaseous mercury concentrations in the flue gas of the reactor are measured on-line, and ash is analyzed for particulate mercury along with other elemental and surface properties. Animal waste like chicken manure behaves very differently from plant waste. The higher chlorine contents of chicken manure cause higher ionic mercury concentrations whereas even with high unburnt carbon, particulate mercury reduces with increase in the chicken manure share. This might be a problem due to coarse fuel particles, low surface area, and iron contents. B-wood and olive residue cofiring reduces the emission of total gaseous mercury and increases particulate mercury capture due to unburnt carbon formed, fine particles, and iron contents of the ash. Calcium in chicken manure does not show any effect on particulate or gaseous mercury. It is probably due to a higher calcium sulfation rate in the presence of high sulfur and chlorine contents. However, in plant waste cofiring, calcium may have reacted with chlorine to reduce ionic mercury to its elemental form. According to thermodynamic predictions, almost 50% of the total ash is melted to form slag at 1300{sup o}C in cofiring because of high calcium, iron, and potassium and hence mercury and other remaining metals are concentrated in small amounts of ash and show an increase at higher temperatures. No slag formation was predicted at 1000{sup o}C. 24 refs., 8 figs., 4 tabs.

  20. Nonresidential buildings energy consumption survey: 1979 consumption and expenditures. Part 2. Steam, fuel oil, LPG, and all fuels

    Energy Technology Data Exchange (ETDEWEB)

    Patinkin, L.

    1983-12-01

    This report presents data on square footage and on total energy consumption and expenditures for commercial buildings in the contiguous United States. Also included are detailed consumption and expenditures tables for fuel oil or kerosene, liquid petroleum gas (LPG), and purchased steam. Commercial buildings include all nonresidential buildings with the exception of those where industrial activities occupy more of the total square footage than any other type of activity. 7 figures, 23 tables.

  1. Air quality and climate change, Topic 3 of the Model Inter-Comparison Study for Asia Phase III (MICS-Asia III) - Part 1: Overview and model evaluation

    Science.gov (United States)

    Gao, Meng; Han, Zhiwei; Liu, Zirui; Li, Meng; Xin, Jinyuan; Tao, Zhining; Li, Jiawei; Kang, Jeong-Eon; Huang, Kan; Dong, Xinyi; Zhuang, Bingliang; Li, Shu; Ge, Baozhu; Wu, Qizhong; Cheng, Yafang; Wang, Yuesi; Lee, Hyo-Jung; Kim, Cheol-Hee; Fu, Joshua S.; Wang, Tijian; Chin, Mian; Woo, Jung-Hun; Zhang, Qiang; Wang, Zifa; Carmichael, Gregory R.

    2018-04-01

    Topic 3 of the Model Inter-Comparison Study for Asia (MICS-Asia) Phase III examines how online coupled air quality models perform in simulating high aerosol pollution in the North China Plain region during wintertime haze events and evaluates the importance of aerosol radiative and microphysical feedbacks. A comprehensive overview of the MICS-Asia III Topic 3 study design, including descriptions of participating models and model inputs, the experimental designs, and results of model evaluation, are presented. Six modeling groups from China, Korea and the United States submitted results from seven applications of online coupled chemistry-meteorology models. Results are compared to meteorology and air quality measurements, including data from the Campaign on Atmospheric Aerosol Research Network of China (CARE-China) and the Acid Deposition Monitoring Network in East Asia (EANET). The correlation coefficients between the multi-model ensemble mean and the CARE-China observed near-surface air pollutants range from 0.51 to 0.94 (0.51 for ozone and 0.94 for PM2.5) for January 2010. However, large discrepancies exist between simulated aerosol chemical compositions from different models. The coefficient of variation (SD divided by the mean) can reach above 1.3 for sulfate in Beijing and above 1.6 for nitrate and organic aerosols in coastal regions, indicating that these compositions are less consistent from different models. During clean periods, simulated aerosol optical depths (AODs) from different models are similar, but peak values differ during severe haze events, which can be explained by the differences in simulated inorganic aerosol concentrations and the hygroscopic growth efficiency (affected by varied relative humidity). These differences in composition and AOD suggest that future models can be improved by including new heterogeneous or aqueous pathways for sulfate and nitrate formation under hazy conditions, a secondary organic aerosol (SOA) formation chemical

  2. Emission factors of air pollutants from CNG-gasoline bi-fuel vehicles: Part II. CO, HC and NOx.

    Science.gov (United States)

    Huang, Xiaoyan; Wang, Yang; Xing, Zhenyu; Du, Ke

    2016-09-15

    The estimation of emission factors (EFs) is the basis of accurate emission inventory. However, the EFs of air pollutants for motor vehicles vary under different operating conditions, which will cause uncertainty in developing emission inventory. Natural gas (NG), considered as a "cleaner" fuel than gasoline, is increasingly being used to reduce combustion emissions. However, information is scarce about how much emission reduction can be achieved by motor vehicles burning NG (NGVs) under real road driving conditions, which is necessary for evaluating the environmental benefits for NGVs. Here, online, in situ measurements of the emissions from nine bi-fuel vehicles were conducted under different operating conditions on the real road. A comparative study was performed for the EFs of black carbon (BC), carbon monoxide (CO), hydrocarbons (HCs) and nitrogen oxides (NOx) for each operating condition when the vehicles using gasoline and compressed NG (CNG) as fuel. BC EFs were reported in part I. The part II in this paper series reports the influence of operating conditions and fuel types on the EFs of CO, HC and NOx. Fuel-based EFs of CO showed good correlations with speed when burning CNG and gasoline. The correlation between fuel-based HC EFs and speed was relatively weak whether burning CNG or gasoline. The fuel-based NOx EFs moderately correlated with speed when burning CNG, but weakly correlated with gasoline. As for HC, the mileage-based EFs of gasoline vehicles are 2.39-12.59 times higher than those of CNG vehicles. The mileage-based NOx EFs of CNG vehicles are slightly higher than those of gasoline vehicles. These results would facilitate a detailed analysis of the environmental benefits for replacing gasoline with CNG in light duty vehicles. Copyright © 2016 Elsevier B.V. All rights reserved.

  3. Development of metal fuel and study of construction materials (I-IV), Part II

    International Nuclear Information System (INIS)

    Mihajlovic, A.

    1965-11-01

    The studies were devoted to problems related to application of metal uranium as fuel in heavy water reactors. Influence of thermal treatment on material texture and recrystallization of cast uranium was investigated. Structural changes of uranium alloys with molybdenum and niobium were tested during different heat treatments. A review of the possibilities for using metal uranium fuel in heavy water reactors is included

  4. Environmental impact data for fuels. Part 2: Background information and technical appendix (New revised edition)

    International Nuclear Information System (INIS)

    Uppenberg, S.; Almemark, M.; Brandel, M.; Lindfors, L.G.; Marcus, H.O.; Stripple, H.; Wachtmeister, A.; Zetterberg, L.

    2001-05-01

    This report is a compilation of data concerning environmental impacts from the utilization of different fuels. The entire life cycle is studied, from the extraction of raw materials to combustion. The fuels under study are gasoline, gasoline with MTBE, diesel, fuel oil, LPG, coal, natural gas, peat, refuse, ethanol, RME, DME, methane and wood fuels (forestry residues, Salix, pellets/briquettes). Utilization areas studied are heating plants, cogeneration plants, power plants, domestic boilers, and light and heavy vehicles. In this new edition, the following changes were made: New life cycle analyses have been included, a few new fuels added, electricity from hydroelectric plants, wind power plants and nuclear power plants have been included and some other minor changes

  5. A Main Steam Safety Valve (MSSV) With Fixed Blowdown According to ASME Section III,Part NC-7512

    International Nuclear Information System (INIS)

    Follmer, Bernhard; Schnettler, Armin

    2002-01-01

    In 1986, the NRC issued the Information Notice (IN) 86-05 'Main Steam Safety Valve test failures and ring setting adjustments'. Shortly after this IN was issued, the Code was revised to require that a full flow test has to be performed on each CL.2 MSSV by the manufacturer to verify that the valve was adjusted so that it would reach full lift and thus full relieving capacity and would re-close at a pressure as specified in the valve Design Specification. In response to the concern discussed in the IN, the Westinghouse Owners Group (WOG) performed extensive full flow testing on PWR MSSVs and found that each valve required a unique setting of a combination of two rings in order to achieve full lift at accumulation of 3% and re-closing at a blowdown of 5%. The Bopp and Reuther MSSV type SiZ 2507 has a 'fixed blowdown' i.e. without any adjusting rings to adjust the 'blowdown' so that the blowdown is 'fixed'. More than 1000 pieces of this type are successfully in nuclear power plants in operation. Many of them since about 25 years. Therefore it can be considered as a proven design. It is new that an optimization of this MSSV type SiZ 2507 fulfill the requirements of part NC-7512 of the ASME Section III although there are still no adjusting rings in the flow part. In 2000, for the Qinshan Candu unit 1 and 2 full flow tests were performed with 32 MSSV type SiZ 2507 size 8'' x 12'' at 51 bar saturated steam in only 6 days. In all tests the functional performance was very stable. It was demonstrated by recording the signals lift and system pressure that all valves had acceptable results to achieve full lift at accumulation of 3% and to re-close at blowdown of 5%. This is an advantage which gives a reduction in cost for flow tests and which gives more reliability after maintenance work during outage compared to the common MSSV design with an individual required setting of the combination of the two rings. The design of the type SiZ 2507 without any adjusting rings in the

  6. AUTOMOTIVE DIESEL MAINTENANCE 1. UNIT XI, PART I--MAINTAINING THE FUEL SYSTEM (PART I), CUMMINS DIESEL ENGINES, PART II--UNIT REPLACEMENT (ENGINE).

    Science.gov (United States)

    Human Engineering Inst., Cleveland, OH.

    THIS MODULE OF A 30-MODULE COURSE IS DESIGNED TO DEVELOP AN UNDERSTANDING OF DIFFERENCES BETWEEN TWO AND FOUR CYCLE ENGINES, THE OPERATION AND MAINTENANCE OF THE DIESEL ENGINE FUEL SYSTEM, AND THE PROCEDURES FOR DIESEL ENGINE REMOVAL. TOPICS ARE (1) REVIEW OF TWO CYCLE AND FOUR CYCLE CONCEPT, (2) SOME BASIC CHARACTERISTICS OF FOUR CYCLE ENGINES,…

  7. Evaluation of biodiesel fuel and a diesel oxidation catalyst in an underground metal mine : Part 3 : Biological and chemical characterization

    Energy Technology Data Exchange (ETDEWEB)

    Bagley, S.T. [Michigan Technological Univ., Houghton, MI (United States). Dept. of Biological Sciences; Gratz, L.D. [Michigan Technological Univ., Houghton, MI (United States). Dept. of Mechanical Engineering-Engineering Mechanics

    1998-07-24

    A collaborative, international, multidisciplinary effort led to the evaluation of the effects of using a 50 per cent biodiesel fuel blend and an advanced-type diesel oxidation catalyst (DOC) on underground metal mine air quality. The location selected for the field trials was the Creighton Mine 3 in Sudbury, Ontario, operated by Inco. Specifically, part 3 of the study evaluated the effects of using a biodiesel blend fuel on potentially health-related diesel particulate matter (DPM) components, with a special emphasis on polynuclear aromatic hydrocarbons (PAH), nitro-PAH, and mutagenic activity. High volume sampler filters containing submicrometer particles were examined, and comparisons made for DPM and DPM component concentrations. The downwind concentrations of DPM were reduced by 20 per cent with the use of the blend biodiesel fuel as compared with the number 2 diesel fuel with an advanced-type DOC. Significant reductions in solids (up to 30 per cent) and up to 75 per cent in the case of mutagenic activity were noted. Significant reductions in the DPM components potentially harmful to human health should result from the use of this blended fuel combined with an advanced-type DOC in an underground environment. 23 refs., 19 tabs.

  8. Fuel depletion analyses for the HEU core of GHARR-1: Part II: Fission product inventory

    International Nuclear Information System (INIS)

    Anim-Sampong, S.; Akaho, E.H.K.; Boadu, H.O.; Intsiful, J.D.K.; Osae, S.

    1999-01-01

    The fission product isotopic inventories have been estimated for a 90.2% highly enriched uranium (HEU) fuel lattice cell of the Ghana Research Reactor-1 (GHARR-1) using the WIMSD/4 transport lattice code. The results indicate a gradual decrease in the Xe 135 inventory, and saturation trend for Sm 149 , Cs 134 and Cs 135 inventories as the fuel is depleted to 10,000 MWd/tU. (author)

  9. Fuel cell-gas turbine hybrid system design part II: Dynamics and control

    Science.gov (United States)

    McLarty, Dustin; Brouwer, Jack; Samuelsen, Scott

    2014-05-01

    Fuel cell gas turbine hybrid systems have achieved ultra-high efficiency and ultra-low emissions at small scales, but have yet to demonstrate effective dynamic responsiveness or base-load cost savings. Fuel cell systems and hybrid prototypes have not utilized controls to address thermal cycling during load following operation, and have thus been relegated to the less valuable base-load and peak shaving power market. Additionally, pressurized hybrid topping cycles have exhibited increased stall/surge characteristics particularly during off-design operation. This paper evaluates additional control actuators with simple control methods capable of mitigating spatial temperature variation and stall/surge risk during load following operation of hybrid fuel cell systems. The novel use of detailed, spatially resolved, physical fuel cell and turbine models in an integrated system simulation enables the development and evaluation of these additional control methods. It is shown that the hybrid system can achieve greater dynamic response over a larger operating envelope than either individual sub-system; the fuel cell or gas turbine. Results indicate that a combined feed-forward, P-I and cascade control strategy is capable of handling moderate perturbations and achieving a 2:1 (MCFC) or 4:1 (SOFC) turndown ratio while retaining >65% fuel-to-electricity efficiency, while maintaining an acceptable stack temperature profile and stall/surge margin.

  10. Safety analysis of RA reactor operation, I-II, Part I - RA reactor technical and operation characteristics; Analiza sigurnosti rada reaktora RA - I-III, I deo - Tehnicke i pogonske karakteristike reaktora RA

    Energy Technology Data Exchange (ETDEWEB)

    Raisic, N [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1963-02-15

    RA research reactor is a thermal, heavy water moderated system with graphite reflector having nominal power 6.5 MW. The 2% enriched metal uranium fuel in the reactor core produces mean thermal neutron flux of 2.9 10{sup 13} neutrons/cm{sup 2} s, and maximum neutron flux 5.5 10{sup 13} neutrons/cm{sup 2} s. main components of the reactor described in this report are: rector core, reflector, biological shield, heavy water cooling system, ordinary water cooling system, helium system, reactor control system, reactor safety system, dosimetry system, power supply system, and fuel transport system. Detailed reactor properties and engineering drawings of all the system are part of this volume.

  11. [Operative treatment of traumatic fractures of the thoracic and lumbar spinal column: Part III: Follow up data].

    Science.gov (United States)

    Reinhold, M; Knop, C; Beisse, R; Audigé, L; Kandziora, F; Pizanis, A; Pranzl, R; Gercek, E; Schultheiss, M; Weckbach, A; Bühren, V; Blauth, M

    2009-03-01

    In this third and final part, the Spine Study Group (AG WS) of the German Trauma Association (DGU) presents the follow-up (NU) data of its second, prospective, internet-based multicenter study (MCS II) for the treatment of thoracic and lumbar spinal injuries including 865 patients from 8 trauma centers. Part I described in detail the epidemiologic data of the patient collective and the subgroups, whereas part II analyzed the different methods of treatment and radiologic findings. The study period covered the years 2002 to 2006 including a 30-month follow-up period from 01.01.2004 until 31.05.2006. Follow-up data of 638 (74%) patients were collected with a new internet-based database system and analyzed. Results in part III will be presented on the basis of the same characteristic treatment subgroups (OP, KONS, PLASTIE) and surgical treatment subgroups (Dorsal, Ventral, Kombi) in consideration of the level of injury (thoracic spine, thoracolumbar junction, lumbar spine). After the initial treatment and discharge from hospital, the average duration of subsequent inpatient rehabilitation was 4 weeks, which lasted significantly longer in patients with persistent neurologic deficits (mean 10.9 weeks) or polytraumatized patients (mean 8.6 weeks). Following rehabilitation on an inpatient basis, subsequent outpatient rehabilitation lasted on average 4 months. Physical therapy was administered significantly longer to patients with neurologic deficits (mean 8.7 months) or type C injuries (mean 8.6 months). The level of injury had no influence of the duration of the inpatient or outpatient rehabilitation. A total of 382 (72.2%) patients who were either operated from posterior approach only or in a combined postero-anterior approach had an implant removal after an average 12 months. During the follow-up period 56 (8.8%) patients with complications were registered and of these 18 (2.8%) had to have surgical revision. The most common complications reported were infection, loss

  12. Nuclear power, nuclear fuel cycle and waste management: Status and trends 1995. Part C of the IAEA Yearbook 1995

    International Nuclear Information System (INIS)

    1995-09-01

    This report was jointly prepared by the Division of Nuclear Power and the Division of Nuclear Fuel Cycle and Waste Management as part of an annual overview of both global nuclear industry activities and related IAEA programmes. This year's report focuses on activities during 1994 and the status at the end of that year. The trends in the industry are projected to 2010. Special events and highlights of IAEA activities over the past year are also presented. Refs, figs and tabs

  13. Nuclear Power, nuclear fuel cycle and waste management: Status and trends 1996. Part C of the IAEA yearbook 1996

    International Nuclear Information System (INIS)

    1996-09-01

    This report was jointly prepared by the Division of Nuclear Power and the Division of Nuclear Fuel Cycle and Waste Management as part of an annual overview of both global nuclear industry activities and related IAEA programmes. This year's report focuses on activities during 1995 and the status at the end of that year. The trends in the industry are projected to the year 2010. Special events and highlights of IAEA activities over the past year are also presented. Refs, figs, tabs

  14. SIMMER-III parametric studies of fuel-steel mixing and radial mesh effects on power excursion in ESFR ULOF transients - 15033

    International Nuclear Information System (INIS)

    Chen, X.N.; Rineiski, A.; Gabrielli, F.; Andriolo, L.; Li, R.; Maschek, W.

    2015-01-01

    This paper deals with SIMMER-III once-through simulations of the first power excursion initiated by an unprotected loss of flow (ULOF) in the Working Horse design of the European Sodium Cooled Fast Reactor (ESFR). Since the sodium void effect is strictly positive in this core and dominant in the transient, a power excursion is initiated by sodium boiling in the ULOF case. Two major effects, namely (1) reactivity effects due to fuel-steel mixing after melting and (2) the radial mesh size, which were not considered initially in SIMMER simulations for ESFR, are studied. The first effect concerns the reactivity difference between the heterogeneous fuel/clad/wrapper configuration and the homogeneous mixture of steel and fuel. The full core homogenization (due to melting) effect is ∼ 2 dollars, though a smaller effect takes place in case of partial core melting. The second effect is due to the SIMMER sub-assembly coarse mesh treatment, where a simultaneous sodium boiling onset in all sub-assemblies belonging to one ring leads to an overestimated reactivity ramp. For investigating the influence of fuel/steel mixing effects, a lumped 'homogenization' reactivity feedback has been introduced, being proportional to the molten steel mass. For improving the coarse mesh treatment, we employ finer radial meshes to take the subchannel effects into account, where the side and interior channels have different coolant velocities and temperatures. The simulation results show that these two effects have significant impacts on the first power excursion after the sodium boiling: both effects delay the power excursion and significantly reduce the height of the power peaks in case of a ULOF

  15. High-Uranium-Loaded U3O8-Al fuel element development program. Part 1

    International Nuclear Information System (INIS)

    Martin, M.M.

    1993-01-01

    The High-Uranium-Loaded U 3 O 8 -Al Fuel Element Development Program supports Argonne National Laboratory efforts to develop high-uranium-density research and test reactor fuel to accommodate use of low-uranium enrichment. The goal is to fuel most research and test reactors with uranium of less than 20% enrichment for the purpose of lowering the potential for diversion of highly-enriched material for nonpeaceful usages. The specific objective of the program is to develop the technological and engineering data base for U 3 O 8 -Al plate-type fuel elements of maximal uranium content to the point of vendor qualification for full scale fabrication on a production basis. A program and management plan that details the organization, supporting objectives, schedule, and budget is in place and preparation for fuel and irradiation studies is under way. The current programming envisions a program of about four years duration for an estimated cost of about two million dollars. During the decades of the fifties and sixties, developments at Oak Ridge National Laboratory led to the use of U 3 O 8 -Al plate-type fuel elements in the High Flux Isotope Reactor, Oak Ridge Research Reactor, Puerto Rico Nuclear Center Reactor, and the High Flux Beam Reactor. Most of the developmental information however applies only up to a uranium concentration of about 55 wt % (about 35 vol % U 3 O 8 ). The technical issues that must be addressed to further increase the uranium loading beyond 55 wt % U involve plate fabrication phenomena of voids and dogboning, fuel behavior under long irradiation, and potential for the thermite reaction between U 3 O 8 and aluminum

  16. Field Surveys, IOC Valleys. Volume III, Part II. Cultural Resources Survey, Pine and Wah Wah Valleys, Utah.

    Science.gov (United States)

    1981-08-01

    including horse, camel, mammoth, Ertm E-TR-48-III-II 20 musk ox, and certain species of bison, goat, and bear, which had previously inhabited the marsh and...34 - - -9,$.. 𔄃 Im I I I Si to * Location lype/Contents Affiliation 42B@644 rid e over cr ek - P/J depression, cleared areas, Fr elon (f4-5-18-92) ground

  17. Part-load performance and emissions of a spark ignition engine fueled with RON95 and RON97 gasoline: Technical viewpoint on Malaysia’s fuel price debate

    International Nuclear Information System (INIS)

    Mohamad, Taib Iskandar; How, Heoy Geok

    2014-01-01

    Highlights: • Recent Malaysia’s gasoline price hike affects mass perception and vehicle sales. • Effects of RON95 and RON97 on a representative engine was experimentally studied. • RON95 produced better torque, power, fuel efficiency and lower NO x . • RON97 gasoline resulted in lower BSFC and lower emissions of CO 2 , CO and HC. • Performance-emission-price cross-analysis indicated RON95 as the better option. - Abstract: Due to world crude oil price hike in the recent years, many countries have experienced increase in gasoline price. In Malaysia, where gasoline are sold in two grades; RON95 and RON97, and fuel price are regulated by the government, gasoline price have been gradually increased since 2009. Price rise for RON97 is more significant. By 2014, its per liter price is 38% more than that of RON95. This has resulted in escalated dissatisfaction among the mass. People argued they were denied from using a better fuel (RON97). In order to evaluate the claim, there is a need to investigate engine response to these two gasoline grades. The effect of gasoline RON95 and RON97 on performance and exhaust emissions in spark ignition engine was investigated on a representative engine: 1.6L, 4-cylinder Mitsubishi 4G92 engine with CR 11:1. The engine was run at constant speed between 1500 and 3500 rpm with 500 rpm increment at various part-load conditions. The original engine ECU, a hydraulic dynamometer and control, a combustion analyzer and an exhaust gas analyzer were used to determine engine performance, cylinder pressure and emissions. Results showed that RON95 produced higher engine performance for all part-load conditions within the speed range. RON95 produced on average 4.4% higher brake torque, brake power, brake mean effective pressure as compared to RON97. The difference in engine performance was more significant at higher engine speed and loads. Cylinder pressure and ROHR were evaluated and correlated with engine output. With RON95, the engine

  18. Mechanical analysis of cylindrical part of canisters for spent nuclear fuel

    International Nuclear Information System (INIS)

    Ikonen, K.

    2005-06-01

    This report describes mechanical analyses of cylindrical part of the VVER 440-, BWR and EPR-type canisters for spent nuclear fuel. The task was first to evaluate the stresses at maximum design pressure and further by increasing pressure load to determine the limit collapse load and corresponding safety factor. Maximum design pressure 44 MPa is a sum of the hydrostatic pressure 30 MPa caused by 3 km ice layer, 7 MPa caused by ground water pressure at the deepest disposal depth of 700 m and 7 MPa from bentonite swelling pressure. The analysis presented in this report concern the middle area of the canisters, where the cast iron insert is considered to be more critical than in the ends of the canister. For the model a piece from the middle area of the canister was separated by two planes perpendicular to the axis of the canister. This piece was studied first by two-dimensional plane strain model, where the planes are constrained and no elongation of the canister takes place. In the second model one of the planes was constrained and the other plane was allowed to displace in axial direction, which remains as a plane during deformation and to which axial pressure force is directed. This analysis, which corresponds better the real condition in the canister, was performed as threedimensional. The analyses gave however practically equal results due to plastic deformation. Thus the analysis can be done by two-dimensional plane strain model leading to same accuracy with less computation effort. Analyses were performed as large displacement and large strain analyses by the PASULA computing package, which has been developed at VTT for a variety of structural analysis and for heat conduction calculations. A special routine was developed for automatic mesh generation. Before the analysis of the VVER 440-, BWR- and EPR-type canisters the calculation methodology was validated with test results, which were received from pressure tests performed with a short BWR canister in Germany

  19. Fabrication and testing of the sintered ceramic UO2 fuel - I - III, Part III - testing of sintered uranium dioxide properties dependent on the fabrication procedure

    International Nuclear Information System (INIS)

    Novakovic, M.; Ristic, M.M.

    1961-12-01

    The objective of this task was testing the influence of some parameters on the properties of sintered UO 2 . The influence of parameters tested were as follows: adhesives; pressure in the pressing procedure; temperature of sintering of the UO 2 powder. Other parameters were chosen according to the theoretical study. Sintering was done in argon atmosphere. Characterization of the UO 2 powder was performed meaning determining the needed chemical, physical and physico-chemical properties. Some new methods were developed within this task: SET method for measuring the specific surfaces, DTA, TGA, high-temperature torsion

  20. LIFE Materials: Phase Formation and Transformations in Transmutation Fuel Materials for the LIFE Engine Part I - Path Forward Volume 3

    Energy Technology Data Exchange (ETDEWEB)

    Turchi, P A; Kaufman, L; Fluss, M

    2008-12-19

    The current specifications of the LLNL fusion-fission hybrid proposal, namely LIFE, impose severe constraints on materials, and in particular on the nuclear fissile or fertile nuclear fuel and its immediate environment. This constitutes the focus of the present report with special emphasis on phase formation and phase transformations of the transmutation fuel and their consequences on particle and pebble thermal, chemical, and mechanical integrities. We first review the work that has been done in recent years to improve materials properties under the Gen-IV project, and with in particular applications to HTGR and MSR, and also under GNEP and AFCI in the USA. Our goal is to assess the nuclear fuel options that currently exist together with their issues. Among the options, it is worth mentioning TRISO, IMF, and molten salts. The later option will not be discussed in details since an entire report (Volume 8 - Molten-salt Fuels) is dedicated to it. Then, in a second part, with the specific LIFE specifications in mind, the various fuel options with their most critical issues are revisited with a path forward for each of them in terms of research, both experimental and theoretical. Since LIFE is applicable to very high burn-up of various fuels, distinctions will be made depending on the mission, i.e., energy production or incineration. Finally a few conclusions are drawn in terms of the specific needs for integrated materials modeling and the in depth knowledge on time-evolution thermo-chemistry that controls and drastically affects the performance of the nuclear materials and their immediate environment. Although LIFE demands materials that very likely have not yet been fully optimized, the challenges are not insurmountable, and a well concerted experimental-modeling effort should lead to dramatic advances that should well serve other fission programs such as Gen-IV, GNEP, AFCI as well as the international fusion program, ITER.

  1. On LMFBR corrosion. Part II: Consideration of the in-reactor fuel-cladding system

    International Nuclear Information System (INIS)

    Bradbury, M.H.; Pickering, S.; Walker, C.T.; Whitlow, W.H.

    1976-05-01

    The scientific and technological aspects of LMFBR cladding corrosion are discussed in detail. Emphasis is placed on the influence of the irradiation environment and the effect of fuel and filler-gas impurities on the corrosion process. These studies are complemented by a concise review of out-of-pile simulation experiments that endeavour to clarify the role of the aggressive fission products cesium, tellurium and iodine. The principal models for cladding corrosion are presented and critically assessed. Areas of uncertainty are exposed and some pertinent experiments are suggested. Consideration is also given to some new observations regarding the role of stress in fuel-cladding reactions and the formation of ferrite in the corrosion zone of the cladding during irradiation. Finally, two technological solutions to the problem of cladding corrosion are proposed. These are based on the use of an oxygen buffer in the fuel and the application of a protective coating to the inner surface of the cladding

  2. The back end of the nuclear fuel cycle: technical and economic analysis-Part 1

    International Nuclear Information System (INIS)

    Roglans-Ribas, J.; Spinrad, B.I.

    1990-01-01

    The back end of the nuclear fuel cycle has been analyzed under current conditions in the United States, taking into consideration the framework defined by the Nuclear Waste Policy Act of 1982 and its amendments. The different steps of the back end of the fuel cycle are studied and different alternatives are compared under technical and economic criteria. Several technical issues have been analyzed for their impact on the economics of the fuel cycle. The bases for the analysis are explained, and the results for a once-through cycle are presented. The results show that a repository in tuff represents the minimum cost situation. The economic model appears very sensitive to several parameters, in particular the period of retrievability and the storage costs

  3. Environmental systems analysis of biogas systems-Part I: Fuel-cycle emissions

    International Nuclear Information System (INIS)

    Boerjesson, Pal; Berglund, Maria

    2006-01-01

    Fuel-cycle emissions of carbon dioxide (CO 2 ), carbon oxide (CO), nitrogen oxides (NO x ), sulphur dioxide (SO 2 ), hydrocarbons (HC), methane (CH 4 ), and particles are analysed from a life-cycle perspective for different biogas systems based on six different raw materials. The gas is produced in large- or farm-scale biogas plants, and is used in boilers for heat production, in turbines for co-generation of heat and electricity, or as a transportation fuel in light- and heavy-duty vehicles. The analyses refer mainly to Swedish conditions. The levels of fuel-cycle emissions vary greatly among the biogas systems studied, and are significantly affected by the properties of the raw material digested, the energy efficiency of the biogas production, and the status of the end-use technology. For example, fuel-cycle emission may vary by a factor of 3-4, and for certain gases by up to a factor of 11, between two biogas systems that provide an equivalent energy service. Extensive handling of raw materials, e.g. ley cropping or collection of waste-products such as municipal organic waste, is often a significant source of emissions. Emission from the production phase of the biogas exceeds the end-use emissions for several biogas systems and for specific emissions. Uncontrolled losses of methane, e.g. leakages from stored digestates or from biogas upgrading, increase the fuel-cycle emissions of methane considerably. Thus, it is necessary to clearly specify the biogas production system and end-use technology being studied in order to be able to produce reliable and accurate data on fuel-cycle emission

  4. Evaluation of Future Fuels in a High Pressure Common Rail System - Part 1 Cummins XPI

    Science.gov (United States)

    2012-10-01

    with iso -octane. The larger components on the stand, such as heaters and heat exchangers, were drained and flushed with new test fuel. Typical...Removed with Debris Unclassified 35 The plunger was washed with iso -octane and the particles collected on filter paper. Examination of the...during this, or any other, time. 27000 27500 28000 28500 29000 29500 30000 0 5 10 15 20 25 30 35 40 Ra il  Pr es su re , p si Cycle Fuel Pressure in

  5. Combustion chemistry and flame structure of furan group biofuels using molecular-beam mass spectrometry and gas chromatography - Part III: 2,5-Dimethylfuran.

    Science.gov (United States)

    Togbé, Casimir; Tran, Luc-Sy; Liu, Dong; Felsmann, Daniel; Oßwald, Patrick; Glaude, Pierre-Alexandre; Sirjean, Baptiste; Fournet, René; Battin-Leclerc, Frédérique; Kohse-Höinghaus, Katharina

    2014-03-01

    This work is the third part of a study focusing on the combustion chemistry and flame structure of furan and selected alkylated derivatives, i.e. furan in Part I, 2-methylfuran (MF) in Part II, and 2,5-dimethylfuran (DMF) in the present work. Two premixed low-pressure (20 and 40 mbar) flat argon-diluted (50%) flames of DMF were studied with electron-ionization molecular-beam mass spectrometry (EI-MBMS) and gas chromatography (GC) under two equivalence ratios (φ=1.0 and 1.7). Mole fractions of reactants, products, and stable and radical intermediates were measured as a function of the distance to the burner. Kinetic modeling was performed using a reaction mechanism that was further developed in the present series, including Part I and Part II. A reasonable agreement between the present experimental results and the simulation is observed. The main reaction pathways of DMF consumption were derived from a reaction flow analysis. Also, a comparison of the key features for the three flames is presented, as well as a comparison between these flames of furanic compounds and those of other fuels. An a priori surprising ability of DMF to form soot precursors (e.g. 1,3-cyclopentadiene or benzene) compared to less substituted furans and to other fuels has been experimentally observed and is well explained in the model.

  6. Combustion chemistry and flame structure of furan group biofuels using molecular-beam mass spectrometry and gas chromatography – Part III: 2,5-Dimethylfuran

    Science.gov (United States)

    Togbé, Casimir; Tran, Luc-Sy; Liu, Dong; Felsmann, Daniel; Oßwald, Patrick; Glaude, Pierre-Alexandre; Sirjean, Baptiste; Fournet, René; Battin-Leclerc, Frédérique; Kohse-Höinghaus, Katharina

    2013-01-01

    This work is the third part of a study focusing on the combustion chemistry and flame structure of furan and selected alkylated derivatives, i.e. furan in Part I, 2-methylfuran (MF) in Part II, and 2,5-dimethylfuran (DMF) in the present work. Two premixed low-pressure (20 and 40 mbar) flat argon-diluted (50%) flames of DMF were studied with electron-ionization molecular-beam mass spectrometry (EI-MBMS) and gas chromatography (GC) under two equivalence ratios (φ=1.0 and 1.7). Mole fractions of reactants, products, and stable and radical intermediates were measured as a function of the distance to the burner. Kinetic modeling was performed using a reaction mechanism that was further developed in the present series, including Part I and Part II. A reasonable agreement between the present experimental results and the simulation is observed. The main reaction pathways of DMF consumption were derived from a reaction flow analysis. Also, a comparison of the key features for the three flames is presented, as well as a comparison between these flames of furanic compounds and those of other fuels. An a priori surprising ability of DMF to form soot precursors (e.g. 1,3-cyclopentadiene or benzene) compared to less substituted furans and to other fuels has been experimentally observed and is well explained in the model. PMID:24518851

  7. Burn-up Credit Criticality Safety Benchmark Phase III-C. Nuclide Composition and Neutron Multiplication Factor of a Boiling Water Reactor Spent Fuel Assembly for Burn-up Credit and Criticality Control of Damaged Nuclear Fuel

    International Nuclear Information System (INIS)

    Suyama, K.; Uchida, Y.; Kashima, T.; Ito, T.; Miyaji, T.

    2016-01-01

    Criticality control of damaged nuclear fuel is one of the key issues in the decommissioning operation of the Fukushima Daiichi Nuclear Power Station accident. The average isotopic composition of spent nuclear fuel as a function of burn-up is required in order to evaluate criticality parameters of the mixture of damaged nuclear fuel with other materials. The NEA Expert Group on Burn-up Credit Criticality (EGBUC) has organised several international benchmarks to assess the accuracy of burn-up calculation methodologies. For BWR fuel, the Phase III-B benchmark, published in 2002, was a remarkable landmark that provided general information on the burn-up properties of BWR spent fuel based on the 8x8 type fuel assembly. Since the publication of the Phase III-B benchmark, all major nuclear data libraries have been revised; in Japan from JENDL-3.2 to JENDL-4, in Europe from JEF-2.2 to JEFF-3.1 and in the US from ENDF/B-VI to ENDF/B-VII.1. Burn-up calculation methodologies have been improved by adopting continuous-energy Monte Carlo codes and modern neutronics calculation methods. Considering the importance of the criticality control of damaged fuel in the Fukushima Daiichi Nuclear Power Station accident, a new international burn-up calculation benchmark for the 9 x 9 STEP-3 BWR fuel assemblies was organised to carry out the inter-comparison of the averaged isotopic composition in the interest of the burnup credit criticality safety community. Benchmark specifications were proposed and approved at the EGBUC meeting in September 2012 and distributed in October 2012. The deadline for submitting results was set at the end of February 2013. The basic model for the benchmark problem is an infinite two-dimensional array of BWR fuel assemblies consisting of a 9 x 9 fuel rod array with a water channel in the centre. The initial uranium enrichment of fuel rods without gadolinium is 4.9, 4.4, 3.9, 3.4 and 2.1 wt% and 3.4 wt% for the rods using gadolinium. The burn-up conditions are

  8. The Analysis of the Available Technology of Exploiting and Applying Biohydrocarbons for Fuel Production Part I

    Directory of Open Access Journals (Sweden)

    Gielo-Klepacz Halina

    2017-08-01

    Full Text Available The article shows the current state of knowledge in the area of applying biohydrocarbons for fuel production, especially in aeronautical applications and to power compression-ignition engines. The technologies based on biochemical and thermal/chemical conversion of biomass are described. Technological potential of these technologies is evaluated. The article is based on the literature review.

  9. Nuclear fuel reprocessing and high level waste disposal: informational hearings. Volume V. Reprocessing. Part 2

    Energy Technology Data Exchange (ETDEWEB)

    None

    1977-03-08

    Testimony was presented by a four member panel on the commercial future of reprocessing. Testimony was given on the status of nuclear fuel reprocessing in the United States. The supplemental testimony and materials submitted for the record are included in this report. (LK)

  10. Development of metal uranium fuel and testing of construction materials (I-VI); Part I

    International Nuclear Information System (INIS)

    Mihajlovic, A.

    1965-11-01

    This project includes the following tasks: Study of crystallisation of metal melt and beta-alpha transforms in uranium and uranium alloys; Study of the thermal treatment influence on phase transformations and texture in uranium alloys; Radiation damage of metal uranium; Project related to irradiation of metal uranium in the reactor; Development of fuel element for nuclear reactors

  11. Evaluation of methods for seismic analysis of nuclear fuel reprocessing plants, part 1

    International Nuclear Information System (INIS)

    Tokarz, F.J.; Murray, R.C.; Arthur, D.F.; Feng, W.W.; Wight, L.H.; Zaslawsky, M.

    1975-01-01

    Currently, no guidelines exist for choosing methods of structural analysis to evaluate the seismic hazard of nuclear fuel reprocessing plants. This study examines available methods and their applicability to fuel reprocessing plant structures. The results of this study should provide a basis for establishing guidelines recommending methods of seismic analysis for evaluating future fuel reprocessing plants. The approach taken is: (1) to identify critical plant structures and place them in four categories (structures at or near grade; deeply embedded structures; fully buried structures; equipment/vessels/attachments/piping), (2) to select a representative structure in each of the first three categories and perform static and dynamic analysis on each, and (3) to evaluate and recommend method(s) of analysis for structures within each category. The Barnwell Nuclear Fuel Plant is selected as representative of future commercial reprocessing plants. The effect of site characteristics on the structural response is also examined. The response spectra method of analysis combined with the finite element model for each category is recommended. For structures founded near or at grade, the lumped mass model could also be used. If a time history response is required, a time-history analysis is necessary. (U.S.)

  12. Stationary liquid fuel fast reactor SLFFR — Part II: Safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Jing, T.; Jung, Y.S.; Yang, W.S., E-mail: yang494@purdue.edu

    2016-12-15

    Highlights: • A multi-channel safety analysis code named MUSA is developed for SLFFR transient analyses. • MUSA is verified against the SYS4A/SASSYS-1 code by simulating the ULOF accident for the advanced burner test reactor. • It is shown that SLFFR has a passive shutdown capability for double-fault, beyond-design-basis accidents UTOP, ULOHS and ULOF. - Abstract: Safety characteristics have been evaluated for the stationary liquid fuel fast reactor (SLFFR) proposed for effective burning of hazardous TRU elements of used nuclear fuel. In order to model the geometrical configuration and reactivity feedback mechanisms unique to SLFFR, a multi-channel safety analysis code named MUSA was developed. MUSA solves the time-dependent coupled neutronics and thermal-fluidic problems. The thermal-fluidic behavior of the core is described by representing the core with one-dimensional parallel channels. The primary heat transport system is modeled by connecting compressible volumes by liquid segments. A point kinetics model with six delayed neutron groups is used to represent the fission power transients. The reactivity feedback is estimated by combining the temperature and density variations of liquid fuel, structural material and sodium coolant with the corresponding axial distributions of reactivity worth in each individual thermal-fluidic channel. Preliminary verification tests with a conventional solid fuel reactor agreed well with the reference solutions obtained with the SAS4A/SASSYS-1 code. Transient analyses of SLFFR were performed for unprotected transient over-power (UTOP), unprotected loss of heat sink (ULOHS) and unprotected loss of flow (ULOF) accidents. The results showed that the thermal expansion of liquid fuel provides sufficiently large negative feedback reactivity for passive shutdown of UTOP and ULOHS. The ULOF transient is also terminated passively with the negative reactivity introduced by the gas expansion modules installed at the core periphery

  13. The Bacterial Flagellar Type III Export Gate Complex Is a Dual Fuel Engine That Can Use Both H+ and Na+ for Flagellar Protein Export.

