WorldWideScience

Sample records for fissile material transfers

  1. The mass transfer mechanism of fissile material due to fission

    International Nuclear Information System (INIS)

    Shafrir, N.H.

    1975-01-01

    A thin 252 Cf source of a mean thickness of an approXimately mono-atomic layer was used as an experimental model for the study of the basic mechanism of the knock-on process taking place in fissile material. Because of the thinness of the source it can be assumed that mainly primary knock-ons are formed. The ejection rate of knock-ons created by direct collisions between fission fragments and source atoms was measured as follows: the ejected atoms were collected in high vacuum on a catcher foil and 252 Cf determined by alpha spectroscopy using a silicon surface barrier detector. The number of 252 Cf ejected from the source in unit time could thus be determined while considering the anisotropy of ejection, geometry and counting efficiency. Taking into account the chemical composition of the source, eta(theor.) = 252 Cf atoms/fission was obtained. This result can be considered in reasonable agreement with experiment confirming that under the experimental conditions described, practically no knock-on cascade is formed. (B.G.)

  2. Canyon transfer neutron absorber to fissile material ratio analysis. Revision 1

    International Nuclear Information System (INIS)

    Clemmons, J.S.

    1994-01-01

    Waste tank fissile material and non-fissile material estimates are used to evaluate criticality safety for the existing sludge inventory and batches of sludge sent to Extended Sludge Processing (ESP). This report documents the weight ratios of several non-fissile waste constituents to fissile waste constituents from canyon reprocessing waste streams. Weight ratios of Fe, Mn, Al, Mi, and U-238 to fissile material are calculated from monthly loss estimates from the F and H Canyon Low Heat Waste (LHW) and High Heat Waste (HHW) streams. The monthly weight ratios for Fe, Mn and U-238 are then compared to calculated minimum safe weight ratios. Documented minimum safe weight ratios for Al and Ni to fissile material are currently not available. Total mass data for the subject sludge constituents is provided along with scatter plots of the monthly weight ratios for each waste stream

  3. Fissile material proliferation risk

    International Nuclear Information System (INIS)

    Dreicer, J.S.; Rutherford, D.A.

    1996-01-01

    The proliferation risk of a facility depends on the material attractiveness, level of safeguards, and physical protection applied to the material in conjunction with an assessment of the impact of the socioeconomic circumstances and threat environment. Proliferation risk is a complementary extension of proliferation resistance. The authors believe a better determination of nuclear proliferation can be achieved by establishing the proliferation risk for facilities that contain nuclear material. Developing a method that incorporates the socioeconomic circumstances and threat environment inherent to each country enables a global proliferation assessment. To effectively reduce the nuclear danger, a broadly based set of criteria is needed that provides the capability to relatively assess a wide range of nuclear related sites and facilities in different countries and still ensure a global decrease in proliferation risk for fissile material (plutonium and highly enriched uranium)

  4. Transfer of fissile material through shielding coatings in emergency heating of HTGR coated particles

    International Nuclear Information System (INIS)

    Gudkov, A.N.; Zhuravkov, S.G.; Koptev, M.A.; Kurepin, A.D.

    1990-01-01

    The measurement results of leakage dynamics of fissile material from the coated particles within a temperature range of 1200 + 2000 deg. C are given. The methods of carrying out the experiments are briefly described. The relation of the leakage rate of uranium-235 from CP (coated particles) with the pyrocarbonic coatings has been obtained. (author)

  5. Fissile materials detection

    International Nuclear Information System (INIS)

    Dumesnil, P.

    1977-03-01

    Description is given of three types of apparatus intended for controlling fossile materials in view of avoiding their diversion or preventing said products to be mixed to less dangerous radioactive wastes. The gantry-type apparatus is intended for the detection of small masses of fissile materials moving through a crossing place; the neutron gantry consists of helium 3 detectors of the type 150NH100, located inside polyethylene blocks; as for the gamma gantry, it consists of two large plastic scintillators integrated to the vertical legs of said gantry. The second apparatus is a high-efficiency detector intended for controlling Pu inside waste casks. It can detect 10mg of Pu inside a 100 liters drum for one minute counting. The third apparatus intended for persons and things monitoring is still on study. Such as the gantries it is based on sampled measurement of the background noise [fr

  6. Repository for fissile materials

    International Nuclear Information System (INIS)

    Gablin, K.A.

    1976-01-01

    A repository for holding and storing fissile or other hazardous materials either under or above the ground is provided by enclosing one or more inner containers, such as standard steel drums, in a larger, corrosion-resistant outer shell, with a layer of foamed polyurethane occupying the space therebetween. The polyurethane foam is free of voids at its interfaces with the inner container and outer shell, and adheres to and reinforces same to provide a stress skin structure. Protection is afforded by the chemical and physical characteristics of the polyurethane foam against destructive influences such as water vapor intrusion, package leakage and damaging effects of the environment, such as freezing, electrolysis, chemical and bacterial action. The outer shell is shaped to conform generally to the shape of the inner container and is made of a tube of bituminized fiber material with endcaps of exterior grade plywood treated with wood preservative. A quantity of fluorescein dye is positioned within the inner container for monitoring each package for leakage

  7. Recovery of fissile materials from plutonium residues, miscellaneous spent nuclear fuel, and uranium fissile wastes

    International Nuclear Information System (INIS)

    Forsberg, C.W.

    1997-01-01

    A new process is proposed that converts complex feeds containing fissile materials into a chemical form that allows the use of existing technologies (such as PUREX and ion exchange) to recover the fissile materials and convert the resultant wastes to glass. Potential feed materials include (1) plutonium scrap and residue, (2) miscellaneous spent nuclear fuel, and (3) uranium fissile wastes. The initial feed materials may contain mixtures of metals, ceramics, amorphous solids, halides, and organics. 14 refs., 4 figs

  8. Electronuclear conversion of fertile to fissile material

    International Nuclear Information System (INIS)

    Van Atta, C.M.; Lee, J.D.; Heckrotte, W.

    1976-01-01

    The electronuclear conversion of fertile to fissile material by accelerator-produced neutrons is discussed. Experimental and theoretical results obtained in the MTA program (1949--1954) on the production of low-energy (less than 20-MeV) neutrons by high-energy proton, deuteron, and neutron bombardment of target materials are briefly reviewed. More recent calculations of the cascade process, by which the low-energy neutrons are produced, are discussed. A system is described by which 500- to 600-MeV deuterons incident on a lithium primary target can be converted to high-energy neutrons, which can be multiplied by spallation cascades and nuclear excitation to produce low-energy neutrons in a depleted-uranium or thorium secondary target. Fission events producing heat and additional neutrons are produced. The evaporation and fission neutrons would be captured, and fissile material would be produced. The production rates for 239 Pu and 233 U are estimated for 0.25-A and 0.375-A deuteron beams from an Alvarez linac. The capital and operating costs are estimated, and the resulting costs of fissile materials are calculated. The cost of generating power in reactors using the fissile material so produced as make-up fuel is also estimated. The energy multiplication (power generated in reactors so fueled/power consumed by the accelerator) ranges from about 10 to about 50 depending upon the make-up of the secondary target; depleted uranium, thorium, or a combination of the two. An experimental and theoretical program to facilitate optimization of the parameters of a production installation is described. 13 figures, 14 tables

  9. Warhead and fissile-material declarations

    International Nuclear Information System (INIS)

    von Hippel, F.

    1992-01-01

    Until recently, arms control agreements were limited by the fact that the only available verification capabilities were national technical means, which involved instruments in space or beyond national borders. As a result, the SALT II treaty constrained only the construction of large missile silos, ballistic-missile submarines and long-range bombers - and limited the flight testing of long-range ballistic missiles. Recently, however, on-site verification has been accepted, making it possible in the INF treaty to extend controls to small mobile missiles and their launchers. This paper therefore outlines a comprehensive system of verifiable limits on nuclear warheads. The authors discuss in some detail the verifiability of a halt in the production of fissile materials for nuclear warheads, the verifiability of declarations of the amounts of fissile material produced for warheads prior to the production cutoff, and the establishment of a verifiable accounting system for the numbers and types of nuclear warheads possessed by each side

  10. Recovery of fissile materials from nuclear wastes

    Science.gov (United States)

    Forsberg, Charles W.

    1999-01-01

    A process for recovering fissile materials such as uranium, and plutonium, and rare earth elements, from complex waste feed material, and converting the remaining wastes into a waste glass suitable for storage or disposal. The waste feed is mixed with a dissolution glass formed of lead oxide and boron oxide resulting in oxidation, dehalogenation, and dissolution of metal oxides. Carbon is added to remove lead oxide, and a boron oxide fusion melt is produced. The fusion melt is essentially devoid of organic materials and halogens, and is easily and rapidly dissolved in nitric acid. After dissolution, uranium, plutonium and rare earth elements are separated from the acid and recovered by processes such as PUREX or ion exchange. The remaining acid waste stream is vitrified to produce a waste glass suitable for storage or disposal. Potential waste feed materials include plutonium scrap and residue, miscellaneous spent nuclear fuel, and uranium fissile wastes. The initial feed materials may contain mixtures of metals, ceramics, amorphous solids, halides, organic material and other carbon-containing material.

  11. Revisited. Euratom's ownership of special fissile materials

    International Nuclear Information System (INIS)

    Pelzer, Norbert

    2015-01-01

    Among all Treaties on the Foundation of the European Community, seemingly, the Euratom Treaty ist the most unobtrusive one having even nearly been declared dead occasionally. For the opponents of nuclear energy the treaty is a thorn in their side because it aims for the peaceful exploitation of nuclear energy. Actually, the treaty likewise aims for the protection of dangers of nuclear energy and encloses a bundle of collective control instruments. The protective purpose provides the community with a strong position in numerous fields towards nuclear energy users including the right to intervene in the operations of nuclear facilities. The communitie's position is further strengthened by the communitie's ownership on special fissile materials. The EAEC Treaty determines: 'Special fissile materials are owned by the community'. The material content of Euratom's ownership is limited by Article 87 of the EAEC Treaty: Unlimited right of use and consumption is granted to the properly possessors unless obligations of the Euratom Treaty oppose. Inherently, the community does not have these rights. It was asked what would be left to the owner Euratom if the properly possessor is entitled to unlimited right of use and even right of consumption.

  12. Fissile material disposition program: Screening of alternate immobilization candidates for disposition of surplus fissile materials

    International Nuclear Information System (INIS)

    Gray, L.W.

    1996-01-01

    With the end of the Cold War, the world faces for the first time the need to dismantle vast numbers of ''excess'' nuclear weapons and dispose of the fissile materials they contain, together with fissile residues in the weapons production complex left over from the production of these weapons. If recently agreed US and Russian reductions are fully implemented, tens of thousands of nuclear weapons, containing a hundred tons or more of plutonium and hundreds of tonnes* of highly enriched uranium (HEU), will no longer be needed worldwide for military purposes. These two materials are the essential ingredients of nuclear weapons, and limits on access to them are the primary technical barrier to prospective proliferants who might desire to acquire a nuclear weapons capability. Theoretically, several kilograms of plutonium, or several times that amount of HEU, is sufficient to make a nuclear explosive device. Therefore, these materials will continue to be a potential threat to humanity for as long as they exist

  13. Measurement of inventories with mixed fissile materials

    International Nuclear Information System (INIS)

    Rinard, P.M.; Krick, M.S.; Kelley, T.; Schneider, C.M.

    1997-01-01

    An inventory with a large number of diverse items containing mixtures of uranium and plutonium has been measured with two nondestructive assay (NDA) instruments used in four modes. A segmented gamma scanner (SGS) was used to find the number of cans and the positions of the fissile materials by scanning each item in front of a transmissions source; at each position, uranium and plutonium isotopics were measured with the passive gamma rays emitted. A shuffler was then used in both the passive and active modes to measure the masses of the two elements. The measured masses for the inventory items were generally in agreement with the declared values, but anomalies were identified for a small fraction of the inventory

  14. Nonintrusive verification attributes for excess fissile materials

    International Nuclear Information System (INIS)

    Nicholas, N.J.; Eccleston, G.W.; Fearey, B.L.

    1997-10-01

    Under US initiatives, over two hundred metric tons of fissile materials have been declared to be excess to national defense needs. These excess materials are in both classified and unclassified forms. The US has expressed the intent to place these materials under international inspections as soon as practicable. To support these commitments, members of the US technical community are examining a variety of nonintrusive approaches (i.e., those that would not reveal classified or sensitive information) for verification of a range of potential declarations for these classified and unclassified materials. The most troublesome and potentially difficult issues involve approaches for international inspection of classified materials. The primary focus of the work to date has been on the measurement of signatures of relevant materials attributes (e.g., element, identification number, isotopic ratios, etc.), especially those related to classified materials and items. The authors are examining potential attributes and related measurement technologies in the context of possible verification approaches. The paper will discuss the current status of these activities, including their development, assessment, and benchmarking status

  15. Assessment of the U.S. regulations for fissile exemptions and fissile material general licenses

    International Nuclear Information System (INIS)

    Parks, C.V.; Hopper, C.M.; Lichtenwalter, J.J.; Easton, E.P.; Brochman, P.G.

    1998-05-01

    The paragraphs for general licenses for fissile material and exemptions (often termed exceptions in the international community) for fissile material have long been a part of the US Code of Federal Regulations (CFR) 10 CFR Part 71, Packaging and Transportation of Radioactive Material. More recently, the Nuclear Regulatory Commission (NRC) issued a final rule on Part 71 via emergency rule-making procedures in order to address an identified deficiency related to one of the fissile exemptions. To address the specified deficiency in a general fashion, the emergency rule adopted the approach of the 1996 Edition of the IAEA: Regulations for the Safe Transport of Radioactive Material (IAEA 1996), which places restrictions on certain moderating materials and limits the quantity of fissile material in a consignment. The public comments received by the NRC indicated general agreement with the need for restrictions on certain moderators (beryllium, deuterium, and graphite). The comments indicated concern relative to both the degree of restriction imposed (not more than 0.1% of fissile material mass) and the need to limit the fissile material mass of the consignment, particularly in light of the subsequent NRC staff position that the true intent was to provide control for limiting the fissile mass of the conveyance. The purpose of the review is to identify potential deficiencies that might be adverse to maintaining adequate subcriticality under normal conditions of transport and hypothetical accident conditions. In addition, ORNL has been asked to identify changes that would address any identified safety issues, enable inherently safe packages to continue to be unencumbered in transport, and seek to minimize the impact on current safe practices

  16. Enhanced safety in the storage of fissile materials

    International Nuclear Information System (INIS)

    Williams, G.E.; Alvares, N.J.

    1979-01-01

    A ''plastic-like'' supporting material impregnated with a neutron-absorbing agent that is suitable for ''lining'' the inner surfaces of fissile-material storage containers was fabricated. The material consists, by weight, of 50% food-grade borax, 25% coal tar, and 25% epoxy resin. It costs much less than commercially available materials, can absorb enough neutrons to isolate units of fissile material, and possesses such structural qualities as flexibility and machinability. Properties and performance of the material are discussed

  17. Disposition of surplus fissile materials via immobilization

    International Nuclear Information System (INIS)

    Gray, L.W.; Kan, T.; Sutcliffe, W.G.; McKibben, J.M.; Danker, W.

    1995-01-01

    In the Cold War aftermath, the US and Russia have agreed to large reductions in nuclear weapons. To aid in the selection of long-term management options, the USDOE has undertaken a multifaceted study to select options for storage and disposition of surplus plutonium (Pu). One disposition alternative being considered is immobilization. Immobilization is a process in which surplus Pu would be embedded in a suitable material to produce an appropriate form for ultimate disposal. To arrive at an appropriate form, we first reviewed published information on HLW immobilization technologies to identify forms to be prescreened. Surviving forms were screened using multi-attribute utility analysis to determine promising technologies for Pu immobilization. We further evaluated the most promising immobilization families to identify and seek solutions for chemical, chemical engineering, environmental, safety, and health problems; these problems remain to be solved before we can make technical decisions about the viability of using the forms for long-term disposition of Pu. All data, analyses, and reports are being provided to the DOE Office of Fissile Materials Disposition to support the Record of Decision that is anticipated in Summer of 1996

  18. Transportation of fissile materials and the danger of criticity

    International Nuclear Information System (INIS)

    Haon, D.; Leclerc, J.; Maubert, L.