    Directory of Open Access Journals (Sweden)

    Tohru Minamino

    2016-03-01

    Full Text Available The bacterial flagellar type III export apparatus utilizes ATP and proton motive force (PMF to transport flagellar proteins to the distal end of the growing flagellar structure for self-assembly. The transmembrane export gate complex is a H+-protein antiporter, of which activity is greatly augmented by an associated cytoplasmic ATPase complex. Here, we report that the export gate complex can use sodium motive force (SMF in addition to PMF across the cytoplasmic membrane to drive protein export. Protein export was considerably reduced in the absence of the ATPase complex and a pH gradient across the membrane, but Na+ increased it dramatically. Phenamil, a blocker of Na+ translocation, inhibited protein export. Overexpression of FlhA increased the intracellular Na+ concentration in the presence of 100 mM NaCl but not in its absence, suggesting that FlhA acts as a Na+ channel. In wild-type cells, however, neither Na+ nor phenamil affected protein export, indicating that the Na+ channel activity of FlhA is suppressed by the ATPase complex. We propose that the export gate by itself is a dual fuel engine that uses both PMF and SMF for protein export and that the ATPase complex switches this dual fuel engine into a PMF-driven export machinery to become much more robust against environmental changes in external pH and Na+ concentration.

  14. The Bacterial Flagellar Type III Export Gate Complex Is a Dual Fuel Engine That Can Use Both H+ and Na+ for Flagellar Protein Export.

    Science.gov (United States)

    Minamino, Tohru; Morimoto, Yusuke V; Hara, Noritaka; Aldridge, Phillip D; Namba, Keiichi

    2016-03-01

    The bacterial flagellar type III export apparatus utilizes ATP and proton motive force (PMF) to transport flagellar proteins to the distal end of the growing flagellar structure for self-assembly. The transmembrane export gate complex is a H+-protein antiporter, of which activity is greatly augmented by an associated cytoplasmic ATPase complex. Here, we report that the export gate complex can use sodium motive force (SMF) in addition to PMF across the cytoplasmic membrane to drive protein export. Protein export was considerably reduced in the absence of the ATPase complex and a pH gradient across the membrane, but Na+ increased it dramatically. Phenamil, a blocker of Na+ translocation, inhibited protein export. Overexpression of FlhA increased the intracellular Na+ concentration in the presence of 100 mM NaCl but not in its absence, suggesting that FlhA acts as a Na+ channel. In wild-type cells, however, neither Na+ nor phenamil affected protein export, indicating that the Na+ channel activity of FlhA is suppressed by the ATPase complex. We propose that the export gate by itself is a dual fuel engine that uses both PMF and SMF for protein export and that the ATPase complex switches this dual fuel engine into a PMF-driven export machinery to become much more robust against environmental changes in external pH and Na+ concentration.

  15. Present status of reactor physics in the United States and Japan-III. 2. Nuclear Fuel Management Optimization Capabilities

    International Nuclear Information System (INIS)

    Karve, Atul A.; Keller, Paul M.; Turinsky, Paul J.; Maldonado, G. Ivan

    2001-01-01

    Nuclear fuel management is a very difficult design optimization problem in that decisions ranging from the microscopic level, e.g., pin enrichment, to the macroscopic level, e.g., core flow rate, and spanning time horizons of several reload cycles are strongly coupled. Added to these attributes are the highly constrained design, disjointed decision space, multimodal objective function, mixed integer type decision variables, highly nonlinear objective and constraint functions, and computationally demanding evaluation of the objective and constraint functions. Not surprisingly, after years of research on nuclear fuel management optimization, only limited progress has been made. The traditional approach to partially overcome these difficulties involves constraining the search space via heuristic rules, decomposing the problem into sub-optimization problems, and utilizing simplified core physics models. These approaches have sometimes proven effective, but to claim that the design decisions are global optimum decisions would not be appropriate. Given the increasingly tight constraints and design complexities of nuclear cores, and stronger desire to reduce generating costs, the nuclear fuel management design optimization problem has grown more challenging and important with the passage of time. In this paper, we summarize our research on this design optimization problem. A suite of computer codes that aid in making nuclear fuel management decisions has been developed. From Table I, it is obvious that decomposition of the global optimization problem into suboptimum problems has been employed. All of these computer codes utilize stochastic optimization techniques to search the decision space for determining the family of near-optimum decisions in the sub-optimization problem being solved. A stochastic optimization approach has been selected since it is well suited to address the problems' attributes noted earlier. The drawback of employing a stochastic optimization

  16. Part 1: Logging residues in piles - Needle loss and fuel quality. Part 2: Nitrogen leaching under piles of logging residues

    International Nuclear Information System (INIS)

    Lehtikangas, P.; Lundkvist, H.

    1991-01-01

    Part 1: Experimental piles were built in three geographical locations during May-Sept. 1989. Logging residues consisted of 95% spruce and 5% pine. Height of the piles varied between 80 and 230 cm. Needles were collected by placing drawers under 40 randomely chosen piles. The drawers were emptied every two weeks during the storage period. Natural needle loss was between 18 and 32% of the total amount of needles after the first two months of storage. At the end of the storage period, 24-42% of the needles had fallen down to the drawers. At the end of the experiment the total needle fall was 95-100% in the shaken piles. According to the results of this study piles smaller than 150 cm had the most effective needle fall. Piles should be placed on open places where the air and sun heat penetrate and dry them. Needles were the most sensitive fraction to variations in precipitation compared to the other components, such as branches. Piles usually dried quickly, but they also rewet easily. This was especially true in the smaller piles. The lowest moisture content was measured at the end of June. The ash content in needles varied between 4 and 8%. 16 refs., 15 figs. Part 2: Three field experiments were equipped with no-tension humus lysimeters. Pairs of lysimeters with the same humus/field layer vegetation material were placed in pairs, one under a pile of felling residues and another in the open clear felling. Leaching of nitrogen as well as pH and electric conductivity in the leachate was followed through sampling of the leachate at regular intervals. The results from the investigation show that: * the amount of leachate was higher in lysimeters in the open clear felling, * pH in the leachate was initially lower under piles of felling residues, * the amount of nitrogen leached was higher in the open clear felling. Thus, storing of felling residues in piles during the summer season did not cause any increase in nitrogen leaching, which had been considered to be a risk

  17. Review of the micro-tubular solid oxide fuel cell. Part I. Stack design issues and research activities

    Energy Technology Data Exchange (ETDEWEB)

    Lawlor, V. [Department of Eco-Energy Engineering, Upper Austrian University of Applied Sciences, A-4600 Wels (Austria); Department of Manufacturing and Mechanical Engineering, Dublin City University, Dublin 9 (Ireland); Griesser, S. [Department of Eco-Energy Engineering, Upper Austrian University of Applied Sciences, A-4600 Wels (Austria); Buchinger, G. [eZelleron GmbH, Collenbusch str. 22, 01324 Dresden (Germany); Olabi, A.G. [Department of Manufacturing and Mechanical Engineering, Dublin City University, Dublin 9 (Ireland); Cordiner, S. [Dipartimento di Ingegneria Meccanica - Universita di Roma Tor Vergata (Italy); Meissner, D. [Department of Eco-Energy Engineering, Upper Austrian University of Applied Sciences, A-4600 Wels (Austria); Department of Material Science, Tallinn University of Technology, Ehitajate 19086 (Estonia)

    2009-09-05

    Fuel cells are devices that convert chemical energy in hydrogen enriched fuels into electricity electrochemically. Micro-tubular solid oxide fuel cells (MT-SOFCs), the type pioneered by K. Kendall in the early 1990s, are a variety of SOFCs that are on the scale of millimetres compared to their much larger SOFC relatives that are typically on the scale of tens of centimetres. The main advantage of the MT-SOFC, over its larger predecessor, is that it is smaller in size and is more suitable for rapid start up. This may allow the SOFC to be used in devices such as auxiliary power units, automotive power supplies, mobile electricity generators and battery re-chargers. The following paper is Part I of a two part series. Part I will introduce the reader to the MT-SOFC stack and its applications, indicating who is researching what in this field and also specifically investigate the design issues related to multi-cell reactor systems called stacks. Part II will review in detail the combinations of materials and methods used to produce the electrodes and electrolytes of MT-SOFC's. Also the role of modelling and validation techniques used in the design and improvement of the electrodes and electrolytes will be investigated. A broad range of scientific and engineering disciplines are involved in a stack design. Scientific and engineering content has been discussed in the areas of thermal-self-sustainability and efficiency, sealing technologies, manifold design, electrical connections and cell performance optimisation. (author)

  18. Application of contact mechanics for fretting damage of fuel rod: part 1 influence functions and numerical method

    International Nuclear Information System (INIS)

    Kim, H. K.; Yoon, K. H.; Kang, H. S.; Song, G. N.

    1998-01-01

    For the analysis of the fretting problem of the fuel rods, present paper(Part I) shows the numerical method developed for evaluating the stresses on the contact surfaces between the fuel rods and the spacer grids. Theory of Contact Mechanics was incorporated. Contact area was regarded as a plane strain condition, so plane problem was taken into consideration. Normal stress profile on the contact surface was assumed to be Hertzian. As for the direction of the shear load, a closed load path, e.g. load increase in transverse increase in axial decrease in transverse decrease in axial increase in transverse increase in axial direction was considered for simulating the rod vibration in a reactor core. Partial slip problem was consulted. As for the numerical method, a triangular traction element was utilized and the corresponding influence functions were evaluated. Numerical program has been implemented for the present analysis, of which the validity was verified by comparing the Mindlin-Cattaneo solution

  19. Applying hot wire anemometry to directly measure the water balance in a proton exchange membrane fuel cell - Part 1

    DEFF Research Database (Denmark)

    Berning, Torsten; Al Shakhshir, Saher

    2015-01-01

    In order to accurately determine the water balance of a proton exchange membrane fuel cell it has recently been suggested to employ constant temperature anemometry (CTA), a frequently used method to measure the velocity of a fluid stream. CTA relies on convective heat transfer around a heated wire...... the equations required to calculate the heat transfer coefficient and the resulting voltage signal as function of the fuel cell water balance. The most critical and least understood part is the determination of the Nusselt number to calculate the heat transfer between the wire and the gas stream. Different...... expressions taken from the literature will be examined in detail, and it will be demonstrated that the power-law approach suggested by Hilpert is the only useful one for the current purposes because in this case the voltage response from the hot-wire sensor E/E0 shows the same dependency to the water balance...

  20. Recovery of actinides from actinide-aluminium alloys by chlorination: Part III - Chlorination with HCl(g)

    Science.gov (United States)

    Meier, Roland; Souček, Pavel; Walter, Olaf; Malmbeck, Rikard; Rodrigues, Alcide; Glatz, Jean-Paul; Fanghänel, Thomas

    2018-01-01

    Two steps of a pyrochemical route for the recovery of actinides from spent metallic nuclear fuel are being investigated at JRC-Karlsruhe. The first step consists in electrorefining the fuel in molten salt medium implying aluminium cathodes. The second step is a chlorination process for the separation of actinides (An) from An-Al alloys formed on the cathodes. The chlorination process, in turn, consists of three steps; the distillation of adhered salt (1), the chlorination of An-Al by HCl/Cl2 under formation of AlCl3 and An chlorides (2), and the subsequent sublimation of AlCl3 (3). In the present work UAl2, UAl3, NpAl2, and PuAl2 were chlorinated with HCl(g) in a temperature range between 300 and 400 °C forming UCl4, NpCl4 or PuCl3 as the major An containing phases, respectively. Thermodynamic calculations were carried out to support the experimental work. The results showed a high chlorination efficiency for all used starting materials and indicated that the sublimation step may not be necessary when using HCl(g).

  1. Development of Induction Brazing System for Sealing Instrumentation Feed through Part of Nuclear Fuel Test Rig

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Jintae; Kim, Kahye; Heo, Sungho; Ahn, Sungho; Joung, Changyoung; Son, Kwangjae; Jung, Yangil [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-12-15

    To test the performance of nuclear fuels, coolant needs to be circulated through the test rig installed in the test loop. Because the pressure and temperature of the coolant is 15.5 MPa and 300 .deg. C respectively, coolant sealing is one of the most important processes in fabricating a nuclear fuel test rig. In particular, 15 instrumentation cables installed in a test rig pass through the pressure boundary, and brazing is generally applied as a sealing method. In this study, an induction brazing system has been developed using a high frequency induction heater including a vacuum chamber. For application in the nuclear field, BNi2 should be used as a paste, and optimal process variables for Ni brazing have been found by several case studies. The performance and soundness of the brazed components has been verified by a tensile test, cross section test, and sealing performance test.

  2. Current activities on improving storage conditions of the research reactor RA spent fuel - Part II

    International Nuclear Information System (INIS)

    Matausek, M.V.; Kopecni, M.; Vukadin, Z.; Plecas, I.; Pavlovic, R.; Sotic, O.; Marinkovic, N.

    1998-01-01

    To minimize further corrosion and preserve integrity of aluminum barrels and the stainless steel channel-type containers that were found to contain leaking spent fuel, actions to improve conditions in the existing spent fuel storage pool at the RA research reactor were initiated. Technology was elaborated and equipment was produced and applied for removal of sludge and other debris from the bottom of the pool, filtration of the pool water, sludge conditioning in cement matrix and disposal at the low and medium waste repository at VINCA site. More sophisticated operations are to be performed together with foreign experts. Safety measures and precautions were determined. Subcriticality was proved under normal and/or possible abnormal conditions. (author)

  3. Uranium accountability for ATR fuel fabrication. Part I. A description of the existing system

    International Nuclear Information System (INIS)

    Dolan, C.A.; Nieschmidt, E.B.; Vegors, S.H. Jr.; Wagner, E.P. Jr.

    1977-06-01

    An evaluation of the materials accountability program at the Atomics International fuel fabrication facility in Canoga Park, California, with regard to the fabrication of highly enriched uranium fuel for the Advanced Test Reactor is presented. An analysis is given of the existing standards program, the existing measurements program and the existing statistical analysis procedures. In addition a short discussion is given of our evaluation of the safeguards procedures at Atomics International together with suggestions for possible modifications and improvements. Appendices of this report contain a rather complete description of the Atomics International plant and the flow of highly enriched uranium through the plant as well as the principal documents used for material accountability records

  4. The back end of the nuclear fuel cycle: Technical and economic analysis-Part 2

    International Nuclear Information System (INIS)

    Roglans-Ribas, J.; Spinrad, B.I.

    1990-01-01

    The back end of the nuclear fuel cycle has been analyzed under current conditions in the United States, including the constraints imposed by the Nuclear Waste Policy Act of 1982 and its amendments. The scenarios for two closed cycles, a regular reprocessing cycle and a reprocessing scheme with cesium and strontium fractionation, are described. The storage of spent fuel discharged from the reactors and the disposal of reprocessed waste are studied for both reprocessing cycles. The economics of waste storage and disposal for the two closed cycles are compared with each other and with the reference once-through cycle. The results show that a standard reprocessing cycle results in the minimum cost for storage and disposal. When reporcessing costs are considered, the closed cycles can compete economically with the once-through cycle only if net reprocessing costs are very low

  5. Probabilistic risk assessment on maritime spent nuclear fuel transportation (Part II: Ship collision probability)

    International Nuclear Information System (INIS)

    Christian, Robby; Kang, Hyun Gook

    2017-01-01

    This paper proposes a methodology to assess and reduce risks of maritime spent nuclear fuel transportation with a probabilistic approach. Event trees detailing the progression of collisions leading to transport casks’ damage were constructed. Parallel and crossing collision probabilities were formulated based on the Poisson distribution. Automatic Identification System (AIS) data were processed with the Hough Transform algorithm to estimate possible intersections between the shipment route and the marine traffic. Monte Carlo simulations were done to compute collision probabilities and impact energies at each intersection. Possible safety improvement measures through a proper selection of operational transport parameters were investigated. These parameters include shipment routes, ship's cruise velocity, number of transport casks carried in a shipment, the casks’ stowage configuration and loading order on board the ship. A shipment case study is presented. Waters with high collision probabilities were identified. Effective range of cruising velocity to reduce collision risks were discovered. The number of casks in a shipment and their stowage method which gave low cask damage frequencies were obtained. The proposed methodology was successful in quantifying ship collision and cask damage frequency. It was effective in assisting decision making processes to minimize risks in maritime spent nuclear fuel transportation. - Highlights: • Proposes a probabilistic framework on the safety of spent nuclear fuel transportation by sea. • Developed a marine traffic simulation model using Generalized Hough Transform (GHT) algorithm. • A transportation case study on South Korean waters is presented. • Single-vessel risk reduction method is outlined by optimizing transport parameters.

  6. Biotechnology and genetic engineering in the new drug development. Part III. Biocatalysis, metabolic engineering and molecular modelling.

    Science.gov (United States)

    Stryjewska, Agnieszka; Kiepura, Katarzyna; Librowski, Tadeusz; Lochyński, Stanisław

    2013-01-01

    Industrial biotechnology has been defined as the use and application of biotechnology for the sustainable processing and production of chemicals, materials and fuels. It makes use of biocatalysts such as microbial communities, whole-cell microorganisms or purified enzymes. In the review these processes are described. Drug design is an iterative process which begins when a chemist identifies a compound that displays an interesting biological profile and ends when both the activity profile and the chemical synthesis of the new chemical entity are optimized. Traditional approaches to drug discovery rely on a stepwise synthesis and screening program for large numbers of compounds to optimize activity profiles. Over the past ten to twenty years, scientists have used computer models of new chemical entities to help define activity profiles, geometries and relativities. This article introduces inter alia the concepts of molecular modelling and contains references for further reading.

  7. SCANAIR a transient fuel performance code Part two: Assessment of modelling capabilities

    Energy Technology Data Exchange (ETDEWEB)

    Georgenthum, Vincent, E-mail: vincent.georgenthum@irsn.fr; Moal, Alain; Marchand, Olivier

    2014-12-15

    Highlights: • The SCANAIR code is devoted to the study of irradiated fuel rod behaviour during RIA. • The paper deals with the status of the code validation for PWR rods. • During the PCMI stage there is a good agreement between calculations and experiments. • The boiling crisis occurrence is rather well predicted. • The code assessment during the boiling crisis has still to be improved. - Abstract: In the frame of their research programmes on fuel safety, the French Institut de Radioprotection et de Sûreté Nucléaire develops the SCANAIR code devoted to the study of irradiated fuel rod behaviour during reactivity initiated accident. A first paper was focused on detailed modellings and code description. This second paper deals with the status of the code validation for pressurised water reactor rods performed thanks to the available experimental results. About 60 integral tests carried out in CABRI and NSRR experimental reactors and 24 separated tests performed in the PATRICIA facility (devoted to the thermal-hydraulics study) have been recalculated and compared to experimental data. During the first stage of the transient, the pellet clad mechanical interaction phase, there is a good agreement between calculations and experiments: the clad residual elongation and hoop strain of non failed tests but also the failure occurrence and failure enthalpy of failed tests are correctly calculated. After this first stage, the increase of cladding temperature can lead to the Departure from Nucleate Boiling. During the film boiling regime, the clad temperature can reach a very high temperature (>700 °C). If the boiling crisis occurrence is rather well predicted, the calculation of the clad temperature and the clad hoop strain during this stage have still to be improved.

  8. Inventory of aerosol and sulphur dioxide emissions from India. Part 1 - Fossil fuel combustion

    International Nuclear Information System (INIS)

    Shekar Reddy, M.; Venkataraman, C.

    2002-01-01

    A comprehensive, spatially resolved (0.25 o x 0.25 o ) fossil fuel consumption database and emissions inventory was constructed, for India, for the first time. Emissions of sulphur dioxide and aerosol chemical constituents were estimated for 1996-1997 and extrapolated to the Indian Ocean Experiment (INDOEX) study period (1998-1999). District level consumption of coal/lignite, petroleum and natural gas in power plants, industrial, transportation and domestic sectors was 9411 PJ, with major contributions from coal (54%) followed by diesel (18%). Emission factors for various pollutants were derived using India specific fuel characteristics and information on combustion/air pollution control technologies for the power and industrial sectors. Domestic and transportation emission factors, appropriate for Indian source characteristics, were compiled from literature. SO 2 emissions from fossil fuel combustion for 1996-1997 were 4.0Tg SO 2 yr -1 , with 756 large point sources (e.g. utilities, iron and steel, fertilisers, cement, refineries and petrochemicals and non-ferrous metals), accounting for 62%. PM 2.5 emitted was 0.5 and 2.0Tgyr -1 for the 100% and the 50% control scenario, respectively, applied to coal burning in the power and industrial sectors. Coal combustion was the major source of PM 2.5 (92%) primarily consisting of fly ash, accounting for 98% of the 'inorganic fraction' emissions (difference between PM 2.5 and black carbon + organic matter) of 1.6Tgyr -1 . Black carbon emissions were estimated at 0.1Tgyr -1 , with 58% from diesel transport, and organic matter emissions at 0.3Tgyr -1 , with 48% from brick-kilns. Fossil fuel consumption and emissions peaked at the large point industrial sources and 22 cities, with elevated area fluxes in northern and western India. The spatial resolution of this inventory makes it suitable for regional-scale aerosol-climate studies. These results are compared to previous studies and differences discussed. Measurements of

  9. Radioactive waste and the back part of fuel cycle of nuclear installations in Slovakia

    International Nuclear Information System (INIS)

    Koprda, V.

    2004-01-01

    This article is devoted to radioactive waste (RAW) management, an integrated system starting with collection and sorting of RAW through its storage, treatment, conditioning, handling and transport up to its disposal. Some notes will touch also the near surface depository of low level and intermediate level radioactive waste in Mochovce, and the long-term storage of waste improper for such type of disposal, and also some words will be addressed to the development and research of a deep geological depository for disposal spent fuel from nuclear power plant and long-lived radioactive waste. (author)

  10. Atomics International fuel fabrication facility and low enrichment program. Part 2

    International Nuclear Information System (INIS)

    Hassel, H.W.

    1993-01-01

    Most of you know our company from the last meeting in May in Vienna, so I won't steal your time with explaining and demonstrating the same techniques that we have heard this morning f rom the other speakers. I would just take some words to explain the order of business with highly enriched uranium. NUKEM handles around almost two tons of highly enriched uranium a year and it was necessary to satisfy all the new physical protection philosophies. That means that we have to install storage and safe fabrication sites for a lot of money, 2.5 meter thick concrete walls, and different alarm systems. So just to demonstrate how silly this business is, we have just overcome this for highly enriched uranium, and now we speak about low enriched uranium for which we don't need all of these investments to make this business safe. I would just like to concentrate my words on the status of fabrication and considerations in my company concerning the medium enriched uranium and low enriched uranium. In TABLE I are the different fuel types (see column 1) and then we have the fabrication in column 2; (The reason that I use the blackboard this morning is that I try to demonstrate all the techniques. However, all the speakers before me did this and in theory we are not so far away from each other.) the experience of my company in kg. In column 3 is the irradiation experience of these fuels types. Column 4 shows the studies and calculations made in our company for lower and medium enriched fuels. The preliminary fabrication tests and calculations are in column 5, and in column 6 we have the delivery time for a prototype core in months after UF 6 supply. Column 7 shows the time for the development of specifications including irradiation time in years for 6 and 7, and column 8 is the estimated cost of 6 and 7. There is just one fuel that is not in this summary and that is U-Zr

  11. High-Uranium-Loaded U3O8-Al fuel element development program. Part 2

    International Nuclear Information System (INIS)

    Knight, R.

    1993-01-01

    Texas Instruments is a product intensive company that manufactures very high volumes of different products, and because of this, their technique in manufacturing is what we call hard tooling. So all of the tools we use at this site whether it is for HFIR, ORR or HFBR are hard tooling. A fuel plate never sees a lathe, milling machine, or any other tool of that nature. I have just a few viewgraphs here that will illustrate some of the types of tooling we use to keep away from machining and get high production at as low as possible cost. Figure I shows weighing aluminum powder. It's done in a glove box more to keep air flow away from the balance than any other reason. The weighing of the U 3 O 8 is similar and the glove box is for personnel protection. Figure 2 shows our blender, and I won't try to explain why it works. This is the only one we have ever found that really blends our powder and does a good job. Figure 3 shows our powder die on the press, and you can see the rectangular compact being extracted. Here is the way we make our frames in a blanking die Figure 4. You will notice there are two holes in the frame. We start off with two cores in a frame. Our lot size is 24, but twelve billets go into the furnace for preheating, at the seventh pass, we cut the two cores apart and at that point they become individual fuel plates. Figure 5 shows the loading of the compacts into the frame. We use a loose fit. We can just drop the cores into the frame with, I think, about 2 mils side clearance and it works very satisfactorily. Figure 6 shows a forming die. Once you make the investment for the fuel plate blanking die shown in Figure 7, you can blank out a fuel plate on the order of about one per minute, to size and to the tolerances required. Figure 8 shows a unique tool developed at Oak Ridge. It's a Homogeneity Scanner. It works on the principal of x-ray attenuation going through an electronic analysis

  12. A tri-generation system based on polymer electrolyte fuel cell and desiccant wheel – Part A: Fuel cell system modelling and partial load analysis

    International Nuclear Information System (INIS)

    Najafi, Behzad; De Antonellis, Stefano; Intini, Manuel; Zago, Matteo; Rinaldi, Fabio; Casalegno, Andrea

    2015-01-01

    Highlights: • A mathematical model for a PEMFC based cogeneration system is developed. • Developed model is validated using the available experimental data. • Performance of the plant at full load conditions is investigated. • Performance indices while applying two different modifications are determined. • System’s performance with and without modifications at partial loads is investigated. - Abstract: Polymer Electrolyte Membrane Fuel Cell (PEMFC) based systems have recently received increasing attention as a viable alternative for meeting the residential electrical and thermal demands. However, as the intermittent demand profiles of a building can only be addressed by a tri-generative unit which can operate at partial loads, the variation of performance of the system at partial loads might affect its corresponding potential benefits significantly. Nonetheless, no previous study has been carried out on assessing the performance of this type of tri-generative systems in such conditions. The present paper is the first of a two part study dedicated to the investigation of the performance of a tri-generative system in which a PEMFC based system is coupled with a desiccant wheel unit. This study is focused on evaluating the performance of the PEMFC subsystem while operating at partial loads. Accordingly, a detailed mathematical model of the fuel cell subsystem is first developed and validated using the experimental data obtained from the plant’s and the fuel cell stack’s manufacturer. Next, in order to increase the performance of the plant, two modifications have been proposed and the resulting performance at partial load have been determined. The obtained results demonstrate that applying both modifications results in increasing the electrical efficiency of the plant by 5.5%. It is also shown that, while operating at partial loads, the electrical efficiency of the plant does not significantly change; the fact which corresponds to the trade-off between

  13. RA reactor safety analysis, Part II - Accident analysis; Analiza sigurnosti rada Reaktora RA I-III, Deo II - Analiza akcidenta

    Energy Technology Data Exchange (ETDEWEB)

    Raisic, N; Radanovic, Lj; Milovanovic, M; Afgan, N; Kulundzic, P [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1963-02-15

    This part of the RA reactor safety analysis includes analysis of possible accidents caused by failures of the reactor devices and errors during reactor operation. Two types of accidents are analyzed: accidents resulting from uncontrolled reactivity increase, and accidents caused by interruption of cooling.

  14. [Blood-brain barrier part III: therapeutic approaches to cross the blood-brain barrier and target the brain].

    Science.gov (United States)

    Weiss, N; Miller, F; Cazaubon, S; Couraud, P-O

    2010-03-01

    Over the last few years, the blood-brain barrier has come to be considered as the main limitation for the treatment of neurological diseases caused by inflammatory, tumor or neurodegenerative disorders. In the blood-brain barrier, the close intercellular contact between cerebral endothelial cells due to tight junctions prevents the passive diffusion of hydrophilic components from the bloodstream into the brain. Several specific transport systems (via transporters expressed on cerebral endothelial cells) are implicated in the delivery of nutriments, ions and vitamins to the brain; other transporters expressed on cerebral endothelial cells extrude endogenous substances or xenobiotics, which have crossed the cerebral endothelium, out of the brain and into the bloodstream. Recently, several strategies have been proposed to target the brain, (i) by by-passing the blood-brain barrier by central drug administration, (ii) by increasing permeability of the blood-brain barrier, (iii) by modulating the expression and/or the activity of efflux transporters, (iv) by using the physiological receptor-dependent blood-brain barrier transport, and (v) by creating new viral or chemical vectors to cross the blood-brain barrier. This review focuses on the illustration of these different approaches. Copyright (c) 2009 Elsevier Masson SAS. All rights reserved.

  15. Impact of Fe(III) as an effective electron-shuttle mediator for enhanced Cr(VI) reduction in microbial fuel cells: Reduction of diffusional resistances and cathode overpotentials

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Qiang [Key Laboratory of Industrial Ecology and Environmental Engineering, Ministry of Education (MOE), School of Environmental Science and Technology, Dalian University of Technology, Dalian 116024 (China); Huang, Liping, E-mail: lipinghuang@dlut.edu.cn [Key Laboratory of Industrial Ecology and Environmental Engineering, Ministry of Education (MOE), School of Environmental Science and Technology, Dalian University of Technology, Dalian 116024 (China); Pan, Yuzhen [College of Chemistry, Dalian University of Technology, Dalian 116024 (China); Quan, Xie [Key Laboratory of Industrial Ecology and Environmental Engineering, Ministry of Education (MOE), School of Environmental Science and Technology, Dalian University of Technology, Dalian 116024 (China); Li Puma, Gianluca, E-mail: g.lipuma@lboro.ac.uk [Environmental Nanocatalysis & Photoreaction Engineering, Department of Chemical Engineering, Loughborough University, Loughborough LE11 3TU (United Kingdom)

    2017-01-05

    Highlights: • Fe(III) shuttles electrons for enhanced reduction of Cr(VI) in MFCs. • The coulombic efficiency increases by 1.6 fold in the presence of Fe(III). • The reduction of Cr(VI) occurs via an indirect Fe(III) mediation mechanism. • Fe(III) decreases the diffusional resistances and the cathode overpotentials. - Abstract: The role of Fe(III) was investigated as an electron-shuttle mediator to enhance the reduction rate of the toxic heavy metal hexavalent chromium (Cr(VI)) in wastewaters, using microbial fuel cells (MFCs). The direct reduction of chromate (CrO{sub 4}{sup −}) and dichromate (Cr{sub 2}O{sub 7}{sup 2−}) anions in MFCs was hampered by the electrical repulsion between the negatively charged cathode and Cr(VI) functional groups. In contrast, in the presence of Fe(III), the conversion of Cr(VI) and the cathodic coulombic efficiency in the MFCs were 65.6% and 81.7%, respectively, 1.6 times and 1.4 folds as those recorded in the absence of Fe(III). Multiple analytical approaches, including linear sweep voltammetry, Tafel plot, cyclic voltammetry, electrochemical impedance spectroscopy and kinetic calculations demonstrated that the complete reduction of Cr(VI) occurred through an indirect mechanism mediated by Fe(III). The direct reduction of Cr(VI) with cathode electrons in the presence of Fe(III) was insignificant. Fe(III) played a critical role in decreasing both the diffusional resistance of Cr(VI) species and the overpotential for Cr(VI) reduction. This study demonstrated that the reduction of Cr(VI) in MFCs was effective in the presence of Fe(III), providing an alternative and environmentally benign approach for efficient remediation of Cr(VI) contaminated sites with simultaneous production of renewable energy.

  16. Impact of Fe(III) as an effective electron-shuttle mediator for enhanced Cr(VI) reduction in microbial fuel cells: Reduction of diffusional resistances and cathode overpotentials

    International Nuclear Information System (INIS)

    Wang, Qiang; Huang, Liping; Pan, Yuzhen; Quan, Xie; Li Puma, Gianluca

    2017-01-01

    Highlights: • Fe(III) shuttles electrons for enhanced reduction of Cr(VI) in MFCs. • The coulombic efficiency increases by 1.6 fold in the presence of Fe(III). • The reduction of Cr(VI) occurs via an indirect Fe(III) mediation mechanism. • Fe(III) decreases the diffusional resistances and the cathode overpotentials. - Abstract: The role of Fe(III) was investigated as an electron-shuttle mediator to enhance the reduction rate of the toxic heavy metal hexavalent chromium (Cr(VI)) in wastewaters, using microbial fuel cells (MFCs). The direct reduction of chromate (CrO_4"−) and dichromate (Cr_2O_7"2"−) anions in MFCs was hampered by the electrical repulsion between the negatively charged cathode and Cr(VI) functional groups. In contrast, in the presence of Fe(III), the conversion of Cr(VI) and the cathodic coulombic efficiency in the MFCs were 65.6% and 81.7%, respectively, 1.6 times and 1.4 folds as those recorded in the absence of Fe(III). Multiple analytical approaches, including linear sweep voltammetry, Tafel plot, cyclic voltammetry, electrochemical impedance spectroscopy and kinetic calculations demonstrated that the complete reduction of Cr(VI) occurred through an indirect mechanism mediated by Fe(III). The direct reduction of Cr(VI) with cathode electrons in the presence of Fe(III) was insignificant. Fe(III) played a critical role in decreasing both the diffusional resistance of Cr(VI) species and the overpotential for Cr(VI) reduction. This study demonstrated that the reduction of Cr(VI) in MFCs was effective in the presence of Fe(III), providing an alternative and environmentally benign approach for efficient remediation of Cr(VI) contaminated sites with simultaneous production of renewable energy.

  17. Thulium oxide fuel characterization study: Part 2, Environmental behavior and mechanical, thermal and chemical stability enhancement

    International Nuclear Information System (INIS)

    Nelson, C.A.

    1970-12-01

    A study was performed of the correlation between fuel form stability and exposure environment of (temperature and atmosphere). 100% Tm 2 O 3 , 80% Tm 2 O 3 /20% Yb 2 O 3 and 100% Yb 2 O 3 wafers were subjected to air, dynamic vacuum and static vacuum at temperatures to 2000 0 C for times to 100 hours. Results showed the Tm 2 O 3 /Yb 2 O 3 cubic structure to be unaffected by elemental levels of iron, aluminum, magnesium and silicon and unaffected by the environmental conditions imposed on the wafers. A second task emphasized the optimization of the thermal, mechanical and chemical stability of Tm 2 O 3 fuel forms. Enhancement was sought through process variable optimization and the addition of metal oxides to Tm 2 O 3 . CaO, TiO 2 and Al 2 O 3 were added to form a grain boundary precipitate to control fines generation. The presence of 1% additive was inadequate to depress the melting point of Tm 2 O 3 or to change the cubic crystalline structure of Tm 2 O 3 /Yb 2 O 3 . Tm 2 O 3 /Yb 2 O 3 wafers containing CaO developed a grain boundary phase that improved the resistance to fines generation. The presence of Yb 2 O 3 did not appear to measurably influence behavior

  18. Mechanical behaviour of a fuel cell stack under vibrating conditions linked to aircraft applications part II: Three-dimensional modelling

    Energy Technology Data Exchange (ETDEWEB)

    Rouss, Vicky; Charon, Willy [M3M, University of Technology Belfort - Montbeliard (France); FCLAB, Rue Thierry Mieg, F 90010 Belfort, Cedex (France); Candusso, Denis [INRETS, The French National Institute for Transport and Safety Research (France); FCLAB, Rue Thierry Mieg, F 90010 Belfort, Cedex (France)

    2008-11-15

    The implementation of fuel cells (FC) in transportation systems such as airplanes requires better understanding of their mechanical behaviour in vibrating environment. To this end, a FC stack was tested on a vibrating platform for all three orthogonal axes. The experimental procedure is described in the first part of the paper. This second part of the paper demonstrates how the experimental data collected can be used to create a three-dimensional, multi-input and multi-output model based on the Artificial Neural Network (ANN) approach. Indeed FCs are nonlinear mechanical systems, difficult to be physically modelled. The ANN methodology which depends strictly on raw data is a particularly interesting alternative solution to model FCs, for example, for monitoring purpose. The ANN model is described along with the training, pruning and validation stages. The results are exposed and commented. (author)

  19. Schinus terebinthifolius countercurrent chromatography (Part III): Method transfer from small countercurrent chromatography column to preparative centrifugal partition chromatography ones as a part of method development.

    Science.gov (United States)

    das Neves Costa, Fernanda; Hubert, Jane; Borie, Nicolas; Kotland, Alexis; Hewitson, Peter; Ignatova, Svetlana; Renault, Jean-Hugues

    2017-03-03

    Countercurrent chromatography (CCC) and centrifugal partition chromatography (CPC) are support free liquid-liquid chromatography techniques sharing the same basic principles and features. Method transfer has previously been demonstrated for both techniques but never from one to another. This study aimed to show such a feasibility using fractionation of Schinus terebinthifolius berries dichloromethane extract as a case study. Heptane - ethyl acetate - methanol -water (6:1:6:1, v/v/v/v) was used as solvent system with masticadienonic and 3β-masticadienolic acids as target compounds. The optimized separation methodology previously described in Part I and II, was scaled up from an analytical hydrodynamic CCC column (17.4mL) to preparative hydrostatic CPC instruments (250mL and 303mL) as a part of method development. Flow-rate and sample loading were further optimized on CPC. Mobile phase linear velocity is suggested as a transfer invariant parameter if the CPC column contains sufficient number of partition cells. Copyright © 2017 Elsevier B.V. All rights reserved.

  20. Normal and sonographic anatomy of selected peripheral nerves. Part III: Peripheral nerves of the lower limb

    Directory of Open Access Journals (Sweden)

    Berta Kowalska

    2012-06-01

    Full Text Available The ultrasonographic examination is currently increasingly used in imaging peripheral nerves, serving to supplement the physical examination, electromyography and magnetic resonance imaging. As in the case of other USG imaging studies, the examination of peripheral nerves is non-invasive and well-tolerated by patients. The typical ultrasonographic picture of peripheral nerves as well as the examination technique have been discussed in part I of this article series, following the example of the median nerve. Part II of the series presented the normal anatomy and the technique for examining the peripheral nerves of the upper limb. This part of the article series focuses on the anatomy and technique for examining twelve normal peripheral nerves of the lower extremity: the iliohypogastric and ilioinguinal nerves, the lateral cutaneous nerve of the thigh, the pudendal, sciatic, tibial, sural, medial plantar, lateral plantar, common peroneal, deep peroneal and superficial peroneal nerves. It includes diagrams showing the proper positioning of the sonographic probe, plus USG images of the successively discussed nerves and their surrounding structures. The ultrasonographic appearance of the peripheral nerves in the lower limb is identical to the nerves in the upper limb. However, when imaging the lower extremity, convex probes are more often utilized, to capture deeply-seated nerves. The examination technique, similarly to that used in visualizing the nerves of upper extremity, consists of locating the nerve at a characteristic anatomic reference point and tracking it using the “elevator technique”. All 3 parts of the article series should serve as an introduction to a discussion of peripheral nerve pathologies, which will be presented in subsequent issues of the “Journal of Ultrasonography”.

  1. Fuel and nuclear fuel cycle

    International Nuclear Information System (INIS)

    Prunier, C.

    1998-01-01

    The nuclear fuel is studied in detail, the best choice and why in relation with the type of reactor, the properties of the fuel cans, the choice of fuel materials. An important part is granted to the fuel assembly of PWR type reactor and the performances of nuclear fuels are tackled. The different subjects for research and development are discussed and this article ends with the particular situation of mixed oxide fuels ( materials, behavior, efficiency). (N.C.)

  2. Organizational Infrastructure in the Collegiate Athletic Training Setting, Part III: Benefits of and Barriers in the Medical and Academic Models

    Science.gov (United States)

    Eason, Christianne M.; Mazerolle, Stephanie M.; Goodman, Ashley

    2017-01-01

    Context: Academic and medical models are emerging as alternatives to the athletics model, which is the more predominant model in the collegiate athletic training setting. Little is known about athletic trainers' (ATs') perceptions of these models. Objective: To investigate the perceived benefits of and barriers in the medical and academic models. Design: Qualitative study. Setting: National Collegiate Athletic Association Divisions I, II, and III. Patients or Other Participants: A total of 16 full-time ATs (10 men, 6 women; age = 32 ± 6 years, experience = 10 ± 6 years) working in the medical (n = 8) or academic (n = 8) models. Data Collection and Analysis: We conducted semistructured telephone interviews and evaluated the qualitative data using a general inductive approach. Multiple-analyst triangulation and peer review were completed to satisfy data credibility. Results: In the medical model, role congruency and work-life balance emerged as benefits, whereas role conflict, specifically intersender conflict with coaches, was a barrier. In the academic model, role congruency emerged as a benefit, and barriers were role strain and work-life conflict. Subscales of role strain included role conflict and role ambiguity for new employees. Role conflict stemmed from intersender conflict with coaches and athletics administrative personnel and interrole conflict with fulfilling multiple overlapping roles (academic, clinical, administrative). Conclusions: The infrastructure in which ATs provide medical care needs to be evaluated. We found that the medical model can support better alignment for both patient care and the wellbeing of ATs. Whereas the academic model has perceived benefits, role incongruence exists, mostly because of the role complexity associated with balancing teaching, patient-care, and administrative duties. PMID:27977302

  3. Organizational Infrastructure in the Collegiate Athletic Training Setting, Part III: Benefits of and Barriers in the Medical and Academic Models.

    Science.gov (United States)

    Eason, Christianne M; Mazerolle, Stephanie M; Goodman, Ashley

    2017-01-01

     Academic and medical models are emerging as alternatives to the athletics model, which is the more predominant model in the collegiate athletic training setting. Little is known about athletic trainers' (ATs') perceptions of these models.  To investigate the perceived benefits of and barriers in the medical and academic models.  Qualitative study.  National Collegiate Athletic Association Divisions I, II, and III.  A total of 16 full-time ATs (10 men, 6 women; age = 32 ± 6 years, experience = 10 ± 6 years) working in the medical (n = 8) or academic (n = 8) models.  We conducted semistructured telephone interviews and evaluated the qualitative data using a general inductive approach. Multiple-analyst triangulation and peer review were completed to satisfy data credibility.  In the medical model, role congruency and work-life balance emerged as benefits, whereas role conflict, specifically intersender conflict with coaches, was a barrier. In the academic model, role congruency emerged as a benefit, and barriers were role strain and work-life conflict. Subscales of role strain included role conflict and role ambiguity for new employees. Role conflict stemmed from intersender conflict with coaches and athletics administrative personnel and interrole conflict with fulfilling multiple overlapping roles (academic, clinical, administrative).  The infrastructure in which ATs provide medical care needs to be evaluated. We found that the medical model can support better alignment for both patient care and the wellbeing of ATs. Whereas the academic model has perceived benefits, role incongruence exists, mostly because of the role complexity associated with balancing teaching, patient-care, and administrative duties.