    1981-01-01

    The authors examine the risk of criticity that can arise during the transportation of fissile matter. They then outline the regulations and studies made in the field of criticity-safety and the computation methods used. They discuss the applications that are reflected in the concept and design of fissile material packagings [fr

  19. Criticality Control Fissile of Materials. Proceedings of the Symposium on Criticality Control of Fissile Materials

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1966-05-15

    Criticality control comprises all the administrative and technical procedures which enable the storage and processing of fissile material to be carried out under conditions of nuclear safety. It is of particular importance in the safe design and operation of chemical and metallurgical plants processing fissile material, in the handling and storage of enriched fuel for reactors, and in transportation of fissile material. The growth of nuclear power, with its increasing use of fissile material and production of plutonium, is leading to an ever widening need for this discipline. This Symposium was held 4 Vulgar-Fraction-One-Half years after the only other international meeting on this topic, at which the first broad exchange of ideas and theories enabled a comparison to be drawn between the various ways in which the subject is handled in the different countries. That meeting showed that criticality safety was often achieved by procedures known to be ultra-safe, as there was a great lack of useful experimental data with which to check theoretical models. Since that time the quantities of material being processed have increased, and with the now urgent necessity of achieving economic, and hence commercially competitive, operation, the procedure of using arbitrary factors of safety is no longer adequate. Plant Managers now require good data on the basis of which they can choose a suitable factor of safety, and design a process to be safe under any foreseeable circumstances. The present Symposium showed the great increase in the amount of available experimental data and its importance in checking the now highly sophisticated computer calculations. There are many diagrams in these Proceedings with curves from which critical parameters for various configurations can be taken. The dearth of data for plutonium systems is causing some difficulty in plutonium processing plants, which are becoming commercially important. The excellent safety record of the atomic energy industry

  20. Fissile materials principles of criticality safety in handling and processing

    International Nuclear Information System (INIS)

    1976-01-01

    This Swedish Standard consists of the English version of the International Standard ISO 1709-1975-Nuclear energy. Fissile materials. Principles of criticality safety in handling and processing. (author)

  1. A 252Cf based nondestructive assay system for fissile material

    International Nuclear Information System (INIS)

    Menlove, H.O.; Crane, T.W.

    1978-01-01

    A modulated 252 Cf source assay system 'Shuffler' based on fast-or-thermal-neutron interrogation combined with delayed-neutron counting has been developed for the assay of fissile material. The 252 Cf neutron source is repetitively transferred from the interrogation position to a shielded position while the delayed neutrons are counted in a high efficiency 3 He neutron well-counter. For samples containing plutonium, this well-counter is also used in the passive coincidence mode to assay the effective 240 Pu content. The design of an optimized neutron tailoring assembly for fast-neutron interrogation using a Monte Carlo Neutron Computer Code is described. The Shuffler system has been applied to the assay of fuel pellets, inventory samples, irradiated fuel and plutonium mixed-oxide fuel. The system can assay samples with fissile contents from a few milligrams up to several kilograms using thermal-neutron interrogation for the low mass samples and fast-neutron interrogation for the high mass samples. Samples containing 235 U- 238 U, or 233 U-Th, or UO 2 -PuO 2 fuel mixtures have been assayed with the Shuffler system. (Auth.)

  2. Enhanced safety in the storage of fissile materials

    International Nuclear Information System (INIS)

    Williams, G.E.; Alvares, N.J.

    1978-01-01

    An inexpensive boron-loaded liner of epoxy resin for fissile-material storage containers was developed that can be easily fabricated of readily available, low-cost materials. Computer calculations indicate reactivity will be reduced substantially if this neutron-absorbing liner is added to containers in a typical storage array. These calculations compare favorably with neutron-attenuation experiments with thermal and fission neutron spectra, and tests at the Fire Test Facility indicate the epoxy resin will survive extreme environmental and accident conditions. The fire-resistant and insulating properties of the epoxy-resin liner further augment its ability to protect fissile materials. Boron-loaded epoxy resin is adaptable to many tasks but is particularly useful for providing enhanced criticality safety in the packaging and storage of fissile materials

  3. 1980 Annual status report: fissile materials control and management

    International Nuclear Information System (INIS)

    1981-01-01

    The R and D activities of the JRC in the field of Fissile Material Control and Management are oriented to the development of safeguards systems in the European Community nuclear fuel cycle and to provide means for a more efficient nuclear material management within the nuclear industry

  4. Systems analysis and simulation of fissile materials disposition alternatives

    International Nuclear Information System (INIS)

    Farish, T.J.; Farmen, R.F.; Boerigter, S.T.; DeMuth, N.S.

    1996-01-01

    A detailed process flow model has been developed for use in the Fissile Materials Disposition program. The model calculates fissile material flows and inventories among the various processing and storage facilities over the life of the disposition program. Given existing inventories and schedules for processing, we can estimate the required size of processing and storage facilities, including equipment requirements, plant floorspace, approximate costs, and surge capacities. The model was designed to allow rapid prototyping, parallel and team development of facility and sub-facility models, consistent levels of detail and the use of a library of generic objects representing unit process operations

  5. Accelerating fissile material detection with a neutron source

    Science.gov (United States)

    Rowland, Mark S.; Snyderman, Neal J.

    2018-01-30

    A neutron detector system for discriminating fissile material from non-fissile material wherein a digital data acquisition unit collects data at high rate, and in real-time processes large volumes of data directly to count neutrons from the unknown source and detecting excess grouped neutrons to identify fission in the unknown source. The system includes a Poisson neutron generator for in-beam interrogation of a possible fissile neutron source and a DC power supply that exhibits electrical ripple on the order of less than one part per million. Certain voltage multiplier circuits, such as Cockroft-Walton voltage multipliers, are used to enhance the effective of series resistor-inductor circuits components to reduce the ripple associated with traditional AC rectified, high voltage DC power supplies.

  6. A line of defense approach to fissile material control

    International Nuclear Information System (INIS)

    Holloway, S.P.; Holloway, N.J.

    1995-01-01

    A crucial element of the safety policy of the UK Atomic Weapons Establishment (AWE) is that concerned with the safe control of fissile material in order to minimize the potential for unplanned criticality. The principles by which AWE controls fissile material advocate a simple Line of Defense (LOD) approach to assessing criticality-safety related aspects of fissile operations. An LOD assessment provides a measure of the depth of defense available to prevent general types of criticality accident and can be used to demonstrate compliance with the risk-based Basic Safety Limits (BSLs) and Objectives (BSOs) used by the UK Nuclear Installations Inspectorate (NII) to judge the safety of operations in accordance with its Safety Assessment Principles (SAPs) for Nuclear Plants. This paper discusses the LOD concept, the basis of LOD assessment and describes LODs specific to criticality control

  7. Safeguards and security issues for the disposition of fissile materials

    International Nuclear Information System (INIS)

    Jaeger, C.D.; Moya, R.W.; Duggan, R.A.; Mangan, D.L.; Tolk, K.M.; Rutherford, D.; Fearey, B.; Moore, L.

    1995-01-01

    The Department of Energy's Office of Fissile Material Disposition (FMD) is analyzing long-term storage and disposition options for surplus weapons-usable fissile materials, preparing a programmatic environmental impact statement (PEIS), preparing for a record of decision (ROD) regarding this material and conducting other activities. The primary security objectives of this program are to reduce major security risks and strengthen arms reduction and nonproliferation (NP). To help achieve these objectives, a safeguards and security (S ampersand S) team consisting of participants from Sandia, Los Alamos, and Lawrence Livermore National Laboratories was established. The S ampersand S activity for this program is a cross-cutting task which addresses all of the FMD program options. It includes both domestic and international safeguards and includes areas such as physical protection, nuclear materials accountability and material containment and surveillance. This paper will discuss the activities of the Fissile Materials Disposition Program (FMDP) S ampersand S team as well as some specific S ampersand S issues associated with various FMDP options/facilities. Some of the items to be discussed include the threat, S ampersand S requirements, S ampersand S criteria for assessing risk, S ampersand S issues concerning fissile material processing/facilities, and international and domestic safeguards

  8. IAEA safeguards for the Fissile Materials Disposition Project

    International Nuclear Information System (INIS)

    Close, D.A.

    1995-06-01

    This document is an overview of International Atomic Energy Agency (IAEA) safeguards and the basic requirements or elements of an IAEA safeguards regime. The primary objective of IAEA safeguards is the timely detection of the diversion of a significant quantity of material and the timely detection of undeclared activities. The two important components of IAEA safeguards to accomplish their primary objective are nuclear material accountancy and containment and surveillance. This overview provides guidance to the Fissile Materials Disposition Project for IAEA inspection requirements. IAEA requirements, DOE Orders, and Nuclear Regulatory Commission regulations will be used as the basis for designing a safeguards and security system for the facilities recommended by the Fissile Materials Disposition Project

  9. Proliferation resistance criteria for fissile material disposition

    International Nuclear Information System (INIS)

    Close, D.A.; Fearey, B.L.; Markin, J.T.; Rutherford, D.A.; Duggan, R.A.; Jaeger, C.D.; Mangan, D.L.; Moya, R.W.; Moore, L.R.; Strait, R.S.

    1995-04-01

    The 1994 National Academy of Sciences study open-quotes Management and Disposition of Excess Weapons Plutoniumclose quotes defined options for reducing the national and international proliferation risks of materials declared excess to the nuclear weapons program. This report proposes criteria for assessing the proliferation resistance of these options. The criteria are general, encompassing all stages of the disposition process from storage through intermediate processing to final disposition including the facilities, processing technologies and materials, the level of safeguards for these materials, and the national/subnational threat to the materials

  10. Storage and processing system for fissile materials

    International Nuclear Information System (INIS)

    Bubowskij, B.G.; Bogatyrew, W.K.; Wladykow, G.M.; Swiridenko, W.J.

    1976-01-01

    The invention concerns the construction of a radiation protection wall by which the reflection of neutrons in a container arranged in the vicinity of the wall is reduced. The radiation protection wall has a coating of neutron-retarding material on top of which there is a layer of neutron absorbing material, the former having a surface structured with regular projections and recesses spaced at 1/8 to 3 neutron ranges. The recesses may be filled with porous material or take up neutron radiation detectors. Other construction features are described. (UWI) [de

  11. Security of fissile materials in Russia

    International Nuclear Information System (INIS)

    Bukharin, O.

    1996-01-01

    The problem of security of huge stocks of weapons-usable highly enriched uranium and plutonium in Russia against theft or diversion remains a serious nonproliferation concern. During the Cold War, the security of Soviet nuclear materials was based on centralization and discipline, protection by the military, and intrusive political oversight of the people. The recent fundamental societal changes have rendered these arrangements inadequate, and the security of nuclear materials has decreased. Safeguarding nuclear materials in Russia is particularly difficult because of their very large inventories and the size and complexity of the nation's nuclear infrastructure. Russia needs a reliable and more objective technology-based system of nuclear safeguards designed to control nuclear materials. The Russian government and the international community are working towards this goal

  12. Fissile material disposition and proliferation risk

    Energy Technology Data Exchange (ETDEWEB)

    Dreicer, J.S.; Rutherford, D.A. [Los Alamos National Lab., NM (United States). NIS Div.

    1996-05-01

    The proliferation risk of a facility is dependent on the material attractiveness, level of safeguards, and physical protection applied to the material in conjunction with an assessment of the impact of the socioeconomic circumstances and threat environment. Proliferation risk is a complementary extension of proliferation resistance. The authors believe a better determination of nuclear material proliferation can be achieved by establishing the proliferation risk for facilities that contain nuclear material. Developing a method that incorporates the socioeconomic circumstances and threat environment inherent to each country enables a global proliferation assessment. In order to effectively reduce the nuclear danger, a broadly based set of criteria is needed that provides the capability to relatively assess a wide range of disposition options/facilities in different countries and still ensure a global decrease in proliferation risk for plutonium.

  13. Fissile material disposition and proliferation risk

    International Nuclear Information System (INIS)

    Dreicer, J.S.; Rutherford, D.A.

    1996-01-01

    The proliferation risk of a facility is dependent on the material attractiveness, level of safeguards, and physical protection applied to the material in conjunction with an assessment of the impact of the socioeconomic circumstances and threat environment. Proliferation risk is a complementary extension of proliferation resistance. The authors believe a better determination of nuclear material proliferation can be achieved by establishing the proliferation risk for facilities that contain nuclear material. Developing a method that incorporates the socioeconomic circumstances and threat environment inherent to each country enables a global proliferation assessment. In order to effectively reduce the nuclear danger, a broadly based set of criteria is needed that provides the capability to relatively assess a wide range of disposition options/facilities in different countries and still ensure a global decrease in proliferation risk for plutonium

  14. Fissile material ban: global and non-discriminatory?

    International Nuclear Information System (INIS)

    Datt, Savita

    1995-01-01

    With the indefinite and unconditional extension of the nuclear Non-Proliferation Treaty (NPT) now out of the way, the next issue on the non-proliferation agenda is that of the existing stocks and further production of plutonium and weapons grade uranium. More than the existing stocks and the surplus fissile materials made available through arms control and disarmament measures, it is the further production of such materials which is sought to be tackled urgently. Of prime concern are the nuclear programmes of threshold countries like India, Pakistan and Israel (countries out of the NPT fold) which need to be capped at all costs. The best method of achieving this, it is believed can be through a global ban on the production of fissile materials. 15 refs

  15. Modeling of fissile material diversion in solvent extraction cascades

    International Nuclear Information System (INIS)

    Schneider, A.; Carlson, R.W.

    1980-01-01

    Changes were calculated for measurable parameters of a solvent extraction section of a reprocessing plant resulting from postulated fissile material diversion actions. The computer program SEPHIS was modified to calculate the time-dependent concentrations of uranium and plutonium in each stage of a cascade. The calculation of the inventories of uranium and plutonium in each contactor was also included. The concentration and inventory histories were computed for a group of four sequential columns during start-up and for postulated diversion conditions within this group of columns. Monitoring of column exit streams or of integrated column inventories for fissile materials could provide qualitative indications of attempted diversions. However, the time delays and resulting changes are complex and do not correlate quantitatively with the magnitude of the initiating event

  16. Gamma ray absorption of cylindrical fissile material with dual shields

    International Nuclear Information System (INIS)

    Wu Chenyan; Cheng Yiying; Huang Yongyi; Lu Fuquan; Yang Fujia

    2005-01-01

    This work analyzed the gamma ray attenuation effect from the self-absorption and shield attenuation perspectively. An exact mathematical equation was given for the geometric factor of the cylindrical fissile material with dual shields. In addition, several approximation approaches suitable for real situation were discussed, especially in the radial and axial directions of the cylinders, since the G-factors have simple forms. Then the space distribution patterns of the G-factor were analyzed based on numerical result and effective ways to solved the geometric information of the cylindrical fissile material, the radii and the heights, were deduced. This method was checked and verified by numerical calculation. Because of the efficiency of the method, it is ideal for application in real situations, such as nuclear safeguards, which demands speed of detection and accuracy of geometric analysis. (authors)

  17. User manual of FUNF code for fissile material data calculation

    International Nuclear Information System (INIS)

    Zhang, Jingshang

    2006-03-01

    The FUNF code (2005 version) is used to calculate fast neutron reaction data of fissile materials with incident energies from about 1 keV up to 20 MeV. The first version of the FUNF code was completed in 1994. the code has been developed continually since that time and has often been used as an evaluation tool for setting up CENDL and for analyzing the measurements of fissile materials. During these years many improvements have been made. In this manual, the format of the input parameter files and the output files, as well as the functions of flag used in FUNF code, are introduced in detail, and the examples of the format of input parameters files are given. FUNF code consists of the spherical optical model, the Hauser-Feshbach model, and the unified Hauser-Feshbach and exciton model. (authors)

  18. Covariance Spectroscopy for Fissile Material Detection

    International Nuclear Information System (INIS)

    Trainham, Rusty; Tinsley, Jim; Hurley, Paul; Keegan, Ray

    2009-01-01

    Nuclear fission produces multiple prompt neutrons and gammas at each fission event. The resulting daughter nuclei continue to emit delayed radiation as neutrons boil off, beta decay occurs, etc. All of the radiations are causally connected, and therefore correlated. The correlations are generally positive, but when different decay channels compete, so that some radiations tend to exclude others, negative correlations could also be observed. A similar problem of reduced complexity is that of cascades radiation, whereby a simple radioactive decay produces two or more correlated gamma rays at each decay. Covariance is the usual means for measuring correlation, and techniques of covariance mapping may be useful to produce distinct signatures of special nuclear materials (SNM). A covariance measurement can also be used to filter data streams because uncorrelated signals are largely rejected. The technique is generally more effective than a coincidence measurement. In this poster, we concentrate on cascades and the covariance filtering problem

  19. Screening of IAEA environmental samples for fissile material content

    International Nuclear Information System (INIS)

    Hembree, Doyle M. Jr.; Carter, Joel A.; Devault, Gerald L.; Whitaker, J. Michael; Glasgow, David

    2001-01-01

    Full text: Analysis of environmental samples for the International Atomic Energy Agency (IAEA) Strengthened Safeguards Systems program requires that stringent measures be taken to control contamination. To facilitate contamination control, it is extremely useful to have some estimate of the fissile content of a given sample prior to beginning sample preparation and analysis. This is particularly true for laboratories that employ clean rooms during sample preparation. A review of the analytical results for samples submitted between January 1, 1999 and September 1, 2000 revealed that the total uranium content values ranged from 0.2 to greater than 500,000 ng/sample. Poor estimates of the uranium or plutonium content in the samples have caused some of the laboratories in the IAEA Network of Analytical Laboratories (NWAL) to experience clean laboratory contamination, sample cross contamination, and non-ideal uranium spike additions. This has led to significant increases in analysis costs (e.g., recertification of clean rooms after removing contamination, and rerunning samples) and degradation in data quality. A number of methods have been proposed for screening environmental samples for fissile material content, including gamma spectrometry, x-ray fluorescence, kinetic phosphorimetry (KPA), and inductively coupled plasma-mass spectrometry (ICP-MS). Gamma spectrometry and x-ray fluorescence are suitable for screening samples with microgram or greater quantities of uranium. ICP-MS and KPA are used successfully in some DOE NWAL laboratories to screen environmental samples. A neutron activation analysis (NAA) method that offers numerous advantages over other screening techniques for environmental samples has recently been proposed. Fissile materials such as 239 Pu and 235 U can be made to undergo fission in the intense neutron field to which they are exposed during neutron activation analysis (NAA). Some of the fission products emit neutrons referred to as 'delayed

  20. Automated monitoring of fissile and fertile materials in incinerator residue

    International Nuclear Information System (INIS)

    Schoenig, F.C. Jr.; Glendinning, S.G.; Tunnell, G.W.; Zucker, M.S.