  4. Analysis of the second part of the fuel cycle of nuclear spanish park using module TREVOL of EVOLCODE2

    International Nuclear Information System (INIS)

    Merino Rodriguez, I.; Alvarez-Velarde, F.; Martin-Fuertes, F.

    2011-01-01

    This paper describes the application of the code TR E VOL an associated fuel cycle Spanish nuclear park, with the objective of estimating the mass of nuclear fuel manufactured by reactor and the mass generated of irradiated fuel.

  5. Uranium accountability for ATR fuel fabrication: Part II. A computer simulation

    International Nuclear Information System (INIS)

    Dolan, C.A.; Nieschmidt, E.B.; Vegors, S.H. Jr.; Wagner, E.P. Jr.

    1977-08-01

    A stochastic computer model has been designed to simulate the material control system used during the production of fuel plates for the Advanced Test Reactor. Great care has been taken to see that this model follows the manufacturing and measuring processes used. The model is designed so that manufacturing process and measurement parameters are fed in as input; hence, changes in the manufacturing process and measurement procedures are easily simulated. Individual operations in the plant are described by program subroutines. By varying the calling sequence of these subroutines, variations in the manufacturing process may be simulated. By using this model values for MUF and LEMUF may be calculated for predetermined plant operating conditions. Furthermore the effect on MUF and LEMUF produced by changing plant operating procedures and measurement techniques may also be examined. A sample calculation simulating one inventory period of the plant's operation is included

  6. Field test corrosion experiments in Denmark with biomass fuels Part I Straw firing

    DEFF Research Database (Denmark)

    Montgomery, Melanie; Karlsson, A; Larsen, OH

    2002-01-01

    plants. The type of corrosion attack can be directly ascribed to the composition of the deposit and the metal surface temperature. A series of field tests have been undertaken in the various straw-fired power plants in Denmark, namely Masnedø, Rudkøbing and Ensted. Three types of exposure were undertaken......In Denmark, straw and other types of biomass are used for generating energy in power plants. Straw has the advantage that it is a "carbon dioxide neutral fuel" and therefore environmentally acceptable. Straw combustion is associated with corrosion problems which are not encountered in coal-fired...... to investigate corrosion: a) the exposure of metal rings on water/air cooled probes, b) the exposure of test tubes in a test superheater, and c) the exposure of test tubes in existing superheaters. Thus both austenitic steels and ferritic steels were exposed in the steam temperature range of 450-600°C...

  7. New 20-cm radio-continuum study of the small Magellanic cloud - part III: Compact Hii regions

    Directory of Open Access Journals (Sweden)

    Wong G.F.

    2012-01-01

    Full Text Available We present and discuss a new catalogue of 48 compact Hii regions in the Small Magellanic Cloud (SMC and a newly created deep 1420 MHz (λ=20 cm radio-continuum image of the N19 region located in the southwestern part of the SMC. The new images were created by merging 1420 MHz radiocontinuum archival data from the Australian Telescope Compact Array. The majority of these detected radio compact Hii regions have rather flat spectral indices which indicates, as expected, that the dominant emission mechanism is of thermal nature.

  8. Mammalian Toxicity of Munition Compounds. Phase II. Effects of Multiple Doses. Part III. 2,6-Dinitrotoluene

    Science.gov (United States)

    1976-07-01

    and the neuromuscular effects in these dogs were not due to hypocalcemia . The lowest serum calcium concen- tration in these dogs was 4.2 meq/liter...motor end plate might produce a local hypocalcemia . Such a mechanism is purely speculative. Qualitatively and quantitavely, most of the effects of 2,6...I ýNw,- -MIM I/ MIDWEST RESEARCH INS14ITUTE H0q .3L I LU -_ MAMMALIAN TOXICITY OF MUNITIONS COMPOUlNDSPHASE II: EFFECTS OF MiULTIPLE DOSES C* •PART

  9. Life-cycle analysis of energy and greenhouse gas emissions of automotive fuels in India: Part 1 – Tank-to-Wheel analysis

    International Nuclear Information System (INIS)

    Gupta, S.; Patil, V.; Himabindu, M.; Ravikrishna, R.V.

    2016-01-01

    As part of a two-part life cycle efficiency and greenhouse gas emission analysis for various automotive fuels in the Indian context, this paper presents the first part, i.e., Tank-to-Wheel analysis of various fuel/powertrain configurations for a subcompact passenger car. The Tank-to-Wheel analysis was applied to 28 fuel/powertrain configurations using fuels such as gasoline, diesel, compressed natural gas, liquefied petroleum gas and hydrogen with various conventional and hybrid electric powertrains. The gasoline-equivalent fuel economy and carbon dioxide emission results for individual fuel/powertrain configuration are evaluated and compared. It is found that the split hybrid configuration is best among hybrids as it leads to fuel economy improvement and carbon dioxide emissions reduction by 20–40% over the Indian drive cycle. Further, the engine efficiency, engine on-off time and regenerative braking energy assessment is done to evaluate the causes for higher energy efficiency of hybrid electric vehicles. The hybridization increases average engine efficiency by 10–60% which includes 19–23% of energy recovered at wheel through regenerative braking over the drive cycle. Overall, the Tank-to-Wheel energy use and efficiency results are evaluated for all fuel/powertrain configurations which show Battery Electric Vehicle, fuel cell vehicles and diesel hybrids are near and long term energy efficient vehicle configurations. - Highlights: • Tank-to-Wheel energy use & CO_2 emissions for subcompact car on Indian driving cycle. • Gasoline, diesel, CNG, LPG, hydrogen and electric vehicles are evaluated in this study. • First comprehensive Tank-to-Wheel analysis for India on small passenger car platform. • Parallel, series and split hybrid electric vehicles with various fuels are analysed.

  10. Reprocessing of the spent nuclear fuel, I-VIII, Part VIII

    International Nuclear Information System (INIS)

    Gal, I.

    1963-02-01

    This volume includes the following three reports: Separation of uranium, plutonium and fission products on zirconium phosphate; Separation of uranium, plutonium and fission products on zirconium phosphate, Part 1 - Adsorption equilibria and kinetics; and Adsorption of dibutyl phosphates and monobutyl phosphates on inorganic oxides

  11. A-Part Gel, an adhesion prophylaxis for abdominal surgery: a randomized controlled phase I-II safety study [NCT00646412].

    Science.gov (United States)

    Lang, Reinhold; Baumann, Petra; Schmoor, Claudia; Odermatt, Erich K; Wente, Moritz N; Jauch, Karl-Walter

    2015-01-01

    Intra-abdominal surgical intervention can cause the development of intra-peritoneal adhesions. To reduce this problem, different agents have been tested to minimize abdominal adhesions; however, the optimal adhesion prophylaxis has not been found so far. Therefore, the A-Part(®) Gel was developed as a barrier to diminish postsurgical adhesions; the aim of this randomized controlled study was a first evaluation of its safety and efficacy. In this prospective, controlled, randomized, patient-blinded, monocenter phase I-II study, 62 patients received either the hydrogel A-Part-Gel(®) as an anti-adhesive barrier or were untreated after primary elective median laparotomy. Primary endpoint was the occurrence of peritonitis and/or wound healing impairment 28 ± 10 days postoperatively. As secondary endpoints anastomotic leakage until 28 days after surgery, adverse events and adhesions were assessed until 3 months postoperatively. A lower rate of wound healing impairment and/or peritonitis was observed in the A-Part Gel(®) group compared to the control group: (6.5 vs. 13.8 %). The difference between the two groups was -7.3%, 90 % confidence interval [-20.1, 5.4 %]. Both treatment groups showed similar frequency of anastomotic leakage but incidence of adverse events and serious adverse events were slightly lower in the A-Part Gel(®) group compared to the control. Adhesion rates were comparable in both groups. A-Part Gel(®) is safe as an adhesion prophylaxis after abdominal wall surgery but no reduction of postoperative peritoneal adhesion could be found in comparison to the control group. This may at least in part be due to the small sample size as well as to the incomplete coverage of the incision due to the used application. NCT00646412.

  12. Dental compensation for skeletal Class III malocclusion by isolated extraction of mandibular teeth. Part 1: Occlusal situation 12 years after completion of active treatment.

    Science.gov (United States)

    Zimmer, Bernd; Schenk-Kazan, Sarah

    2015-05-01

    The purpose of this work was to statistically evaluate the outcomes achieved by isolated extraction of mandibular teeth (second premolars or first molars) for Class III compensation. Part A of the study dealt with the quality of outcomes at the end of active treatment, using weighted Peer Assessment Rating (PAR) scores determined on the basis of casts for 25 (14 female and 11 male) consecutive patients aged 16 ± 1.7 years at the time of debonding. These results were compared to the scores in a randomly selected control group of 25 (14 female and 11 male) patients who were 14.7 ± 1.9 years old at debonding. Part B evaluated the long-term stability of the outcomes based on 12 (all of them female) patients available for examination after a mean of 11.8 years. The mean weighted PAR scores obtained in both study parts were analyzed for statistical differences using a two-tailed paired Student's t-test at a significance level of p ≤ 0.05. Mean weighted PAR scores of 4.76 ± 3.94 and 3.92 ± 3.44 were obtained in the Class III extraction group and the control group, respectively, at the end of active treatment. This difference was not significant (p = 0.49). Among the 12 longitudinal patients, the mean score increased from 4 ± 3.46 at debonding to 6.25 ± 3.67 by the end of the 11.8-year follow-up period. This difference was significant (p = 0.0008). Treatment of Class III anomalies by isolated extraction of lower premolars or molars can yield PAR scores similar to those achieved by standard therapies. These scores, while increasing significantly, remained at a clinically acceptable level over 11.8 years. Hence this treatment modality--intended for cases that border on requiring orthognathic surgery--may also be recommended from a long-term point of view.

  13. Thermal conductivity of catalyst layer of polymer electrolyte membrane fuel cells: Part 1 - Experimental study

    Science.gov (United States)

    Ahadi, Mohammad; Tam, Mickey; Saha, Madhu S.; Stumper, Jürgen; Bahrami, Majid

    2017-06-01

    In this work, a new methodology is proposed for measuring the through-plane thermal conductivity of catalyst layers (CLs) in polymer electrolyte membrane fuel cells. The proposed methodology is based on deconvolution of bulk thermal conductivity of a CL from measurements of two thicknesses of the CL, where the CLs are sandwiched in a stack made of two catalyst-coated substrates. Effects of hot-pressing, compression, measurement method, and substrate on the through-plane thermal conductivity of the CL are studied. For this purpose, different thicknesses of catalyst are coated on ethylene tetrafluoroethylene (ETFE) and aluminum (Al) substrates by a conventional Mayer bar coater and measured by scanning electron microscopy (SEM). The through-plane thermal conductivity of the CLs is measured by the well-known guarded heat flow (GHF) method as well as a recently developed transient plane source (TPS) method for thin films which modifies the original TPS thin film method. Measurements show that none of the studied factors has any effect on the through-plane thermal conductivity of the CL. GHF measurements of a non-hot-pressed CL on Al yield thermal conductivity of 0.214 ± 0.005 Wṡm-1ṡK-1, and TPS measurements of a hot-pressed CL on ETFE yield thermal conductivity of 0.218 ± 0.005 Wṡm-1ṡK-1.

  14. Use of rumination and activity monitoring for the identification of dairy cows with health disorders: Part III. Metritis.

    Science.gov (United States)

    Stangaferro, M L; Wijma, R; Caixeta, L S; Al-Abri, M A; Giordano, J O

    2016-09-01

    The objectives of this study were to evaluate (1) the performance of an automated health-monitoring system (AHMS) to identify cows with metritis based on an alert system (health index score, HIS) that combines rumination time and physical activity; (2) the number of days between the first HIS alert and clinical diagnosis (CD) of metritis by farm personnel; and (3) the daily rumination time, physical activity, and HIS patterns around CD. In this manuscript, the overall performance of HIS to detect cows with all disorders of interest in this study [ketosis, displaced abomasum, indigestion (companion paper, part I), mastitis (companion paper, part II), and metritis] is also reported. Holstein cattle (n=1,121; 451 nulliparous and 670 multiparous) were fitted with a neck-mounted electronic rumination and activity monitoring tag (HR Tags, SCR Dairy, Netanya, Israel) from at least -21 to 80 d in milk (DIM). Raw data collected in 2-h periods were summarized per 24 h as daily rumination and activity. An HIS (0 to 100 arbitrary units) was calculated daily for individual cows with an algorithm that used rumination and activity. A positive HIS outcome was defined as an HIS of cows (n=459) at -11±3, -4±3, 0, 3±1, 7±1, 14±1, and 28±1 DIM. The overall sensitivity of HIS was 55% for all cases of metritis (n=349), but it was greater for cows with metritis and another disorder (78%) than for cows with metritis only (53%). Cows diagnosed with metritis and flagged based on HIS had substantial alterations in their rumination, activity, and HIS patterns around CD, alterations of blood markers of metabolic and health status around calving, reduced milk production, and were more likely to exit the herd than cows not flagged based on the HIS and cows without disease, suggesting that cows flagged based on the HIS had a more severe episode of metritis. Including all disorders of interest for this study, the overall sensitivity was 59%, specificity was 98%, positive predictive value was

  15. Controlled production of camembert-type cheeses: part III role of the ripening microflora on free fatty acid concentrations.

    Science.gov (United States)

    Leclercq-Perlat, Marie-Noëlle; Corrieu, Georges; Spinnler, Henry-Eric

    2007-05-01

    Phenomena generating FFAs, important flavour precursors, are significant in cheese ripening. In Camembert-like cheeses, it was intended to establish the relationships between the dynamics of FFA concentrations changes and the succession of ripening microflora during ripening. Experimental Camembert-type cheeses were prepared in duplicate from pasteurised milk inoculated with Kluyveromyces lactis, Geotrichum candidum, Penicillium camemberti, and Brevibacterium aurantiacum under aseptic conditions. For each cheese and each cheesy medium, concentrations of FFAs with odd-numbered carbons, except for 9:0 and 13:0, did not change over time. For long-chain FFAs, concentrations varied with the given cheese part (rind or core). K. lactis produced only short or medium-chain FFAs during its growth and had a minor influence on caproic, caprylic, capric, and lauric acids in comparison with G. candidum, the most lipolytic of the strains used here. It generated all short or medium-chain FFAs (4:0-12:0) during its exponential and slowdown growth periods and only long-chain ones (14:0-18:0) during its stationary phase. Pen. camemberti produced more long-chain FFAs (14:0-18:0) during its sporulation. Brev. aurantiacum did not generate any FFAs. The evidence of links between specific FFAs and the growth of a given microorganism is shown.

  16. Ash behavior during hydrothermal treatment for solid fuel applications. Part 1: Overview of different feedstock

    International Nuclear Information System (INIS)

    Mäkelä, Mikko; Fullana, Andrés; Yoshikawa, Kunio

    2016-01-01

    Highlights: • Ash behavior of 29 different feedstock interpreted using multivariate data analysis. • Two different groups identified based on char ash content and ash yield. • Solubility of individual elements evaluated based on a smaller data set. • Ash from industrial sludge contained anthropogenic metals with low solubility. - Abstract: Differences in ash behavior during hydrothermal treatment were identified based on multivariate data analysis of literature information on 29 different feedstock. In addition, the solubility of individual elements was evaluated based on a smaller data set. As a result two different groups were distinguished based on char ash content and ash yield. Virgin terrestrial and aquatic biomass, such as different types of wood and algae, in addition to herbaceous and agricultural biomass, bark, brewer’s spent grain, compost and faecal waste showed lower char ash content than municipal solid wastes, anaerobic digestion residues and municipal and industrial sludge. Lower char ash content also correlated with lower ash yield indicating differences in chemical composition and ash solubility. Further evaluation of available data showed that ash in industrial sludge mainly contained anthropogenic Al, Fe and P or Ca and Si with low solubility during hydrothermal treatment. Char from corn stover, miscanthus, switch grass, rice hulls, olive, artichoke and orange wastes and empty fruit bunch had generally higher contents of K, Mg, S and Si than industrial sludge although differences existed within the group. In the future information on ash behavior should be used for enhancing the fuel properties of char based on feedstock type and hydrothermal treatment conditions.

  17. Transcontinental methane measurements: Part 2. Mobile surface investigation of fossil fuel industrial fugitive emissions

    Science.gov (United States)

    Leifer, Ira; Culling, Daniel; Schneising, Oliver; Farrell, Paige; Buchwitz, Michael; Burrows, John P.

    2013-08-01

    The potent greenhouse gas, methane, CH4, has a wide variety of anthropogenic and natural sources. Fall, continental-scale (Florida to California) surface CH4 data were collected to investigate the importance of fossil fuel industrial (FFI) emissions in the South US. A total of 6600 measurements along 7020-km of roadways were made by flame ion detection gas chromatography onboard a nearly continuously moving recreational vehicle in 2010. A second, winter survey in Southern California measured CH4 at 2 Hz with a cavity ring-down spectrometer in 2012. Data revealed strong and persistent FFI CH4 sources associated with refining, oil/gas production, a presumed major pipeline leak, and a coal loading plant. Nocturnal CH4 mixing ratios tended to be higher than daytime values for similar sources, sometimes significantly, which was attributed to day/night meteorological differences, primarily changes in the boundary layer height. The highest CH4 mixing ratio (39 ppm) was observed near the Kern River Oil Field, California, which uses steam reinjection. FFI CH4 plume signatures were distinguished as stronger than other sources on local scales. On large (4°) scales, the CH4 trend was better matched spatially with FFI activity than wetland spatial patterns. Qualitative comparison of surface data with SCIAMACHY and GOSAT satellite retrievals showed agreement of the large-scale CH4 spatial patterns. Comparison with inventory models and seasonal winds suggests for some seasons and some portions of the Gulf of Mexico a non-negligible underestimation of FFI emissions. For other seasons and locations, qualitative interpretation is not feasible. Unambiguous quantitative source attribution is more complex, requiring transport modeling.

  18. Hybridization of powertrain and downsizing of IC engine - A way to reduce fuel consumption and pollutant emissions - Part 1

    International Nuclear Information System (INIS)

    Katrasnik, Tomaz

    2007-01-01

    The aim of this two part paper is to present the results of extensive simulation and analytical analysis of the energy conversion efficiency in parallel hybrid powertrains. The simulation approach is based on an accurate and fast forward facing simulation model of a parallel hybrid powertrain and a conventional internal combustion engine powertrain. The model of the ICE is based on a verified dynamic model that provides sufficiently small time steps to model adequately the dynamics of electric systems during transient test cycles. Models of the electrical devices enable computation of the instantaneous energy consumption, production and storage as well as computation of the instantaneous energy losses and component efficiencies. Moreover, the paper offers an analytical approach based on the energy balance in order to analyze and predict the energy conversion efficiency of hybrid powertrains. The analysis covers a broad range of parallel hybrid powertrain configurations from mild to full hybrids. Combined simulation and analytical analysis enables deep insight into the energy conversion phenomena in hybrid powertrains. The paper reveals the conditions and influences that lead to improved fuel economy of hybrid powertrains with the emphasis on determining the optimum hybridization ratio. The theoretical background, simulation program and brief analysis of one test cycle are presented in Part 1, whereas the extensive analysis and parametric study is presented in the companion paper, Part 2

  19. Performance limits of coated particle fuel. Part I. The significance of empirical performance diagrams and mathematical models in fuel development and power reactor studies

    Energy Technology Data Exchange (ETDEWEB)

    Graham, L. W.; Hick, H.

    1973-06-15

    This report introduces a general survey of our present knowledge and understanding of coated particle fuel performance. It defines first the reference power reactor conditions and the reference coated particle design on which the survey is centred. It describes then the typical strategy which has been followed in coated particle fuel development by the Dragon Project R & D Branch. Finally it shows the priorities which have governed the time scale and scope of fuel development and of the present review.

  20. Evaluation of pump pulsation in respirable size-selective sampling: Part III. Investigation of European standard methods.

    Science.gov (United States)

    Soo, Jhy-Charm; Lee, Eun Gyung; Lee, Larry A; Kashon, Michael L; Harper, Martin

    2014-10-01

    Lee et al. (Evaluation of pump pulsation in respirable size-selective sampling: part I. Pulsation measurements. Ann Occup Hyg 2014a;58:60-73) introduced an approach to measure pump pulsation (PP) using a real-world sampling train, while the European Standards (EN) (EN 1232-1997 and EN 12919-1999) suggest measuring PP using a resistor in place of the sampler. The goal of this study is to characterize PP according to both EN methods and to determine the relationship of PP between the published method (Lee et al., 2014a) and the EN methods. Additional test parameters were investigated to determine whether the test conditions suggested by the EN methods were appropriate for measuring pulsations. Experiments were conducted using a factorial combination of personal sampling pumps (six medium- and two high-volumetric flow rate pumps), back pressures (six medium- and seven high-flow rate pumps), resistors (two types), tubing lengths between a pump and resistor (60 and 90 cm), and different flow rates (2 and 2.5 l min(-1) for the medium- and 4.4, 10, and 11.2 l min(-1) for the high-flow rate pumps). The selection of sampling pumps and the ranges of back pressure were based on measurements obtained in the previous study (Lee et al., 2014a). Among six medium-flow rate pumps, only the Gilian5000 and the Apex IS conformed to the 10% criterion specified in EN 1232-1997. Although the AirChek XR5000 exceeded the 10% limit, the average PP (10.9%) was close to the criterion. One high-flow rate pump, the Legacy (PP=8.1%), conformed to the 10% criterion in EN 12919-1999, while the Elite12 did not (PP=18.3%). Conducting supplemental tests with additional test parameters beyond those used in the two subject EN standards did not strengthen the characterization of PPs. For the selected test conditions, a linear regression model [PPEN=0.014+0.375×PPNIOSH (adjusted R2=0.871)] was developed to determine the PP relationship between the published method (Lee et al., 2014a) and the EN methods

  1. Management strategies to effect change in intensive care units: lessons from the world of business. Part III. Effectively effecting and sustaining change.

    Science.gov (United States)

    Gershengorn, Hayley B; Kocher, Robert; Factor, Phillip

    2014-03-01

    Reaping the optimal rewards from any quality improvement project mandates sustainability after the initial implementation. In Part III of this three-part ATS Seminars series, we discuss strategies to create a culture for change, improve cooperation and interaction between multidisciplinary teams of clinicians, and position the intensive care unit (ICU) optimally within the hospital environment. Coaches are used throughout other industries to help professionals assess and continually improve upon their practice; use of this strategy is as of yet infrequent in health care, but would be easily transferable and potentially beneficial to ICU managers and clinicians alike. Similarly, activities focused on improving teamwork are commonplace outside of health care. Simulation training and classroom education about key components of successful team functioning are known to result in improvements. In addition to creating an ICU environment in which individuals and teams of clinicians perform well, ICU managers must position the ICU to function well within the hospital system. It is important to move away from the notion of a standalone ("siloed") ICU to one that is well integrated into the rest of the institution. Creating a "pull-system" (in which participants are active in searching out needed resources and admitting patients) can help ICU managers both provide better care for the critically ill and strengthen relationships with non-ICU staff. Although not necessary, there is potential upside to creating a unified critical care service to assist with achieving these ends.

  2. Determining the axial power profile of partly flooded fuel in a compact core assembled in reactor LR-0

    International Nuclear Information System (INIS)

    Košťál, Michal; Švadlenková, Marie; Baroň, Petr; Rypar, Vojtěch; Milčák, Ján

    2016-01-01

    Highlights: • Fission density in partly flooded compact core. • Calculation of fission density axial profile. • Significant calculational under prediction of experimental axial profile. - Abstract: Measurement and calculation of the axial power profile near the boundary of a moderated and non-moderated core is used to analyze the suitability of the neutron-physical process description, mainly the angular cross-section of a water-moderated uranium system. This is also an important issue because it affects the radiation situation above the partly flooded core of a water-moderated reactor. Axial power profiles of various fuel pins irradiated on reactor LR-0 were measured and the results were compared with MCNP6 code calculations using the ENDF/B-VII.0 nuclear data library. The calculated power profile in positions above the moderator level significantly underestimates experimental results. This might be caused by an improper description of the angular distribution of scattered neutrons in a water-moderated uranium system.

  3. Extracción de cromo con disolventes orgánicos. III parte. Aplicación al tratamiento de residuos polimetálicos industriales

    Directory of Open Access Journals (Sweden)

    de Juan, D.

    1998-10-01

    Full Text Available The use of Primene 81R as extraction agent of chromium present in solid wastes containing nickel, iron and copper has been studied. The waste was leached with a sulphuric acid solution up to pH 3 and oxidation of Cr(III to Cr(VI with Caro acid was also studied. Because of the negative result of oxidation, the treatment was applied on Cr(III directly. Extraction/scrubbing/stripping process was studied in the leach. The composition of organic phase used in the extraction step was 10 % v/v Primene 81R, 10 % isodecanol and kerosene. All the iron, 91 % Cr, 10 % Ni and large part of the copper contained in the initial leach solution are recovered in the organic phase. In the scrubbing stage (with a sulphuric acid solution at pH 1,4, all the copper and nickel and 30 % Cr go to the washing liquor, while all the iron and 70 % Cr remained in the organic phase. In the stripping stage (with a 2N NH4OH or 2N NaOH solution all the iron and chromium are recovered as a precipitate of highly absorbent hydroxides. After the treatment mentioned, 63 % Cr and 100 % Fe are recovered as a mixture of hydroxides, and 28 % of the initial chromium, all the nickel and the copper are found in the washing liquor.

    Se estudia la aplicación del Primene 81R como agente de extracción del cromo presente en residuos sólidos que poseen níquel, hierro y cobre. El residuo se lixivió con disolución de ácido sulfúrico hasta pH 3 y se investigó la oxidación del Cr(III a Cr(VI con ácido de Caro. Ante el nulo resultado en la oxidación, se actuó directamente sobre el Cr(III. Se estudió el proceso de extracción/lavado/reextracción sobre la lejía de lixiviación. La fase orgánica empleada en la extracción estaba constituida por 10 % v/v de Primene 81R, 10 % de isodecanol y queroseno. En la fase orgánica se recupera todo el hierro, el 91 % Cr, el 10 % Ni y gran parte del cobre contenidos en la lejía de partida. En la etapa de lavado (con disolución de

  4. Prototypical spent nuclear fuel rod consolidation equipment: Phase 2, Final design report: Volume 4, Appendices: Part 3

    International Nuclear Information System (INIS)

    Ciez, A.P.

    1987-01-01

    The purpose of this manual is to provide assembly, installation, operation, maintenance, and off-normal recovery procedures for the Consolidation Equipment. The Consolidation System is a horizontal, dry system capable of processing one Pressurized Water Reactor (PWR) fuel assembly or one Boiling Water Reactor (BWR) fuel assembly at a time. The system will process all spent PWR and BWR fuels from the commercial US nuclear power reactor industry. Component changeouts for various fuel types have been minimized to reduce costs, required in-cell module storage space, and to increase efficiency by decreasing set-up time between fuel consolidation campaigns. The most important feature of the Westinghouse system is the ability to control the fuel rods at all times during the consolidation process from rod extraction, through canister loading. This features assures that the rods from two PWR fuel assemblies or four BWR fuel assemblies (minimum) can be loaded into one consolidated rods canister

  5. Nuclear fuel technology - Tank calibration and volume determination for nuclear materials accountancy - Part 1: Procedural overview

    International Nuclear Information System (INIS)

    2007-01-01

    Accurate determinations of volume are a fundamental component of any measurement-based system of control and accountability in a facility that processes or stores nuclear materials in liquid form. Volume determinations are typically made with the aid of a calibration or volume measurement equation that relates the response of the tank's measurement system to some independent measure of tank volume. The ultimate purpose of the calibration exercise is to estimate the tank's volume measurement equation (the inverse of the calibration equation), which relates tank volume to measurement system response. The steps carried out to acquire data for estimating the tank's calibration or volume measurement equation are collectively described as the process of tank calibration. This part of ISO 18213 describes procedures for tank calibration and volume determination for nuclear process tanks equipped with pressure-measurement systems for determining liquid content. Specifically, overall guidance is provided for planning a calibration exercise undertaken to obtain the data required for the measurement equation to estimate a tank's volume. The key steps in the procedure are also presented for subsequently using the estimated volume-measurement equation to determine tank liquid volumes. The procedures presented apply specifically to tanks equipped with bubbler probe systems for measuring liquid content. Moreover, these procedures produce reliable results only for clear (i.e. without suspended solids), homogeneous liquids that are at both thermal and static equilibrium. The paper elaborates on scope, physical principles involved, the calibration model, equipment required, a typical tank calibration procedure, calibration planning and pre-calibration activities, and volume determination. A bibliography is provided

  6. Thorium utilization in a small long-life HTR. Part I: Th/U MOX fuel blocks

    Energy Technology Data Exchange (ETDEWEB)

    Ding, Ming, E-mail: dingm2005@gmail.com [Delft University of Technology, Reactor Institute Delft, Mekelweg 15, 2629 JB, Delft (Netherlands); Harbin Engineering University, Nantong Street 145, 150001 Harbin (China); Kloosterman, Jan Leen, E-mail: j.l.kloosterman@tudelft.nl [Delft University of Technology, Reactor Institute Delft, Mekelweg 15, 2629 JB, Delft (Netherlands)

    2014-02-15

    Highlights: • We propose thorium MOX (TMOX) fuel blocks for a small block-type HTR. • The TMOX fuel blocks with low-enriched uranium are recommended. • More thorium decreases the reactivity swing of the TMOX fuel blocks. • Thorium reduces the negative temperature coefficient of the TMOX fuel blocks. • Thorium increases the conversion ratio of the TMOX fuel blocks. - Abstract: The U-Battery is a small, long-life and transportable high temperature gas-cooled reactor (HTR). The neutronic features of a typical fuel block with uranium and thorium have been investigated for a application of the U-Battery, by parametrically analyzing the composition and geometric parameters. The type of fuel block is defined as Th/U MOX fuel block because uranium and thorium are assumed to be mixed in each fuel kernel as a form of (Th,U)O{sub 2}. If the initially loaded mass of U-235 is mostly consumed in the early period of the lifetime of Th/U MOX fuel block, low-enriched uranium (LEU) as ignited fuel will not largely reduce the neutronic performance of the Th/U MOX fuel block, compared with high-enriched uranium. The radii of fuel kernels and fuel compacts and packing fraction of TRISO particles determine the atomic ratio of the carbon to heavy metal. When the ratio is smaller than 400, the difference among them due to double heterogeneous effects can be neglected for the Th/U MOX fuel block. In the range between 200 and 400, the reactivity swing of the Th/U MOX fuel block during 10 years is sufficiently small. The magnitude of the negative reactivity temperature coefficients of the Th/U MOX fuel block decreases by 20–45%, which is positive to reduce temperature defect of the Th/U MOX fuel block. The conversion ratio (CR) of the fuel increases from 0.48 (typical CR of the LEU-fueled U-Battery) to 0.78. The larger conversion ratio of the Th/U MOX fuel block reduces the reactivity swing during 10 years for the U-Battery.

  7. Milliped Miscellany — Part III

    NARCIS (Netherlands)

    Jeekel, C.A.W.

    1956-01-01

    MATERIAL: Eritrea: Gula (15° 36’ N., 38° 21’ E.), 500 m., 19 July 1953, Coll. W. J. STOWER, 1 ♂ (holotype). COLOUR : Head, except the lower portion of the clypeus and a sharply demarcated spot at the medio-posterior side of the antennal sockets which are yellowish, very dark brown. Antennae and legs

  8. Tarski Geometry Axioms. Part III

    Directory of Open Access Journals (Sweden)

    Coghetto Roland

    2017-12-01

    Full Text Available In the article, we continue the formalization of the work devoted to Tarski’s geometry - the book “Metamathematische Methoden in der Geometrie” by W. Schwabhäuser, W. Szmielew, and A. Tarski. After we prepared some introductory formal framework in our two previous Mizar articles, we focus on the regular translation of underlying items faithfully following the abovementioned book (our encoding covers first seven chapters. Our development utilizes also other formalization efforts of the same topic, e.g. Isabelle/HOL by Makarios, Metamath or even proof objects obtained directly from Prover9. In addition, using the native Mizar constructions (cluster registrations the propositions (“Satz” are reformulated under weaker conditions, i.e. by using fewer axioms or by proposing an alternative version that uses just another axioms (ex. Satz 2.1 or Satz 2.2.

  9. Workshop 97. Part III. Proceedings

    International Nuclear Information System (INIS)

    1996-12-01

    This volume of the Proceedings covers the following branches of science and technology: power systems and electrical engineering, electronics and measuring and communication engineering, optics, quantum electronics and photonics, microelectronics, and biomedical engineering. Out of the contributions, 2 have been input to INIS. (P.A.)

  10. ASIST 2003: Part III: Posters.

    Science.gov (United States)

    Proceedings of the ASIST Annual Meeting, 2003

    2003-01-01

    Twenty-three posters address topics including access to information; metadata; personal information management; scholarly information communication; online resources; content analysis; interfaces; Web queries; information evaluation; informatics; information needs; search effectiveness; digital libraries; diversity; automated indexing; e-commerce;…

  11. MicroVent (part III)

    DEFF Research Database (Denmark)

    Dreau, Jerome Le; Heiselberg, Per Kvols; Jensen, Rasmus Lund

    This study aims at using the InVentilate unit in the cooling case, without heat recovery. It results in a relatively low inlet air temperature. Different solutions have been tested to decrease the risk of draught in the occupied zone: ‐ Using a mixer (2 designs) ‐ Using an inlet grille ‐ Using...

  12. Estimation of the development possibility of the ABC/ATW fuel cycle based on LiF-BeF2 fuel salt. Part 2

    International Nuclear Information System (INIS)

    Bychkov, A.V.; Naumov, V.S.

    1994-01-01

    The aim of the first chapter was generalization of data on solubility and equilibrium states of fission product and actinide fluorides in fluoride salt melts-solvents and fuel composition melts based on LiF-BeF 2 mixture which was proposed as fuel basis for ABC/ATW facility. The second chapter is devoted to description of processes proposed for the chemical-technological complex of the ABC/ATW facility and their physico-chemical peculiarities. The complex is responsible for the removal of fission products and actinides from irradiated fuel salt

  13. Fuel processing

    International Nuclear Information System (INIS)

    Allardice, R.H.

    1990-01-01

    The technical and economic viability of the fast breeder reactor as an electricity generating system depends not only upon the reactor performance but also on a capability to recycle plutonium efficiently, reliably and economically through the reactor and fuel cycle facilities. Thus the fuel cycle is an integral and essential part of the system. Fuel cycle research and development has focused on demonstrating that the challenging technical requirements of processing plutonium fuel could be met and that the sometimes conflicting requirements of the fuel developer, fuel fabricator and fuel reprocessor could be reconciled. Pilot plant operation and development and design studies have established both the technical and economic feasibility of the fuel cycle but scope for further improvement exists through process intensification and flowsheet optimization. These objectives and the increasing processing demands made by the continuing improvement to fuel design and irradiation performance provide an incentive for continuing fuel cycle development work. (author)

  14. Material control in nuclear fuel fabrication facilities. Part II. Accountability, instrumntation, and measurement techniques in fuel fabrication facilities, P.O.1236909. Final report

    International Nuclear Information System (INIS)

    Borgonovi, G.M.; McCartin, T.J.; McDaniel, T.; Miller, C.L.; Nguyen, T.

    1978-12-01

    This report describes the measurement techniques, the instrumentation, and the procedures used in accountability and control of nuclear materials, as they apply to fuel fabrication facilities. Some of the material included has appeared elswhere and it has been summarized. An extensive bibliography is included. A spcific example of application of the accountability methods to a model fuel fabrication facility which is based on the Westinghouse Anderson design

  15. Material control in nuclear fuel fabrication facilities. Part II. Accountability, instrumntation, and measurement techniques in fuel fabrication facilities, P. O. 1236909. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Borgonovi, G.M.; McCartin, T.J.; McDaniel, T.; Miller, C.L.; Nguyen, T.

    1978-12-01

    This report describes the measurement techniques, the instrumentation, and the procedures used in accountability and control of nuclear materials, as they apply to fuel fabrication facilities. Some of the material included has appeared elswhere and it has been summarized. An extensive bibliography is included. A spcific example of application of the accountability methods to a model fuel fabrication facility which is based on the Westinghouse Anderson design.

  16. 10 CFR Appendix F to Part 50 - Policy Relating to the Siting of Fuel Reprocessing Plants and Related Waste Management Facilities

    Science.gov (United States)

    2010-01-01

    ... and Related Waste Management Facilities F Appendix F to Part 50 Energy NUCLEAR REGULATORY COMMISSION... Relating to the Siting of Fuel Reprocessing Plants and Related Waste Management Facilities 1. Public health... facilities for the temporary storage of highlevel radioactive wastes, may be located on privately owned...

  17. Uranium (III) precipitation in molten chloride by wet argon sparging

    Energy Technology Data Exchange (ETDEWEB)

    Vigier, Jean-François, E-mail: jean-francois.vigier@ec.europa.eu [CEA, Nuclear Energy Division, Radiochemistry & Processes Department, F-30207 Bagnols sur Cèze (France); Unité de Catalyse et de Chimie du Solide, UCCS UMR CNRS 8181, Univ. Lille Nord de France, ENSCL-USTL, B.P. 90108, 59652 Villeneuve d' Ascq Cedex (France); Laplace, Annabelle [CEA, Nuclear Energy Division, Radiochemistry & Processes Department, F-30207 Bagnols sur Cèze (France); Renard, Catherine [Unité de Catalyse et de Chimie du Solide, UCCS UMR CNRS 8181, Univ. Lille Nord de France, ENSCL-USTL, B.P. 90108, 59652 Villeneuve d' Ascq Cedex (France); Miguirditchian, Manuel [CEA, Nuclear Energy Division, Radiochemistry & Processes Department, F-30207 Bagnols sur Cèze (France); Abraham, Francis [Unité de Catalyse et de Chimie du Solide, UCCS UMR CNRS 8181, Univ. Lille Nord de France, ENSCL-USTL, B.P. 90108, 59652 Villeneuve d' Ascq Cedex (France)

    2016-06-15

    In the context of pyrochemical processes for nuclear fuel treatment, the precipitation of uranium (III) in molten salt LiCl-CaCl{sub 2} (30–70 mol%) at 705 °C is studied. First, this molten chloride is characterized with the determination of the water dissociation constant. With a value of 10{sup −4.0}, the salt has oxoacid properties. Then, the uranium (III) precipitation using wet argon sparging is studied. The salt is prepared using UCl{sub 3} precursor. At the end of the precipitation, the salt is totally free of solubilized uranium. The main part is converted into UO{sub 2} powder but some uranium is lost during the process due to the volatility of uranium chloride. The main impurity of the resulting powder is calcium. The consequences of oxidative and reductive conditions on precipitation are studied. Finally, coprecipitation of uranium (III) and neodymium (III) is studied, showing a higher sensitivity of uranium (III) than neodymium (III) to precipitation. - Highlights: • Precipitation of Uranium (III) is quantitative in molten salt LiCl-CaCl{sub 2} (30–70 mol%). • The salt is oxoacid with a water dissociation constant of 10{sup −4.0} at 705 °C. • Volatility of uranium chloride is strongly reduced in reductive conditions. • Coprecipitation of U(III) and Nd(III) leads to a consecutive precipitation of the two elements.

  18. Comparison of thermal, radical and chemical effects of EGR gases using availability analysis in dual-fuel engines at part loads

    International Nuclear Information System (INIS)

    Hosseinzadeh, A.; Khoshbakhti Saray, R.; Seyed Mahmoudi, S.M.

    2010-01-01

    Dual-fuel engines at part load inevitably suffer from lower thermal efficiency and higher emission of carbon monoxide and unburned fuel. A quasi-two-zone combustion model has been developed for studying the second-law analysis of a dual-fuel (diesel-gas) engine operating under part-load conditions. The model is composed of two divisions: a single-zone combustion model with chemical kinetics for combustion of natural gas fuel and a subsidiary zone for combustion of pilot fuel. In the latter zone, the pilot fuel is considered as a heat source derived from two superposed Wiebe's combustion functions to account for contribution of pilot fuel in ignition of gaseous fuel and the rest of the total released energy. This quasi-two-zone combustion model is able to establish the development of combustion process with time and associated important operating parameters, such as pressure, temperature, heat release rate (HRR) and species concentration. The present work is an attempt to investigate the combustion phenomenon from second-law point of view at part load and using exhaust gas recirculation (EGR) to improve the aforementioned problems. Therefore, the availability analysis is applied to the engine from inlet valve closing (IVC) until exhaust valve opening (EVO). Various availability components are identified and calculated separately with crank position. In this paper, the various availability components are identified and calculated separately with crank position. Then the different cases of EGR (chemical, radical and thermal cases) are applied to the availability analysis in dual-fuel engines at part loads. It is found that the chemical case of EGR has negative effect and in this case the unburned chemical availability is increased and the work availability decreases in comparison with baseline engine (without EGR). While the thermal and radical cases have positive effects on the availability terms especially on the unburned chemical availability and work availability

  19. Thorium utilization in a small long-life HTR. Part II: Seed-and-blanket fuel blocks

    Energy Technology Data Exchange (ETDEWEB)

    Ding, Ming, E-mail: dingming@hrbeu.edu.cn [Delft University of Technology, Reactor Institute Delft, Mekelweg 15, 2629 JB Delft (Netherlands); Harbin Engineering University, Nantong Street 145, 150001 Harbin (China); Kloosterman, Jan Leen [Delft University of Technology, Reactor Institute Delft, Mekelweg 15, 2629 JB Delft (Netherlands)

    2014-02-15

    Highlights: • Seed-and-blanket (S and B) fuel blocks are proposed for a small block-type HTR. • S and B fuel blocks consist of a seed region (UO{sub 2}) and a blanket region (ThO{sub 2}). • The neutronic performance of S and B fuel blocks are analyzed using SCALE 6. • Three S and B fuel blocks with a reactivity swing of 0.1 Δk are recommended. • S and B fuel blocks are compared with thorium MOX fuel blocks. - Abstract: In order to utilize thorium in high temperature gas-cooled reactors (HTRs), the concept of seed-and-blanket (S and B) fuel block is introduced into the U-Battery, which is a long-life block-type HTR with a thermal power of 20 MWth. A S and B fuel block consists of a seed region with uranium in the center, and a blanket region with thorium. The neutronic performance, such as the multiplication factor, conversion ratio and reactivity swing, of a typical S and B fuel block was investigated by SCALE 6.0 by parametric analysis of the composition parameters and geometric parameters of the fuel block for the U-Battery application. Since the purpose of U-235 in the S and B fuel block is to ignite the fission reactions in the fuel block, 20% enriched uranium is recommended for the S and B fuel block. When the ratio of the number of carbon to heavy metal atoms changes with the geometric parameters of the fuel block in the range of 200–250, the reactivity swing reaches very small values. Furthermore, for a reactivity swing of 0.1 Δk during 10 effective full power years, three configurations with 36, 54 and 78 UO{sub 2} fuel rods are recommended for the application of the U-Battery. The comparison analysis of the S and B fuel block with the Th/U MOX fuel block shows that the former has a longer lifetime and a lower reactivity swing.