    1986-01-01

    This patent describes an apparatus for determining the fissile and fertile material content of incinerator residue contained in a manipulatable container. The apparatus comprises a main body member formed of neutron moderating material and formed with a well for receiving the container; a first plug formed of neutron reflecting material for closing the top of the well; and a second plug containing a first neutron source for alternatively closing the top of the well and for directing neutrons into the well. It also includes a second neutron source selectively positionable in the bottom of the well for directing neutrons into the well; manipulating means for placing the container in the well and removing the container therefrom and for selectively placing one of the first and second plugs in the top of the well. Neutron detectors are positioned within the neutron moderating material of the main body member around the sides of the well. At least one gamma ray detector is positioned adjacent the bottom of the well. A means receives and processes the signals from the neutron and gamma ray detectors when the container is in the well for determining the fissile and fertile material content of the incinerator residue in the container

  1. Contribution to fissile materials transportation in transit storage

    International Nuclear Information System (INIS)

    Silva, Teresinha de Moraes da

    2005-01-01

    The national and international standards for the transportation of fissile materials establish two indexes: Transport Index (Tl) and Criticality Safety Index (ISC). Besides, in non-exclusive transit, the largest of these indexes cannot overtake the value 50. Considering several groups to be transported, the sum of the transportation indexes cannot overtake 200 and the distance between them should be 6 meters This work aimed, as a primary target, to verify when an index is superior to another, in relation to the fissile materials studied, i.e., uranium oxides UO 2 , U 3 O 8 and uranium silicide U 3 Si 2 , taking into account the different enrichment grades. The result found is that the criticality safety index is always greater. As a second goal, it was tried to verify if there is any alteration in the case of these compounds aging process, i.e., alteration in transport index (Tl) due to gamma radiation of daughters radioisotopes in secular equilibrium. No alteration, was verified as the daughters contribution although considerable related to the transport index is very small concerning the criticality safety index. As a third target, it was tried to justify a distance equal to 6 meters, between each group of fissile material. The result showed that, in air media, the distance of 1 meter is sufficient, except for the UO 2 compound at 100% of enrichment, which reaches 2 meter while in the water means the distance of 40cm is enough for the compounds studied. This fact is of great importance when the cost of the necessary area in the transportation and storage is taken into consideration. (author)

  2. Fissile material disposition program final immobilization form assessment and recommendation

    International Nuclear Information System (INIS)

    Cochran, S.G.; Dunlop, W.H.; Edmunds, T.A.; MacLean, L.M.; Gould, T.H.

    1997-01-01

    Lawrence Livermore National Laboratory (LLNL), in its role as the lead laboratory for the development of plutonium immobilization technologies for the Department of Energy's Office of Fissile Materials Disposition (MD), has been requested by MD to recommend an immobilization technology for the disposition of surplus weapons- usable plutonium. The recommendation and supporting documentation was requested to be provided by September 1, 1997. This report addresses the choice between glass and ceramic technologies for immobilizing plutonium using the can-in-canister approach. Its purpose is to provide a comparative evaluation of the two candidate technologies and to recommend a form based on technical considerations

  3. Fissile materials in solution concentration measured by active neutron interrogation

    International Nuclear Information System (INIS)

    Romeyer Dherbey, J.; Passard, Ch.; Cloue, J.; Bignan, G.

    1993-01-01

    The use of the active neutron interrogation to measure the concentration of plutonium contained in flow solutions is particularly interesting for fuel reprocessing plants. Indeed, this method gives a signal which is in a direct relation with the fissile materials concentration. Moreover, it is less sensitive to the gamma dose rate than the other nondestructive methods. Two measure methods have been evolved in CEA. Their principles are given into details in this work. The first one consists to detect fission delayed neutrons induced by a 252 Cf source. In the second one fission prompt neutrons induced by a neutron generator of 14 MeV are detected. (O.M.)

  4. Warheads and Fissile Materials:Declarations and Counting

    International Nuclear Information System (INIS)

    Sutcliffe, W.G.

    1991-01-01

    This paper reviews some of the issues about verifying the dismantlement of nuclear warheads and controlling nuclear materials in the context of arms control objectives. It is asserted that information about the stockpiles of nuclear warheads and materials is necessary to analyze the impacts and verification requirements of arms control measures including warhead dismantlement and fissile material controls. It is proposed that the US and the Soviets engage in a series of declarations about their stockpiles of nuclear weapons and materials. It is also asserted that currently it is more important to verify that warheads are retired to safe, secure facilities than to verify their dismantlement. It is proposed that production of new or rebuilt warheads be limited to less than the number retired each year. Verifying the number of new and rebuilt warheads deployed and the number retired avoids many of the difficulties in verifying dismantlement and material controls

  5. Ensuring the 50 year life of a fissile material container

    International Nuclear Information System (INIS)

    Glass, R.E.; Towne, T.L.

    1997-12-01

    Sandia was presented with an opportunity in 1993 to design containers for the long term storage and transport of fissile material. This program was undertaken at the direction of the US Department of Energy and in cooperation with Lawrence Livermore National Laboratory and Los Alamos National Laboratory which were tasked with developing the internal fixturing for the contents. The hardware is being supplied by Allied Signal Federal Manufacturing and Technologies, and the packaging will occur at Mason and Hangar Corporation's Pantex Plant. The unique challenge was to design a container that could be sealed with the fissile material contents; and, anytime during the next 50 years, the container could be transported with only the need for the pre-shipment leak test. This required not only a rigorous design capable of meeting the long term storage and transportation requirements, but also resulted in development of a surveillance program to ensure that the container continues to perform as designed over the 50-year life. This paper addresses the design of the container, the testing that was undertaken to demonstrate compliance with US radioactive materials transport regulations, and the surveillance program that has been initiated to ensure the 50-year performance

  6. Disposition scenarios and safeguardability of fissile materials under START Treaty

    International Nuclear Information System (INIS)

    Pillay, K.K.S.

    1993-01-01

    Under the Strategic Arms Reduction Treaty (START-I) signed in 1991 and the Lisbon Protocol of 1992, a large inventory of fissile materials will be removed from the weapons fuel cycles of the United States and the Former Soviet Union (FSU). The Lisbon Protocol calls for Ukraine, Kazakstan, and Byelarus to become nonnuclear members of the treaty and for Russia to assume the responsibility of the treaty as a nuclear weapons state. In addition, the START-II Treaty, which was signed in 1993 by the United States and Russia, further reduces deployed nuclear warheads and adds to the inventory of excess special nuclear materials (SNM). Because storage of in-tact warheads has the potential for a open-quotes breakout,close quotes it would be desirable to dismantle the warheads and properly dispose of the SNMs under appropriate safeguards to prevent their reentry into the weapons fuel cycle. The SNM recovered from dismantled warheads can be disposed of in several ways, and the final choices may be up to the country having the title to the SNM. Current plans are to store them indefinitely, leaving serious safeguards concerns. Recognizing that the underlying objective of these treaties is to prevent the fissile materials from reentering the weapons fuel cycle, it is necessary to establish a verifiable disposal scheme that includes safeguards requirements. This paper identifies some realistic scenarios for the disposal of SNM from the weapons fuel cycle and examines the safeguardability of those scenarios

  7. Materials technology for accelerator production of fissile isotopes

    International Nuclear Information System (INIS)

    Horak, J.A.

    1978-02-01

    The materials used for the accelerator production of fissile isotopes must enable the facility to achieve maximum fuel production at a minimum cost. Neutron production in the target would be maximized by use of thorium cooled with Pb--56 percent Bi or with sodium. The thorium should be ion-plated with approximately 1 mil of nickel or stainless steel for retention of fission products. The target container will have to be replaced at frequent intervals because of the copious quantities of neutronically produced helium and hydrogen in the container. Replacement would coincide with shutdown of the facility for the removal of the fissile material produced. If sodium is used to cool both the target and fertile blanket, a simple basket-type target container could be used. This would greatly reduce radiation effects in the target container. Type 316 stainless steel or V--20 wt percent Ti should perform satisfactorily as a target container. The fertile blanket should be 233 Th or 238 U that is coated with approximately 1 mil of nickel or stainless steel and cooled with sodium. The blanket container could be an austenitic stainless steel such as type 304 or 316; some ferritic alloys may also provide a satisfactory blanket container. 31 references

  8. Mathematical model for choosing the nuclear safe matrix compositions for fissile material immobilization

    International Nuclear Information System (INIS)

    Gorshtein, A.I.; Matyunin, Yu.I.; Poluehktov, P.P.

    2000-01-01

    A mathematical model is proposed for preliminary choice of the nuclear safe matrix compositions for fissile material immobilization. The IBM PC computer software for nuclear safe matrix composition calculations is developed. The limiting concentration of fissile materials in the some used and perspective nuclear safe matrix compositions for radioactive waste immobilization is calculated [ru

  9. Verification of classified fissile material using unclassified attributes

    International Nuclear Information System (INIS)

    Nicholas, N.J.; Fearey, B.L.; Puckett, J.M.; Tape, J.W.

    1998-01-01

    This paper reports on the most recent efforts of US technical experts to explore verification by IAEA of unclassified attributes of classified excess fissile material. Two propositions are discussed: (1) that multiple unclassified attributes could be declared by the host nation and then verified (and reverified) by the IAEA in order to provide confidence in that declaration of a classified (or unclassified) inventory while protecting classified or sensitive information; and (2) that attributes could be measured, remeasured, or monitored to provide continuity of knowledge in a nonintrusive and unclassified manner. They believe attributes should relate to characteristics of excess weapons materials and should be verifiable and authenticatable with methods usable by IAEA inspectors. Further, attributes (along with the methods to measure them) must not reveal any classified information. The approach that the authors have taken is as follows: (1) assume certain attributes of classified excess material, (2) identify passive signatures, (3) determine range of applicable measurement physics, (4) develop a set of criteria to assess and select measurement technologies, (5) select existing instrumentation for proof-of-principle measurements and demonstration, and (6) develop and design information barriers to protect classified information. While the attribute verification concepts and measurements discussed in this paper appear promising, neither the attribute verification approach nor the measurement technologies have been fully developed, tested, and evaluated

  10. Portal monitoring for detecting fissile materials and chemical explosives

    International Nuclear Information System (INIS)

    Albright, D.

    1992-01-01

    The portal monitoring of pedestrians, packages, equipment, and vehicles entering or leaving areas of high physical security has been common for many years. Many nuclear facilities rely on portal monitoring to prevent the theft or diversion of plutonium and highly enriched uranium. At commercial airports, portals are used to prevent firearms and explosives from being smuggled onto airplanes. An August 1989 Federal Aviation Administration (FAA) regulation requires US airlines to screen luggage on international flights for chemical explosives. This paper reports that portal monitoring is now being introduced into arms-control agreements. Because some of the portal-monitoring equipment that would be useful in verifying arms-control agreements is already widely used as part of the physical security systems at nuclear facilities and commercial airports, the authors review these uses of portal monitoring, as well as its role in verifying the INF treaty. Then the authors survey the major types of portal-monitoring equipment that would be most useful in detecting nuclear warheads or fissile material

  11. Immobilization as a route to surplus fissile materials disposition

    International Nuclear Information System (INIS)

    Gray, L.W.; Kan, T.

    1995-01-01

    In the aftermath of the Cold War, the US and Russia have agreed to large reductions in nuclear weapons. To aid in the selection of long-term management options, DOE has undertaken a multifaceted study to select options for storage and disposition of plutonium (Pu) in keeping with the national policy that Pu must be subjected to the highest standards of safety, security, and accountability. One alternative being considered is immobilization. To arrive at a suitable immobilization form, the authors first reviewed published information on high-level waste (HLW) immobilization technologies in order to identify 72 possible Pu immobilization forms to be prescreened. Surviving forms were screened using multiattribute analysis to determine the most promising technologies. Promising immobilization families were further evaluated to identify chemical, engineering, environmental, safety, and health problems that remain to be solved prior to making technical decisions as to the viability of using the form for long-term disposition of plutonium. All data, analyses, and reports are being provided to the DOE Fissile Materials Disposition Project Office to support the Record of Decision that is anticipated in the fourth quarter of FY96

  12. R ampersand D plan for immobilization technologies: fissile materials disposition program. Revision 1.0

    International Nuclear Information System (INIS)

    Shaw, H.F.; Armantrout, G.A.

    1996-09-01

    In the aftermath of the Cold War, the US and Russia have agreed to large reductions in nuclear weapons. To aid in the selection of long- term fissile material management options, the Department of Energy's Fissile Materials Disposition Program (FMDP) is conducting studies of options for the storage and disposition of surplus plutonium (Pu). One set of alternatives for disposition involve immobilization. The immobilization alternatives provide for fixing surplus fissile materials in a host matrix in order to create a solid disposal form that is nuclear criticality-safe, proliferation-resistant and environmentally acceptable for long-term storage or disposal

  13. Fuel costs of a light water reactor with fissile material recycling

    International Nuclear Information System (INIS)

    Clauss, J.

    1984-01-01

    In the light of the present prices of natural uranium and separative work and fabrication costs, savings can be achieved by reloading recycled fissile material. As in all recycling techniques, the product recovered cannot meet the whole new requirement. No excessive economic expectations should be associated with fissile material recycling in ligth water reactors. The main advantages of the procedure are the conservation of resources and the safety against proliferation. Besides, the original purpose of reprocessing should not be forgotten, i.e., in addition to the recycling of fissile material, to have a safe and easy method of secular disposal of high level waste (concentrated fission products). (orig.) [de

  14. Non-proliferation, safeguards, and security for the fissile materials disposition program immobilization alternatives

    Energy Technology Data Exchange (ETDEWEB)

    Duggan, R.A.; Jaeger, C.D.; Tolk, K.M. [Sandia National Labs., Albuquerque, NM (United States); Moore, L.R. [Lawrence Livermore National Lab., CA (United States)

    1996-05-01

    The Department of Energy is analyzing long-term storage and disposition alternatives for surplus weapons-usable fissile materials. A number of different disposition alternatives are being considered. These include facilities for storage, conversion and stabilization of fissile materials, immobilization in glass or ceramic material, fabrication of fissile material into mixed oxide (MOX) fuel for reactors, use of reactor based technologies to convert material into spent fuel, and disposal of fissile material using geologic alternatives. This paper will focus on how the objectives of reducing security and proliferation risks are being considered, and the possible facility impacts. Some of the areas discussed in this paper include: (1) domestic and international safeguards requirements, (2) non-proliferation criteria and measures, (3) the threats, and (4) potential proliferation, safeguards, and security issues and impacts on the facilities. Issues applicable to all of the possible disposition alternatives will be discussed in this paper. However, particular attention is given to the plutonium immobilization alternatives.

  15. Detection of tiny amounts of fissile materials in large-sized containers with radioactive waste

    Science.gov (United States)

    Batyaev, V. F.; Skliarov, S. V.

    2018-01-01

    The paper is devoted to non-destructive control of tiny amounts of fissile materials in large-sized containers filled with radioactive waste (RAW). The aim of this work is to model an active neutron interrogation facility for detection of fissile ma-terials inside NZK type containers with RAW and determine the minimal detectable mass of U-235 as a function of various param-eters: matrix type, nonuniformity of container filling, neutron gen-erator parameters (flux, pulse frequency, pulse duration), meas-urement time. As a result the dependence of minimal detectable mass on fissile materials location inside container is shown. Nonu-niformity of the thermal neutron flux inside a container is the main reason of the space-heterogeneity of minimal detectable mass in-side a large-sized container. Our experiments with tiny amounts of uranium-235 (<1 g) confirm the detection of fissile materials in NZK containers by using active neutron interrogation technique.