  20. Fuel gases

    International Nuclear Information System (INIS)

    Anon.

    1996-01-01

    This paper gives a brief presentation of the context, perspectives of production, specificities, and the conditions required for the development of NGV (Natural Gas for Vehicle) and LPG-f (Liquefied Petroleum Gas fuel) alternative fuels. After an historical presentation of 80 years of LPG evolution in vehicle fuels, a first part describes the economical and environmental advantages of gaseous alternative fuels (cleaner combustion, longer engines life, reduced noise pollution, greater natural gas reserves, lower political-economical petroleum dependence..). The second part gives a comparative cost and environmental evaluation between the available alternative fuels: bio-fuels, electric power and fuel gases, taking into account the processes and constraints involved in the production of these fuels. (J.S.)

  1. Reflecting Equity and Diversity. Part I: Guidelines and Procedure for Evaluating Bias in Instructional Materials. Part II: Bias Awareness Training Worksheets. Part III: Bias Awareness and Procedure Training Course.

    Science.gov (United States)

    Bebermeyer, Jim; Edmond, Mary, Ed.

    Reflecting a need to prepare students for working in diverse organizations, this document was developed to increase school officials' awareness of bias in instructional materials and help them select bias-free materials. A number of the examples illustrate situations dealing with diversity in the workplace. The guide is divided into three parts:…

  2. Scientific issues in fuel behaviour

    International Nuclear Information System (INIS)

    1995-01-01

    The current limits on discharge burnup in today's nuclear power stations have proven the fuel to be very reliable in its performance, with a negligibly small rate of failure. However, for reasons of economy, there are moves to increase the fuel enrichment in order to extend both the cycle time and the discharge burnup. But, longer periods of irradiation cause increased microstructural changes in the fuel and cladding, implying a larger degradation of physical and mechanical properties. This degradation may well limit the plant life, hence the NSC concluded that it is of importance to develop a predictive capability of fuel behaviour at extended burnup. This can only be achieved through an improved understanding of the basic underlying phenomena of fuel behaviour. The Task Force on Scientific Issues Related to Fuel Behaviour of the NEA Nuclear Science Committee has identified the most important scientific issues on the subject and has assigned priorities. Modelling aspects are listed in Appendix A and discussed in Part II. In addition, quality assurance process for performing and evaluating new integral experiments is considered of special importance. Main activities on fuel behaviour modelling, as carried out in OECD Member countries and international organisations, are listed in Part III. The aim is to identify common interests, to establish current coverage of selected issues, and to avoid any duplication of efforts between international agencies. (author). refs., figs., tabs

  3. Comparison Between Conventional Design and Cathode Gas Recirculation Design of a Direct-Syngas Solid Oxide Fuel Cell–Gas Turbine Hybrid Systems Part I: Design Performance

    Directory of Open Access Journals (Sweden)

    Vahid Azami

    2017-06-01

    Keywords: Solid oxide fuel cell, Gas turbine, Cathode gas recirculation, Exergy. Article History: Received Feb 23rd 2017; Received in revised form May 26th 2017; Accepted June 1st 2017; Available online How to Cite This Article: Azami, V, and Yari, M. (2017 Comparison between conventional design and cathode gas recirculation design of a direct-syngas solid oxide fuel cell–gas turbine hybrid systems part I: Design performance. International Journal of Renewable Energy Develeopment, 6(2, 127-136. https://doi.org/10.14710/ijred.6.2.127-136

  4. Nuclear fuel technology - Determination of uranium in solutions, uranium hexafluoride and solids - Part 2: Iron(II) reduction/cerium(IV) oxidation titrimetric method

    International Nuclear Information System (INIS)

    2004-01-01

    This first edition of ISO 7097-1 together with ISO 7097-2:2004 cancels and replaces ISO 7097:1983, which has been technically revised, and ISO 9989:1996. ISO 7097 consists of the following parts, under the general title Nuclear fuel technology - Determination of uranium in solutions, uranium hexafluoride and solids: Part 1: Iron(II) reduction/potassium dichromate oxidation titrimetric method; Part 2: Iron(II) reduction/cerium(IV) oxidation titrimetric method. This part 2. of ISO 7097 describes procedures for determination of uranium in solutions, uranium hexafluoride and solids. The procedures described in the two independent parts of this International Standard are similar: this part uses a titration with cerium(IV) and ISO 7097-1 uses a titration with potassium dichromate

  5. Nuclear fuel technology - Determination of uranium in solutions, uranium hexafluoride and solids - Part 1: Iron(II) reduction/potassium dichromate oxidation titrimetric method

    International Nuclear Information System (INIS)

    2004-01-01

    This first edition of ISO 7097-1 together with ISO 7097-2:2004 cancels and replaces ISO 7097:1983, which has been technically revised, and ISO 9989:1996. ISO 7097 consists of the following parts, under the general title Nuclear fuel technology - Determination of uranium in solutions, uranium hexafluoride and solids: Part 1: Iron(II) reduction/potassium dichromate oxidation titrimetric method; Part 2: Iron(II) reduction/cerium(IV) oxidation titrimetric method. This part 1. of ISO 7097 describes procedures for the determination of uranium in solutions, uranium hexafluoride and solids. The procedures described in the two independent parts of this International Standard are similar: this part uses a titration with potassium dichromate and ISO 7097-2 uses a titration with cerium(IV)

  6. Nuclear power fuel cycle

    International Nuclear Information System (INIS)

    Havelka, S.; Jakesova, L.

    1982-01-01

    Economic problems are discussed of the fuel cycle (cost of the individual parts of the fuel cycle and the share of the fuel cycle in the price of 1 kWh), the technological problems of the fuel cycle (uranium ore mining and processing, uranium isotope enrichment, the manufacture of fuel elements, the building of long-term storage sites for spent fuel, spent fuel reprocessing, liquid and gaseous waste processing), and the ecologic aspects of the fuel cycle. (H.S.)

  7. Final Report on Utilization of TRU TRISO Fuel as Applied to HTR Systems Part I: Pebble Bed Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Brian Boer; Abderrafi M. Ougouag

    2011-03-01

    The Deep-Burn (DB) concept [ ] focuses on the destruction of transuranic nuclides from used light water reactor (LWR) fuel. These transuranic nuclides are incorporated into tri-isotopic (TRISO) coated fuel particles and used in gas-cooled reactors with the aim of a fractional fuel burnup of 60 to 70% in fissions per initial metal atom (FIMA). This high performance is expected through the use of multiple recirculation passes of the fuel in pebble form without any physical or chemical changes between passes. In particular, the concept does not call for reprocessing of the fuel between passes. In principle, the DB pebble bed concept employs the same reactor designs as the presently envisioned low-enriched uranium core designs, such as the 400 MWth Pebble Bed Modular Reactor (PBMR-400) [ ]. Although it has been shown in the previous Fiscal Year (FY) (2009) that a PuO2 fueled pebble bed reactor concept is viable, achieving a high fuel burnup while remaining within safety-imposed prescribed operational limits for fuel temperature, power peaking, and temperature reactivity feedback coefficients for the entire temperature range, is challenging. The presence of the isotopes 239Pu, 240Pu, and 241Pu that have resonances in the thermal energy range significantly modifies the neutron thermal energy spectrum as compared to a standard, UO2-fueled core. Therefore, the DB pebble bed core exhibits a relatively hard neutron energy spectrum. However, regions within the pebble bed that are near the graphite reflectors experience a locally softer spectrum. This can lead to power and temperature peaking in these regions. Furthermore, a shift of the thermal energy spectrum with increasing temperature can lead to increased absorption in the resonances of the fissile Pu isotopes. This can lead to a positive temperature reactivity coefficient for the graphite moderator under certain operating conditions. Regarding the coated particle performance, the FY 2009 investigations showed that no

  8. Final Report on Utilization of TRU TRISO Fuel as Applied to HTR Systems Part I: Pebble Bed Reactors

    International Nuclear Information System (INIS)

    Boer, Brian; Ougouag, Abderrafi M.

    2011-01-01

    The Deep-Burn (DB) concept focuses on the destruction of transuranic nuclides from used light water reactor (LWR) fuel. These transuranic nuclides are incorporated into tri-isotopic (TRISO) coated fuel particles and used in gas-cooled reactors with the aim of a fractional fuel burnup of 60 to 70% in fissions per initial metal atom (FIMA). This high performance is expected through the use of multiple recirculation passes of the fuel in pebble form without any physical or chemical changes between passes. In particular, the concept does not call for reprocessing of the fuel between passes. In principle, the DB pebble bed concept employs the same reactor designs as the presently envisioned low-enriched uranium core designs, such as the 400 MWth Pebble Bed Modular Reactor (PBMR-400). Although it has been shown in the previous Fiscal Year (FY) (2009) that a PuO2 fueled pebble bed reactor concept is viable, achieving a high fuel burnup while remaining within safety-imposed prescribed operational limits for fuel temperature, power peaking, and temperature reactivity feedback coefficients for the entire temperature range, is challenging. The presence of the isotopes 239Pu, 240Pu, and 241Pu that have resonances in the thermal energy range significantly modifies the neutron thermal energy spectrum as compared to a standard, UO2-fueled core. Therefore, the DB pebble bed core exhibits a relatively hard neutron energy spectrum. However, regions within the pebble bed that are near the graphite reflectors experience a locally softer spectrum. This can lead to power and temperature peaking in these regions. Furthermore, a shift of the thermal energy spectrum with increasing temperature can lead to increased absorption in the resonances of the fissile Pu isotopes. This can lead to a positive temperature reactivity coefficient for the graphite moderator under certain operating conditions. Regarding the coated particle performance, the FY 2009 investigations showed that no significant

  9. Part I. Fuel-motion diagnostics in support of fast-reactor safety experiments. Part II. Fission product detection system in support of fast reactor safety experiments

    International Nuclear Information System (INIS)

    Devolpi, A.; Doerner, R.C.; Fink, C.L.; Regis, J.P.; Rhodes, E.A.; Stanford, G.S.; Braid, T.H.; Boyar, R.E.

    1986-05-01

    In all destructive fast-reactor safety experiments at TREAT, fuel motion and cladding failure have been monitored by the fast-neutron/gamma-ray hodoscope, providing experimental results that are directly applicable to design, modeling, and validation in fast-reactor safety. Hodoscope contributions to the safety program can be considered to fall into several groupings: pre-failure fuel motion, cladding failure, post-failure fuel motion, steel blockages, pretest and posttest radiography, axial-power-profile variations, and power-coupling monitoring. High-quality results in fuel motion have been achieved, and motion sequences have been reconstructed in qualitative and quantitative visual forms. A collimated detection system has been used to observe fission products in the upper regions of a test loop in the TREAT reactor. Particular regions of the loop are targeted through any of five channels in a rotatable assembly in a horizontal hole through the biological shield. A well-type neutron detector, optimized for delayed neutrons, and two GeLi gamma ray spectrometers have been used in several experiments. Data are presented showing a time history of the transport of Dn emitters, of gamma spectra identifying volatile fission products deposited as aerosols, and of fission gas isotopes released from the coolant

  10. Nuclear proliferation and civilian nuclear power. Report of the Nonproliferation Alternative Systems Assessment Program. Volume III. Resources and fuel cycle facilities

    International Nuclear Information System (INIS)

    1980-06-01

    The ability of uranium supply and the rest of the nuclear fuel cycle to meet the demand for nuclear power is an important consideration in future domestic and international planning. Accordingly, the purpose of this assessment is to evaluate the adequacy of potential supply for various nuclear resources and fuel cycle facilities in the United States and in the world outside centrally planned economy areas (WOCA). Although major emphasis was placed on uranium supply and demand, material resources (thorium and heavy water) and facility resources (separative work, spent fuel storage, and reprocessing) were also considered

  11. Release to the Gas Phase of Inorganic Elements during Wood Combustion. Part 2: Influence of Fuel Composition

    DEFF Research Database (Denmark)

    van Lith, Simone Cornelia; Jensen, Peter Arendt; Frandsen, Flemming

    2008-01-01

    temperatures in the range of 500–1150 °C in a laboratory-scale tube reactor and by performing mass balance calculations based on the weight measurements and chemical analyses of the wood fuels and the residual ash samples. Four wood fuels with different ash contents and inorganic compositions were investigated...... of the alkali metals K and Na was, however, strongly dependent on both the temperature and the fuel composition under the investigated conditions. The release of the heavy metals Zn and Pb started around 500 °C and increased sharply to more than 85% at 850 °C in the case of spruce, beech, and bark...

  12. Development of multi-component diesel surrogate fuel models – Part I: Validation of reduced mechanisms of diesel fuel constituents in 0-D kinetic simulations

    DEFF Research Database (Denmark)

    Poon, Hiew Mun; Pang, Kar Mun; Ng, Hoon Kiat

    2016-01-01

    In the present work, development and validation of reduced chemical kinetic mechanisms for several different hydrocarbons are performed. These hydrocarbons are potential representative for practical diesel fuel constituents. n-Hexadecane (HXN), 2,2,4,4,6,8,8-heptamethylnonane (HMN), cyclohexane...... (CHX) and toluene are selected to represent straight-alkane, branched-alkane, cyclo-alkane and aromatic compounds in the diesel fuel. A five-stage chemical kinetic mechanism reduction scheme formulated in the previous work is applied to develop the reduced HMN and CHX models based on their respective...... detailed mechanisms. Alongside with the development of the reduced CHX model, a skeletal toluene sub-mechanism is constructed since the elementary reactions for toluene are subset of the detailed CHX mechanism. The final reduced HMN mechanism comprises 89 species with 319 elementary reactions, while...

  13. Environmental Radiation Protection Standards for Management and Disposal of Spent Nuclear Fuel and Transuranic Radioactive Wastes (40 CFR Part 191)

    Science.gov (United States)

    This regulation sets environmental standards for public protection from the management and disposal of spent nuclear fuel, high-level wastes and wastes that contain elements with atomic numbers higher than uranium (transuranic wastes).

  14. The history, genotoxicity, and carcinogenicity of carbon-based fuels and their emissions. Part 3: diesel and gasoline.

    Science.gov (United States)

    Claxton, Larry D

    2015-01-01

    Within this review the genotoxicity of diesel and gasoline fuels and emissions is placed in an historical context. New technologies have changed the composition of transportation methods considerably, reducing emissions of many of the components of health concern. The similarity of modern diesel and gasoline fuels and emissions to other carbonaceous fuels and emissions is striking. Recently an International Agency for Research on Cancer (IARC) Working Group concluded that there was sufficient evidence in humans for the carcinogenicity of diesel exhaust (Group 1). In addition, the Working Group found that diesel exhaust has "a positive association (limited evidence) with an increased risk of bladder cancer." Like most other carbonaceous fuel emissions, diesel and gasoline exhausts contain toxic levels of respirable particles (PM gasoline emissions has declined in certain regions over time because of changes in engine design, the development of better aftertreatment devices (e.g., catalysts), increased fuel economy, changes in the fuels and additives used, and greater regulation. Additional research and better exposure assessments are needed so that decision makers and the public can decide to what extent diesel and gasoline engines should be replaced. Copyright © 2014 Elsevier B.V. All rights reserved.

  15. Impact of alternative fuels on emissions characteristics of a gas turbine engine - part 2: volatile and semivolatile particulate matter emissions.

    Science.gov (United States)

    Williams, Paul I; Allan, James D; Lobo, Prem; Coe, Hugh; Christie, Simon; Wilson, Christopher; Hagen, Donald; Whitefield, Philip; Raper, David; Rye, Lucas

    2012-10-02

    The work characterizes the changes in volatile and semivolatile PM emissions from a gas turbine engine resulting from burning alternative fuels, specifically gas-to-liquid (GTL), coal-to-liquid (CTL), a blend of Jet A-1 and GTL, biodiesel, and diesel, to the standard Jet A-1. The data presented here, compares the mass spectral fingerprints of the different fuels as measured by the Aerodyne high resolution time-of-flight aerosol mass spectrometer. There were three sample points, two at the exhaust exit plane with dilution added at different locations and another probe located 10 m downstream. For emissions measured at the downstream probe when the engine was operating at high power, all fuels produced chemically similar organic PM, dominated by C(x)H(y) fragments, suggesting the presence of long chain alkanes. The second largest contribution came from C(x)H(y)O(z) fragments, possibly from carbonyls or alcohols. For the nondiesel fuels, the highest loadings of organic PM were from the downstream probe at high power. Conversely, the diesel based fuels produced more organic material at low power from one of the exit plane probes. Differences in the composition of the PM for certain fuels were observed as the engine power decreased to idle and the measurements were made closer to the exit plane.

  16. Theoretical analysis of nuclear reactors (Phase II), I-V, Part III, Reactor poisoning; Razrada metoda teorijske analize nuklearnih reaktora (II faza) I-V, III Deo, Zatrovanje reaktora, II faza

    Energy Technology Data Exchange (ETDEWEB)

    Pop-Jordanov, J [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1962-10-15

    This phase is dealing with influence of all the fission products except Xe{sup 135} on the reactivity of a reactor, usually named as reactor poisoning. The first part of the report is a review of methods for calculation of reactor poisoning. The second part shows the most frequently used method for calculation of cross sections and yields of pseudo products (for thermal neutrons). The system of equations was adopted dependent on the conditions of the available computer system. It is described in part three. Detailed method for their application is described in part four and results obtained are presented in part five.

  17. State of the art of the welding method for sealing spent nuclear fuel canister made of copper. Part 2 - EBW

    International Nuclear Information System (INIS)

    Salonen, T.

    2014-05-01

    This report consist the results of the development of the electron beam welding (EBW) method for sealing spent nuclear fuel (SNF) disposal canister. This report has been used as background material for selection of the sealing method for the SNF canister. Report contains the state of the art knowledge of the EBW method and research and development (R and D) results done by Posiva. Relevant R and D results of EB-welds done by SKB are also reviewed in this report. Requirements set for the welding and weld are present. These requirements are based on the long term safety and also some part of requirements are set by other processes like non-destructive testing (NDT) and manufacturing processes of components. Initial state of the weld is described in this report. Initial state has significant effect on the long term safety issues like corrosion resistance and creep ductility. Also short and long term mechanical properties as well as corrosion properties are described. Microstructure and residual stresses of the weld is represented in this report. Report consists also imperfections of the weld and statistical analysis of the evaluation of the probability of the largest defect size on the weld. Results of corrosion and creep tests of EB-welds are reviewed in this report. EBW process and machine are described. Preliminary designing of the EBW-machine has been done including component handling equipments. Preliminary welding procedure specification (pWPS) has drawn up and qualification of the personnel is described briefly. In-line process and quality control system including seam tracking system is implemented in modern EBW machine. Also NDT methods for inspection of the weld are described in this report. Concerning the results from the research and development work it can be concluded that EB welding method is suitable method for sealing SNF canister. Weld material fulfils requirements set by the long term safety. The welding system is robust and reliable and it is based

  18. Final Report on Utilization of TRU TRISO Fuel as Applied to HTR Systems Part II: Prismatic Reactor Cross Section Generation

    Energy Technology Data Exchange (ETDEWEB)

    Vincent Descotes

    2011-03-01

    The deep-burn prismatic high temperature reactor is made up of an annular core loaded with transuranic isotopes and surrounded in the center and in the periphery by reflector blocks in graphite. This disposition creates challenges for the neutronics compared to usual light water reactor calculation schemes. The longer mean free path of neutrons in graphite affects the neutron spectrum deep inside the blocks located next to the reflector. The neutron thermalisation in the graphite leads to two characteristic fission peaks at the inner and outer interfaces as a result of the increased thermal flux seen in those assemblies. Spectral changes are seen at least on half of the fuel blocks adjacent to the reflector. This spectral effect of the reflector may prevent us from successfully using the two step scheme -lattice then core calculation- typically used for light water reactors. We have been studying the core without control mechanisms to provide input for the development of a complete calculation scheme. To correct the spectrum at the lattice level, we have tried to generate cross-sections from supercell calculations at the lattice level, thus taking into account part of the graphite surrounding the blocks of interest for generating the homogenised cross-sections for the full-core calculation. This one has been done with 2 to 295 groups to assess if increasing the number of groups leads to more accurate results. A comparison with a classical single block model has been done. Both paths were compared to a reference calculation done with MCNP. It is concluded that the agreement with MCNP is better with supercells, but that the single block model remains quite close if enough groups are kept for the core calculation. 26 groups seems to be a good compromise between time and accu- racy. However, some trials with depletion have shown huge variations of the isotopic composition across a block next to the reflector. It may imply that at least an in- core depletion for the

  19. Main routes for the thermo-conversion of biomass into fuels and chemicals. Part 1: Pyrolysis systems

    International Nuclear Information System (INIS)

    Balat, Mustafa; Balat, Mehmet; Kirtay, Elif; Balat, Havva

    2009-01-01

    Since the energy crises of the 1970s, many countries have become interest in biomass as a fuel source to expand the development of domestic and renewable energy sources and reduce the environmental impacts of energy production. Biomass is used to meet a variety of energy needs, including generating electricity, heating homes, fueling vehicles and providing process heat for industrial facilities. The methods available for energy production from biomass can be divided into two main categories: thermo-chemical and biological conversion routes. There are several thermo-chemical routes for biomass-based energy production, such as direct combustion, liquefaction, pyrolysis, supercritical water extraction, gasification, air-steam gasification and so on. The pyrolysis is thermal degradation of biomass by heat in the absence of oxygen, which results in the production of charcoal (solid), bio-oil (liquid), and fuel gas products. Pyrolysis liquid is referred to in the literature by terms such as pyrolysis oil, bio-oil, bio-crude oil, bio-fuel oil, wood liquid, wood oil, liquid smoke, wood distillates, pyroligneous tar, and pyroligneous acid. Bio-oil can be used as a fuel in boilers, diesel engines or gas turbines for heat and electricity generation.

  20. Fuel temperature influence on the performance of a last generation common-rail diesel ballistic injector. Part II: 1D model development, validation and analysis

    International Nuclear Information System (INIS)

    Payri, R.; Salvador, F.J.; Carreres, M.; De la Morena, J.

    2016-01-01

    Highlights: • A 1D model of a solenoid common-rail ballistic injector is implemented in AMESim. • A detailed dimensional and a hydraulic characterization lead to a fair validation. • Fuel temperature influence on injector dynamics is assessed through 1D simulations. • Temperature impacts through changes in inlet orifice regime and viscous friction. • Cold fuel temperature leads to a slower injection opening due to high viscosity. - Abstract: A one-dimensional model of a solenoid-driven common-rail diesel injector has been developed in order to study the influence of fuel temperature on the injection process. The model has been implemented after a thorough characterization of the injector, both from the dimensional and the hydraulic point of view. In this sense, experimental tools for the determination of the geometry of the injector lines and orifices have been described in the paper, together with the hydraulic setup introduced to characterize the flow behaviour through the calibrated orifices. An extensive validation of the model has been performed by comparing the modelled mass flow rate against the experimental results introduced in the first part of the paper, which were performed for different engine-like operating conditions involving a wide range of fuel temperatures, injection pressures and energizing times. In that first part of the study, an important influence of the fuel temperature was reported, especially in terms of the dynamic behaviour of the injector, due to its ballistic nature. The results from the model have allowed to explain and further extend the findings of the experimental study by analyzing key features of the injector dynamics, such as the pressure drop established in the control volume due to the control orifices performance or the forces due to viscous friction, also assessing their influence on the needle lift laws.

  1. The removal of toxic metals from liquid effluents by ion exchange resins. Part IV: Chromium(III)/H+ /Lewatit SP112

    International Nuclear Information System (INIS)

    Alguacil, F.J.

    2017-01-01

    This investigation presented results on the removal of chromium(III), from aqueous solution in the 0-5 pH range, using Lewatit SP112 cationic exchange resin. Several aspects affecting the ion exchange process were evaluated, including: the influence of the stirring speed, temperature, pH of the solution, resin dosage and aqueous ionic strength. The selectivity of the system was tested against the presence of other metals in the aqueous solution, whereas the removal of chromium(III) from solutions was compared with results obtained using multiwalled carbon nanotubes as adsorbents. From the batch experimental data, best fit of the results is obtained with the Langmuir model, whereas the ion exchange process is best explained by the pseudo-second order model, moreover, experimental data responded well to the film-diffusion controlled model. Elution of the chromium(III) loaded into the resin is well accomplished by the use of sodium hydroxide solutions. [es

  2. Extracción de cromo con disolventes orgánicos. III parte. Aplicación al tratamiento de residuos polimetálicos industriales

    OpenAIRE

    de Juan, D.; Meseguer, V.; Lozano, L. J.

    1998-01-01

    The use of Primene 81R as extraction agent of chromium present in solid wastes containing nickel, iron and copper has been studied. The waste was leached with a sulphuric acid solution up to pH 3 and oxidation of Cr(III) to Cr(VI) with Caro acid was also studied. Because of the negative result of oxidation, the treatment was applied on Cr(III) directly. Extraction/scrubbing/stripping process was studied in the leach. The composition of organic phase used in the extraction step was 10 % v/v Pr...

  3. Fission products and nuclear fuel behaviour under severe accident conditions part 1: Main lessons learnt from the first VERDON test

    Science.gov (United States)

    Pontillon, Y.; Geiger, E.; Le Gall, C.; Bernard, S.; Gallais-During, A.; Malgouyres, P. P.; Hanus, E.; Ducros, G.

    2017-11-01

    This paper describes the first VERDON test performed at the end of September 2011 with special emphasis on the behaviour of fission products (FP) and actinides during the accidental sequence itself. Two other papers discuss in detail the post-test examination results (SEM, EPMA and SIMS) of the VERDON-1 sample. The first VERDON test was devoted to studying UO2 fuel behaviour and fission product releases under reducing conditions at very high temperature (∼2883 K), which was able to confirm the very good performance of the VERDON loop. The fuel sample did not lose its integrity during this test. According to the FP behaviour measured by the online gamma station (fuel sight), the general classification of the FP in relation to their released fraction is very accurate, and the burn-up effect on the release rate is clearly highlighted.

  4. MCTP, a code for the thermo-mechanical analysis of a fuel rod of BWR type reactors (Neutron part)

    International Nuclear Information System (INIS)

    Hernandez L, H.; Ortiz V, J.

    2003-01-01

    In the National Institute of Nuclear Research of Mexico a code for the thermo-mechanical analysis of the fuel rods of the BWR type reactors of the Nucleo electric Central of Laguna Verde is developed. The code solves the diffusion equation in cylindrical coordinates with several energy groups. The code, likewise, calculates the temperature distribution and power distribution in those fuel rods. The code is denominated Multi groups With Temperatures and Power (MCTP). In the code, the energy with which the fission neutrons are emitted it is divided in six groups. They are also considered the produced perturbations by the changes in the temperatures of the materials that constitute the fuel rods, the content of fission products, the uranium consumption and in its case the gadolinium, as well as the plutonium production. In this work there are present preliminary results obtained with the code, using data of operation of the Nucleo electric Central of Laguna Verde. (Author)

  5. Flow sheet development for the dissolution of unirradiated Mark 42 fuel tubes in F-Canyon, Part II

    International Nuclear Information System (INIS)

    Murray, A.M.

    1999-01-01

    Two dissolution flow sheets were tested for the desorption of unirradiated Mark 42 fuel tubes. Both the aluminum (from the can, cladding, and fuel core) and the plutonium oxide (PuO 2 ) are dissolved simultaneously, i.e., a co-dissolution flow sheet. In the first series of tests, 0.15 and 0.20 molar (M) potassium fluoride (KF) solutions were used and the dissolution extended over several days. In the other series of tests, solutions with higher concentrations of fluoride (0.25 to 0.30 M) were used. Calcium fluoride (CaF 2 ) was used in those tests as the fluoride source

  6. RA Reactor operation and maintenance (I-IX), Part III, Task 3.08/04-02 Refurbishment of the electrical equipment; Pogon i odrzavanje reaktora RA (I-IX), III Deo, Zadatak 3.08/04-02 Remont elektro opreme

    Energy Technology Data Exchange (ETDEWEB)

    Zecevic, V; Nikolic, M; Poznanovic, B; Rajic, M [Institute of Nuclear Sciences Boris Kidric, Reaktor RA, Vinca, Beograd (Serbia and Montenegro)

    1963-12-15

    This volume contains detailed action plan for repair of electrical equipment of the RA reactor, the list of electrical equipment parts which were either repaired or exchanged for improvement of their performance. Detailed work describing the repair and maintenance work done of the listed equipment is part of this report. Equipment related to dosimetry and control systems are included as well.

  7. Nuclear fuels

    International Nuclear Information System (INIS)

    Beauvy, M.; Berthoud, G.; Defranceschi, M.; Ducros, G.; Guerin, Y.; Limoge, Y.; Madic, Ch.; Santarini, G.; Seiler, J.M.; Sollogoub, P.; Vernaz, E.; Guillet, J.L.; Ballagny, A.; Bechade, J.L.; Bonin, B.; Brachet, J.Ch.; Delpech, M.; Dubois, S.; Ferry, C.; Freyss, M.; Gilbon, D.; Grouiller, J.P.; Iracane, D.; Lansiart, S.; Lemoine, P.; Lenain, R.; Marsault, Ph.; Michel, B.; Noirot, J.; Parrat, D.; Pelletier, M.; Perrais, Ch.; Phelip, M.; Pillon, S.; Poinssot, Ch.; Vallory, J.; Valot, C.; Pradel, Ph.; Bonin, B.; Bouquin, B.; Dozol, M.; Lecomte, M.; Vallee, A.; Bazile, F.; Parisot, J.F.; Finot, P.; Roberts, J.F.

    2009-01-01

    Fuel is one of the essential components in a reactor. It is within that fuel that nuclear reactions take place, i.e. fission of heavy atoms, uranium and plutonium. Fuel is at the core of the reactor, but equally at the core of the nuclear system as a whole. Fuel design and properties influence reactor behavior, performance, and safety. Even though it only accounts for a small part of the cost per kilowatt-hour of power provided by current nuclear power plants, good utilization of fuel is a major economic issue. Major advances have yet to be achieved, to ensure longer in-reactor dwell-time, thus enabling fuel to yield more energy; and improve ruggedness. Aside from economics, and safety, such strategic issues as use of plutonium, conservation of resources, and nuclear waste management have to be addressed, and true technological challenges arise. This Monograph surveys current knowledge regarding in-reactor behavior, operating limits, and avenues for R and D. It also provides illustrations of ongoing research work, setting out a few noteworthy results recently achieved. Content: 1 - Introduction; 2 - Water reactor fuel: What are the features of water reactor fuel? 9 (What is the purpose of a nuclear fuel?, Ceramic fuel, Fuel rods, PWR fuel assemblies, BWR fuel assemblies); Fabrication of water reactor fuels (Fabrication of UO 2 pellets, Fabrication of MOX (mixed uranium-plutonium oxide) pellets, Fabrication of claddings); In-reactor behavior of UO 2 and MOX fuels (Irradiation conditions during nominal operation, Heat generation, and removal, The processes involved at the start of irradiation, Fission gas behavior, Microstructural changes); Water reactor fuel behavior in loss of tightness conditions (Cladding, the first containment barrier, Causes of failure, Consequences of a failure); Microscopic morphology of fuel ceramic and its evolution under irradiation; Migration and localization of fission products in UOX and MOX matrices (The ceramic under irradiation

  8. Nuclear fuels

    Energy Technology Data Exchange (ETDEWEB)

    Beauvy, M.; Berthoud, G.; Defranceschi, M.; Ducros, G.; Guerin, Y.; Limoge, Y.; Madic, Ch.; Santarini, G.; Seiler, J.M.; Sollogoub, P.; Vernaz, E.; Guillet, J.L.; Ballagny, A.; Bechade, J.L.; Bonin, B.; Brachet, J.Ch.; Delpech, M.; Dubois, S.; Ferry, C.; Freyss, M.; Gilbon, D.; Grouiller, J.P.; Iracane, D.; Lansiart, S.; Lemoine, P.; Lenain, R.; Marsault, Ph.; Michel, B.; Noirot, J.; Parrat, D.; Pelletier, M.; Perrais, Ch.; Phelip, M.; Pillon, S.; Poinssot, Ch.; Vallory, J.; Valot, C.; Pradel, Ph.; Bonin, B.; Bouquin, B.; Dozol, M.; Lecomte, M.; Vallee, A.; Bazile, F.; Parisot, J.F.; Finot, P.; Roberts, J.F

    2009-07-01

    Fuel is one of the essential components in a reactor. It is within that fuel that nuclear reactions take place, i.e. fission of heavy atoms, uranium and plutonium. Fuel is at the core of the reactor, but equally at the core of the nuclear system as a whole. Fuel design and properties influence reactor behavior, performance, and safety. Even though it only accounts for a small part of the cost per kilowatt-hour of power provided by current nuclear power plants, good utilization of fuel is a major economic issue. Major advances have yet to be achieved, to ensure longer in-reactor dwell-time, thus enabling fuel to yield more energy; and improve ruggedness. Aside from economics, and safety, such strategic issues as use of plutonium, conservation of resources, and nuclear waste management have to be addressed, and true technological challenges arise. This Monograph surveys current knowledge regarding in-reactor behavior, operating limits, and avenues for R and D. It also provides illustrations of ongoing research work, setting out a few noteworthy results recently achieved. Content: 1 - Introduction; 2 - Water reactor fuel: What are the features of water reactor fuel? 9 (What is the purpose of a nuclear fuel?, Ceramic fuel, Fuel rods, PWR fuel assemblies, BWR fuel assemblies); Fabrication of water reactor fuels (Fabrication of UO{sub 2} pellets, Fabrication of MOX (mixed uranium-plutonium oxide) pellets, Fabrication of claddings); In-reactor behavior of UO{sub 2} and MOX fuels (Irradiation conditions during nominal operation, Heat generation, and removal, The processes involved at the start of irradiation, Fission gas behavior, Microstructural changes); Water reactor fuel behavior in loss of tightness conditions (Cladding, the first containment barrier, Causes of failure, Consequences of a failure); Microscopic morphology of fuel ceramic and its evolution under irradiation; Migration and localization of fission products in UOX and MOX matrices (The ceramic under

  9. Numerical analysis of a downsized spark-ignition engine fueled by butanol/gasoline blends at part-load operation

    International Nuclear Information System (INIS)

    Scala, F.; Galloni, E.; Fontana, G.

    2016-01-01

    Highlights: • Bio-fuels will reduce the overall CO_2 emission. • The properties of butanol/gasoline–air mixtures have been determined. • A 1-D model of a SI engine has been calibrated and validated. • The butanol content reduces the combustion duration. • The optimal ignition timing slightly changes. - Abstract: In this paper, the performance of a turbocharged SI engine, firing with butanol/gasoline blends, has been investigated by means of numerical simulations of the engine behavior. When engine fueling is switched from gasoline to alcohol/gasoline mixture, engine control parameters must be adapted. The main necessary modifications in the Electronic Control Unit have been highlighted in the paper. Numerical analyses have been carried out at partial load operation and at two different engine speeds (3000 and 4000 rpm). Several n-butanol/gasoline mixtures, differing for the alcohol contents, have been analyzed. Such engine performances as torque and indicated efficiency have been evaluated. Both these characteristics decrease with the alcohol contents within the mixtures. On the contrary, when the engine is fueled by neat n-butanol, torque and efficiency reach values about 2% higher than those obtained with neat gasoline. Furthermore, the optimal spark timing, for alcohol/gasoline mixture operation, must be retarded (up to 13%) in comparison with the correspondent values of the gasoline operation. In general, engine performance and operation undergo little variations when fuel supplying is switched from gasoline to alcohol/gasoline blends.

  10. Novel electrospun gas diffusion layers for polymer electrolyte membrane fuel cells: Part I. Fabrication, morphological characterization, and in situ performance

    Science.gov (United States)

    Chevalier, S.; Lavielle, N.; Hatton, B. D.; Bazylak, A.

    2017-06-01

    In this first of a series of two papers, we report an in-depth analysis of the impact of the gas diffusion layer (GDL) structure on the polymer electrolyte membrane (PEM) fuel cell performance through the use of custom GDLs fabricated in-house. Hydrophobic electrospun nanofibrous gas diffusion layers (eGDLs) are fabricated with controlled fibre diameter and alignment. The eGDLs are rendered hydrophobic through direct surface functionalization, and this molecular grafting is achieved in the absence of structural alteration. The fibre diameter, chemical composition, and electrical conductivity of the eGDL are characterized, and the impact of eGDL structure on fuel cell performance is analysed. We observe that the eGDL facilitates higher fuel cell power densities compared to a commercial GDL (Toray TGP-H-60) at highly humidified operating conditions. The ohmic resistance of the fuel cell is found to significantly increase with increasing inter-fiber distance. It is also observed that the addition of a hydrophobic treatment enhances membrane hydration, and fibres perpendicularly aligned to the channel direction may enhance the contact area between the catalyst layer and the GDL.

  11. SEU blending project, concept to commercial operation, Part 3: production of powder for demonstration irradiation fuel bundles

    International Nuclear Information System (INIS)

    Ioffe, M.S.; Bhattacharjee, S.; Oliver, A.J.; Ozberk, E.

    2005-01-01

    The processes for production of Slightly Enriched Uranium (SEU) dioxide powder and Blended Dysprosium and Uranium (BDU) oxide powder that were developed at laboratory scale at Cameco Technology Development (CTD), were implemented and further optimized to supply to Zircatec Precision Industries (ZPI) the quantities required for manufacturing twenty six Low Void Reactivity (LVRF) CANFLEX fuel bundles. The production of this new fuel was a challenge for CTD and involved significant amount of work to prepare and review documentation, develop and approve new analytical procedures, and go through numerous internal reviews and audits by Bruce Power, CNSC and third parties independent consultants that verified the process and product quality. The audits were conducted by Quality Assurance specialists as well as by Human Factor Engineering experts with the objective to systematically address the role of human errors in the manufacturing of New Fuel and confirm whether or not a credible basis had been established for preventing human errors. The project team successfully passed through these audits. The project management structure that was established during the SEU and BDU blending process development, which included a cross-functional project team from several departments within Cameco, maintained its functionality when Cameco Technology Development was producing the powder for manufacturing Demonstration Irradiation fuel bundles. Special emphasis was placed on the consistency of operating steps and product quality certification, independent quality surveillance, materials segregation protocol, enhanced safety requirements, and accurate uranium accountability. (author)

  12. Water activities in Forsmark (Part II). The final disposal facility for spent fuel: water activities above ground

    International Nuclear Information System (INIS)

    Werner, Kent; Hamren, Ulrika; Collinder, Per; Ridderstolpe, Peter

    2010-09-01

    The construction of the repository for spent nuclear fuel in Forsmark is associated with a number of measures above ground that constitute water operations according to Chapter 11 in the Swedish Environmental Code. This report, which is an appendix to the Environmental Impact Assessment, describes these water operations, their effects and consequences, and planned measures

  13. Geology of quadrangles H-12, H-13, and parts of I-12 and I-13, (zone III) in northeastern Santander Department, Colombia

    Science.gov (United States)

    Ward, Dwight Edward; Goldsmith, Richard; Cruz, Jaime B.; Restrepo, Hernan A.

    1974-01-01

    A program of geologic mapping and mineral investigation in Colombia was undertaken cooperatively by the Colombian Instituto Nacional de Investigaciones Geologico-Mineras (formerly known as the Inventario Minero Nacional), and the U. S. Geological Survey; by the Government of Colombia and the Agency for International Development, U. S. Department of State. The purpose was to study, and evaluate mineral resources (excluding of petroleum, coal, emeralds, and alluvial gold) of four selected areas, designated Zones I to IV, that total about 70,000 km2. The work in Zone III, in the Cordillera Oriental, was done from 1965 to 1968. The northeast trend of the Cordillera Oriental of Colombia swings abruptly to north-northwest in the area of this report, and divides around the southern end of the Maracaibo Basin. This section of the Cordillera Oriental is referred to as the Santander Massif. Radiometric age determinations indicate that the oldest rocks of the Santander massif are Precambrian and include high-grade gneiss, schist, and migmatite of the Bucaramanga Formation. These rocks were probably part of the Precambrian Guayana Shield. Low- to medium-grade metamorphic rocks of late Precambrian to Ordovician age .include phyllite, schist, metasiltstone, metasandstone, and marble of the Silgara Formation, a geosynclinal series of considerable extent in the Cordillera Oriental and possibly the Cordillera de Merida of Venezuela. Orthogneiss ranging from granite to tonalite is widely distributed in the high- and medium-grade metamorphic rocks of the central core of the massif and probably represents rocks of two ages, Precambrian and Ordovician to Early Devonian. Younger orthogneiss and the Silgara are overlain by Middle Devonian beds of the Floresta Formation which show a generally low but varying degree of metamorphism. Phyllite and argillite are common, and infrequent marble and other calcareous beds are fossiliferous. Except for recrystallization in limestones of !the

  14. Synthesis and structural characterisation of mixed An(IV)-An(III) actinide oxalates used as precursors for dedicated fuel or target

    International Nuclear Information System (INIS)

    Tamain, Christelle; Grandjean, Stephane; Arab Chapelet, Benedicte; Abraham, Francis

    2010-01-01

    Oxalic co-conversion process plays an important role by producing mixed-actinide compounds used as starting materials as they are particularly suitable precursors of actinide oxide solid solutions. In these oxalate compounds, a mixed crystallographic site which accommodates both elements in spite of their different oxidation states has been established. The charge compensation is ensured by monovalent cations present in the acidic solution. This communication reviews the various mixed-actinide oxalates obtained by crystallization from acidic solution. First, crystallographic structures determined by X-ray diffraction from single crystals are described. Then completing data obtained by powder X-ray diffraction are presented on various systems. The different supramolecular arrangements underline the complexity of An(IV)-An(III)/Ln(III) oxalate system and the need to pursue studies on single crystals. (authors)

  15. Theoretical analysis of nuclear reactors (Phase I), I-V, Part IV, Nuclear fuel depletion; Razrada metoda teorijske analize nuklearnih reaktora (I faza) I-V, IV Deo, Promena izotopnog sastava goriva

    Energy Technology Data Exchange (ETDEWEB)

    Pop-Jordanov, J [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1962-07-15

    Nuclear fuel depletion is analyzed in order to estimate the qualitative and quantitative fuel property changes during irradiation and the influence of changes on the reactivity during long-term reactor operation. The changes of fuel properties are described by changes of neutron absorption and fission cross sections. Part one of this report covers the economic significance of fuel burnup and the review of fuel isotopic changes during depletion. Pat two contains the analysis of the U{sup 235} chain, analytical expressions for the concentrations of U{sup 235}, U{sup 236} and Np{sup 237} as a function of burnup. Part three contains the analysis of neutron spectrum influence on the Westcott method for calculating the cross sections. Part four contains the calculation method applied on Calder Hall type reactor. The results were obtained by applying ZUSE-22 R digital computer.

  16. Industrial Fuel Gas Demonstration Plant Program. Conceptual design and evaluation of commercial plant. Volume III. Economic analyses (Deliverable Nos. 15 and 16)

    Energy Technology Data Exchange (ETDEWEB)

    None

    1978-01-01

    This report presents the results of Task I of Phase I in the form of a Conceptual Design and Evaluation of Commercial Plant report. The report is presented in four volumes as follows: I - Executive Summary, II - Commercial Plant Design, III - Economic Analyses, IV - Demonstration Plant Recommendations. Volume III presents the economic analyses for the commercial plant and the supporting data. General cost and financing factors used in the analyses are tabulated. Three financing modes are considered. The product gas cost calculation procedure is identified and appendices present computer inputs and sample computer outputs for the MLGW, Utility, and Industry Base Cases. The results of the base case cost analyses for plant fenceline gas costs are as follows: Municipal Utility, (e.g. MLGW), $3.76/MM Btu; Investor Owned Utility, (25% equity), $4.48/MM Btu; and Investor Case, (100% equity), $5.21/MM Btu. The results of 47 IFG product cost sensitivity cases involving a dozen sensitivity variables are presented. Plant half size, coal cost, plant investment, and return on equity (industrial) are the most important sensitivity variables. Volume III also presents a summary discussion of the socioeconomic impact of the plant and a discussion of possible commercial incentives for development of IFG plants.

  17. Presentation of safety after closure of the repository for spent nuclear fuel. Main report of the project SR-Site. Part I

    International Nuclear Information System (INIS)

    2011-01-01

    The purpose of the safety assessment SR-Site is to investigate whether a safe repository for spent nuclear fuel by KBS-3 type can be constructed at Forsmark in Oesthammar in Sweden. The location of the Forsmark has been selected based on results of several surveys from surface conditions at depth in Forsmark and in Laxemar in Oskarshamn. The choice of location is not justified in SR-Site Report, but in other attachments to SKB's permit applications. SR-Site Report is an important part of SKB's permit applications to construct and operate a repository for spent nuclear fuel at Forsmark in Oesthammar. The purpose of the report in the applications is to show that a repository at Forsmark is safe after closure

  18. Fuel and fission product behaviour in early phases of a severe accident. Part II: Interpretation of the experimental results of the PHEBUS FPT2 test

    Energy Technology Data Exchange (ETDEWEB)

    Dubourg, R. [Institut de Radioprotection et de Sûreté Nucléaire, B.P. 3, 13115 Saint Paul-lez-Durance Cedex (France); Barrachin, M., E-mail: marc.barrachin@irsn.fr [Institut de Radioprotection et de Sûreté Nucléaire, B.P. 3, 13115 Saint Paul-lez-Durance Cedex (France); Ducher, R. [Institut de Radioprotection et de Sûreté Nucléaire, B.P. 3, 13115 Saint Paul-lez-Durance Cedex (France); Gavillet, D. [Paul Scherrer Institute, CH-5232 Villigen PSI (Switzerland); De Bremaecker, A. [Institute for Nuclear Materials Sciences, SCK-CEN, Boeretang 200, B-2400 Mol (Belgium)

    2014-10-15

    One objective of the FPT2 test of the PHEBUS FP Program was to study the degradation of an irradiated UO{sub 2} fuel bundle and the fission product behaviour under conditions of low steam flow. The results of the post-irradiation examinations (PIE) at the upper levels (823 mm and 900 mm) of the test section previously reported are interpreted in the present paper. Solid state interactions between fuel and cladding have been compared with the characteristics of interaction identified in the previous separate-effect tests. Corium resulting from the interaction between fuel and cladding was formed. The uranium concentration in the corium is compared to analytical tests and a scenario for the corium formation is proposed. The analysis showed that, despite the rather low fuel burn up, the conditions of temperature and oxygen potential reached during the starvation phase are able to give an early very significant release fraction of caesium. A significant part (but not all) of the molybdenum was segregated at grain boundaries and trapped in metallic inclusions from which they were totally removed in the final part of the experiment. During the steam starvation phase, the conditions of oxygen potential were favourable for the formation of simple Ba and BaO chemical forms but the temperature was too low to provoke their volatility. This is one important difference with out-of-pile experiments such as VERCORS for which only a combination of high temperature and low oxygen potential induced a significant barium release. Finally another significant difference with analytical out-of-pile experiments comes from the formation of foamy zones due to the fission gas presence in FPT2-type experiments which give an additional possibility for the formation of stable fission product compounds.

  19. Theoretical analysis of nuclear reactors (Phase I), I-V, Part III, Reactor poisoning; Razrada metoda teorijske analize nuklearnih reaktora (I faza) I-V, III Deo, Zatrovanje reaktora, I faza

    Energy Technology Data Exchange (ETDEWEB)

    Pop-Jordanov, J [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1962-07-15

    Method was developed for calculation of Xe{sup 135} static effect and kinetic effects of Xe{sup 135} and Sm{sup 149} with separate treatment of iodine effect and influence of reactor poisoning during power increase. Mentioned effects are treated first for uranium fuel and then the basic formulae were generalized for a mixture of fissile material. The annex contains a table with data needed for calculations and the Xe{sup 135} microscopic capture cross section dependent on temperature. Razradjen je metod proracuna statickog efekta Xe{sup 135} zatim kinetickog efekta Xe{sup 135} i Sm{sup 149} sa posebnim tretiranjem jodne jame i promene zatrovanja pri prelazu sa jedne snage na drugu. Navedeni efekti su tretirani prvo za uransko gorivo, a zatim su glavni obrasci uopsteni za smesu fisibilnih materijala. U prilogu su dati u vidu tabele, podaci potrebni za proracun i grafik zavisnosti mikroskopskog preseka zahvata Xe{sup 135} od temperature.

  20. Asymmetrical Interleaved DC/DC Switching Converters for Photovoltaic and Fuel Cell Applications—Part 2: Control-Oriented Models

    Directory of Open Access Journals (Sweden)

    Sergio Ignacio Serna-Garces

    2013-10-01

    Full Text Available A previous article has presented the members of the asymmetrical interleaved dc/dc switching converters family as very appropriate candidates to interface between photovoltaic or fuel cell generators and their loads because of their reduced ripple and increased current processing capabilities. After a review of the main modeling methods suitable for high-order converters operating, as the asymmetrical interleaved converters (AIC ones, in discontinuous current conduction mode a full-order averaged model has been adapted and improved to describe the dynamic behavior of AIC. The excellent agreement between the mathematical model predictions, the switched simulations and the experimental results has allowed for satisfactory design of a linear-quadratic regulator (LQR in a fuel-cell application example, which demonstrates the usefulness of the improved control-oriented modeling approach when the switching converters operate in discontinuous conduction mode.

  1. CORROSION RESISTANCE OF ORGANOMETALLIC COATING APLICATED IN FUEL TANKS USING ELECTROCHEMICAL IMPEDANCE SPECTROSCOPY IN BIOFUEL – PART I

    Directory of Open Access Journals (Sweden)

    Milene Adriane Luciano

    2014-10-01

    Full Text Available Nowadays, the industry has opted for more sustainable production processes, and the planet has also opted for new energy sources. From this perspective, automotive tanks with organometallic coatings as well as a partial substitution of fossil fuels by biofuels have been developed. These organometallic coated tanks have a zinc layer, deposited by a galvanizing process, formed between the steel and the organometallic coating. This work aims to characterize the organometallic coating used in metal automotive tanks and evaluate their corrosion resistance in contact with hydrated ethyl alcohol fuel (AEHC. For this purpose, the resistance of all layers formed between Zinc and EEP steel and also the tin coated steel, which has been used for over thirty years, were evaluated. The technique chosen was the Electrochemical Impedance Spectroscopy. The results indicated an increase on the corrosion resistance when organometallic coatings are used in AEHC medium. In addition to that, these coatings allow an estimated 25% reduction in tanks production costs.

  2. Drying of encapsulated parts (nuclear fuel rods) in applying vacuum, by introducing dehydratings, vacuum, and filling with an inert gas

    International Nuclear Information System (INIS)

    Johnson, C.R.

    1976-01-01

    This invention concerns a decontamination technique, in particular a process and equipment for extracting the water contained in fuel rods and other similar components of a nuclear reactor. The extraction of the contaminants contained in the fuel rods is carried out by a standard method by drilling a small hole in the surface of the cladding and applying a vacuum to bleed the rod of its impurities (moisture and gas). The invention consists for example in applying a vacuum at the hole drilled in the cladding to extract the contaminants and introducing spirit into the rod through the same orifice. The spirit absorbs the remaining liquid and other impurities. The spirit charged with the impurities is then pumped out by the same aperture by means of a regulated atmosphere inside a closed receptacle. This receptacle is then filled with an inert gas cooled to ambient temperature. The rods are then pressurised and the small orifice is sealed [fr

  3. Nuclear fuel reprocessing and high level waste disposal: informational hearings. Volume XII. Public and private roles, Part 2

    International Nuclear Information System (INIS)

    1977-01-01

    Presentations were made on institutional experiences at Nuclear Fuel Services, the framework for an acceptable nuclear future, the Price-Anderson Indemnity Act, Congress and nuclear energy policy, human dimension, and risk perception. The supplemental testimony and materials submitted for the record included information of the nuclear waste at West Valley, New York, the perception and acceptability of risk from nuclear and alternative energy sources, and psychological determinants of perceived and acceptable risk

  4. Nuclear fuel element design and thermal-hydraulic analysis of Wolsung-1, 600 MWe CANDU-PHWR (Part II)

    International Nuclear Information System (INIS)

    Suk, H.C; Lee, J.C.; Suh, K.S.; Yuk, K.E.; Whang, W.; Park, J.S.; Eim, J.S.; Bang, K.H.; Eim, M.S.; Rim, C.S.

    1982-01-01

    The main objective of the present thermal hydraulic analysis is to determine the thermal hydraulic characteristics of Wolsung-1 600 MWe CANDU-PHW reactor under normal operation. This is to verify and expedite the development of the nuclear fuel design and fabrication as well as the management. The computer program package developed for the stated objective are DOD81, CANREPP, PLOC81 and COBRA-CANDU. (Author)

  5. Development of methods for theoretical analysis of nuclear reactors (Phase II), I-V, Part IV, Fuel depletion

    International Nuclear Information System (INIS)

    Pop-Jordanov, J.

    1962-10-01

    This report includes the analysis of plutonium isotopes from U 238 depletion chain. Two theoretical approaches for solving the depletion of fuel are shown. One results in the system of differential equations that can be solved only by using electronic calculators and the second, Machinari-Goto method enables obtaining analytical equations for approximative values of particular nuclei. In addition, differential equations are given for different approximation levels in calculating Pu 239 , as well as relations between the released energy and irradiation [sr

  6. Final disposal of spent fuels and high activity waste: the European model for a shared regional repository. Part 3

    International Nuclear Information System (INIS)

    Herscovich de Pahissa, Marta

    2009-01-01

    Geological disposal is a essential element and the only available approach to the management strategy for spent nuclear fuel and high level radioactive waste from reprocessing and also for other long-lived waste from nuclear technology applications. It is technically feasible and offers the required long term safety. The growth of existing nuclear programmes and the expansion of nuclear technology to new countries will have effects on the fuel cycle because of the increased concern on proliferation and waste management. The crucial task is to ensure that all countries that use nuclear energy now or will do it in the future, have defined and agreed safety and security standards for all facilities and a credible waste disposal strategy , accepted by the community, when this become necessary. Multinational cooperation on essential aspects of fuel cycle, particularly the geological disposal, is required for several countries with relatively small nuclear energy programmes or small quantities of radioactive waste. For these countries, that can be in different stages of development, the possibility to share a deep geological repository could be convenient. The European Union SAPIERR project is described in this paper as an example of a regional multinational cooperation. (author) [es

  7. Air quality and climate change, Topic 3 of the Model Inter-Comparison Study for Asia Phase III (MICS-Asia IIIPart 1: Overview and model evaluation

    Directory of Open Access Journals (Sweden)

    M. Gao

    2018-04-01

    Full Text Available Topic 3 of the Model Inter-Comparison Study for Asia (MICS-Asia Phase III examines how online coupled air quality models perform in simulating high aerosol pollution in the North China Plain region during wintertime haze events and evaluates the importance of aerosol radiative and microphysical feedbacks. A comprehensive overview of the MICS-Asia III Topic 3 study design, including descriptions of participating models and model inputs, the experimental designs, and results of model evaluation, are presented. Six modeling groups from China, Korea and the United States submitted results from seven applications of online coupled chemistry–meteorology models. Results are compared to meteorology and air quality measurements, including data from the Campaign on Atmospheric Aerosol Research Network of China (CARE-China and the Acid Deposition Monitoring Network in East Asia (EANET. The correlation coefficients between the multi-model ensemble mean and the CARE-China observed near-surface air pollutants range from 0.51 to 0.94 (0.51 for ozone and 0.94 for PM2.5 for January 2010. However, large discrepancies exist between simulated aerosol chemical compositions from different models. The coefficient of variation (SD divided by the mean can reach above 1.3 for sulfate in Beijing and above 1.6 for nitrate and organic aerosols in coastal regions, indicating that these compositions are less consistent from different models. During clean periods, simulated aerosol optical depths (AODs from different models are similar, but peak values differ during severe haze events, which can be explained by the differences in simulated inorganic aerosol concentrations and the hygroscopic growth efficiency (affected by varied relative humidity. These differences in composition and AOD suggest that future models can be improved by including new heterogeneous or aqueous pathways for sulfate and nitrate formation under hazy conditions, a secondary organic aerosol (SOA

  8. Reprocessing of the spent nuclear fuel, I-VIII, Part IV, Engineering drawings, C - Sampling equipment; Prerada isluzenog nuklearnog goriva, I-VIII, IV Deo, Konstruktivni crtezi, C - Uredjaj za uzimanje uzoraka

    Energy Technology Data Exchange (ETDEWEB)

    Gal, I [Institute of Nuclear Sciences Boris Kidric, Laboratorija za hemiju visoke aktivnosti, Vinca, Beograd (Serbia and Montenegro)

    1963-02-15

    This volume includes the engineering drawings of the sampling equipment which is part of the pilot device for for extracting uranium, plutonium and fission products from the fuel irradiated in the reactor.

  9. A world's dilemma upon which the sun never sets. The nuclear waste management strategy: Western European nation states and the United States of America. Part I of III

    International Nuclear Information System (INIS)

    Sanders, Mark Callis; Sanders, Charlotta E.

    2016-01-01

    The management of spent nuclear fuel (SNF) and nuclear wastes demands a strategy to provide for the safe, secure, and permanent disposal of radioactive material from power generation, defense uses, and other activities. Nation states have taken different paths to nuclear waste management and are at various stages of the development of a nuclear waste management strategy. A strategy may include developing a geological repository, nuclear fuel reprocessing, interim storage, as well as discussions of the creation of a multinational storage facility. The paper provides an overview of the strategy used (or being developed) and its place within the legal framework. The paper concludes that though each nation state must look outward to its shared international obligations, there must also be an inward reflection of a nation state to its own traditions, customs, and legal/law making regimes.

  10. Radiation protection at the RA reactor in 1984, Part III Removal of the liquid radioactive effluents for the needs of the RA reactor

    International Nuclear Information System (INIS)

    Mandic, M.; Plecas, I.; Vukovic, Z.; Knezevic, Lj.; Jankovic, O.; Kostadinovic, A.; Mihailovic, B.

    1984-01-01

    Contaminated water originates from: hot cells, heavy water distillation device, storage pools for cooling and cutting of fuel elements, water biological shield of the reactor. During 1984, 400 liters of water contaminated by 60 Co was treated. Most recent measurements showed that the VR-1 pool contains 280 m 3 of effluents having specific activity of 3.3 10 4 Bq/ml, and VR-2 contains 30 m 3 with specific activity of 4 10 3 Bq/ml

  11. Fiscal 2000 survey report. Refuse-fueled power generation introduction technology, etc. Part 2. Survey of general refuse-fueled power generation; 2000 nendo chosa hokokusho. Haikibutsu hatsuden donyu gijutsu chosa to - Ippan haikibutsu hatsuden chosa Part 2

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-03-01

    Surveys were conducted of the effect on waste quality of the Law for Promotion of Sorted Collection and Recycling of Containers and Packaging, diffusion of PFI (private finance initiative) involving refuse-fueled power generation, recycling of slag, etc. Questionnairing was conducted for the survey of the effect on waste of the effectuation of the recycling law, size of population supplying the waste and the amount actually incinerated, actually measured data of the composition of incinerated waste, assorted collection programs, amount reducing measures, and the like. Studies were made as to if the empirical formulas applied in the survey of the preceding fiscal year would remain applicable after the coming into force of the recycling law. In the survey of PFI popularization, it was found that the rate it was taken into account for concrete projects was so low as 13.8% though most of the autonomous bodies had cognizance of PFI. It was requested that it be clearly stipulated that 'the conventional subsidies and grants apply also to PFI projects.' In the survey of conversion of residue into molten slag in refuse-fueled power generation and its reuse, four kinds of slag specimens were examined for physical properties, elution, and analyzed for ingredients, and then it was found that they posed no problems like heavy metal elution. (NEDO)

  12. Full and part load exergetic analysis of a hybrid micro gas turbine fuel cell system based on existing components

    International Nuclear Information System (INIS)

    Bakalis, Diamantis P.; Stamatis, Anastassios G.

    2012-01-01

    Highlights: ► Hybrid SOFC/GT system based on existing components. ► Exergy analysis using AspenPlus™ software. ► Greenhouse gases emission is significantly affected by SOFC stack temperature. ► Comparison with a conventional GT of similar power. ► SOFC/GT is almost twice efficient in terms of second low efficiency and CO 2 emission. - Abstract: The paper deals with the examination of a hybrid system consisting of a pre-commercially available high temperature solid oxide fuel cell and an existing recuperated microturbine. The irreversibilities and thermodynamic inefficiencies of the system are evaluated after examining the full and partial load exergetic performance and estimating the amount of exergy destruction and the efficiency of each hybrid system component. At full load operation the system achieves an exergetic efficiency of 59.8%, which increases during the partial load operation, as a variable speed control method is utilized. Furthermore, the effects of the various performance parameters such as fuel cell stack temperature and fuel utilization factor are assessed. The results showed that the components in which chemical reactions occur have the higher exergy destruction rates. The exergetic performance of the system is affected significantly by the stack temperature. Based on the exergetic analysis, suggestions are given for reducing the overall system irreversibility. Finally, the environmental impact of the operation of the hybrid system is evaluated and compared with a similarly rated conventional gas turbine plant. From the comparison it is apparent that the hybrid system obtains nearly double exergetic efficiency and about half the amount of greenhouse gas emissions compared with the conventional plant.

  13. Control of anode supported SOFCs (solid oxide fuel cells): Part I. mathematical modeling and state estimation within one cell

    International Nuclear Information System (INIS)

    Amedi, Hamid Reza; Bazooyar, Bahamin; Pishvaie, Mahmoud Reza

    2015-01-01

    In this paper, a 3-dimensional mathematical model for one cell of an anode-supported SOFC (solid oxide fuel cells) is presented. The model is derived from the partial differential equations representing the conservation laws of ionic and electronic charges, mass, energy, and momentum. The model is implemented to fully characterize the steady state operation of the cell with countercurrent flow pattern of fuel and air. The model is also used for the comparison of countercurrent with concurrent flow patterns in terms of thermal stress (temperature distribution) and quality of operation (current density). Results reveal that the steady-state cell performance curve and output of simulations qualitatively match experimental data of the literature. Results also demonstrate that countercurrent flow pattern leads to an even distribution of temperature, more uniform current density along the cell and thus is more enduring and superior to the concurrent flow pattern. Afterward, the thorough 3-dimensional model is used for state estimation instead of a real cell. To estimate states, the model is simplified and changed to a 1-dimensional model along flow streams. This simplified model includes uncertainty (because of simplifying assumptions of the model), noise, and disturbance (because of measurements). The behaviors of extended and ensemble Kalman filter as an observer are evaluated in terms of estimating the states and filtering the noises. Results demonstrate that, like extended Kalman filter, ensemble Kalman filter properly estimates the states with 20 sets. - Highlights: • A 3-dimensional model for one cell of SOFC (solid oxide fuel cells) is presented. • Higher voltages and thermal stress in countercurrent than concurrent flow pattern. • State estimation of the cell is examined by ensemble and extended Kalman filters. • Ensemble with 20 sets is as good as extended Kalman filter.

  14. Part I. Alternative fuel-cycle and deployment strategies: their influence on long-term energy supply and resource usage

    International Nuclear Information System (INIS)

    Till, C.E.; Chang, Y.I.; Rudolph, R.R.

    1980-01-01

    This report examines the implications of alternative fast breeder fuel cycles and deployment strategies on long-term energy supply and uranium resource utilization. An international-aggregate treatment for nuclear energy demand and resource base assumptions was adopted where specific assumptions were necessary for system analyses, but the primary emphasis was placed on understanding the general relationships between energy demand, uranium resource and breeder deployment option. The fast breeder deployment options studied include the reference Pu/U cycle as well as alternative cycles with varying degrees of thorium utilization

  15. Part II: Oxidative Thermal Aging of Pd/Al2O3 and Pd/CexOy-ZrO2 in Automotive Three Way Catalysts: The Effects of Fuel Shutoff and Attempted Fuel Rich Regeneration

    Directory of Open Access Journals (Sweden)

    Qinghe Zheng

    2015-10-01

    Full Text Available The Pd component in the automotive three way catalyst (TWC experiences deactivation during fuel shutoff, a process employed by automobile companies for enhancing fuel economy when the vehicle is coasting downhill. The process exposes the TWC to a severe oxidative aging environment with the flow of hot (800 °C–1050 °C air. Simulated fuel shutoff aging at 1050 °C leads to Pd metal sintering, the main cause of irreversible deactivation of 3% Pd/Al2O3 and 3% Pd/CexOy-ZrO2 (CZO as model catalysts. The effect on the Rh component was presented in our companion paper Part I. Moderate support sintering and Pd-CexOy interactions were also experienced upon aging, but had a minimal effect on the catalyst activity losses. Cooling in air, following aging, was not able to reverse the metallic Pd sintering by re-dispersing to PdO. Unlike the aged Rh-TWCs (Part I, reduction via in situ steam reforming (SR of exhaust HCs was not effective in reversing the deactivation of aged Pd/Al2O3, but did show a slight recovery of the Pd activity when CZO was the carrier. The Pd+/Pd0 and Ce3+/Ce4+ couples in Pd/CZO are reported to promote the catalytic SR by improving the redox efficiency during the regeneration, while no such promoting effect was observed for Pd/Al2O3. A suggestion is made for improving the catalyst performance.

  16. PACE. A Program for Acquiring Competence in Entrepreneurship. Part III: Being an Entrepreneur. Unit C: Keeping the Business Records. Research and Development Series No. 194 C-3.

    Science.gov (United States)

    Ohio State Univ., Columbus. National Center for Research in Vocational Education.

    This three-part curriculum for entrepreneurship education is primarily for postsecondary level, including four-year colleges and adult education, but can be adapted for special groups or vocational teacher education. The emphasis of the eight instructional units in Part II is operating a business. Unit C focuses on record keeping. It introduces…

  17. Contributions to Economic Geology, 1913: Part II - Mineral Fuels - Oil and Gas in the Western Part of the Olympic Peninsula, Washington

    Science.gov (United States)

    Lupton, Charles T.

    1915-01-01

    High-grade paraffin oil is reported to have been discovered in the western part of the Olympic Peninsula, Wash., as early as 1881. Since then attempts to obtain oil or gas in commercial quantities by drilling have been made from time to time in different localities in this region, but without success. Within the past few years interest has been aroused in oil seeps near the mouth of Hoh River and in gas vents in other parts of the field to such an extent that many persons have been attracted to this country to search for oil and gas. As a result of this interest and on account of the fact that efforts had been made to lease tracts of land for this purpose in the Queniult Indian Reservation, an examination of this region was made by the United States Geological Survey at the request of the Office of Indian Affairs. The results of the investigation, which are enumerated below and which are discussed in detail throughout this report, suggest that certain parts of the field are worthy of careful consideration by oil operators. The following summary includes the most important facts regarding the area examined: High-grade paraffin oil issues from two seeps near the mouth of Hoh River, and at other localities oil-saturated sandy clay ('smell mud' of the Indians) is exposed. Natural gas containing about 95 per cent methane escapes from a conical mound just north of the mouth of Queniult River and also from an inverted cone-shaped water-filled depression on Hoh River a short distance west of Spruce post office. Other minor gas vents are also known in this field and are described in detail in this report. Three wells - one in the reservation about 1 mile north and slightly west from Taholah, another near the mouth of Hoh River, and the third about 1 mile south of Forks - are being drilled for oil and gas. So far as drilling has progressed none of these wells have encountered oil in paying quantities, but all of them have struck small amounts of gas. A study of the structure

  18. Peatlands as Filters for Polluted Mine Water?—A Case Study from an Uranium-Contaminated Karst System in South Africa—Part III: Quantifying the Hydraulic Filter Component

    Directory of Open Access Journals (Sweden)

    Frank Winde

    2011-03-01

    Full Text Available As Part III of a four-part series on the filter function of peat for uranium (U, this paper focuses on the hydraulic component of a conceptual filter model introduced in Part II. This includes the quantification of water flow through the wetland as a whole, which was largely unknown and found to be significantly higher that anticipated. Apart from subaquatic artesian springs associated with the underlying karst aquifer the higher flow volumes were also caused by plumes of polluted groundwater moving laterally into the wetland. Real-time, quasi-continuous in situ measurements of porewater in peat and non-peat sediments indicate that rising stream levels (e.g., during flood conditions lead to the infiltration of stream water into adjacent peat deposits and thus allow for a certain proportion of flood water to be filtered. However, changes in porewater quality triggered by spring rains may promote the remobilization of possibly sorbed U.

  19. Pelletised fuel production from coal tailings and spent mushroom compost - Part II. Economic feasibility based on cost analysis

    International Nuclear Information System (INIS)

    Ryu, Changkook; Khor, Adela; Sharifi, Vida N.; Swithenbank, Jim

    2008-01-01

    Due to the growing market for sustainable energy, in order to increase the quality of the fuels, pellets are being produced from various materials such as wood and other biomass energy crops, and municipal waste. This paper presents the results from an economic feasibility study for pellet production using blends of two residue materials: coal tailings from coal cleaning and spent mushroom compost (SMC) from mushroom production. Key variables such as the mixture composition, raw material haulage and plant scale were considered and the production costs were compared to coal and biomass energy prices. For both wet materials, the moisture content was the critical parameter that influenced the fuel energy costs. The haulage distance of the raw materials was another factor that can pose a high risk. The results showed that the pellet production from the above two materials can be viable when a less energy-intensive drying process is utilised. Potential market outlets and ways to lower the costs are also discussed in this paper. (author)

  20. The Effect of early physiotherapy on the recovery of mandibular function after orthognathic surgery for Class III correction: part I--jaw-motion analysis.

    Science.gov (United States)

    Teng, Terry Te-Yi; Ko, Ellen Wen-Ching; Huang, Chiung Shing; Chen, Yu-Ray

    2015-01-01

    The aim of this prospective study was to compare the mandibular range of motion in Class III patients with and without early physiotherapy after orthognathic surgery (OGS). This study consisted of 63 Class III patients who underwent 2-jaw OGS. The experimental group comprised 31 patients who received early systematic physical rehabilitation. The control group consisted of 32 patients who did not have physical rehabilitation. Twelve variables of 3-dimensional (3D) jaw-motion analysis (JMA) were recorded before surgery (T1) and 6 weeks (T2) and 6 months (T3) after surgery. A 2-sample t test was conducted to compare the JMA results between the two groups at different time points. At T2, the JMA data were measured to be 77.5%-145.7% of presurgical values in the experimental group, and 60.3%-90.6% in the control group. At T3, the measurements were 112.2%-179.2% of presurgical values in the experimental group, and 77.6%-157.2% in the control group. The patients in the experimental group exhibited more favorable recovery than did those in the control group, from T1 to T2 and T1 to T3. However, after termination of physiotherapy, no significant difference in the extent of recovery was observed between groups up to 6 months after OGS. Copyright © 2014 European Association for Cranio-Maxillo-Facial Surgery. Published by Elsevier Ltd. All rights reserved.

  1. Superfund TIO videos. Set A. Regulatory overview - CERCLA's relationship to other programs: RCRA, Title III, UST, CWA, SDWA. Part 1. Audio-Visual

    International Nuclear Information System (INIS)

    1990-01-01

    The videotape is divided into five sections. Section 1 provides definitions and historical information on both the Resource Conservation and Recovery Act (RCRA) and the Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA). The four types of RCRA regulatory programs - Subtitles C, D, I, and J - are described. Treatment, storage, and disposal (TSD) and recycling facilities are also discussed. Section 2 discusses the history behind the Emergency Planning and Community Right-to-Know Act (Title III). The four major provisions of Title III, which are emergency planning, emergency release notification, community right-to-know reporting, and the toxic chemical release inventory are covered. Section 3 outlines the UST program covering notification, record keeping, and the UST Trust Fund. Section 4 outlines the six major provisions of the Clean Water Act (CWA): water quality, pretreatment, prevention of oil and hazardous substance discharges, responses to oil and hazardous substance discharges, discharges of hazardous substances into the ocean, and dredge and fill. Section 5 explains the purpose, regulations, and standards of the Safe Drinking Water Act (SDWA). Specific issues such as underground injection, sole source aquifers, and lead contamination are discussed

  2. Numerical study of the THM effects on the near-field safety of a hypothetical nuclear waste repository - BMT1 of the DECOVALEX III project. Part 1: conceptualization and characterization of the problems and summary of results

    International Nuclear Information System (INIS)

    Chijimatsu, M.; Nguyen, T.S.; Jing, L.; De Jonge, J.; Kohlmeier, M.; Millard, A.; Rejeb, A.; Rutqvist, J.; Souley, M.; Sugita, Y.

    2004-01-01

    Geological disposal of the spent nuclear fuel uses often the concept of multiple barrier systems. In order to predict the performance of these barriers, mathematical models have been developed, verified and validated against analytical solutions, laboratory tests and field experiments within the international DECOVALEX III project. These models in general consider the full coupling of thermal (T), hydraulic (H) and mechanical (M) processes that would prevail in the geological media around the repository. For Bench Mark Test no. 1 (BMT1) of the DECOVALEX III project, seven multinational research teams studied the implications of coupled THM processes on the safety of a hypothetical nuclear waste repository at the near-field and are presented in three accompany papers in this issue. This paper is the first of the three companion papers, which provides the conceptualization and characterization of the BMT1 as well as some general conclusions based on the findings of the numerical studies. It also shows the process of building confidence in the mathematical models by calibration with a reference T-H-M experiment with realistic rock mass conditions and bentonite properties and measured outputs of thermal, hydraulic and mechanical variables

  3. Water Pollution, and Treatments Part III: Biodegradation of Oil in Refineries Waste Water and Oils Adsorbed in Agricultural Wastes by Selected Strains of Cyanobacteria

    International Nuclear Information System (INIS)

    El-Emary, M.M.; Ali, N.A.; Naguib, M.M.

    2011-01-01

    The main objective of this study is to determine the biological degradation of oil hydrocarbons and sulfur compounds of Marine Balayim crude oil and its refined products by selected indigenous Cyanobacteria strains. The oils used were Marine Balayim crude oil, skimmed oil and some refined products such as gasoline, kerosene, gas oil, fuel oil and petroleum coke. The selected organisms in the current study are the Blue-Green Algae Cyanobacteria, Oscillatoria limentica. This organism was collected from the hyper saline environment of the solar lake in Taba, Sinai, Egypt. The results obtained revealed that the utilization of such strains can be used for the bioremediation of oily waste water.

  4. The fuel cycle scoping system

    International Nuclear Information System (INIS)

    Dooley, G.D.; Malone, J.P.

    1986-01-01

    The Fuel Cycle Scoping System (FCSS) was created to fill the need for a scoping tool which provides the utilities with the ability to quickly evaluate alternative fuel management strategies, tails assay choices, fuel fabrication quotes, fuel financing alternatives, fuel cycle schedules, and other fuel cycle perturbations. The FCSS was specifically designed for PC's that support dBASE-III(TM), a relational data base software system by Ashton-Tate. However, knowledge of dBASE-III is not necessary in order to utilize the FCSS. The FCSS is menu driven and can be utilized as a teaching tool as well as a scoping tool

  5. Part-Load Performance Prediction and Operation Strategy Design of Organic Rankine Cycles with a Medium Cycle Used for Recovering Waste Heat from Gaseous Fuel Engines

    Directory of Open Access Journals (Sweden)

    Xuan Wang

    2016-07-01

    Full Text Available The Organic Rankine Cycle (ORC is regarded as a suitable way to recover waste heat from gaseous fuel internal combustion engines. As waste heat recovery systems (WHRS have always been designed based on rated working conditions, while engines often work under part-load conditions, it is quite significant to analyze the part-load performance and corresponding operation strategy of ORC systems. This paper presents a dynamic model of ORC with a medium cycle used for a large gaseous fuel engine and analyzes the effect of adjustable parameters on the system performance, giving effective control directions under various conditions. The results indicate that the intermediary fluid mass flow rate has nearly no effect on the output power and thermal efficiency of the ORC, while the mass flow rate of working fluid has a great effect on them. In order to get a better system performance under different working conditions, the system should be operated with the working fluid mass flow rate as large as possible, but with a slight degree of superheating. Then, with the control of constant superheat degree at the end of the heating process, the performance of the combined system that consists of ORC and the engine at steady state under seven typical working conditions is also analyzed. The results indicate that the energy-saving effect of WHRS becomes worse and worse as the working condition decreases. Especially at 40% working condition the WHRS nearly has no energy-saving effect anymore.

  6. Modern State and Efficiency Analysis of Heat Recovery in Fuel Furnaces Using High Temperature Recuperators. Part 2

    Directory of Open Access Journals (Sweden)

    B. S. Soroka

    2013-01-01

    Full Text Available The paper analyzes various factors that affect upon heat transfer in high temperature recuperators, namely: heat transfer enhancement, heat exchange surface increase and rise of temperature head between primary and secondary heat transfer fluids. Comparison of experimental data with the results of mathematical and computational fluid dynamics (CFD modeling has been performed in the paper. The paper considers some new designs of high temperature heat recovery plants: tube recuperator equipped with internal inserts – secondary emitters inside tubes for metallurgical furnaces and high-efficient two-way radiative recuperators for machinery engineering furnaces.  Advantages of new recuperators in comparison with existing analogues have been estimated in the paper. These advantages are:  provision of additional fuel saving due to increase of preheating temperature of the combustion air and improvement of design stability by decrease of tube wall temperature.

  7. Richard III

    DEFF Research Database (Denmark)

    Lauridsen, Palle Schantz

    2017-01-01

    Kort analyse af Shakespeares Richard III med fokus på, hvordan denne skurk fremstilles, så tilskuere (og læsere) langt henad vejen kan føle sympati med ham. Med paralleller til Netflix-serien "House of Cards"......Kort analyse af Shakespeares Richard III med fokus på, hvordan denne skurk fremstilles, så tilskuere (og læsere) langt henad vejen kan føle sympati med ham. Med paralleller til Netflix-serien "House of Cards"...

  8. Methanol Fuel Cell

    Science.gov (United States)

    Voecks, G. E.

    1985-01-01

    In proposed fuel-cell system, methanol converted to hydrogen in two places. External fuel processor converts only part of methanol. Remaining methanol converted in fuel cell itself, in reaction at anode. As result, size of fuel processor reduced, system efficiency increased, and cost lowered.

  9. Rise, fall and resurrection of chromosome territories: a historical perspective Part II. Fall and resurrection of chromosome territories during the 1950s to 1980s. Part III. Chromosome territories and the functional nuclear architecture: experiments and m

    OpenAIRE

    T Cremer; C Cremer

    2009-01-01

    Part II of this historical review on the progress of nuclear architecture studies points out why the original hypothesis of chromosome territories from Carl Rabl and Theodor Boveri (described in part I) was abandoned during the 1950s and finally proven by compelling evidence forwarded by laser-uvmicrobeam studies and in situ hybridization experiments. Part II also includes a section on the development of advanced light microscopic techniques breaking the classical Abbe limit written for reade...

  10. Simulation of thermal stresses in anode-supported solid oxide fuel cell stacks. Part II: Loss of gas-tightness, electrical contact and thermal buckling

    Science.gov (United States)

    Nakajo, Arata; Wuillemin, Zacharie; Van herle, Jan; Favrat, Daniel

    Structural stability issues in planar solid oxide fuel cells arise from the mismatch between the coefficients of thermal expansion of the components. The stress state at operating temperature is the superposition of several contributions, which differ depending on the component. First, the cells accumulate residual stresses due to the sintering phase during the manufacturing process. Further, the load applied during assembly of the stack to ensure electric contact and flatten the cells prevents a completely stress-free expansion of each component during the heat-up. Finally, thermal gradients cause additional stresses in operation. The temperature profile generated by a thermo-electrochemical model implemented in an equation-oriented process modelling tool (gPROMS) was imported into finite-element software (ABAQUS) to calculate the distribution of stress and contact pressure on all components of a standard solid oxide fuel cell repeat unit. The different layers of the cell in exception of the cathode, i.e. anode, electrolyte and compensating layer were considered in the analysis to account for the cell curvature. Both steady-state and dynamic simulations were performed, with an emphasis on the cycling of the electrical load. The study includes two different types of cell, operation under both thermal partial oxidation and internal steam-methane reforming and two different initial thicknesses of the air and fuel compressive sealing gaskets. The results generated by the models are presented in two papers: Part I focuses on cell cracking. In the present paper, Part II, the occurrences of loss of gas-tightness in the compressive gaskets and/or electrical contact in the gas diffusion layer were identified. In addition, the dependence on temperature of both coefficients of thermal expansion and Young's modulus of the metallic interconnect (MIC) were implemented in the finite-element model to compute the plastic deformation, while the possibilities of thermal buckling

  11. Description and exploitation of benchmarks involving {sup 149}Sm, a fission product taking part of the burn up credit in spent fuels

    Energy Technology Data Exchange (ETDEWEB)

    Anno, J.; Poullot, G. [CEA Centre d`Etudes de Fontenay-aux-Roses, 92 (France). Inst. de Protection et de Surete Nucleaire; Fouillaud, P.; Grivot, P. [CEA Centre d`Etudes de Valduc, 21 - Is-sur-Tille (France)

    1995-12-31

    Up to now, there was no benchmark to validate the Fission Products (FPs) cross sections in criticality safety calculations. The protection and nuclear safety institute (IPSN) has begun an experimental program on 6 FPs ({sup 103}Rh, {sup 133}Cs, {sup 143}Nd, {sup 149}Sm, {sup 152}Sm, and {sup 155}Gd daughter of {sup 155}Eu) giving alone a decrease of reactivity equal to half the whole FPs in spent fuels (except Xe and I). Here are presented the experiments with the {sup 149}Sm and the results obtained with the APOLLO I-MORET III calculations codes. 11 experiments are carried out in a zircaloy tank of 3.5 1 containing slightly nitric acid solutions of Samarium (96,9% in weight of {sup 149S}m) at 0.1048 -0.2148 - 0.6262 g/l concentrations. It was placed in the middle of arrays of UO{sub 2} rods (4.742 % U5 weight %) at square pitch of 13 mm. The underwater height of the rods is the critical parameter. In addition, 7 experiments were performed with the same apparatus with water and boron proving a good experimental representativeness and a good accuracy of the calculations. As the reactivity worth of the Sm tank is between 2000 and 6000 10{sup -5}, the benchmarks are well representative and the cumulative absorption ratios show that {sup 149}Sm is well qualified under 1 eV. (authors). 8 refs., 7 figs., 6 tabs.