  16. Update to the Fissile Materials Disposition program SST/SGT transportation estimation

    International Nuclear Information System (INIS)

    John Didlake

    1999-01-01

    This report is an update to ''Fissile Materials Disposition Program SST/SGT Transportation Estimation,'' SAND98-8244, June 1998. The Department of Energy Office of Fissile Materials Disposition requested this update as a basis for providing the public with an updated estimation of the number of transportation loads, load miles, and costs associated with the preferred alternative in the Surplus Plutonium Disposition Final Environmental Impact Statement (EIS)

  17. Exploiting Fission Chain Reaction Dynamics to Image Fissile Materials

    Science.gov (United States)

    Chapman, Peter Henry

    Radiation imaging is one potential method to verify nuclear weapons dismantlement. The neutron coded aperture imager (NCAI), jointly developed by Oak Ridge National Laboratory (ORNL) and Sandia National Laboratories (SNL), is capable of imaging sources of fast (e.g., fission spectrum) neutrons using an array of organic scintillators. This work presents a method developed to discriminate between non-multiplying (i.e., non-fissile) neutron sources and multiplying (i.e., fissile) neutron sources using the NCAI. This method exploits the dynamics of fission chain-reactions; it applies time-correlated pulse-height (TCPH) analysis to identify neutrons in fission chain reactions. TCPH analyzes the neutron energy deposited in the organic scintillator vs. the apparent neutron time-of-flight. Energy deposition is estimated from light output, and time-of-flight is estimated from the time between the neutron interaction and the immediately preceding gamma interaction. Neutrons that deposit more energy than can be accounted for by their apparent time-of-flight are identified as fission chain-reaction neutrons, and the image is reconstructed using only these neutron detection events. This analysis was applied to measurements of weapons-grade plutonium (WGPu) metal and 252Cf performed at the Nevada National Security Site (NNSS) Device Assembly Facility (DAF) in July 2015. The results demonstrate it is possible to eliminate the non-fissile 252Cf source from the image while preserving the fissileWGPu source. TCPH analysis was also applied to additional scenes in which theWGPu and 252Cf sources were measured individually. The results of these separate measurements further demonstrate the ability to remove the non-fissile 252Cf source and retain the fissileWGPu source. Simulations performed using MCNPX-PoliMi indicate that in a one hour measurement, solid spheres ofWGPu are retained at a 1sigma level for neutron multiplications M -˜ 3.0 and above, while hollowWGPu spheres are

  18. Requirements for timber and cadmium used in shielding for fissile material transport packaging

    International Nuclear Information System (INIS)

    1982-02-01

    This Code of Practice has been prepared as a guide for designers who require packaging for fissile materials. It should be noted that this document covers design requirements only and it is not a manufacturing specification which can be quoted on a manufacturing contract without qualification. Compliance with the regulations regarding the safe transport of fissile materials may be achieved by the provision of an effective shield embodying:- (a) a moderating material -usually one rich in hydrogen, such as wood - in order to thermalise incoming neutrons, and (b) a material - such as cadmium - with a large absorption cross-section for thermal neutrons, located between the moderator and the fissile material, in order to capture the incoming neutrons. This Code describes the requirements in two sections, one for each of these materials. (author)

  19. Standard problem exercise to validate criticality codes for large arrays of packages of fissile materials

    International Nuclear Information System (INIS)

    Whitesides, G.E.; Stephens, M.E.

    1986-01-01

    A study has been conducted by an Office of Economic Cooperation and Development-Committee on the Safety of Nuclear Installations (OECD-CSNI) Working Group that examined computational methods used to compute k/sub eff/ for large greater than or equal to5 3 arrays of fissile material (in which each unit is a substantial fraction of a critical mass). Five fissile materials that might typically be transported were used in the study. The ''packages'' used for this exercise were simplified to allow studies unperturbed by the variety of structural materials which would exist in an actual package. The only material present other than the fissile material was a variation in the moderator (water) surrounding the fissile material. Consistent results were obtained from calculations using several computational methods. That is, when the bias demonstrated by each method for actual critical experiments was used to ''correct'' the results obtained for systems for which there were no experimental data, there was good agreement between the methods. Two major areas of concern were raised by this exercise. First, the lack of experimental data for arrays with size greater than 5 3 limits validation for large systems. Second, there is a distinct possibility that the comingling of two shipments of unlike units could result in a reduction of the safety margins. Additional experiments and calculations will be required to satisfactorily resolve the remaining questions regarding the safe transport of large arrays of fissile materials

  20. Fissile material detection and control facility with pulsed neutron sources and digital data processing

    International Nuclear Information System (INIS)

    Romodanov, V.L.; Chernikova, D.N.; Afanasiev, V.V.

    2010-01-01

    Full text: In connection with possible nuclear terrorism, there is long-felt need of devices for effective control of radioactive and fissile materials in the key points of crossing the state borders (airports, seaports, etc.), as well as various customs check-points. In International Science and Technology Center Projects No. 596 and No. 2978, a new physical method and digital technology have been developed for the detection of fissile and radioactive materials in models of customs facilities with a graphite moderator, pulsed neutron source and digital processing of responses from scintillation PSD detectors. Detectability of fissile materials, even those shielded with various radiation-absorbing screens, has been shown. The use of digital processing of scintillation signals in this facility is a necessary element, as neutrons and photons are discriminated in the time dependence of fissile materials responses at such loads on the electronic channels that standard types of spectrometers are inapplicable. Digital processing of neutron and photon responses practically resolves the problem of dead time and allows implementing devices, in which various energy groups of neutrons exist for some time after a pulse of source neutrons. Thus, it is possible to detect fissile materials deliberately concealed with shields having a large cross-section of absorption of photons and thermal neutrons. Two models of detection and the control of fissile materials were advanced: 1. the model based on graphite neutrons moderator and PSD scintillators with digital technology of neutrons and photons responses separation; 2. the model based on plastic scintillators and detecting of time coincidences of fission particles by digital technology. Facilities that count time coincidences of neutrons and photons occurring in the fission of fissile materials can use an Am Li source of neutrons, e.g. that is the case with the AWCC system. The disadvantages of the facility are related to the issues

  1. Underground autocatalytic-criticality potential and its implications to weapons fissile- material disposition

    International Nuclear Information System (INIS)

    Choi, J.-S.

    1998-01-01

    Several options for weapons fissile-material disposition, such as once-through mixed- oxide (MOX) fuel in reactors or immobilisation in waste glass, would result in end products requiring geologic disposal. The criticality potential of the fissile end products containing U-235 and Pu-239 and the associated consequences in a geologic setting are important considerations for the final disposal of these materials. The possibility of underground criticality, and especially autocatalytic criticality, is affected by (1) groundwater leaking into a failed waste container, (2) preferential leaching of neutron absorbers or of fissile material from a failed container, and (3) preferential deposition of fissile material in the surrounding rock. Bowman and Venneri have pointed out that fissile material mixed with varying compositions of water and silica can undergo a nuclear chain reaction. Some configurations can become autocatalytically supercritical resulting in considerable energy release, terminated finally by disassembly. Some reviews rejected the Bowman and Venneri warning as implausible because of low probabilities of scenarios that could lead to such configurations. Sanchez et al. reported possible supercritical conditions in systems of Pu-SiO 2 -H 2 O and Pu-tuff-H 2 O but concluded that the probability of forming such combinations is extremely low. Kastenberg et al. studied the potential for autocatalytic criticality of plutonium or highly enriched uranium in the proposed Yucca Mountain geologic repository. They concluded that plutonium or uranium could, theoretically, become supercritical, but that such criticality is unlikely given the hydrology, geology and geochemistry of the Yucca Mountain site. These studies are not definitive. The possibility of criticality exists. Detailed mechanisms have not been sufficiently studied for clear conclusions on the probabilities of occurrence. More technical analysis is needed to understand the potential for underground

  2. The role of congress in future disposal of fissile materials from dismantled nuclear weapons

    International Nuclear Information System (INIS)

    Donnelly, W.H.; Davis, Z.S.

    1991-01-01

    Assuming the Soviet Union remains intact as a major power and the superpowers do not retrogress to a new Cold War era, it is likely that the United States and the Soviet Union will eventually agree to deep cuts in their nuclear arsenals. Future arms control agreements may be coupled with companion agreements to stop production of fissile materials for nuclear weapons, to dismantle the warheads of the nuclear weapons, and to dispose of their fissile materials to prevent reuse in new warheads. Such agreements would be negotiated by the U.S. executive branch but probably would require ratification, funding, and enabling legislation from the U.S. Congress if they are to succeed. There follows a brief review of the ideas for disposal of fissile materials from dismantled nuclear warheads and the potential role and influence of the Congress in the negotiation, ratification, and implementation of U.S.-Soviet agreements for such disposal

  3. Verification arrangements for the proposed fissile material cut-off treaty

    International Nuclear Information System (INIS)

    Bragin, V.