  12. Description and exploitation of benchmarks involving 149Sm, a fission product taking part of the burn up credit in spent fuels

    International Nuclear Information System (INIS)

    Anno, J.; Poullot, G.

    1995-01-01

    Up to now, there was no benchmark to validate the Fission Products (FPs) cross sections in criticality safety calculations. The protection and nuclear safety institute (IPSN) has begun an experimental program on 6 FPs ( 103 Rh, 133 Cs, 143 Nd, 149 Sm, 152 Sm, and 155 Gd daughter of 155 Eu) giving alone a decrease of reactivity equal to half the whole FPs in spent fuels (except Xe and I). Here are presented the experiments with the 149 Sm and the results obtained with the APOLLO I-MORET III calculations codes. 11 experiments are carried out in a zircaloy tank of 3.5 1 containing slightly nitric acid solutions of Samarium (96,9% in weight of 149S m) at 0.1048 -0.2148 - 0.6262 g/l concentrations. It was placed in the middle of arrays of UO 2 rods (4.742 % U5 weight %) at square pitch of 13 mm. The underwater height of the rods is the critical parameter. In addition, 7 experiments were performed with the same apparatus with water and boron proving a good experimental representativeness and a good accuracy of the calculations. As the reactivity worth of the Sm tank is between 2000 and 6000 10 -5 , the benchmarks are well representative and the cumulative absorption ratios show that 149 Sm is well qualified under 1 eV. (authors). 8 refs., 7 figs., 6 tabs

  13. Reprocessing of the spent nuclear fuel, I-VIII, Part VIII; Prerada isluzenog nuklearnog goriva, I-VIII, VIII Deo

    Energy Technology Data Exchange (ETDEWEB)

    Gal, I [Institute of Nuclear Sciences Boris Kidric, Laboratorija za hemiju visoke aktivnosti, Vinca, Beograd (Serbia and Montenegro)

    1963-02-15

    This volume includes the following three reports: Separation of uranium, plutonium and fission products on zirconium phosphate; Separation of uranium, plutonium and fission products on zirconium phosphate, Part 1 - Adsorption equilibria and kinetics; and Adsorption of dibutyl phosphates and monobutyl phosphates on inorganic oxides.

  14. Standing on shaky ground- US patent-eligibility of isolated DNA and genetic diagnostics after AMP v. USPTO - Part III (unsolved questions & subsequent case law)

    DEFF Research Database (Denmark)

    Minssen, Timo; Nilsson, David

    2012-01-01

    invigorated U.S. debate on patent eligibility, referring inter alia to the 2010 U.S. Supreme Court decision in Bilski v. Kappos and the pending certiorari in Prometheus v. Mayo (1). Before this background, Part I recited the complex procedural history of AMP v. USPTO (2) and summarized the underpinnings...... of the outcome, i.e. the three different opinions of the Federal Circuit judges Lourie, Moore & Bryson who comprised the panel (3). Part II continued the tale with a detailed analysis of the decision's practical implications (4), which is followed by a closer look on the chances for an ultimate Supreme Court...... decision in Prometheus v. Mayo. Part IV, which is to be published in issue 4, will finally offer a broader discussion of the recent US patent-eligibility developments from an innovation policy perspective including brief references to recent European developments (7). This will provide the basis...

  15. A complex study on the reliability assessment of the containment of a PWR. Part III.- Structural reliability assessment under internal and external loading conditions

    International Nuclear Information System (INIS)

    Bauer, J.; Schueller, G.I.

    1977-01-01

    The first part of the analysis is concerned with the determination of the failure probability of the steel hull under internal load conditions. Two independent failure criteria are the basis for this calculation; the first one being the ultimate yield which is actually an instability condition and the second one being the fracture condition as described in Part II of the paper. Both the global and the local failure probabilities are investigated. The second part of the analysis is concerned with the external load case of earthquake. As it has already been described in Part I the probability of occurrence of a LOCA, given an earthquake has been considered in connection with the probable damage which the steel hull might experience during the earthquake. In other words the survival probability of the hull with deteriorated resistance is calculated, taking into account the frequencies of occurrence of the various events. The third part of the analysis is concerned with the reliability determination of the reinforced concrete dome structure, which is supposed to protect, the steel hull against external load conditions such as airplane crash and external pressure waves (the latter covering the load case of tornado occurrence). The reliability analysis of the reinforced concrete structure under earthquake loading is performed by utilizing the time-history method. Some aspects of the drawbacks of the response spectra method -when used in a risk analysis- are pointed out. The probability distribution of the concrete strength as determined under intermediate strain rate as described in Part II is utilized in the analysis. Finally the remaining two external load cases are discussed in light of their use in a reliability analysis and with respect to their frequency of occurrence and the probability distribution of their load intensities. The reliability demonstration is performed using the containment structure of the PWR-plant 'Biblis B' which is locate

  16. PARDISEKO III

    International Nuclear Information System (INIS)

    Jordan, H.; Sack, C.

    1975-05-01

    This report gives a detailed description of the latest version of the PARDISEKO code, PARDISEKO III, with particular emphasis on the numerical and programming methods employed. The physical model and its relation to nuclear safety as well as a description and the results of confirming experiments are treated in detail in the Karlsruhe Nuclear Research Centre report KFK-1989. (orig.) [de

  17. Proceedings of the Scientific Meeting and Presentation on Basic Research in Nuclear of the Science and Technology part III : Radioactive Waste Management and Environment

    International Nuclear Information System (INIS)

    Kamsul Abraha; Yateman Arryanto; Sri Jauhari S; Agus Taftazani; Kris Tri Basuki; Djoko Sardjono, Ign.; Sukarsono, R.; Samin; Syarip; Suryadi, MS; Sardjono, Y.; Tri Mardji Atmono; Dwiretnani Sudjoko; Tjipto Sujitno, BA.

    2007-08-01

    The Scientific Meeting and Presentation on Basic Research in Nuclear Science and Technology is a routine activity held by Centre for Accelerator Technology and Material Process, National Nuclear Energy Agency, for monitoring the research activity which achieved in National Nuclear Energy Agency. The Meeting was held in Yogyakarta on July 10, 2007. The proceedings contains papers presented on the meeting about Radioactive Waste Management and Environment and there are 25 papers which have separated index. The proceedings is the third part of the three parts which published in series. (PPIN)

  18. Deep Desulfurization of Diesel Fuels with Plasma/Air as Oxidizing Medium, Diperiodatocuprate (III) as Catalyzer and Ionic Liquid as Extraction Solvent

    Science.gov (United States)

    Ban, Lili; Liu, Ping; Ma, Cunhua; Dai, Bin

    2013-12-01

    In this paper, the oxidative desulfurization (ODS) system is directly applied to deal with the catalytic oxidation of sulfur compounds of sulfur-containing model oil by dielectric barrier discharge (DBD) plasma in the presence of air plus an extraction step with the oxidation-treated fuel put over ionic liquid [BMIM]FeCl4 (1-butyl-3-methylimidazolium tetrachloroferrate). This new system exhibited an excellent desulfurization effect. The sulfur content of DBT in diesel oil decreased from 200 ppm to 4.92 ppm (S removal rate up to 97.5%) under the following optimal reaction conditions: air flow rate (ν) of 60 mL/min, amplitude of applied voltage (U) on DBD of 16 kV, input frequency (f) of 79 kHz, catalyst amount (ω) of 1.25 wt%, reaction time (t) of 10 min. Moreover, a high desulfurization rate was obtained during oxidation of benzothiophene (BT) or 4,6-DMDBT (4,6-dimethyl-dibenzothiophene) under the aforementioned conditions. The oxidation reactivity of different S compounds was decreased in the order of DBT, 4,6-DMDBT and BT. The remarkable advantage of the novel ODS system is that the desulfurization condition applies in the presence of air at ambient conditions without peroxides, aqueous solvent or biphasic oil-aqueous solution system.

  19. Deep Desulfurization of Diesel Fuels with Plasma/Air as Oxidizing Medium, Diperiodatocuprate (III) as Catalyzer and Ionic Liquid as Extraction Solvent

    International Nuclear Information System (INIS)

    Ban Lili; Liu Ping; Ma Cunhua; Dai Bin

    2013-01-01

    In this paper, the oxidative desulfurization (ODS) system is directly applied to deal with the catalytic oxidation of sulfur compounds of sulfur-containing model oil by dielectric barrier discharge (DBD) plasma in the presence of air plus an extraction step with the oxidation-treated fuel put over ionic liquid [BMIM]FeCl 4 (1-butyl-3-methylimidazolium tetrachloroferrate). This new system exhibited an excellent desulfurization effect. The sulfur content of DBT in diesel oil decreased from 200 ppm to 4.92 ppm (S removal rate up to 97.5%) under the following optimal reaction conditions: air flow rate (ν) of 60 mL/min, amplitude of applied voltage (U) on DBD of 16 kV, input frequency (f) of 79 kHz, catalyst amount (ω) of 1.25 wt%, reaction time (t) of 10 min. Moreover, a high desulfurization rate was obtained during oxidation of benzothiophene (BT) or 4,6-DMDBT (4,6-dimethyl-dibenzothiophene) under the aforementioned conditions. The oxidation reactivity of different S compounds was decreased in the order of DBT, 4,6-DMDBT and BT. The remarkable advantage of the novel ODS system is that the desulfurization condition applies in the presence of air at ambient conditions without peroxides, aqueous solvent or biphasic oil-aqueous solution system. (plasma technology)

  20. Physical and welding metallurgy of Gd-enriched austenitic alloys for spent nuclear fuel applications. Part II, nickel base alloys

    International Nuclear Information System (INIS)

    Mizia, Ronald E.; Michael, Joseph Richard; Williams, David Brian; Dupont, John Neuman; Robino, Charles Victor

    2004-01-01

    The physical and welding a metallurgy of gadolinium- (Gd-) enriched Ni-based alloys has been examined using a combination of differential thermal analysis, hot ductility testing. Varestraint testing, and various microstructural characterization techniques. Three different matrix compositions were chosen that were similar to commercial Ni-Cr-Mo base alloys (UNS N06455, N06022, and N06059). A ternary Ni-Cr-Gd alloy was also examined. The Gd level of each alloy was ∼2 wt-%. All the alloys initiated solidification by formation of primary austenite and terminated solidification by a Liquid γ + Ni 5 Gd eutectic-type reaction at ∼1270 C. The solidification temperature ranges of the alloys varied from ∼100 to 130 C (depending on alloy composition). This is a substantial reduction compared to the solidification temperature range to Gd-enriched stainless steels (360 to 400 C) that terminate solidification by a peritectic reaction at ∼1060 C. The higher-temperature eutectic reaction that occurs in the Ni-based alloys is accompanied by significant improvements in hot ductility and solidification cracking resistance. The results of this research demonstrate that Gd-enriched Ni-based alloys are excellent candidate materials for nuclear criticality control in spent nuclear fuel storage applications that require production and fabrication of large amounts of material through conventional ingot metallurgy and fusion welding techniques

  1. Ash behavior during hydrothermal treatment for solid fuel applications. Part 2: Effects of treatment conditions on industrial waste biomass

    International Nuclear Information System (INIS)

    Mäkelä, Mikko; Yoshikawa, Kunio

    2016-01-01

    Highlights: • Effect of treatment conditions on composition and solubility of ash. • Ash dissolution and yield governed by liquid pH and calcium carbonate solubility. • Dissolution of calcium carbonate decreases ash fusion temperature during combustion. • Decreasing the ash content of sludge can weaken ash properties for combustion. - Abstract: This second half of our work on ash behavior concentrates on the effects of hydrothermal treatment conditions on paper sludge. Ash composition and solubility were determined based on treatment temperature, reactor solid load and liquid pH using experimental design and univariate regression methods. In addition, ash properties for combustion were evaluated based on recent developments on ash classification. Based on the results, all experimental variables had a statistically significant effect on ash yields. Only reactor solid load was statistically insignificant for char ash content, which increased based on increasing treatment temperature due to the decomposition of organic components. Ash dissolution and ash yield were governed by liquid pH and the generation of acids mainly due to the solubility of calcium carbonate identified as the main mineral species of paper sludge. Dissolution of calcium carbonate however decreased ash fusion temperatures more likely causing problems during char incineration. This indicated that decreasing the ash content of sludge during hydrothermal treatment can actually weaken ash properties for solid fuel applications.

  2. Fuel assemblies

    International Nuclear Information System (INIS)

    Echigoya, Hironori; Nomata, Terumitsu.

    1983-01-01

    Purpose: To render the axial distribution relatively flat. Constitution: First nuclear element comprises a fuel can made of zircalloy i.e., the metal with less neutron absorption, which is filled with a plurality of UO 2 pellets and sealed by using a lower end plug, a plenum spring and an upper end plug by means of welding. Second fuel element is formed by substituting a part of the UO 2 pellets with a water tube which is sealed with water and has a space for allowing the heat expansion. The nuclear fuel assembly is constituted by using the first and second fuel elements together. In such a structure, since water reflects neutrons and decrease their leakage to increase the temperature, reactivity is added at the upper portion of the fuel assembly to thereby flatten the axial power distribution. Accordingly, stable operation is possible only by means of deep control rods while requiring no shallow control rods. (Sekiya, K.)

  3. Children Who Desperately Want To Read, but Are Not Working at Grade Level: Use Movement Patterns as "Windows" To Discover Why. Part III: The Frontal Midline.

    Science.gov (United States)

    Corso, Marjorie

    A longitudinal research study observed 30 children between the ages of infancy and elementary age to determine if using large muscle motor patterns to master the three identified midlines that concur with the body planes used in anatomy is reflected in academic classroom learning levels. This third part of the study focused on the frontal midline.…

  4. AUTOMOTIVE DIESEL MAINTENANCE 1. UNIT XXII, I--MAINTAINING THE FUEL SYSTEM (PART I)--CUMMINS DIESEL ENGINE, II--UNDERSTANDING THE DIFFERENTIAL.

    Science.gov (United States)

    Minnesota State Dept. of Education, St. Paul. Div. of Vocational and Technical Education.

    THIS MODULE OF A 30-MODULE COURSE IS DESIGNED TO DEVELOP AN UNDERSTANDING OF THE FUNCTION AND MAINTENANCE OF THE DIESEL ENGINE FUEL SYSTEM AND DIFFERENTIAL DRIVE UNITS USED IN DIESEL POWERED VEHICLES. TOPICS ARE (1) FUEL SYSTEM COMPARISONS, (2) FUEL SYSTEM SUPPLY COMPONENTS, (3) FUEL SUPPLY SECTION MAINTENANCE, (4) FUNCTION OF THE DIFFERENTIAL,…

  5. A simple, fast, and accurate thermodynamic-based approach for transfer and prediction of gas chromatography retention times between columns and instruments Part III: Retention time prediction on target column.

    Science.gov (United States)

    Hou, Siyuan; Stevenson, Keisean A J M; Harynuk, James J

    2018-03-27

    This is the third part of a three-part series of papers. In Part I, we presented a method for determining the actual effective geometry of a reference column as well as the thermodynamic-based parameters of a set of probe compounds in an in-house mixture. Part II introduced an approach for estimating the actual effective geometry of a target column by collecting retention data of the same mixture of probe compounds on the target column and using their thermodynamic parameters, acquired on the reference column, as a bridge between both systems. Part III, presented here, demonstrates the retention time transfer and prediction from the reference column to the target column using experimental data for a separate mixture of compounds. To predict the retention time of a new compound, we first estimate its thermodynamic-based parameters on the reference column (using geometric parameters determined previously). The compound's retention time on a second column (of previously determined geometry) is then predicted. The models and the associated optimization algorithms were tested using simulated and experimental data. The accuracy of predicted retention times shows that the proposed approach is simple, fast, and accurate for retention time transfer and prediction between gas chromatography columns. © 2018 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  6. EOSLT Consortium Biomass Co-firing. WP 4. Biomass co-firing in oxy-fuel combustion. Part 1. Lab- Scale Experiments

    Energy Technology Data Exchange (ETDEWEB)

    Fryda, L.E. [ECN Biomass, Coal and Environmental Research, Petten (Netherlands)

    2011-07-15

    In the frame of WP4 of the EOS LT Co-firing program, the ash formation and deposition of selected coal/biomass blends under oxyfuel and air conditions were studied experimentally in the ECN lab scale coal combustor (LCS). The fuels used were Russian coal, South African coal and Greek Lignite, either combusted separately or in blends with cocoa and olive residue. The first trial period included tests with the Russian and South African coals and their blends with cocoa, the second trial period included Lignite with olive residue tests and a final period firing only Lignite and Russian coal, mainly to check and verify the observed results. During the testing, also enriched air combustion was applied, in order to establish conclusions whether a systematic trend on ash formation and deposition exists, ranging from conventional air, to enriched air (improving post combustion applications) until oxyfuel conditions. A horizontal deposition probe equipped with thermocouples and heat transfer sensors for on line data acquisition, and a cascade impactor (staged filter) to obtain size distributed ash samples including the submicron range at the reactor exit were used. The deposition ratio and the deposition propensity measured for the various experimental conditions were higher in all oxyfuel cases. No significant variations in the ash formation mechanisms and the ash composition were established. Finally the data obtained from the tests performed under air and oxy-fuel conditions were utilised for chemical equilibrium calculations in order to facilitate the interpretation of the measured data; the results indicate that temperature dependence and fuels/blends ash composition are the major factors affecting gaseous compound and ash composition rather than the combustion environment, which seems to affect neither the ash and fine ash (submicron) formation, nor the ash composition. The ash deposition mechanisms were studied in more detail in Part II of this report.

  7. Neutronic studies of the long life core concept: Part 1, Design and performance of 1000 MWe uranium oxide fueled low power density LMR cores

    International Nuclear Information System (INIS)

    Orechwa, Y.

    1987-04-01

    The parametric behavior of some key neutronic performance parameters for low power density LMR cores fueled with uranium oxide is investigated. The results are compared to reference homogeneous and heterogeneous cores with normal fuel management and Pu fueling. It can be concluded that with respect to minimizing the initial fissile mass and thereby economizing on the inventory costs and carrying charges, the superior neutron economy of the LMR fuel cycle is best exploited through normal fuel management with Pu recycling. In the once-through mode the LMR fuel cycle has disadvantages due to a higher fissile inventory and is not competitive with the LWR fuel cycle

  8. Simulación de la fluencia en caliente de un acero microaleado con un contenido medio de carbono. III parte. Ecuaciones constitutivas

    Directory of Open Access Journals (Sweden)

    Cabrera, J. M.

    1997-08-01

    Full Text Available According to the part 1 of this work the constitutive equations of the hot flow behaviour of a commercial microalloyed steel have been obtained. For this purpose, the uniaxial hot compression tests described in the part 2 were employed. Tests were carried out over a range of 5 orders of magnitude in strain rate and 300 °C of temperature. Experimental results are compared with the theoretical model introduced in the first part of this study. It is concluded that deviations between experimental and theoretical curves are lower than 10 %. It is shown that the classical hiperbolic sine constitutive equation described accurately the experimental behaviour provided that stresses are normalized by the Young's modulus and strain rates by the self-diffusion coefficient. An internal stress must also be introduced in the latter equation when the initial grain size is fine enough.

    Siguiendo el planteamiento teórico efectuado en la I parte de este trabajo, se determinaron las ecuaciones constitutivas del comportamiento a la deformación en caliente de un acero comercial microaleado con un contenido medio de carbono. Para este objetivo se emplearon los ensayos de compresión uniaxial en caliente ya descritos en la II parte, los cuales se efectuaron en un rango de cinco órdenes de magnitud en velocidad de deformación y 300 °C de temperatura. Se comparan los resultados experimentales con el modelo teórico introducido en la I parte y se verifica que el error es inferior al 10 %. Se comprobó que la clásica ecuación del seno hiperbólico podía describir con precisión el comportamiento observado siempre y cuando las tensiones se normalicen por el módulo de Young, las velocidades de deformación por el coeficiente de autodifusión de la austenita, y se considere un efecto adicional sobre la tensión cuando el tamaño de grano inicial sea suficientemente fino.

  9. Experience with advanced driver fuels in EBR-II

    International Nuclear Information System (INIS)

    Lahm, C.E.; Koenig, J.F.; Pahl, R.G.; Porter, D.L.; Crawford, D.C.

    1992-01-01

    The Experimental Breeder Reactor II (EBR-II) is a complete nuclear power plant, incorporating a pool-type liquid-metal reactor (LMR) with a fuel-power thermal output of 62.5 MW and an electrical output of 20 MW. Initial criticality was in 1961, utilizing a metallic driver fuel design called the Mark-I. The fuel design has evolved over the last 30 yr, and significant progress has been made on improving performance. The first major innovations were incorporated into the Mark-II design, and burnup then increased dramatically. This design performed successfully, and fuel element lifetime was limited by subassembly hardware performance rather than the fuel element itself. Transient performance of the fuel was also acceptable and demonstrated the ability of EBR-II to survive severe upsets such as a loss of flow without scram. In the mid 1980s, with renewed interest in metallic fuels and Argonne's integral fast reactor (IFR) concept, the Mark-II design was used as the basis for new designs, the Mark-III and Mark-IV. In 1987, the Mark-III design began qualification testing to become a driver fuel for EBR-II. This was followed in 1989 by the Mark-IIIA and Mark-IV designs. The next fuel design, the Mark-V, is being planned to demonstrate the utilization of recycled fuel. The fuel cycle facility attached to EBR-II is being refurbished to produce pyroprocessed recycled fuel as part of the demonstration of the IFR

  10. DETAILS OF OPERATIONS PERFORMED BY THE REMOTE CONTROL ROBOT (CONCEPT TO THE HORIZONTAL FUEL CHANNEL DURING DECOMMISSIONING PHASE OF NUCLEAR REACTOR CALANDRIA STRUCTURE. PART II: INSIDE OPERATIONS

    Directory of Open Access Journals (Sweden)

    Constantin POPESCU

    2017-05-01

    Full Text Available The authors contribution to this paper is to present a concept solution of a remote control robot (RCR used for decommissioning of the horizontal fuel channels pressure tube in the CANDU nuclear reactor. In this paper the authors highlight few details of geometry, operations, constraints by kinematics and dynamics of the robot movement inside of the reactor fuel channel. Inside operations performed has as the main steps of dismantling process the followings: unblock and extract the channel closure plug (from End Fitting - EF, unblock and extract the channel shield plug (from Lattice Tube - LT, cut the ends of the pressure tube, extract the pressure tube and cut it in small parts, sorting and storage extracted items in the safe robot container. All steps are performed in automatic mode. The remote control robot (RCR represents a safety system controlled by sensors and has the capability to analyze any error registered and decide next activities or abort the inside decommissioning procedure in case of any risk rise in order to ensure the environmental and workers protection.

  11. State of the art of the welding method for sealing spent nuclear fuel canister made of copper. Part 1 - FSW

    Energy Technology Data Exchange (ETDEWEB)

    Purhonen, T.

    2014-05-15

    The purpose of this report is to gather together comprehensive information concerning FSW as an optional welding method for welding the nuclear waste copper canister at the disposal facility. This report discusses the current situation, knowledge of the process and information concerning results of the development and research work related to welding thick copper and the special needs of the disposal environment. Most of the research work and development work has been done by Posiva's Swedish partner SKB, Swedish Nuclear Fuel and Waste Management Co. SKB chose FSW as their reference welding method in 2005. FSW (friction stir welding) is a solid-state welding method, invented in 1991, in which frictional heat is generated between the tool and the weld metal, causing the metal to soften, normally without reaching the melting point, and allowing the tool to traverse the joint line. Friction stir welding can be used for joining many types of materials and material combinations, if the tool materials and designs can be found which operate at the forging temperature of the workpiece. The general requirements for the copper canister weld and base material are presented in Posiva's VAHA-system, which sets the most critical values or demands concerning the short- and long-term properties or other needs. The sections in this report are set out in a similar way as in the VAHA-system. Concerning the results from the research and development work, it can be said that FS weld material fulfils the values set by VAHA. The quality of the welds fulfils the set demands for intact weld material and the welding process is robust using an automatic control system. There still remains work concerning the acceptance procedure for the welding process and other open issues which are described in this report. (orig.)

  12. State of the art of the welding method for sealing spent nuclear fuel canister made of copper. Part 1 - FSW

    International Nuclear Information System (INIS)

    Purhonen, T.

    2014-05-01

    The purpose of this report is to gather together comprehensive information concerning FSW as an optional welding method for welding the nuclear waste copper canister at the disposal facility. This report discusses the current situation, knowledge of the process and information concerning results of the development and research work related to welding thick copper and the special needs of the disposal environment. Most of the research work and development work has been done by Posiva's Swedish partner SKB, Swedish Nuclear Fuel and Waste Management Co. SKB chose FSW as their reference welding method in 2005. FSW (friction stir welding) is a solid-state welding method, invented in 1991, in which frictional heat is generated between the tool and the weld metal, causing the metal to soften, normally without reaching the melting point, and allowing the tool to traverse the joint line. Friction stir welding can be used for joining many types of materials and material combinations, if the tool materials and designs can be found which operate at the forging temperature of the workpiece. The general requirements for the copper canister weld and base material are presented in Posiva's VAHA-system, which sets the most critical values or demands concerning the short- and long-term properties or other needs. The sections in this report are set out in a similar way as in the VAHA-system. Concerning the results from the research and development work, it can be said that FS weld material fulfils the values set by VAHA. The quality of the welds fulfils the set demands for intact weld material and the welding process is robust using an automatic control system. There still remains work concerning the acceptance procedure for the welding process and other open issues which are described in this report. (orig.)

  13. Effect of various coal contaminants on the performance of solid oxide fuel cells: Part I. Accelerated testing

    Energy Technology Data Exchange (ETDEWEB)

    Bao, JianEr; Krishnan, Gopala N.; Jayaweera, Palitha; Perez-Mariano, Jordi; Sanjurjo, Angel [SRI International, 333 Ravenswood Ave, Menlo Park, CA 94025 (United States)

    2009-09-05

    The contaminants that are potentially present in the coal-derived gas stream and their thermochemical nature are discussed. Accelerated testing was carried out on Ni-YSZ/YSZ/LSM solid oxide fuel cells (YSZ: yttria stabilized zirconia and LSM: lanthanum strontium manganese oxide) for eight main kind of contaminants: CH{sub 3}Cl, HCl, As, P, Zn, Hg, Cd and Sb at the temperature range of 750-850 C. The As and P species, at 10 and 35 ppm, respectively, resulted in severe power density degradation at temperatures 800 C and below. SEM and EDX analysis indicated that As attacked the Ni region of the anode surface and the Ni current collector, caused the break of the current collector and the eventual cell failure at 800 C. The phosphorous containing species were found in the bulk of the anode, they were segregated and formed ''grain boundary'' like phases separating large Ni patches. These species are presumably nickel phosphide/phosphate and zirconia phosphate, which could break the Ni network for electron transport and inhibit the YSZ network for oxygen ion transport. The presence of 40 ppm CH{sub 3}Cl and 5 ppm Cd only affected the cell power density at above 800 C and Cd caused significant performance loss. Whereas the presence of 9 ppm Zn, 7 ppm Hg and 8 ppm Sb only degraded the cell power density by less than 1% during the 100 h test in the temperature range of 750-850 C. (author)

  14. A Systematic Study of Separators in Air-Breathing Flat-Plate Microbial Fuel Cells—Part 1: Structure, Properties, and Performance Correlations

    Directory of Open Access Journals (Sweden)

    Sona Kazemi

    2016-01-01

    Full Text Available Passive air-breathing microbial fuel cells (MFCs are a promising technology for energy recovery from wastewater and their performance is highly dependent on characteristics of the separator that isolates the anaerobic anode from the air-breathing cathode. The goal of the present work is to systematically study the separator characteristics and its effect on the performance of passive air-breathing flat-plate MFCs (FPMFCs. This was performed through characterization of structure, properties, and performance correlations of eight separators in Part 1 of this work. Eight commercial separators were characterized, in non-inoculated and inoculated setups, and were examined in passive air-breathing FPMFCs with different electrode spacing. The results showed a decrease in the peak power density as the oxygen and ethanol mass transfer coefficients in the separators increased, due to the increase of mixed potentials especially at smaller electrode spacing. Increasing the electrode spacing was therefore desirable for the application of diaphragms. The highest peak power density was measured using Nafion®117 with minimal electrode spacing, whereas using Nafion®117 or Celgard® with larger electrode spacing resulted in similar peak powers. Part 2 of this work focuses on numerical modelling of the FPMFCs based on mixed potential theory, implementing the experimental data from Part 1.

  15. Nuclear fuel technology - Tank calibration and volume determination for nuclear materials accountancy - Part 2: Data standardization for tank calibration

    International Nuclear Information System (INIS)

    2007-01-01

    Measurements of the volume and height of liquid in a process accountancy tank are often made in order to estimate or verify the tank's calibration or volume measurement equation. The calibration equation relates the response of the tank's measurement system to some independent measure of tank volume. The ultimate purpose of the calibration exercise is to estimate the tank's volume measurement equation (the inverse of the calibration equation), which relates tank volume to measurement system response. In this part of ISO 18213, it is assumed that the primary measurement-system response variable is liquid height and that the primary measure of liquid content is volume. This part of ISO 18213 presents procedures for standardizing a set of calibration data to a fixed set of reference conditions so as to minimize the effect of variations in ambient conditions that occur during the measurement process. The procedures presented herein apply generally to measurements of liquid height and volume obtained for the purpose of calibrating a tank (i.e. calibrating a tank's measurement system). When used in connection with other parts of ISO 18213, these procedures apply specifically to tanks equipped with bubbler probe systems for measuring liquid content. The standardization algorithms presented herein can be profitably applied when only estimates of ambient conditions, such as temperature, are available. However, the most reliable results are obtained when relevant ambient conditions are measured for each measurement of volume and liquid height in a set of calibration data. Information is provided on scope, physical principles, data required, calibration data, dimensional changes in the tank, multiple calibration runs and results on standardized calibration data. Four annexes inform about density of water, buoyancy corrections for mass determination, determination of tank heel volume and statistical method for aligning data from several calibration runs. A bibliography is

  16. The effect of early physiotherapy on the recovery of mandibular function after orthognathic surgery for class III correction. Part II: electromyographic activity of masticatory muscles.

    Science.gov (United States)

    Ko, Ellen Wen-Ching; Teng, Terry Te-Yi; Huang, Chiung Shing; Chen, Yu-Ray

    2015-01-01

    The study was conducted to evaluate the effect of early physical rehabilitation by comparing the differences of surface electromyographic (sEMG) activity in the masseter and anterior temporalis muscles after surgical correction of skeletal class III malocclusion. The prospective study included 63 patients; the experimental groups contained 31 patients who received early systematic physical rehabilitation; the control group (32 patients) did not receive physiotherapy. The amplitude of sEMG in the masticatory muscles reached 72.6-121.3% and 37.5-64.6% of pre-surgical values in the experimental and control groups respectively at 6 weeks after orthognathic surgery (OGS). At 6 months after OGS, the sEMG reached 135.1-233.4% and 89.6-122.5% of pre-surgical values in the experimental and control groups respectively. Most variables in the sEMG examination indicated that recovery of the masticatory muscles in the experimental group was better than the control group as estimated in the early phase (T1 to T2) and the total phase (T1 to T3); there were no significant differences between the mean recovery percentages in the later phase (T2 to T3). Early physical rehabilitative therapy is helpful for early recovery of muscle activity in masticatory muscles after OGS. After termination of physical therapy, no significant difference in recovery was indicated in patients with or without early physiotherapy. Copyright © 2014 European Association for Cranio-Maxillo-Facial Surgery. Published by Elsevier Ltd. All rights reserved.

  17. Fabrication and testing of ceramic UO{sub 2} fuel - I-III. Part I; Izrada i ispitivanje keramickog goriva na bazi UO{sub 2}- I-III, I Deo

    Energy Technology Data Exchange (ETDEWEB)

    Novakovic, M [Institute of Nuclear Sciences Boris Kidric, Laboratorija za termotehniku reaktora, Vinca, Beograd (Serbia and Montenegro)

    1961-12-15

    The task described consists of the following: fabrication of UO{sub 2} with different granulation from uranyl nitrate by ammonia diuranate; determination of size and shape distributions of metal and ceramic powders; fabrication of sintered pressed samples UO{sub 2}; investigating the properties of sintered uranium dioxide dependent on the fabrication process; producing a vibrator for compacting UO{sub 2} powder. This volume includes reports on the first two tasks.

  18. The removal of toxic metals from liquid effluents by ion exchange resins. Part IV: Chromium(III)/H+ /Lewatit SP112; La eliminación de metales tóxicos presentes en efluentes líquidos mediante resinas de cambio iónico. Parte IV: cromo(III)/H+/Lewatit SP112

    Energy Technology Data Exchange (ETDEWEB)

    Alguacil, F.J.

    2017-09-01

    This investigation presented results on the removal of chromium(III), from aqueous solution in the 0-5 pH range, using Lewatit SP112 cationic exchange resin. Several aspects affecting the ion exchange process were evaluated, including: the influence of the stirring speed, temperature, pH of the solution, resin dosage and aqueous ionic strength. The selectivity of the system was tested against the presence of other metals in the aqueous solution, whereas the removal of chromium(III) from solutions was compared with results obtained using multiwalled carbon nanotubes as adsorbents. From the batch experimental data, best fit of the results is obtained with the Langmuir model, whereas the ion exchange process is best explained by the pseudo-second order model, moreover, experimental data responded well to the film-diffusion controlled model. Elution of the chromium(III) loaded into the resin is well accomplished by the use of sodium hydroxide solutions. [Spanish] En este trabajo se presentan los resultados obtenidos en la eliminación de cromo(III) de disoluciones acuosas (pH 0-5) mediante la resina de intercambio catiónico Lewatit SP112. Se han investigado algunas variables que pueden afectar al sistema: influencia de la agitación, temperatura, pH y fuerza iónica del medio acuoso y cantidad de resina; también se ha investigado acerca de la selectividad del sistema cuando otros metales están presentes en el medio acuoso, comparándose los resultados de la eliminación del cromo(III) usando la resina con los resultados obtenidos cuando se emplea otro adsorbente como son los nanotubos de carbono de pared múltiple. Los resultados experimentales indican que la carga del cromo(III) en la resina responde mejor al modelo de Langmuir, mientras que los modelos cinéticos indican que la carga del metal en la resina responde al modelo de pseudo-segundo orden y difusión en la capa límite. La elución del cromo(III) se realiza con disoluciones de hidróxid.

  19. A trigeneration system based on polymer electrolyte fuel cell and desiccant wheel – Part B: Overall system design and energy performance analysis

    International Nuclear Information System (INIS)

    Intini, M.; De Antonellis, S.; Joppolo, C.M.; Casalegno, A.

    2015-01-01

    Highlights: • Seasonal simulation of a trigeneration system for building air-conditioning. • Effects of technical constraints on trigeneration system power consumption. • Optimal PEMFC unit size for maximizing trigeneration primary energy savings. - Abstract: This paper represents the second part of a major work focusing on a trigeneration system integrating a low temperature polymer electrolyte fuel cell (PEMFC) and a desiccant wheel-based air handling unit. Low temperature PEMFC systems have a significant potential in combined heating, cooling and power applications. However cogenerated heat temperature is relatively low (up to 65–70 °C), resulting in low efficiency of the cooling process, and the fuel processor is far from being flexible, hindering the operation of the system at low load conditions. Therefore a trigeneration system based on PEMFC should be carefully designed through accurate simulation tools. In the current paper a detailed analysis of the energy performance of the trigenerative system is provided, taking into account constraints of real applications, such as PEMFC part load behavior, desiccant wheel effectiveness, heat storage losses and air handling unit electrical consumptions. The methodology adopted to model system components is deeply described. Energy simulations are performed on yearly basis with variable building air conditioning loads and climate conditions, in order to investigate the optimal trigenerative unit size. A sensitivity analysis on crucial design parameters is provided. It is shown that constrains of actual applications have relevant effects on system energy consumption, which is significantly far from expected values based on a simplified analysis. Primary energy savings can be positive in winter time if the ratio of PEMFC heating capacity to air conditioning peak heating load is close to 0.15. Instead on yearly basis primary energy savings cannot be achieved with present components performance. Positive savings

  20. Fermilab III

    International Nuclear Information System (INIS)

    Anon.

    1990-01-01

    The total ongoing plans for Fermilab are wrapped up in the Fermilab III scheme, centrepiece of which is the proposal for a new Main Injector. The Laboratory has been awarded a $200,000 Illinois grant which will be used to initiate environmental assessment and engineering design of the Main Injector, while a state review panel recommended that the project should also benefit from $2 million of funding

  1. Fermilab III

    Energy Technology Data Exchange (ETDEWEB)

    Anon.

    1990-09-15

    The total ongoing plans for Fermilab are wrapped up in the Fermilab III scheme, centrepiece of which is the proposal for a new Main Injector. The Laboratory has been awarded a $200,000 Illinois grant which will be used to initiate environmental assessment and engineering design of the Main Injector, while a state review panel recommended that the project should also benefit from $2 million of funding.

  2. Systematic Study of Separators in Air-Breathing Flat-Plate Microbial Fuel Cells—Part 2: Numerical Modeling

    Directory of Open Access Journals (Sweden)

    Sona Kazemi

    2016-01-01

    Full Text Available The separator plays a key role on the performance of passive air-breathing flat-plate MFCs (FPMFC as it isolates the anaerobic anode from the air-breathing cathode. The goal of the present work was to study the separator characteristics and its effect on the performance of passive air-breathing FPMFCs. This was performed partially through characterization of structure, properties, and performance correlations of eight separators presented in Part 1. Current work (Part 2 presents a numerical model developed based on the mixed potential theory to investigate the sensitivity of the electrode potentials and the power output to the separator characteristics. According to this numerical model, the decreased peak power results from an increase in the mass transfer coefficients of oxygen and ethanol, but mainly increasing mixed potentials at the anode by oxygen crossover. The model also indicates that the peak power is affected by the proton transport number of the separator, which affects the cathode pH. Anode pH, on the other hand, remains constant due to application of phosphate buffer solution as the electrolyte. Also according to this model, the peak power is not sensitive to the resistivity of the separator because of the overshadowing effect of the oxygen crossover.

  3. A Fully Nonmetallic Gas Turbine Engine Enabled by Additive Manufacturing of Ceramic Composites. Part III; Additive Manufacturing and Characterization of Ceramic Composites

    Science.gov (United States)

    Halbig, Michael C.; Grady, Joseph E.; Singh, Mrityunjay; Ramsey, Jack; Patterson, Clark; Santelle, Tom

    2015-01-01

    This publication is the third part of a three part report of the project entitled "A Fully Nonmetallic Gas Turbine Engine Enabled by Additive Manufacturing" funded by NASA Aeronautics Research Institute (NARI). The objective of this project was to conduct additive manufacturing to produce ceramic matrix composite materials and aircraft engine components by the binder jet process. Different SiC powders with median sizes ranging from 9.3 to 53.0 microns were investigated solely and in powder blends in order to maximize powder packing. Various infiltration approaches were investigated to include polycarbosilane (SMP-10), phenolic, and liquid silicon. Single infiltrations of SMP-10 and phenolic only slightly filled in the interior. When the SMP-10 was loaded with sub-micron sized SiC powders, the infiltrant gave a much better result of filling in the interior. Silicon carbide fibers were added to the powder bed to make ceramic matrix composite materials. Microscopy showed that the fibers were well distributed with no preferred orientation on the horizontal plane and fibers in the vertical plane were at angles as much as 45deg. Secondary infiltration steps were necessary to further densify the material. Two to three extra infiltration steps of SMP-10 increased the density by 0.20 to 0.55 g/cc. However, the highest densities achieved were 2.10 to 2.15 g/cc. Mechanical tests consisting of 4 point bend tests were conducted. Samples from the two CMC panels had higher strengths and strains to failure than the samples from the two nonfiber reinforced panels. The highest strengths were from Set N with 65 vol% fiber loading which had an average strength of 66 MPa. Analysis of the fracture surfaces did not reveal pullout of the reinforcing fibers. Blunt fiber failure suggested that there was not composite behavior. The binder jet additive manufacturing method was used to also demonstrate the fabrication of turbine engine vane components of two different designs and sizes. The

  4. Stochastic foundations of undulatory transport phenomena: generalized Poisson-Kac processes—part III extensions and applications to kinetic theory and transport

    Science.gov (United States)

    Giona, Massimiliano; Brasiello, Antonio; Crescitelli, Silvestro

    2017-08-01

    This third part extends the theory of Generalized Poisson-Kac (GPK) processes to nonlinear stochastic models and to a continuum of states. Nonlinearity is treated in two ways: (i) as a dependence of the parameters (intensity of the stochastic velocity, transition rates) of the stochastic perturbation on the state variable, similarly to the case of nonlinear Langevin equations, and (ii) as the dependence of the stochastic microdynamic equations of motion on the statistical description of the process itself (nonlinear Fokker-Planck-Kac models). Several numerical and physical examples illustrate the theory. Gathering nonlinearity and a continuum of states, GPK theory provides a stochastic derivation of the nonlinear Boltzmann equation, furnishing a positive answer to the Kac’s program in kinetic theory. The transition from stochastic microdynamics to transport theory within the framework of the GPK paradigm is also addressed.

  5. Stochastic foundations of undulatory transport phenomena: generalized Poisson–Kac processes—part III extensions and applications to kinetic theory and transport

    International Nuclear Information System (INIS)

    Giona, Massimiliano; Brasiello, Antonio; Crescitelli, Silvestro

    2017-01-01

    This third part extends the theory of Generalized Poisson–Kac (GPK) processes to nonlinear stochastic models and to a continuum of states. Nonlinearity is treated in two ways: (i) as a dependence of the parameters (intensity of the stochastic velocity, transition rates) of the stochastic perturbation on the state variable, similarly to the case of nonlinear Langevin equations, and (ii) as the dependence of the stochastic microdynamic equations of motion on the statistical description of the process itself (nonlinear Fokker–Planck–Kac models). Several numerical and physical examples illustrate the theory. Gathering nonlinearity and a continuum of states, GPK theory provides a stochastic derivation of the nonlinear Boltzmann equation, furnishing a positive answer to the Kac’s program in kinetic theory. The transition from stochastic microdynamics to transport theory within the framework of the GPK paradigm is also addressed. (paper)

  6. Methodology for the assessment of innovative nuclear reactors and fuel cycles. Report of Phase 1B (first part) of the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO)

    International Nuclear Information System (INIS)

    2004-12-01

    an innovative nuclear energy system (INS) to meet the overall target of sustainable energy supply. As well, the initial development of the INPRO method for the assessment of nuclear energy systems was carried out. The Basic Principles, User Requirements, and Criteria and the INPRO method of assessment, taken together, comprise the INPRO methodology. The INPRO methodology provides the possibility to take into account local, regional and global boundary conditions of IAEA Member States, including those of both developing and developed countries. Phase 1A was completed in June of 2003 with the publication of IAEA-TECDOC-1362, Guidance for the Evaluation of Innovative Nuclear Reactors and Fuel Cycles, which documented the results of the Phase 1A work. The next step of INPRO was immediately launched. In this step, referred to as Phase 1B (first part), INPRO arranged for some 14 case studies to be performed, by national teams or by individual experts from seven countries, to test and provide feedback on the applicability, consistency and completeness of the INPRO methodology. This report documents changes to the basic principles, user requirements, criteria and the method of assessment that resulted from the second step of INPRO (referred to as Phase 1B (first part)), which started in June 2003 and ended in December 2004. During this step, Member States and individual experts performed 14 case studies with the objective of testing and validating the INPRO methodology. Based on the feedback from these case studies and numerous consultancies mostly held at the IAEA, the INPRO methodology has been significantly updated and revised, as documented in this report. The ongoing and future activities of INPRO will lead to further modifications to the INPRO methodology, based on the feedback received from Member States in light of their experience in applying the methodology. Thus, additional reports will be issued, as appropriate, to update the INPRO methodology

  7. Reactor oscillator - I - III, Part III - Electronic device

    International Nuclear Information System (INIS)

    Lolic, B.; Jovanovic, S.

    1961-12-01

    This report describes functioning of the reactor oscillator electronic system. Two methods of oscillator operation were discussed. The first method is so called method of amplitude modulation of the reactor power, and the second newer method is phase method. Both methods are planned for the present reactor oscillator

  8. Solid oxide fuel cell short stack performance testing - Part A: Experimental analysis and μ-combined heat and power unit comparison

    Science.gov (United States)

    Mastropasqua, L.; Campanari, S.; Brouwer, J.

    2017-12-01

    The need to experimentally understand the detailed performance of SOFC stacks under operating conditions typical of commercial SOFC systems has prompted this two-part study. The steady state performance of a 6-cell short stack of yttria (Y2O3) stabilised zirconia (YSZ) with Ni/YSZ anodes and composite Sr-doped lanthanum manganite (LaMnO3, LSM)/YSZ cathodes is experimentally evaluated. In Part A, the stack characterisation is carried out by means of sensitivity analyses on the fuel utilisation factor and the steam-to-carbon ratio. Electrical and environmental performances are assessed and the results are compared with a commercial full-scale micro-CHP system, which comprises the same cells. The results show that the measured temperature dynamics of the short stack in a test stand environment are on the order of many minutes; therefore, one cannot neglect temperature dynamics for a precise measurement of the steady state polarisation behaviour. The overall polarisation performance is comparable to that of the full stack employed in the micro-CHP system, confirming the good representation that short-stack analyses can give of the entire SOFC module. The environmental performance is measured verifying the negligible values of NO emissions (<10 ppb) across the whole polarisation curve.

  9. Concerning major items in government ordinance requiring modification of part of enforcement regulation for law relating to control of nuclear material, nuclear fuel and nuclear reactor

    International Nuclear Information System (INIS)

    1989-01-01

    The report describes major items planned to be incorporated into the enforcement regulations for laws relating to control of nuclear material, nuclear fuel and nuclear reactor. The modifications have become necessary for the nation to conclude a nuclear material protection treaty with other countries. The modification include the definitions of 'special nuclear fuel substances' and 'special nuclear fuel substances' and 'special nuclear fuel substances subject to protection'. The modifications require that protective measures be taken when handling and transporting special nuclear fuel substances subject to protection. Transport of special nuclear fuel substances requires approval from the Prime Minister or Transport Minister. Transport of special nuclear fuel substances subject to protection should be conducted after notifying the prefectural Public Safety Commission. Transport of special nuclear fuel substances subject to protection requires the conclusion of arrangements among responsible persons and approval of them from the Prime Minister. (N.K.)

  10. Importance of characteristics and modalities of physical activity and exercise in the management of cardiovascular health in individuals with cardiovascular disease (Part III).

    Science.gov (United States)

    Vanhees, L; Rauch, B; Piepoli, M; van Buuren, F; Takken, T; Börjesson, M; Bjarnason-Wehrens, B; Doherty, P; Dugmore, D; Halle, M

    2012-12-01

    The beneficial effect of exercise training and exercise-based cardiac rehabilitation on symptom-free exercise capacity,cardiovascular and skeletal muscle function, quality of life, general healthy lifestyle, and reduction of depressive symptoms and psychosocial stress is nowadays well recognized. However, it remains largely obscure, which characteristics of physical activity (PA) and exercise training--frequency, intensity, time (duration), type (mode), and volume (dose: intensity x duration) of exercise--are the most effective. The present paper, therefore, will deal with these exercise characteristics in the management of individuals with cardiovascular disease, i.e. coronary artery disease and chronic heart failure patients, but also in patients with congenital or valvular heart disease. Based on the current literature, and if sufficient evidence is available, recommendations from the European Association on Cardiovascular Prevention and Rehabilitation are formulated regarding frequency, intensity, time and type of PA, and safety aspects during exercise inpatients with cardiovascular disease. This paper is the third in a series of three papers, all devoted to the same theme: the importance of the exercise characteristics in the management of cardiovascular health. Part I is directed to the general population and Part II to individuals with cardiovascular risk factors. In general, PA recommendations and exercise training programmes for patients with coronary artery disease or chronic heart failure need to be tailored to the individual's exercise capacity and risk profile, with the aim to reach and maintain the individually highest fitness level possible and to perform endurance exercise training 30–60 min daily (3–5 days per week) in combination with resistance training 2–3 times a week. Because of the frequently reported dose–response relationship between training effect and exercise intensity, one should seek sufficiently high training intensities

  11. The classification of the finite simple groups, number 7 part III, chapters 7-11 the generic case, stages 3b and 4a

    CERN Document Server

    Gorenstein, Daniel; Solomon, Ronald

    2018-01-01

    The classification of finite simple groups is a landmark result of modern mathematics. The multipart series of monographs which is being published by the AMS (Volume 40.1-40.7 and future volumes) represents the culmination of a century-long project involving the efforts of scores of mathematicians published in hundreds of journal articles, books, and doctoral theses, totaling an estimated 15,000 pages. This part 7 of the series is the middle of a trilogy (Volume 40.5, Volume 40.7, and forthcoming Volume 40.8) treating the Generic Case, i.e., the identification of the alternating groups of degree at least 13 and most of the finite simple groups of Lie type and Lie rank at least 4. Moreover, Volumes 40.4-40.8 of this series will provide a complete treatment of the simple groups of odd type, i.e., the alternating groups (with two exceptions) and the groups of Lie type defined over a finite field of odd order, as well as some of the sporadic simple groups. In particular, this volume completes the construction, be...

  12. Differences between easy- and difficult-to-mill chickpea (Cicer arietinum L.) genotypes. Part III: free sugar and non-starch polysaccharide composition.

    Science.gov (United States)

    Wood, Jennifer A; Knights, Edmund J; Campbell, Grant M; Choct, Mingan

    2014-05-01

    Parts I and II of this series of papers identified several associations between the ease of milling and the chemical compositions of different chickpea seed fractions. Non-starch polysaccharides were implicated; hence, this study examines the free sugars and sugar residues. Difficult milling is associated with: (1) lower glucose and xylose residues (less cellulose and xyloglucans) and more arabinose, rhamnose and uronic acid in the seed coat, suggesting a more flexible seed coat that resists cracking and decortication; (2) a higher content of soluble and insoluble non-starch polysaccharide fractions in the cotyledon periphery, supporting a pectic polysaccharide mechanism comprising arabinogalacturonan, homogalacturonan, rhamnogalalcturonan, and glucuronan backbone structures; (3) higher glucose and mannose residues in the cotyledon periphery, supporting a lectin-mediated mechanism of adhesion; and (4) higher arabinose and glucose residues in the cotyledon periphery, supporting a mechanism involving arabinogalactan-proteins. This series has shown that the chemical composition of chickpea does vary in ways that are consistent with physical explanations of how seed structure and properties relate to milling behaviour. Seed coat strength and flexibility, pectic polysaccharide binding, lectins and arabinogalactan-proteins have been implicated. Increased understanding in these mechanisms will allow breeding programmes to optimise milling performance in new cultivars. © 2013 Society of Chemical Industry.

  13. Winter ecology of the Porcupine caribou herd, Yukon: Part III, Role of day length in determining activity pattern and estimating percent lying

    Directory of Open Access Journals (Sweden)

    D. E. Russell

    1986-06-01

    Full Text Available Data on the activity pattern, proportion of time spent lying and the length of active and lying periods in winter are presented from a 3 year study on the Porcupine caribou herd. Animals were most active at sunrise and sunset resulting in from one (late fall, early and mid winter to two (early fall and late winter to three (spring intervening lying periods. Mean active/lying cycle length decreased from late fall (298 mm to early winter (238 min, increased to a peak in mid winter (340 min then declined in late winter (305 min and again in spring (240 min. Mean length of the lying period increased throughout the 3 winter months from 56 min m early winter to 114 min in mid winter and 153 min in late winter. The percent of the day animals spent lying decreased from fall to early winter, increased throughout the winter and declined in spring. This pattern was related, in part, to day length and was used to compare percent lying among herds. The relationship is suggested to be a means of comparing quality of winter ranges.

  14. 76 FR 18066 - Regulation of Fuels and Fuel Additives: Changes to Renewable Fuel Standard Program

    Science.gov (United States)

    2011-04-01

    ... ENVIRONMENTAL PROTECTION AGENCY 40 CFR Part 80 Regulation of Fuels and Fuel Additives: Changes to Renewable Fuel Standard Program CFR Correction In Title 40 of the Code of Federal Regulations, Parts 72 to...-generating foreign producers and importers of renewable fuels for which RINs have been generated by the...

  15. Acid-base titrations by stepwise addition of equal volumes of titrant with special reference to automatic titrations-III Presentation of a fully automatic titration apparatus and of results supporting the theories given in the preceding parts.

    Science.gov (United States)

    Pehrsson, L; Ingman, F

    1977-02-01

    This paper forms Part III of a series in which the first two parts describe methods for evaluating titrations performed by stepwise addition of equal volumes of titrant. The great advantage of these methods is that they do not require an accurate calibration of the electrode system. This property makes the methods very suitable for routine work. e.g., in automatic analysis. An apparatus for performing such titrations automatically is presented. Further, results of titrations of monoprotic acids, a diprotic acid, an ampholyte, a mixture of an acid with its conjugate base, and mixtures of two acids with a small difference between the stability constants are given. Most of these titrations cannot be evaluated by the Gran or Hofstee methods but yield results having errors of the order of 0.1% if the methods proposed in Parts I and II of this series are employed. The advantages of the method of stepwise addition of equal volumes of titrant combined with the proposed evaluation methods, in comparison with common methods such as titration to a preset pH, are that all the data are used in the evaluation, permitting a statistical treatment and giving better possibilities for tracing systematic errors.

  16. VTT ENIGMA Calculations for FUMEX-III CRP

    International Nuclear Information System (INIS)

    Tulkki, Ville

    2013-01-01

    International Atomic Energy Agency IAEA has initiated a string of Coordinated Research Programmes (CRPs) to enhance co-operation between fuel modellers. One of these CRPs, FUMEX-III, was ongoing during 2008 - 2011 and has provided material and incentive for assessment of the fuel codes. This report presents the Finnish FUMEX-III simulations performed with the VTT-modified ENIGMA v5.9b. The work has been done as a part of SAFIR2010 (SAfety of FInnish Reactors 2010) project POKEVA (years 2008 to 2010) and SAFIR2014 project PALAMA (year 2011). VTT Technical Research Centre of Finland received ENIGMA v5.9b from Nuclear Electric plc of the UK in 1992. Internal development has been on-going since then. 'ENIGMA v5.9b with VTT modifications' (from now on, 'ENIGMA' for short) is a separate and different program from British Energy's ENIGMA 5.14 and UK National Nuclear Laboratory's ENIGMA-B. The FUMEX-III work has been performed in tandem of VTT's internal review work attempting to catalogue the changes done since 1992 and to assess the current state of the code. Several individual internal reports detail the changes made and the individual model assessments done during the years. (author)

  17. Storage of Spent Nuclear Fuel. Specific Safety Guide

    International Nuclear Information System (INIS)

    2012-01-01

    This Safety Guide provides recommendations and guidance on the storage of spent nuclear fuel. It covers all types of storage facilities and all types of spent fuel from nuclear power plants and research reactors. It takes into consideration the longer storage periods that have become necessary owing to delays in the development of disposal facilities and the decrease in reprocessing activities. It also considers developments associated with nuclear fuel, such as higher enrichment, mixed oxide fuels and higher burnup. The Safety Guide is not intended to cover the storage of spent fuel if this is part of the operation of a nuclear power plant or spent fuel reprocessing facility. Guidance is provided on all stages for spent fuel storage facilities, from planning through siting and design to operation and decommissioning, and in particular retrieval of spent fuel. Contents: 1. Introduction; 2. Protection of human health and the environment; 3. Roles and responsibilities; 4. Management system; 5. Safety case and safety assessment; 6. General safety considerations for storage of spent fuel. Appendix I: Specific safety considerations for wet or dry storage of spent fuel; Appendix II: Conditions for specific types of fuel and additional considerations; Annex: I: Short term and long term storage; Annex II: Operational and safety considerations for wet and dry spent fuel storage facilities; Annex III: Examples of sections of operating procedures for a spent fuel storage facility; Annex IV: Site conditions, processes and events for consideration in a safety assessment (external human induced phenomena); Annex V: Site conditions, processes and events for consideration in a safety assessment (external natural phenomena); Annex VI: Site conditions, processes and events for consideration in a safety assessment (external human induced phenomena); Annex VII: Postulated initiating events for consideration in a safety assessment (internal phenomena).

  18. Fuel cells science and engineering. Materials, processes, systems and technology. Vol. 1

    Energy Technology Data Exchange (ETDEWEB)

    Stolten, Detlef; Emonts, Bernd (eds.) [Forschungszentrum Juelich GmbH (DE). Inst. fuer Energieforschung (IEF), Brennstoffzellen (IEF-3)

    2012-07-01

    The first volume is divided in four parts and 22 chapters. It is structured as follows: PART I: Technology. Chapter 1: Technical Advancement of Fuel-Cell Research and Development (Dr. Bernd Emonts, Ludger Blum, Thomas Grube, Werner Lehnert, Juergen Mergel, Martin Mueller and Ralf Peters); 2: Single-Chamber Fuel Cells (Teko W. Napporn and Melanie Kuhn); 3: Technology and Applications of Molten Carbonate Fuel Cells (Barbara Bosio, Elisabetta Arato and Paolo Greppi); 4: Alkaline Fuel Cells (Erich Guelzow); 5: Micro Fuel Cells (Ulf Groos and Dietmar Gerteisen); 6: Principles and Technology of Microbial Fuel Cells (Jan B. A. Arends, Joachim Desloover, Sebastia Puig and Willy Verstraete); 7: Micro-Reactors for Fuel Processing (Gunther Kolb); 8: Regenerative Fuel Cells (Martin Mueller). PART II: Materials and Production Processes. Chapter 9: Advances in Solid Oxide Fuel Cell Development between 1995 and 2010 at Forschungszentrum Juelich GmbH, Germany (Vincent Haanappel); 10: Solid Oxide Fuel Cell Electrode Fabrication by Infiltration (Evren Gunen); 11: Sealing Technology for Solid Oxide Fuel Cells (K. Scott Weil); 12: Phosphoric Acid, an Electrolyte for Fuel Cells - Temperature and Composition Dependence of Vapor Pressure and Proton Conductivity (Carsten Korte); 13: Materials and Coatings for Metallic Bipolar Plates in Polymer Electrolyte Membrane Fuel Cells (Heli Wang and John A. Turner); 14: Nanostructured Materials for Fuel Cells (John F. Elter); 15: Catalysis in Low-Temperature Fuel Cells - An Overview (Sabine Schimpf and Michael Bron). PART III: Analytics and Diagnostics. Chapter 16: Impedance Spectroscopy for High-Temperature Fuel Cells (Ellen Ivers-Tiffee, Andre Leonide, Helge Schichlein, Volker Sonn and Andre Weber); 17: Post-Test Characterization of Solid Oxide Fuel-Cell Stacks (Norbert H. Menzler and Peter Batfalsky); 18: In Situ Imaging at Large-Scale Facilities (Christian Toetzke, Ingo Manke and Werner Lehnert); 19: Analytics of Physical Properties of Low

  19. Fuel assembly

    International Nuclear Information System (INIS)

    Ueda, Sei; Ando, Ryohei; Mitsutake, Toru.

    1995-01-01

    The present invention concerns a fuel assembly suitable to a BWR-type reactor and improved especially with the nuclear characteristic, heat performance, hydraulic performance, dismantling or assembling performance and economical property. A part of poison rods are formed as a large-diameter/multi-region poison rods having a larger diameter than a fuel rod. A large number of fuel rods are disposed surrounding a large diameter water rod and a group of the large-diameter/multi-region poison rods in adjacent with the water rod. The large-diameter water rod has a burnable poison at the tube wall portion. At least a portion of the large-diameter poison rods has a coolant circulation portion allowing coolants to circulate therethrough. Since the large-diameter poison rods are disposed at a position of high neutron fluxes, a large neutron multiplication factor suppression effect can be provided, thereby enabling to reduce the number of burnable poison rods relative to fuels. As a result, power peaking in the fuel assembly is moderated and a greater amount of plutonium can be loaded. In addition the flow of cooling water which tends to gather around the large diameter water rod can be controlled to improve cooling performance of fuels. (N.H.)

  20. Energy Conversion Alternatives Study (ECAS), General Electric Phase 1. Volume 3: Energy conversion subsystems and components. Part 3: Gasification, process fuels, and balance of plant

    Science.gov (United States)

    Boothe, W. A.; Corman, J. C.; Johnson, G. G.; Cassel, T. A. V.

    1976-01-01

    Results are presented of an investigation of gasification and clean fuels from coal. Factors discussed include: coal and coal transportation costs; clean liquid and gas fuel process efficiencies and costs; and cost, performance, and environmental intrusion elements of the integrated low-Btu coal gasification system. Cost estimates for the balance-of-plant requirements associated with advanced energy conversion systems utilizing coal or coal-derived fuels are included.

  1. Fabrication and testing of the sintered ceramic UO{sub 2} fuel - I - III, Part III - testing of sintered uranium dioxide properties dependent on the fabrication procedure; Izrada i ispitivanje keramickog goriva na bazi UO{sub 2}- I-III, III Deo - Ispitivanje osobina sinterovanog urandioksida u zavisnosti od procesa dobijanja

    Energy Technology Data Exchange (ETDEWEB)

    Novakovic, M; Ristic, M M [Institute of Nuclear Sciences Boris Kidric, Laboratorija za termotehniku reaktora, Vinca, Beograd (Serbia and Montenegro)

    1961-12-15

    The objective of this task was testing the influence of some parameters on the properties of sintered UO{sub 2}. The influence of parameters tested were as follows: adhesives; pressure in the pressing procedure; temperature of sintering of the UO{sub 2} powder. Other parameters were chosen according to the theoretical study. Sintering was done in argon atmosphere. Characterization of the UO{sub 2} powder was performed meaning determining the needed chemical, physical and physico-chemical properties. Some new methods were developed within this task: SET method for measuring the specific surfaces, DTA, TGA, high-temperature torsion.

  2. The individual effects of cetane number, oxygen content or fuel properties on the ignition delay, combustion characteristics, and cyclic variation of a turbocharged CRDI diesel engine – Part 1

    International Nuclear Information System (INIS)

    Labeckas, Gvidonas; Slavinskas, Stasys; Kanapkienė, Irena

    2017-01-01

    running at the high speed of 2500 rpm mainly. Whereas fuel-oxygen content should be neither too high nor too low, but just enough to assure complete combustion and low cyclic variation. The differing properties of the fuel involving ethanol or biodiesel were a separate factor strongly affecting diffusive combustion and the coefficient of cyclic variation (COV). Developing trends in the combustion characteristics were used to interpret the resulting changes in engine performance, emissions, and smoke (Part 2).

  3. Fourier Transform Infrared Spectroscopy Part III. Applications.

    Science.gov (United States)

    Perkins, W. D.

    1987-01-01

    Discusses the use of the FT-IR spectrometer in analyses that were previously avoided. Examines some of the applications of this spectroscopy with aqueous solutions, circular internal reflection, samples with low transmission, diffuse reflectance, infrared emission, and the infrared microscope. (TW)

  4. [Medicine in notafilia--Part III].

    Science.gov (United States)

    Babić, Rade R; Babić, Gordana Stanković

    2013-01-01

    Notafilia is the study of paper money. Only a few countries in the world have issued banknotes with portraits of well-known scientists who brought international fame to their own people and medicine. PORTRAITS OF SCIENTISTS ON THE BANKNOTES OF YUGOSLAVIA, SERBIA AND MONTENEGRO AND SERBIA. Nikola Tesla and Mihailo Pupin Idvorski were the ingenious inventors and scientists of our time who made special contributions to radiology. Nikola Tesla (1856-1943) pioneered the use of X-rays for medical purposes, thus effectively laying the foundations of radiology and radiography, and revealed the existence of harmful effects of X-rays on the human body. Mihailo Pupin Idvorski (1854-1935) was worldwide famous for applying physics in practice, as well as in the basis of telephone and telegraph transmissions. He also studied the nature of X-rays and contributed to establishing of radiology. PORTRAITS OF SCIENTISTS ON THE BANKNOTES OF THE WORLD: Maria Sklodowska Curie (1867-1934) was the first woman to gain the academic title of the Academy of Medicine, Paris. Together with her husband Pierre Curie (1859-1906) she gave an outstanding contribution to science and medicine. The discovery of the radioactive elements introduced the concept of "radioactivity" into physics and "radiotherapy" as a new discipline in medicine, thus creating the conditions for the development of nuclear medicine, oncology, and mobile diagnostic radiology. This paper presents the banknotes featuring the portraits of Nikola Tesla, Mihailo Pupin Idvorski, Maria Sklodowska Curie and Pierre Curie, the world renowned scientists, who made enormous contributions to medicine and laid the foundation for radiology.

  5. 'Riesenrad' ion gantry for hadrontherapy: Part III

    International Nuclear Information System (INIS)

    Benedikt, M.; Bryant, P.; Holy, P.; Pullia, M.

    1999-01-01

    When using accelerator beams for cancer therapy, the three-dimensional freedom afforded by a gantry helps the treatment planner to spread out surface doses, avoid directions that intercept vital organs and irradiate a volume that is conformal with the tumour. The general preference is for an iso-centric gantry turning 360 deg. in the vertical plane around the patient bed with sufficient space to be able to orientate the patient through 360 deg. in the horizontal plane. For hadrontherapy, gantries are impressive structures of the order of 10 m in diameter and 100 ton in weight and to date only proton gantries have been demonstrated to operate satisfactorily. The increased magnetic rigidity of say carbon ions will make ion gantries more difficult and costly to build. For this reason, exo-centric gantries and, in particular the so-called 'Riesenrad' gantry with a single 90 deg. bending magnet, merit further attention. The power consumption is reduced and the heavy magnets with their counterbalance weight are reduced and are kept close to the axis. The treatment room, which is lighter, is positioned at a larger radius, but only the patient bed requires careful alignment. An optics module called a 'rotator' is needed to match an incoming dispersion vector to the gantry in order to have an achromatic beam at the patient. A practical design is described that assumes the beam is derived from a slow-extraction scheme in a synchrotron and that the beam sizes are controlled by modules in the transfer line. Magnetic scanning is integrated into the gantry optics for both transverse directions

  6. Reactor oscillator - I - III, Part I

    International Nuclear Information System (INIS)

    Lolic, B.

    1961-12-01

    Project 'Reactor oscillator' covers the following activities: designing reactor oscillators for reactors RA and RB with detailed engineering drawings; constructing and mounting of the oscillator; designing and constructing the appropriate electronic equipment for the oscillator; measurements at the RA and RB reactors needed for completing the oscillator construction

  7. `Riesenrad' ion gantry for hadrontherapy: Part III

    Science.gov (United States)

    Benedikt, M.; Bryant, P.; Holy, P.; Pullia, M.

    1999-07-01

    When using accelerator beams for cancer therapy, the three-dimensional freedom afforded by a gantry helps the treatment planner to spread out surface doses, avoid directions that intercept vital organs and irradiate a volume that is conformal with the tumour. The general preference is for an iso-centric gantry turning 360° in the vertical plane around the patient bed with sufficient space to be able to orientate the patient through 360° in the horizontal plane. For hadrontherapy, gantries are impressive structures of the order of 10 m in diameter and 100 ton in weight and to date only proton gantries have been demonstrated to operate satisfactorily. The increased magnetic rigidity of say carbon ions will make ion gantries more difficult and costly to build. For this reason, exo-centric gantries and, in particular the so-called `Riesenrad' gantry with a single 90° bending magnet, merit further attention. The power consumption is reduced and the heavy magnets with their counterbalance weight are reduced and are kept close to the axis. The treatment room, which is lighter, is positioned at a larger radius, but only the patient bed requires careful alignment. An optics module called a `rotator' is needed to match an incoming dispersion vector to the gantry in order to have an achromatic beam at the patient. A practical design is described that assumes the beam is derived from a slow-extraction scheme in a synchrotron and that the beam sizes are controlled by modules in the transfer line. Magnetic scanning is integrated into the gantry optics for both transverse directions.

  8. Topics in Finance Part III--Leverage

    Science.gov (United States)

    Laux, Judy

    2010-01-01

    This article investigates operating and financial leverage from the perspective of the financial manager, accenting the relationships to stockholder wealth maximization (SWM), risk and return, and potential agency problems. It also covers some of the pertinent literature related specifically to the implications of operating and financial risk and…

  9. Multibarrier waste forms. Part III: Process considerations

    International Nuclear Information System (INIS)

    Lokken, R.O.

    1979-10-01

    The multibarrier concept for the solidification and storage of radioactive waste utilizes up to three barriers to isolate radionuclides from the environment: a solidified waste inner core, an impervious coating, and a metal matrix. The coating and metal matrix give the composite waste form enhanced inertness with improvements in thermal stability, mechanical strength, and leach resistance. Preliminary process flow rates and material costs were evaluated for four multibarrier waste forms with the process complexity increasing thusly: glass marbles, uncoated supercalcine, glass-coated supercalcine, and PyC/Al 2 O 3 -coated supercalcine. This report discusses the process variables and their effect on optimization of product quality, processing simplicity, and material cost. 11 figures, 2 tables

  10. Fuel vapor pressure (FVAPRS)

    International Nuclear Information System (INIS)

    Mason, R.E.

    1979-04-01

    A subcode (FVAPRS) is described which calculates fuel vapor pressure. This subcode was developed as part of the fuel rod behavior modeling task performed at EG and G Idaho, Inc. The fuel vapor pressure subcode (FVAPRS), is presented and a discussion of literature data, steady state and transient fuel vapor pressure equations and estimates of the standard error of estimate to be expected with the FVAPRS subcode are included

  11. Recent operational history of the new Sandia Pulsed Reactor III (SPR III)

    International Nuclear Information System (INIS)

    Schmidt, T.R.; Estes, B.F.; Reuscher, J.A.

    1977-01-01

    The Sandia Pulsed Reactor III (SPR III) is a fast-pulse research reactor which was designed and built at Sandia Laboratories and achieved criticality in August 1975. The reactor is now characterized and is in an operational configuration. The core consists of 18 fuel plates (258 kg fuel mass) of fully enriched uranium alloyed with 10 wt.% molybdenum. It is arranged in an annular configuration with an inside diameter of 17.78 cm, an outside diameter of 29.72 cm, and a height of 35.9 cm. The reactor core uses reflectors of copper and aluminum for control and an external bolting arrangement to secure the fuel plates. SPR III and SPR II are operated on an interchangeable basis using the same facility and control system. As of June 1977, SPR III has had over 240 operations with core temperatures up to 541 0 C

  12. HTGR Fuel performance basis

    International Nuclear Information System (INIS)

    Shamasundar, B.I.; Stansfield, O.M.; Jensen, D.D.

    1982-05-01

    The safety characteristics of the high-temperature gas-cooled reactor (HTGR) during normal and accident conditions are determined in part by HTGR fuel performance. During normal operation, less than 0.1% fuel failure occurs, primarily from defective particles. This low fuel failure fraction limits circulating activity to acceptable levels. During severe accidents, the radiological consequence is influenced by high-temperature fuel particle behavior. An empirical fuel failure model, supported by recent experimental data, is presented. The onset of significant fuel particle failure occurs at temperatures in excess of 1600 0 C, and complete fuel failure occurs at 2660 0 C. This indicates that the fuel is more retentive at higher temperatures than previously assumed. The more retentive nature of the fuel coupled with the high thermal capacitance of the core results in slow release of fission products from the core during severe accidents

  13. Towards an interpretation of the mechanism of the actinides(III)/lanthanides(III) separation by synergistic solvent extraction with nitrogen-containing polydendate ligands

    International Nuclear Information System (INIS)

    Francois, N.

    2000-01-01

    In the field of the separation of long-lived radionuclides from the wastes produced by nuclear fuel reprocessing, aromatic nitrogen-containing polydendate ligands are potential candidates for the selective extraction, alone or in synergistic mixture with acidic extractants, of trivalent actinides from trivalent lanthanides. The first part of this work deals with the complexation of trivalent f cations with various nitrogen-containing ligands (poly-pyridine analogues). Time-resolved laser-induced fluorimetry (TRLIF) and UV-visible spectrophotometry were used to determine the nature and evaluate the stability of each complex. Among the ligands studied, the least basic Me-Btp proved to be highly selective towards americium(III) in acidic solution. In the second part, two synergistic systems (nitrogen-containing polydendate ligand and lipophilic carboxylic acid) are studied and compared in regard to the extraction and separation of lanthanides(III) and actinides(III). TRLIF and gamma spectrometry allowed the nature of the extracted complexes and the optimal conditions of efficiency of both systems to be determined. Comparison between these different studies showed that the selectivity of complexation of trivalent f cations by a given nitrogen-containing polydendate ligand could not always be linked to the Am(III)Eu(III) selectivity reached in synergistic extraction. The latter depends on the 'balance' between the acid-basic properties on the one hand, and on the hard-soft characteristics on the other hand, of both components of synergistic system. (author)

  14. Standard Compliance: Guidelines to Help State and Alternative Fuel Provider Fleets Meet Their Energy Policy Act Requirements, 10 CFR Part 490 (Book)

    Energy Technology Data Exchange (ETDEWEB)

    2014-03-01

    This guidebook addresses the primary requirements of the Alternative Fuel Transportation Program to help state and alternative fuel provider fleets comply with the Energy Policy Act via the Standard Compliance option. It also addresses the topics that covered fleets ask about most frequently.

  15. AUTOMOTIVE DIESEL MAINTENANCE 1. UNIT XXIII, I--MAINTAINING THE FUEL SYSTEM, PART II--CATERPILLAR DIESEL ENGINE, II--UNDERSTANDING STEERING SYSTEMS.

    Science.gov (United States)

    Minnesota State Dept. of Education, St. Paul. Div. of Vocational and Technical Education.

    THIS MODULE OF A 30-MODULE COURSE IS DESIGNED TO DEVELOP AN UNDERSTANDING OF THE OPERATION AND MAINTENANCE OF THE DIESEL ENGINE FUEL INJECTION SYSTEM AND THE STEERING SYSTEM OF DIESEL POWERED VEHICLES. TOPICS ARE FUEL INJECTION SECTION, AND DESCRIPTION OF THE STEERING SYSTEM. THE MODULE CONSISTS OF A SELF-INSTRUCTIONAL BRANCH PROGRAMED TRAINING…

  16. Development of metal fuel and study of construction materials (I-IV), Part II; Razvoj metalnog goriva i ispitivanje konstrukcionih materijala (I-VI deo); II deo

    Energy Technology Data Exchange (ETDEWEB)

    Mihajlovic, A [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1965-11-15

    The studies were devoted to problems related to application of metal uranium as fuel in heavy water reactors. Influence of thermal treatment on material texture and recrystallization of cast uranium was investigated. Structural changes of uranium alloys with molybdenum and niobium were tested during different heat treatments. A review of the possibilities for using metal uranium fuel in heavy water reactors is included.

  17. Environmental impact data for fuels. Part 1: Main report. Resource consumption and emissions from the entire life cycle (New revised edition)

    International Nuclear Information System (INIS)

    Uppenberg, S.; Almemark, M.; Brandel, M.; Lindfors, L.G.; Marcus, H.O.; Stripple, H.; Wachtmeister, A.; Zetterberg, L.

    2001-05-01

    This report is a compilation of data concerning environmental impacts from the utilization of different fuels. The entire life cycle is studied, from the extraction of raw materials to combustion. The fuels under study are gasoline, gasoline with MTBE, diesel, fuel oil, LPG, coal, natural gas, peat, refuse, ethanol, RME, DME, methane and wood fuels (forestry residues, Salix, pellets/briquettes). Utilization areas studied are heating plants, cogeneration plants, power plants, domestic boilers, and light and heavy vehicles. In this new edition, the following changes were made: New life cycle analyses have been included, a few new fuels added, electricity from hydroelectric plants, wind power plants and nuclear power plants have been included and some other minor changes

  18. Department of Energy Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs, Draft Environmental Impact Statement. Volume 1, Appendix D: Part A, Naval Spent Nuclear Fuel Management

    Energy Technology Data Exchange (ETDEWEB)

    1994-06-01

    Volume 1 to the Department of Energy`s Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Management Programs Environmental Impact Statement evaluates a range of alternatives for managing naval spent nuclear fuel expected to be removed from US Navy nuclear-powered vessels and prototype reactors through the year 2035. The Environmental Impact Statement (EIS) considers a range of alternatives for examining and storing naval spent nuclear fuel, including alternatives that terminate examination and involve storage close to the refueling or defueling site. The EIS covers the potential environmental impacts of each alternative, as well as cost impacts and impacts to the Naval Nuclear Propulsion Program mission. This Appendix covers aspects of the alternatives that involve managing naval spent nuclear fuel at four naval shipyards and the Naval Nuclear Propulsion Program Kesselring Site in West Milton, New York. This Appendix also covers the impacts of alternatives that involve examining naval spent nuclear fuel at the Expended Core Facility in Idaho and the potential impacts of constructing and operating an inspection facility at any of the Department of Energy (DOE) facilities considered in the EIS. This Appendix also considers the impacts of the alternative involving limited spent nuclear fuel examinations at Puget Sound Naval Shipyard. This Appendix does not address the impacts associated with storing naval spent nuclear fuel after it has been inspected and transferred to DOE facilities. These impacts are addressed in separate appendices for each DOE site.

  19. Rise, fall and resurrection of chromosome territories: a historical perspective. Part II. Fall and resurrection of chromosome territories during the 1950s to 1980s. Part III. Chromosome territories and the functional nuclear architecture: experiments and models from the 1990s to the present.

    Science.gov (United States)

    Cremer, T; Cremer, C

    2006-01-01

    Part II of this historical review on the progress of nuclear architecture studies points out why the original hypothesis of chromosome territories from Carl Rabl and Theodor Boveri (described in part I) was abandoned during the 1950s and finally proven by compelling evidence forwarded by laser-uv-microbeam studies and in situ hybridization experiments. Part II also includes a section on the development of advanced light microscopic techniques breaking the classical Abbe limit written for readers with little knowledge about the present state of the theory of light microscopic resolution. These developments have made it possible to perform 3D distance measurements between genes or other specifically stained, nuclear structures with high precision at the nanometer scale. Moreover, it has become possible to record full images from fluorescent structures and perform quantitative measurements of their shapes and volumes at a level of resolution that until recently could only be achieved by electron microscopy. In part III we review the development of experiments and models of nuclear architecture since the 1990s. Emphasis is laid on the still strongly conflicting views about the basic principles of higher order chromatin organization. A concluding section explains what needs to be done to resolve these conflicts and to come closer to the final goal of all studies of the nuclear architecture, namely to understand the implications of nuclear architecture for nuclear functions.

  20. Transportation fuel production from gasified biomass integrated with a pulp and paper mill - Part B: Analysis of economic performance and greenhouse gas emissions

    International Nuclear Information System (INIS)

    Isaksson, Johan; Jansson, Mikael; Åsblad, Anders; Berntsson, Thore

    2016-01-01

    This paper presents a comparison between four gasification-based biorefineries integrated with a pulp and paper mill. It is a continuation of 'Transportation fuel production from gasified biomass integrated with a pulp and paper mill - Part A: Heat integration and system performance'. Synthesis into methanol, Fischer-Tropsch crude or synthetic natural gas, or electricity generation in a gas turbine combined cycle, were evaluated. The concepts were assessed in terms of GHG (greenhouse gas) emissions and economic performance. Net annual profits were positive for all biofuel cases for an annuity factor of 0.1 in the year 2030; however, the results are sensitive to biofuel selling prices and CO_2_,_e_q charge. Additionally, GHG emissions from grid electricity are highly influential on the results since all biofuel processes require external power. Credits for stored CO_2 might be necessary for processes to be competitive, i.e. storage of separated CO_2 from the syngas conditioning has an important role to play. Without CO_2 storage, the gas turbine case is better than, or equal to, biofuels regarding GHG emissions. Efficiency measures at the host mill prior to heat integration of a gasification process are beneficial from the perspective of GHG emissions, while having a negative impact on the economy. - Highlights: • Biomass gasification integrated with a pulp and paper mill was evaluated. • Greenhouse gas emission consequences and economic performance were assessed. • CCS has an important role to play, both in terms of emissions and economy. • Green electricity production is competitive compared to biofuel production in terms of GHG. • All biofuel cases are profitable in 2030 with assumed level of future policy instruments.

  1. CONSIDERATIONS FOR THE DEVELOPMENT OF A DEVICE FOR THE DECOMMISSIONING OF THE HORIZONTAL FUEL CHANNELS IN THE CANDU 6 NUCLEAR REACTOR. PART 6 - PRESENTATION OF THE DECOMMISSIONING DEVICE

    Directory of Open Access Journals (Sweden)

    Gabi ROSCA FARTAT

    2015-05-01

    Full Text Available The objective of this paper is to present a possible solution for the designing of a device for the decommissioning of the horizontal fuel channels in the CANDU 6 nuclear reactor. The decommissioning activities are dismantling, demolition, controlled removal of equipment, components, conventional or hazardous waste (radioactive, toxic in compliance with the international basic safety standards on radiation protection. One as the most important operation in the final phase of the nuclear reactor dismantling is the decommissioning of fuel channels. For the fuel channels decommissioning should be taken into account the detailed description of the fuel channel and its components, the installation documents history, adequate radiological criteria for decommissioning guidance, safety and environmental impact assessment, including radiological and non-radiological analysis of the risks that can occur for workers, public and environment, the description of the proposed program for decommissioning the fuel channel and its components, the description of the quality assurance program and of the monitoring program, the equipments and methods used to verify the compliance with the decommissioning criteria, the planning of performing the final radiological assessment at the end of the fuel channel decommissioning. These will include also, a description of the proposed radiation protection procedures to be used during decommissioning. The dismantling of the fuel channel is performed by one device which shall provide radiation protection during the stages of decommissioning, ensuring radiation protection of the workers. The device shall be designed according to the radiation protection procedures. The decommissioning device assembly of the fuel channel components is composed of the device itself and moving platform support for coupling of the selected channel to be dismantled. The fuel channel decommissioning device is an autonomous device designed for

  2. The importance of chemical analysis and quality-assurance for environmental mediation and acceptance in the use of secondary fuels in the cement plant Leube as an essential part of the ecological business strategy

    International Nuclear Information System (INIS)

    Mlekusch, T.

    2000-11-01

    In this paper I tried to combine the fields of environmental technology, environmental analysis, quality-management, environmental mediation and limiting values an the example of the use of secondary-fuels in the cement plant Leube in Gartenau (Austria). This project has proved to be very satisfying for the company and the adjoining owners. The first chapter deals with relations between environment and technology. A chronological description of the use of secondary fuels shows the development in this technology. The reactions of trace-elements in the burning process of cement clinker are described in the second chapter. Limiting values and the protection of human beings by the use of them are explained in the third chapter. The environmental mediation project of the use of secondary fuels in Gartenau is object of the fourth chapter. An essential part to acceptance is contributed by a good working quality-management and laboratory. (author)

  3. A prospective, randomised, controlled, double-blind phase I-II clinical trial on the safety of A-Part® Gel as adhesion prophylaxis after major abdominal surgery versus non-treated group

    Directory of Open Access Journals (Sweden)

    Weis Christine

    2010-07-01

    Full Text Available Abstract Background Postoperative adhesions occur when fibrous strands of internal scar tissue bind anatomical structures to one another. The most common cause of intra-abdominal adhesions is previous intra-abdominal surgical intervention. Up to 74% of intestinal obstructions are caused by post surgical adhesions. Although a variety of methods and agents have been investigated to prevent post surgical adhesions, the problem of peritoneal adhesions remains largely unsolved. Materials serving as an adhesion barrier are much needed. Methods/Design This is a prospective, randomised, controlled, patient blinded and observer blinded, single centre phase I-II trial, which evaluates the safety of A-Part® Gel as an adhesion prophylaxis after major abdominal wall surgery, in comparison to an untreated control group. 60 patients undergoing an elective median laparotomy without prior abdominal surgery are randomly allocated into two groups of a 1:1- ratio. Safety parameter and primary endpoint of the study is the occurrence of wound healing impairment or peritonitis within 28 (+10 days after surgery. The frequency of anastomotic leakage within 28 days after operation, occurrence of adverse and serious adverse events during hospital stay up to 3 months and the rate of adhesions along the scar within 3 months are defined as secondary endpoints. After hospital discharge the investigator will examine the enrolled patients at 28 (+10 days and 3 months (±14 days after surgery. Discussion This trial aims to assess, whether the intra-peritoneal application of A-Part® Gel is safe and efficacious in the prevention of post-surgical adhesions after median laparotomy, in comparison to untreated controls. Trial registration NCT00646412

  4. Cetuximab in combination with irinotecan/5-fluorouracil/folinic acid (FOLFIRI) in the initial treatment of metastatic colorectal cancer: a multicentre two-part phase I/II study

    International Nuclear Information System (INIS)

    Raoul, Jean-Luc; Van Laethem, Jean-Luc; Peeters, Marc; Brezault, Catherine; Husseini, Fares; Cals, Laurent; Nippgen, Johannes; Loos, Anja-Helena; Rougier, Philippe

    2009-01-01

    This study was designed to investigate the efficacy and safety of the epidermal growth factor receptor (EGFR) inhibitor cetuximab combined with irinotecan, folinic acid (FA) and two different doses of infusional 5-fluorouracil (5-FU) in the first-line treatment of EGFR-detectable metastatic colorectal cancer. The 5-FU dose was selected on the basis of dose-limiting toxicities (DLTs) during part I of the study. Patients received cetuximab (400 mg/m 2 initial dose and 250 mg/m 2 /week thereafter) and every 2 weeks irinotecan (180 mg/m 2 ), FA (400 mg/m 2 ) and 5-FU (either low dose [LD], 300 mg/m 2 bolus plus 2,000 mg/m 2 46-hour infusion, n = 7; or, high-dose [HD], 400 mg/m 2 bolus plus 2,400 mg/m 2 ; n = 45). Only two DLTs occurred in the HD group, and HD 5-FU was selected for use in part II. Apart from rash, commonly observed grade 3/4 adverse events such as leucopenia, diarrhoea, vomiting and asthenia occurred within the expected range for FOLFIRI. Among 52 patients, the overall response rate was 48%. Median progression-free survival (PFS) was 8.6 months (counting all reported progressions) and the median overall survival was 22.4 months. Treatment facilitated the resection of initially unresectable metastases in fourteen patients (27%): of these, 10 patients (71%) had no residual tumour after surgery, and these resections hindered the estimation of PFS. The combination of cetuximab and FOLFIRI was active and well tolerated in this setting. Initially unresectable metastases became resectable in one-quarter of patients, with a high number of complete resections, and these promising results formed the basis for the investigation of FOLFIRI with and without cetuximab in the phase III CRYSTAL trial

  5. Cetuximab in combination with irinotecan/5-fluorouracil/folinic acid (FOLFIRI in the initial treatment of metastatic colorectal cancer: a multicentre two-part phase I/II study

    Directory of Open Access Journals (Sweden)

    Cals Laurent

    2009-04-01

    Full Text Available Abstract Background This study was designed to investigate the efficacy and safety of the epidermal growth factor receptor (EGFR inhibitor cetuximab combined with irinotecan, folinic acid (FA and two different doses of infusional 5-fluorouracil (5-FU in the first-line treatment of EGFR-detectable metastatic colorectal cancer. Methods The 5-FU dose was selected on the basis of dose-limiting toxicities (DLTs during part I of the study. Patients received cetuximab (400 mg/m2 initial dose and 250 mg/m2/week thereafter and every 2 weeks irinotecan (180 mg/m2, FA (400 mg/m2 and 5-FU (either low dose [LD], 300 mg/m2 bolus plus 2,000 mg/m2 46-hour infusion, n = 7; or, high-dose [HD], 400 mg/m2 bolus plus 2,400 mg/m2; n = 45. Results Only two DLTs occurred in the HD group, and HD 5-FU was selected for use in part II. Apart from rash, commonly observed grade 3/4 adverse events such as leucopenia, diarrhoea, vomiting and asthenia occurred within the expected range for FOLFIRI. Among 52 patients, the overall response rate was 48%. Median progression-free survival (PFS was 8.6 months (counting all reported progressions and the median overall survival was 22.4 months. Treatment facilitated the resection of initially unresectable metastases in fourteen patients (27%: of these, 10 patients (71% had no residual tumour after surgery, and these resections hindered the estimation of PFS. Conclusion The combination of cetuximab and FOLFIRI was active and well tolerated in this setting. Initially unresectable metastases became resectable in one-quarter of patients, with a high number of complete resections, and these promising results formed the basis for the investigation of FOLFIRI with and without cetuximab in the phase III CRYSTAL trial.

  6. Fuel Exhaling Fuel Cell.

    Science.gov (United States)

    Manzoor Bhat, Zahid; Thimmappa, Ravikumar; Devendrachari, Mruthyunjayachari Chattanahalli; Kottaichamy, Alagar Raja; Shafi, Shahid Pottachola; Varhade, Swapnil; Gautam, Manu; Thotiyl, Musthafa Ottakam

    2018-01-18

    State-of-the-art proton exchange membrane fuel cells (PEMFCs) anodically inhale H 2 fuel and cathodically expel water molecules. We show an unprecedented fuel cell concept exhibiting cathodic fuel exhalation capability of anodically inhaled fuel, driven by the neutralization energy on decoupling the direct acid-base chemistry. The fuel exhaling fuel cell delivered a peak power density of 70 mW/cm 2 at a peak current density of 160 mA/cm 2 with a cathodic H 2 output of ∼80 mL in 1 h. We illustrate that the energy benefits from the same fuel stream can at least be doubled by directing it through proposed neutralization electrochemical cell prior to PEMFC in a tandem configuration.

  7. Fuel trading

    International Nuclear Information System (INIS)

    2015-01-01

    A first part of this report proposes an overview of trends and predictions. After a synthesis on the sector changes and trends, it indicates and comments the most recent predictions for the consumption of refined oil products and for the turnover of the fuel wholesale market, reports the main highlights concerning the sector's life, and gives a dashboard of the sector activity. The second part proposes the annual report on trends and competition. It presents the main operator profiles and fuel categories, the main determining factors of the activity, the evolution of the sector context between 2005 and 2015 (consumptions, prices, temperature evolution). It analyses the evolution of the sector activity and indicators (sales, turnovers, prices, imports). Financial performances of enterprises are presented. The economic structure of the sector is described (evolution of the economic fabric, structural characteristics, French foreign trade). Actors are then presented and ranked in terms of turnover, of added value, and of result

  8. Aspectos da Reologia e da Estabilidade de Suspensões Cerâmicas. Parte III: Mecanismo de Estabilização Eletroestérica de Suspensões com Alumina Aspects of Rheology and Stability of Ceramic Suspensions. Part III: Electrosteric Stabilization Mechanism of Alumina Suspensions

    Directory of Open Access Journals (Sweden)

    F. S. Ortega

    1997-08-01

    Full Text Available Esta terceira e última parte da revisão sobre os aspectos reológicos e de estabilização de suspensões com pós cerâmicos vem reunir a aplicação dos conhecimentos adquiridos nas primeiras duas partes publicadas anteriormente. Aqui, os fenômenos eletrostático devido à dupla camada elétrica, e estérico, relacionado à adsorção de moléculas poliméricas, são combinados para explicar o mecanismo eletroestérico de estabilização de suspensões cerâmicas. Os defloculantes que atuam através desse mecanismo abrangem uma classe específica de polímero denominada polieletrólitos, a qual é constituída por macromoléculas ionizáveis quando em solução. O estudo da forma com que os polieletrólitos atuam justifica-se devido à larga utilização desta classe de polímeros na indústria cerâmica. Os ácidos poliacrílico (PAA e polimetacrílico (PMAA são exemplos de polieletrólitos amplamente utilizados no processo de materiais à base de alumina. Dá-se destaque à influência do pH do meio e da presença de íons, sendo novamente aqui importante o conceito de força iônica da suspensão. Como aplicação prática, apresenta-se a estabilidade do sistema alumina-PMAA, reportando-se sobre o comportamento da viscosidade e da efetiva defloculação da suspensão. Este estudo é concluído apresentando resultados do efeito do peso molecular sobre a viscosidade, chamando atenção para o fato de que não basta definir apenas a classe de polímero a ser usada, sendo também fundamental especificar o peso molecular médio do polímero selecionado.The third and last part of this review about stabilization and rheological aspects of ceramic suspension gathers the knowledge in the two parts previously published. Here, the electrostatic and steric phenomena, related to the electrical double layer and polymeric molecules adsorption, respectively, are combined to explain the electrosteric stabilization mechanism of ceramic suspensions. The

  9. Towards an interpretation of the mechanism of the actinides(III)/lanthanides(III) separation by synergistic solvent extraction with nitrogen-containing polydendate ligands; Vers une interpretation des mecanismes de la separation actinides(III)/lanthanides(III) par extraction liquide-liquide synergique impliquant des ligands polyazotes

    Energy Technology Data Exchange (ETDEWEB)

    Francois, N [CEA/VALRHO - site de Marcoule, Dept. de Recherche en Retraitement et en Vitrification, (DRRV), 30 - Marcoule (France); Universite Henri Poincare, 54 - Vandoeuvre-les-Nancy (France)

    2000-07-01

    In the field of the separation of long-lived radionuclides from the wastes produced by nuclear fuel reprocessing, aromatic nitrogen-containing polydendate ligands are potential candidates for the selective extraction, alone or in synergistic mixture with acidic extractants, of trivalent actinides from trivalent lanthanides. The first part of this work deals with the complexation of trivalent f cations with various nitrogen-containing ligands (poly-pyridine analogues). Time-resolved laser-induced fluorimetry (TRLIF) and UV-visible spectrophotometry were used to determine the nature and evaluate the stability of each complex. Among the ligands studied, the least basic Me-Btp proved to be highly selective towards americium(III) in acidic solution. In the second part, two synergistic systems (nitrogen-containing polydendate ligand and lipophilic carboxylic acid) are studied and compared in regard to the extraction and separation of lanthanides(III) and actinides(III). TRLIF and gamma spectrometry allowed the nature of the extracted complexes and the optimal conditions of efficiency of both systems to be determined. Comparison between these different studies showed that the selectivity of complexation of trivalent f cations by a given nitrogen-containing polydendate ligand could not always be linked to the Am(III)Eu(III) selectivity reached in synergistic extraction. The latter depends on the 'balance' between the acid-basic properties on the one hand, and on the hard-soft characteristics on the other hand, of both components of synergistic system. (author)

  10. Towards an interpretation of the mechanism of the actinides(III)/lanthanides(III) separation by synergistic solvent extraction with nitrogen-containing polydendate ligands; Vers une interpretation des mecanismes de la separation actinides(III)/lanthanides(III) par extraction liquide-liquide synergique impliquant des ligands polyazotes

    Energy Technology Data Exchange (ETDEWEB)

    Francois, N. [CEA/VALRHO - site de Marcoule, Dept. de Recherche en Retraitement et en Vitrification, (DRRV), 30 - Marcoule (France); Universite Henri Poincare, 54 - Vandoeuvre-les-Nancy (France)

    2000-07-01

    In the field of the separation of long-lived radionuclides from the wastes produced by nuclear fuel reprocessing, aromatic nitrogen-containing polydendate ligands are potential candidates for the selective extraction, alone or in synergistic mixture with acidic extractants, of trivalent actinides from trivalent lanthanides. The first part of this work deals with the complexation of trivalent f cations with various nitrogen-containing ligands (poly-pyridine analogues). Time-resolved laser-induced fluorimetry (TRLIF) and UV-visible spectrophotometry were used to determine the nature and evaluate the stability of each complex. Among the ligands studied, the least basic Me-Btp proved to be highly selective towards americium(III) in acidic solution. In the second part, two synergistic systems (nitrogen-containing polydendate ligand and lipophilic carboxylic acid) are studied and compared in regard to the extraction and separation of lanthanides(III) and actinides(III). TRLIF and gamma spectrometry allowed the nature of the extracted complexes and the optimal conditions of efficiency of both systems to be determined. Comparison between these different studies showed that the selectivity of complexation of trivalent f cations by a given nitrogen-containing polydendate ligand could not always be linked to the Am(III)Eu(III) selectivity reached in synergistic extraction. The latter depends on the 'balance' between the acid-basic properties on the one hand, and on the hard-soft characteristics on the other hand, of both components of synergistic system. (author)

  11. Nuclear fuel performance in boiling water reactors

    International Nuclear Information System (INIS)

    Elkins, R.B.; Baily, W.E.; Proebstle, R.A.; Armijo, J.S.; Klepfer, H.H.

    1981-01-01

    A major development program is described to improve the performance of Boiling Water Reactor fuel. This sustained program is described in four parts: 1) performance monitoring, 2) fuel design changes, 3) plant operating recommendations, and 4) advanced fuel programs

  12. Uranium (III)-Plutonium (III) co-precipitation in molten chloride

    Science.gov (United States)

    Vigier, Jean-François; Laplace, Annabelle; Renard, Catherine; Miguirditchian, Manuel; Abraham, Francis

    2018-02-01

    Co-management of the actinides in an integrated closed fuel cycle by a pyrochemical process is studied at the laboratory scale in France in the CEA-ATALANTE facility. In this context the co-precipitation of U(III) and Pu(III) by wet argon sparging in LiCl-CaCl2 (30-70 mol%) molten salt at 705 °C is studied. Pu(III) is prepared in situ in the molten salt by carbochlorination of PuO2 and U(III) is then introduced as UCl3 after chlorine purge by argon to avoid any oxidation of uranium up to U(VI) by Cl2. The oxide conversion yield through wet argon sparging is quantitative. However, the preferential oxidation of U(III) in comparison to Pu(III) is responsible for a successive conversion of the two actinides, giving a mixture of UO2 and PuO2 oxides. Surprisingly, the conversion of sole Pu(III) in the same conditions leads to a mixture of PuO2 and PuOCl, characteristic of a partial oxidation of Pu(III) to Pu(IV). This is in contrast with coconversion of U(III)-Pu(III) mixtures but in agreement with the conversion of Ce(III).

  13. The Nuclear Fuel Cycle

    International Nuclear Information System (INIS)

    2011-08-01

    This brochure describes the nuclear fuel cycle, which is an industrial process involving various activities to produce electricity from uranium in nuclear power reactors. The cycle starts with the mining of uranium and ends with the disposal of nuclear waste. The raw material for today's nuclear fuel is uranium. It must be processed through a series of steps to produce an efficient fuel for generating electricity. Used fuel also needs to be taken care of for reuse and disposal. The nuclear fuel cycle includes the 'front end', i.e. preparation of the fuel, the 'service period' in which fuel is used during reactor operation to generate electricity, and the 'back end', i.e. the safe management of spent nuclear fuel including reprocessing and reuse and disposal. If spent fuel is not reprocessed, the fuel cycle is referred to as an 'open' or 'once-through' fuel cycle; if spent fuel is reprocessed, and partly reused, it is referred to as a 'closed' nuclear fuel cycle.

  14. MCTP, a code for the thermo-mechanical analysis of a fuel rod of BWR type reactors (Neutron part); MCTP, un codigo para el analisis termo-mecanico de una barra combustible de reactores tipo BWR (Parte Neutronica)

    Energy Technology Data Exchange (ETDEWEB)

    Hernandez L, H; Ortiz V, J [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

    2003-07-01

    In the National Institute of Nuclear Research of Mexico a code for the thermo-mechanical analysis of the fuel rods of the BWR type reactors of the Nucleo electric Central of Laguna Verde is developed. The code solves the diffusion equation in cylindrical coordinates with several energy groups. The code, likewise, calculates the temperature distribution and power distribution in those fuel rods. The code is denominated Multi groups With Temperatures and Power (MCTP). In the code, the energy with which the fission neutrons are emitted it is divided in six groups. They are also considered the produced perturbations by the changes in the temperatures of the materials that constitute the fuel rods, the content of fission products, the uranium consumption and in its case the gadolinium, as well as the plutonium production. In this work there are present preliminary results obtained with the code, using data of operation of the Nucleo electric Central of Laguna Verde. (Author)

  15. Dry process fuel performance technology development

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Kweon Ho; Kim, K. W.; Kim, B. K. (and others)

    2006-06-15

    The objective of the project is to establish the performance evaluation system of DUPIC fuel during the Phase III R and D. In order to fulfil this objectives, property model development of DUPIC fuel and irradiation test was carried out in Hanaro using the instrumented rig. Also, the analysis on the in-reactor behavior analysis of DUPIC fuel, out-pile test using simulated DUPIC fuel as well as performance and integrity assessment in a commercial reactor were performed during this Phase. The R and D results of the Phase III are summarized as follows: Fabrication process establishment of simulated DUPIC fuel for property measurement, Property model development for the DUPIC fuel, Performance evaluation of DUPIC fuel via irradiation test in Hanaro, Post irradiation examination of irradiated fuel and performance analysis, Development of DUPIC fuel performance code (KAOS)

  16. Dry process fuel performance technology development

    International Nuclear Information System (INIS)

    Kang, Kweon Ho; Kim, K. W.; Kim, B. K.

    2006-06-01

    The objective of the project is to establish the performance evaluation system of DUPIC fuel during the Phase III R and D. In order to fulfil this objectives, property model development of DUPIC fuel and irradiation test was carried out in Hanaro using the instrumented rig. Also, the analysis on the in-reactor behavior analysis of DUPIC fuel, out-pile test using simulated DUPIC fuel as well as performance and integrity assessment in a commercial reactor were performed during this Phase. The R and D results of the Phase III are summarized as follows: Fabrication process establishment of simulated DUPIC fuel for property measurement, Property model development for the DUPIC fuel, Performance evaluation of DUPIC fuel via irradiation test in Hanaro, Post irradiation examination of irradiated fuel and performance analysis, Development of DUPIC fuel performance code (KAOS)

  17. IAEA activities on nuclear fuel

    International Nuclear Information System (INIS)

    Basak, U.

    2011-01-01

    In this paper a brief description and the main objectives of IAEA Programme B on Nuclear fuel cycle are given. The following Coordinated Research Projects: 1) FUel performance at high burn-up and in ageing plant by management and optimisation of WAter Chemistry Technologies (FUWAC ); 2) Near Term and Promising Long Term Options for Deployment of Thorium Based Nuclear Energy; 3) Fuel Modelling (FUMEX-III) are shortly described. The data collected by the IAEA Expert Group of Fuel Failures in Water Cooled Reactors including information about fuel failure cause for PWR (1994-2006) and failure mechanisms for BWR fuel (1994-2006) are shown. The just published Fuel Failure Handbook as well as preparation of a Monograph on Zirconium including an overview of Zirconium for nuclear applications are presented. The current projects in Sub-programme B2 - Power Reactor Fuel Engineering are also listed

  18. Guidebook on spent fuel storage

    International Nuclear Information System (INIS)

    1984-01-01

    The Guidebook summarizes the experience and information in various areas related to spent fuel storage: technological aspects, the transport of spent fuel, economical, regulatory and institutional aspects, international safeguards, evaluation criteria for the selection of a specific spent fuel storage concept, international cooperation on spent fuel storage. The last part of the Guidebook presents specific problems on the spent fuel storage in the United Kingdom, Sweden, USSR, USA, Federal Republic of Germany and Switzerland

  19. Insight into the Extraction Mechanism of Americium(III) over Europium(III) with Pyridylpyrazole: A Relativistic Quantum Chemistry Study.

    Science.gov (United States)

    Kong, Xiang-He; Wu, Qun-Yan; Wang, Cong-Zhi; Lan, Jian-Hui; Chai, Zhi-Fang; Nie, Chang-Ming; Shi, Wei-Qun

    2018-05-10

    Separation of trivalent actinides (An(III)) and lanthanides (Ln(III)) is one of the most important steps in spent nuclear fuel reprocessing. However, it is very difficult and challenging to separate them due to their similar chemical properties. Recently the pyridylpyrazole ligand (PypzH) has been identified to show good separation ability toward Am(III) over Eu(III). In this work, to explore the Am(III)/Eu(III) separation mechanism of PypzH at the molecular level, the geometrical structures, bonding nature, and thermodynamic behaviors of the Am(III) and Eu(III) complexes with PypzH ligands modified by alkyl chains (Cn-PypzH, n = 2, 4, 8) have been systematically investigated using scalar relativistic density functional theory (DFT). According to the NBO (natural bonding orbital) and QTAIM (quantum theory of atoms in molecules) analyses, the M-N bonds exhibit a certain degree of covalent character, and more covalency appears in Am-N bonds compared to Eu-N bonds. Thermodynamic analyses suggest that the 1:1 extraction reaction, [M(NO 3 )(H 2 O) 6 ] 2+ + PypzH + 2NO 3 - → M(PypzH)(NO 3 ) 3 (H 2 O) + 5H 2 O, is the most suitable for Am(III)/Eu(III) separation. Furthermore, the extraction ability and the Am(III)/Eu(III) selectivity of the ligand PypzH is indeed enhanced by adding alkyl-substituted chains in agreement with experimental observations. Besides this, the nitrogen atom of pyrazole ring plays a more significant role in the extraction reactions related to Am(III)/Eu(III) separation compared to that of pyridine ring. This work could identify the mechanism of the Am(III)/Eu(III) selectivity of the ligand PypzH and provide valuable theoretical information for achieving an efficient Am(III)/Eu(III) separation process for spent nuclear fuel reprocessing.

  20. Engineering product storage under the advanced fuel cycle initiative. Part I: An iterative thermal transport modeling scheme for high-heat-generating radioactive storage forms

    International Nuclear Information System (INIS)

    Kaminski, Michael D.

    2005-01-01

    The US Department of Energy is developing an integrated nuclear fuel cycle technology under its Advanced Fuel Cycle Initiative (AFCI). Under the AFCI, waste minimization is stressed. Engineered product storage materials will be required to store concentrated radioactive cesium, strontium, americium, and curium for periods of tens to hundreds of years. The fabrication of such engineered products has some precedence but the concept is largely novel. We thus present a theoretical model used to calculate the maximum radial dimensions of right cylinder storage forms under several scenarios. Maximum dimensions are small, comparable to nuclear fuel pins in some cases, to avoid centerline melting temperatures; this highlights the need for a careful strategy for engineered product storage fabrication and storage

  1. Department of Energy Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs Draft Environmental Impact Statement. Volume 2, Part A

    Energy Technology Data Exchange (ETDEWEB)

    1994-06-01

    This document analyzes at a programmatic level the potential environmental consequences over the next 40 years of alternatives related to the transportation, receipt, processing, and storage of spent nuclear fuel under the responsibility of the US Department of Energy. It also analyzes the site-specific consequences of the Idaho National Engineering Laboratory sitewide actions anticipated over the next 10 years for waste and spent nuclear fuel management and environmental restoration. For programmatic spent nuclear fuel management this document analyzes alternatives of no action, decentralization, regionalization, centralization and the use of the plans that existed in 1992/1993 for the management of these materials. For the Idaho National Engineering Laboratory, this document analyzes alternatives of no action, ten-year plan, minimum and maximum and maximum treatment, storage, and disposal of US Department of Energy wastes.

  2. Fuel and fission product behaviour in early phases of a severe accident. Part I: Experimental results of the PHEBUS FPT2 test

    Energy Technology Data Exchange (ETDEWEB)

    Barrachin, M., E-mail: marc.barrachin@irsn.fr [Institut de Radioprotection et de Sûreté Nucléaire, B.P. 3, 13115 Saint Paul-lez-Durance Cedex (France); Gavillet, D. [Paul Scherrer Institute, Würenlingen and Villigen, CH-5232 Villigen PSI (Switzerland); Dubourg, R. [Institut de Radioprotection et de Sûreté Nucléaire, B.P. 3, 13115 Saint Paul-lez-Durance Cedex (France); De Bremaecker, A. [Institute for Nuclear Materials Sciences, SCK-CEN, Boeretang 200, B-2400 Mol (Belgium)

    2014-10-15

    One objective of the FPT2 test of the PHEBUS FP Program was to study the degradation of an irradiated UO{sub 2} fuel test section and the fission product behaviour under conditions of low steam flow. The results of the post-irradiation examinations (PIE) at the upper levels (823 and 900 mm) of the 1-m long test section are presented in this paper. Material interactions leading to local corium formation were identified: firstly between fuel and Zircaloy-4 cladding, notably at 823 mm, where the cladding melting temperature was reached, and secondly between fuel and stainless steel oxides. Regarding fission products, molybdenum left so-called metallic precipitates mainly composed of ruthenium. Xenon and caesium behave similarly whereas barium and molybdenum often seems to be associated in precipitates.

  3. The fuel of nuclear reactors

    International Nuclear Information System (INIS)

    1995-03-01

    This booklet is a presentation of the different steps of the preparation of nuclear fuels performed by Cogema. The documents starts with a presentation of the different French reactor types: graphite moderated reactors, PWRs using MOX fuel, fast breeder reactors and research reactors. The second part describes the fuel manufacturing process: conditioning of nuclear materials and fabrication of fuel assemblies. The third part lists the different companies involved in the French nuclear fuel industry while part 4 gives a short presentation of the two Cogema's fuel fabrication plants at Cadarache and Marcoule. Part 5 and 6 concern the quality assurance, the safety and reliability aspects of fuel elements and the R and D programs. The last part presents some aspects of the environmental and personnel protection performed by Cogema. (J.S.)

  4. Oral Assessment Kit, Levels II & III. Draft.

    Science.gov (United States)

    Agrelo-Gonzalez, Maria; And Others

    The assessment packet includes a series of oral tests to help develop speaking as an integral part of second language instruction at levels II and III. It contains: 8 mini-tests for use at level II; 9 mini-tests for use at level III; a rating scale and score sheet masters for evaluating performance on these tests; and a collection of suggested…

  5. The priority cases of the FUMEX-III exercises simulated with the TRANSURANUS code

    International Nuclear Information System (INIS)

    Boneva, S.

    2011-01-01

    The FUMEX-III project provides a good basis for testing common code priorities and the needs for further developments. The GAIN experiment contains results on four Gd 2 O 3 doped UO 2 rods and offers good opportunities for testing of the fuel performance codes in the case of Gd-doped fuel. A good agreement between the TRANSURANUS calculations and the measurements is achieved for the fuel and the cladding deformation. The FUMEX-III priority cases cover two rods from the GINNA reactor experiment: rod2 with fuel solid pellets, and rod4 with annular pellets and standard Zircaloy-4 cladding. Both rods were irradiated 5 cycles up to 52MWd/kgU. The simulations of the GINNA and US PWR experiments are part of the ongoing validation of the TRANSURANUS code - for different pellet design. The simulations of irradiation transients reveal the need for improving the fission gas release model, including burst release and release from the high burn-up structure

  6. Romanian nuclear fuel program

    International Nuclear Information System (INIS)

    Budan, O.

    1999-01-01

    The paper presents and comments the policy adopted in Romania for the production of CANDU-6 nuclear fuel before and after 1990. The CANDU-6 nuclear fuel manufacturing started in Romania in December 1983. Neither AECL nor any Canadian nuclear fuel manufacturer were involved in the Romanian industrial nuclear fuel production before 1990. After January 1990, the new created Romanian Electricity Authority (RENEL) assumed the responsibility for the Romanian Nuclear Power Program. It was RENEL's decision to stop, in June 1990, the nuclear fuel production at the Institute for Nuclear Power Reactors (IRNE) Pitesti. This decision was justified by the Canadian specialists team findings, revealed during a general, but well enough technically founded analysis performed at IRNE in the spring of 1990. All fuel manufactured before June 1990 was quarantined as it was considered of suspect quality. By that time more than 31,000 fuel bundles had already been manufactured. This fuel was stored for subsequent assessment. The paper explains the reasons which provoked this decision. The paper also presents the strategy adopted by RENEL after 1990 regarding the Romanian Nuclear Fuel Program. After a complex program done by Romanian and Canadian partners, in November 1994, AECL issued a temporary certification for the Romanian nuclear fuel plant. During the demonstration manufacturing run, as an essential milestone for the qualification of the Romanian fuel supplier for CANDU-6 reactors, 202 fuel bundles were produced. Of these fuel bundles, 66 were part of the Cernavoda NGS Unit 1 first fuel load (the balance was supplied by Zircatec Precision Industries Inc. ZPI). The industrial nuclear fuel fabrication re-started in Romania in January 1995 under AECL's periodical monitoring. In December 1995, AECL issued a permanent certificate, stating the Romanian nuclear fuel plant as a qualified and authorised CANDU-6 fuel supplier. The re-loading of the Cernavoda NGS Unit 1 started in the middle

  7. Resolution 197/012. It replaces the wording of Annex III of the Regulation of Technical Specifications for Quality of Liquid Fuels, approved by Resolution of the URSEA 150/008

    International Nuclear Information System (INIS)

    2012-01-01

    This resolution is about modifications made in the regulation of technical specifications in the quality of liquid fuels, approved by Resolution of the URSEA 150/008. These modifications concern to the Especial Oil gas

  8. Reprocessing of the spent nuclear fuel, I-VIII, Part I, Building the cell for inverse stream extraction of U and Pu

    International Nuclear Information System (INIS)

    Gal, I.

    1963-02-01

    This report covers the description of the hot cell for extracting uranium, plutonium and fission products from the fuel irradiated in the reactor. The level of activity planned was 10 Ci. The technology of the process is described, followed by the detailed description of the equipment, instrumentation

  9. Development of metal uranium fuel and testing of construction materials (I-VI); Part I; Razvoj metalnog goriva i ispitivanje konstrukcionih materijala (I-VI deo); I deo

    Energy Technology Data Exchange (ETDEWEB)

    Mihajlovic, A [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1965-11-15

    This project includes the following tasks: Study of crystallisation of metal melt and beta-alpha transforms in uranium and uranium alloys; Study of the thermal treatment influence on phase transformations and texture in uranium alloys; Radiation damage of metal uranium; Project related to irradiation of metal uranium in the reactor; Development of fuel element for nuclear reactors.

  10. 11-th International conference Nuclear power safety and nuclear education - 2009. Abstracts. Part 1. Session: Safety of nuclear technology; Innovative nuclear systems and fuel cycle; Nuclear knowledge management

    International Nuclear Information System (INIS)

    2009-01-01

    The book includes abstracts of the 11-th International conference Nuclear power safety and nuclear education - 2009 (29 Sep - 2 Oct, 2009, Obninsk). Problems of safety of nuclear technology are discussed, innovative nuclear systems and fuel cycles are treated. Abstracts on professional education for nuclear power and industry are presented. Nuclear knowledge management are discussed

  11. Iron making technology with fuels and other materials injection in blast furnace tuyeres. Part 1. Auxiliary fuels characteristics and its influence in the blast furnace process; Tecnologia de fabricacion de arrabio con la inyeccion de combustibles y otros materiales por toberas en el horno alto. I parte. Caracteristicas de los combustibles auxiliares y su influencia en el proceso del horno alto

    Energy Technology Data Exchange (ETDEWEB)

    Garcia, L. [Union de Empresas de Recuperacion de Materias Primas. Ciudad de La Habana (Cuba); Cores, A.; Formoso, A. [Centro Nacional de Investigaciones Metalurgicas. Madrid (Spain); Babich, A.; Yaroshevskii, S. [Universidad Estatal Tecnologica de Donetsk. Ucrania (Ukraine)

    1998-06-01

    The injection of fuels by tuyeres in the blast furnace is a used practice in most furnaces with the principal aim to reduce the coke consumption by ton of pig iron produced. The nature of these fuels is very diverse and depends on the resources of each country and of the fuel price. At this moment the coal injection (pulverized and granular) is the most extended practice, and the number of furnaces with facilities for coal injection increases continuously. (Author) 14 refs.

  12. Fuel assemblies

    International Nuclear Information System (INIS)

    Mukai, Hideyuki

    1987-01-01

    Purpose: To prevent bending of fuel rods caused by the difference of irradiation growth between coupling fuel rods and standards fuel rods thereby maintain the fuel rod integrity. Constitution: The f value for a fuel can (the ratio of pole of zirconium crystals in the entire crystals along the axial direction of the fuel can) of a coupling fuel rod secured by upper and lower tie plates is made smaller than the f value for the fuel can of a standard fuel rod not secured by the upper and the lower tie plates. This can make the irradiation growth of the fuel can of the coupling fuel rod greater than the irradiation growth of the fuel can of the standard fuel rod and, accordingly, since the elongation of the standard fuel rod can always by made greater, bending of the standard fuel rod can be prevented. (Yoshihara, M.)

  13. Activation calculation of steel of the control rods of TRIGA Mark III reactor; Calculo de activacion del acero de las barras de control del reactor TRIGA Mark III

    Energy Technology Data Exchange (ETDEWEB)

    Garcia M, T.; Cruz G, H. S.; Ruiz C, M. A.; Angeles C, A., E-mail: teodoro.garcia@inin.gob.mx [ININ, Carretera Mexico-Toluca sn, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2014-10-15

    In the pool of TRIGA Mark III reactor of the Instituto Nacional de Investigaciones Nucleares (ININ), there are control rods that were removed from the core, and which are currently on shelves of decay. These rods were part of the reactor core when only had fuel standard (from 1968-1989). To conduct a proper activation analysis of the rods, is very important to have well-characterized the materials which are built, elemental composition of the same ones, the atomic densities and weight fractions of the elements that constitute them. To determine the neutron activation of the control rods MCNP5 code was used, this code allows us to have well characterized the radionuclides inventory that were formed during irradiation of the control rods. This work is limited to determining the activation of the steel that is part of the shielding of the control rods, the nuclear fuel that is in the fuel follower does not include. The calculation model of the code will be validated with experimental measurements and calculating the activity of fission products of the fuel follower which will take place at the end of 2014. (Author)

  14. Presentation of safety after closure of the repository for spent nuclear fuel. Main report of the project SR-Site. Part I; Redovisning av saekerhet efter foerslutning av slutfoervaret foer anvaent kaernbraensle. Huvudrapport fraan projekt SR-Site. Del I

    Energy Technology Data Exchange (ETDEWEB)

    2011-07-01

    The purpose of the safety assessment SR-Site is to investigate whether a safe repository for spent nuclear fuel by KBS-3 type can be constructed at Forsmark in Oesthammar in Sweden. The location of the Forsmark has been selected based on results of several surveys from surface conditions at depth in Forsmark and in Laxemar in Oskarshamn. The choice of location is not justified in SR-Site Report, but in other attachments to SKB's permit applications. SR-Site Report is an important part of SKB's permit applications to construct and operate a repository for spent nuclear fuel at Forsmark in Oesthammar. The purpose of the report in the applications is to show that a repository at Forsmark is safe after closure

  15. Presentation of safety after closure of the repository for spent nuclear fuel. Main report of the project SR-Site. Part II; Redovisning av saekerhet efter foerslutning av slutfoervaret foer anvaent kaernbraensle. Huvudrapport fraan projekt SR-Site. Del II

    Energy Technology Data Exchange (ETDEWEB)

    2011-07-01

    The purpose of the safety assessment SR-Site is to investigate whether a safe repository for spent nuclear fuel by KBS-3 type can be constructed at Forsmark in Oesthammar in Sweden. The location of the Forsmark has been selected based on results of several surveys from surface conditions at depth in Forsmark and in Laxemar in Oskarshamn. The choice of location is not justified in SR-Site Report, but in other attachments to SKB's permit applications. SR-Site Report is an important part of SKB's permit applications to construct and operate a repository for spent nuclear fuel at Forsmark in Oesthammar. The purpose of the report in the applications is to show that a repository at Forsmark is safe after closure

  16. Impact assessment of biomass-based district heating systems in densely populated communities. Part II: Would the replacement of fossil fuels improve ambient air quality and human health?

    Science.gov (United States)

    Petrov, Olga; Bi, Xiaotao; Lau, Anthony

    2017-07-01

    To determine if replacing fossil fuel combustion with biomass gasification would impact air quality, we evaluated the impact of a small-scale biomass gasification plant (BRDF) at a university campus over 5 scenarios. The overall incremental contribution of fine particles (PM2.5) is found to be at least one order of magnitude lower than the provincial air quality objectives. The maximum PM2.5 emission from the natural gas fueled power house (PH) could adversely add to the already high background concentration levels. Nitrogen dioxide (NO2) emissions from the BRDF with no engineered pollution controls for NOx in place exceeded the provincial objective in all seasons except during summer. The impact score, IS, was the highest for NO2 (677 Disability Adjusted Life Years, DALY) when biomass entirely replaced fossil fuels, and the highest for PM2.5 (64 DALY) and CO (3 DALY) if all energy was produced by natural gas at PH. Complete replacement of fossil fuels by one biomass plant can result in almost 28% higher health impacts (708 DALY) compared to 513 DALY when both the current BRDF and the PH are operational mostly due to uncontrolled NO2 emissions. Observations from this study inform academic community, city planners, policy makers and technology developers on the impacts of community district heating systems and possible mitigation strategies: a) community energy demand could be met either by splitting emissions into more than one source at different locations and different fuel types or by a single source with the least-impact-based location selection criteria with biomass as a fuel; b) advanced high-efficiency pollution control devices are essential to lower emissions for emission sources located in a densely populated community; c) a spatial and temporal impact assessment should be performed in developing bioenergy-based district heating systems, in which the capital and operational costs should be balanced with not only the benefit to greenhouse gas emission

  17. Asymmetrical Interleaved DC/DC Switching Converters for Photovoltaic and Fuel Cell Applications—Part 1: Circuit Generation, Analysis and Design 

    Directory of Open Access Journals (Sweden)

    Sergio Serna

    2012-11-01

    Full Text Available A novel asymmetrical interleaved dc/dc switching converters family intended for photovoltaic and fuel cell applications is presented in this paper. The main requirements on such applications are small ripples in the generator and load, as well as high voltage conversion ratio. Therefore, interleaved structures and voltage multiplier cells have been asymmetrically combined to generate new converters, which inherently operate indiscontinuous conduction mode. The novel family is derived from boost, buck-boost and flyback-based structures. This converter family is analyzed to obtain the design equations and synthesize a design process based on the typical requirements of photovoltaic and fuel cell applications. Finally, the experimental results validate the characteristics and usefulness of the asymmetrical interleaved converter family. 

  18. Fission products and nuclear fuel behaviour under severe accident conditions part 3: Speciation of fission products in the VERDON-1 sample

    Science.gov (United States)

    Le Gall, C.; Geiger, E.; Gallais-During, A.; Pontillon, Y.; Lamontagne, J.; Hanus, E.; Ducros, G.

    2017-11-01

    Qualitative and quantitative analyses on the VERDON-1 sample made it possible to obtain valuable information on fission product behaviour in the fuel during the test. A promising methodology based on the quantitative results of post-test characterisations has been implemented to assess the release fraction of non γ-emitter fission products. The order of magnitude of the estimated release fractions for each fission product was consistent with their class of volatility.

  19. Performance limits of coated particle fuel. Part II. Mechanical failure of coated particles due to internal gas pressure and kernel swelling

    Energy Technology Data Exchange (ETDEWEB)

    Hick, H.; Nabielek, H.; Harrison, T. A.

    1973-10-15

    This report presents a summary of experimental results and their theoretical explanation with regard to the "Pressure Failure" of coated particle fuel. While the experimental results refer mainly to the Dragon Reference Particle as proposed for typical Low Enriched Homogeneous Prismatic Steam Cycle HTR Power Reactors, the theoretical understanding of the phenomena and the mathematical models for their description are not limited to a specific design line.

  20. Theoretical and Experimental Flow Cell Studies of a Hydrogen-Bromine Fuel Cell, Part 1. M.S. Thesis. Final Report

    Science.gov (United States)

    Savinell, R. F.; Fritts, S. D.

    1986-01-01

    There is increasing interest in hydrogen-bromine fuel cells as both primary and regenerative energy storage systems. One promising design for a hydrogen-bromine fuel cell is a negative half cell having only a gas phase, which is separated by a cationic exchange membrane from a positive half cell having an aqueous electrolyte. The hydrogen gas and the aqueous bromide solution are stored external to the cell. In order to calculate the energy storage capacity and to predict and assess the performance of a single cell, the open circuit potential (OCV) must be estimated for different states of change, under various conditions. Theoretical expressions were derived to estimate the OCV of a hydrogen-bromine fuel cell. In these expressions temperature, hydrogen pressure, and bromine and hydrobromic acid concentrations were taken into consideration. Also included are the effects of the Nafion membrance separator and the various bromide complex species. Activity coefficients were taken into account in one of the expressions. The sensitivity of these parameters on the calculated OCV was studied.