    2001-01-01

    Since the mid-1950's, an agreement to terminate the production of fissile material for nuclear weapons has been on the agenda. On December 16, 1993, the UNGA adopted Resolution A/RES/48/75/L which recommends ''the negotiation in the most appropriate international forum of a non-discriminatory, multilateral and internationally and effectively verifiable treaty banning the production of fissile material for nuclear weapons and other nuclear explosive devices''. The proposed Fissile Material Cut-off Treaty (FMCT) is still one of the most important items on the multilateral disarmament and non-proliferation agenda. Successful achievement of the FMCT would be an important step towards the goal of eliminating nuclear weapons. (author)

  4. Criticality Safety in the Handling of Fissile Material. Specific Safety Guide

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2014-05-15

    This Safety Guide provides guidance and recommendations on how to meet the relevant requirements for ensuring subcriticality when dealing with fissile material and for planning the response to criticality accidents. The guidance and recommendations are applicable to both regulatory bodies and operating organizations. The objectives of criticality safety are to prevent a self-sustained nuclear chain reaction and to minimize the consequences of this if it were to occur. The Safety Guide makes recommendations on how to ensure subcriticality in systems involving fissile materials during normal operation, anticipated operational occurrences, and, in the case of accident conditions, within design basis accidents, from initial design through commissioning, operation, and decommissioning and disposal.

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

    International Nuclear Information System (INIS)

    Zoltowski, T.

    1988-01-01

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

  6. Long-term criticality safety concerns associated with surplus fissile material disposition

    International Nuclear Information System (INIS)

    Choi, J.S.

    1995-01-01

    A substantial inventory of surplus fissile material would result from ongoing and planned dismantlement of US and Russian nuclear weapons. This surplus fissile material could be dispositioned by irradiation in nuclear reactors, and the resulting spent MOx fuel would be similar in radiation characteristics to regular LWR spent UO2 fuel. The surplus fissile material could also be immobilized into high-level waste forms, such as borosilicate glass, synroc, or metal-alloy matrix. The MOx spent fuel, or the immobilized waste forms, could then be directly disposed of in a geologic repository. Long-term criticality safety concerns arise because the fissile contents (i.e., Pu-239 and its decay daughter U-235) in these waste forms are higher than in LWR spent UO2 fuel. MOx spent fuel could contain 3 to 4 wt% of reactor-grade plutonium, compared to only 0.9 wt% of plutonium in LWR spent UO2 fuel. At some future time (tens of thousand of years), when the waste forms had deteriorated due to intruding groundwater, the water could mix with the long-lived fissile materials to form into a critical system. If the critical system is self-sustaining, somewhat like the natural-occurring reactor in OKLO, fission products produced could readily be available for dissolution and release out to the accessible environment, adversely affecting public health and safety. This paper will address ongoing activities to evaluate long-term criticality safety concerns associated with disposition of fissile material in a geologic setting. Issues to be addressed include the identification of a worst-case water-intrusion scenario and waste-form geometries which present the most concern for long-term criticality safety; and suggests of technical solutions for such concerns

  7. Implementation of safeguards and security for fissile materials disposition reactor alternative facilities

    International Nuclear Information System (INIS)

    Jaeger, C.D.; Duggan, R.A.; Tolk, K.M.

    1995-01-01

    A number of different disposition alternatives are being considered and include facilities which provide for long-ten-n and interim storage, convert and stabilize fissile materials for other disposition alternatives, immobilize fissile material in glass and/or ceramic material, fabricate fissile material into mixed oxide (MOX) fuel for reactors, use reactor based technologies to convert material into spent fuel, and dispose of fissile material using a number of geologic alternatives. Particular attention will be given to the reactor alternatives which include existing, partially completed, advanced or evolutionary LWRs and CANDU reactors. The various reactor alternatives are all very similar and include processing which converts Pu to a usable form for fuel fabrication, a MOX fuel fab facility located in either the US or in Europe, US LWRs or the CANDU reactors and ultimate disposal of spent fuel in a geologic repository. This paper focuses on how the objectives of reducing security risks and strengthening arms reduction and nonproliferation will be accomplished and the possible impacts of meeting these objectives on facility operations and design. Some of the areas in this paper include: (1) domestic and international safeguards requirements, (2) non-proliferation criteria and measures, (3) the threat, and (4) potential proliferation risks, the impacts on the facilities, and safeguards and security issues unique to the presence of Category 1 or strategic special nuclear material

  8. Nuclear energy - Fissile materials - Principles of criticality safety in storing, handling and processing

    International Nuclear Information System (INIS)

    1995-01-01

    This International Standard specifies the basic principles and limitations which govern operations with fissile materials. It discusses general criticality safety criteria for equipment design and for the development of operating controls, while providing guidance for the assessment of procedures, equipment, and operations. It does not cover quality assurance requirements or details of equipment or operational procedures, nor does it cover the effects of radiation on man or materials, or sources of such radiation, either natural or as the result of nuclear chain reactions. Transport of fissile materials outside the boundaries of nuclear establishments is not within the scope of this International Standard and should be governed by appropriate national and international standards and regulations. These criteria apply to operations with fissile materials outside nuclear reactors but within the boundaries of nuclear establishments. They are concerned with the limitations which must be imposed on operations because of the unique properties of these materials which permit them to support nuclear chain reactions. These principles apply to quantities of fissile materials in which nuclear criticality can be established

  9. The molten salt reactor option for beneficial use of fissile material from dismantled weapons

    International Nuclear Information System (INIS)

    Gat, U.; Engel, J.R.

    1991-01-01

    The Molten Salt Reactor (MSR) option for burning fissile fuel from dismantled weapons is examined and is found very suitable for the beneficial use of this fuel. MSRs can utilize any fissile fuel in continuous operation with no special modifications, as demonstrated in the Molten Salt Reactor Experiment. Thus, MSRs are flexible while maintaining their economy. Furthermore, MSRs require only a minimum of special fuel preparation. They can tolerate denaturing and dilution of their fuel. The size of fuel shipments can be determined to optimize safety and security-all of which supports nonproliferation and resists diversion. In addition, MSRs have inherent safety features that make them acceptable and attractive. They can burn fissile material completely or can convert it to other fuels. MSRs also have the potential for burning the actinides and delivering the waste in an optimal form, thus contributing to the solution of one of the major remaining problems in the deployment of nuclear power

  10. The back-end management of fissile material at SCK-CEN

    International Nuclear Information System (INIS)

    Noynaert, L.; Massaut, V.; Braeckeveldt, M.

    1999-01-01

    The back-end management of fissile materials at SCK-CEN mainly concerns the HEU spent fuel of the BR2 (MTR) and the LEU and MOX spent fuel of the BR3, the first PWR installed in Western Europe and in decommissioning since 1987. It also concerns the experimental fuels tested in the SCK-CEN facilities. Furthermore as a result of its R and D programs in reprocessing and characterisation of spent fuel, considerable amounts of fissile materials in all kinds of forms and characteristics are stored in the different laboratories. For these, six main types of fissile materials are identified: highly enriched uranium, experimental spent fuel from the fast breeder programmes, MOX fuel, low enriched fuel, natural uranium and lab fissile materials. For the BR2 and BR3 spent fuel, various options, i.e. reprocessing, dry storage in casks and dry storage in canisters were evaluated against criteria, e.g. available techniques, safety, waste production, overall costs and policies. As a result of these studies, it was decided to opt in the case of the HEU from the BR2 reactor for the reprocessing without recovery of uranium while for the LEU and MOX fuel from the BR3 reactor, the dry storage in containers was chosen. For the others, the studies are still in progress. (author)

  11. Recommended nuclear criticality safety experiments in support of the safe transportation of fissile material

    International Nuclear Information System (INIS)

    Tollefson, D.A.; Elliott, E.P.; Dyer, H.R.; Thompson, S.A.

    1993-01-01

    Validation of computer codes and nuclear data (cross-section) libraries using benchmark quality critical (or certain subcritical) experiments is an essential part of a nuclear criticality safety evaluation. The validation results establish the credibility of the calculational tools for use in evaluating a particular application. Validation of the calculational tools is addressed in several American National Standards Institute/American Nuclear Society (ANSI/ANS) standards, with ANSI/ANS-8.1 being the most relevant. Documentation of the validation is a required part of all safety analyses involving significant quantities of fissile materials. In the case of transportation of fissile materials, the safety analysis report for packaging (SARP) must contain a thorough discussion of benchmark experiments, detailing how the experiments relate to the significant packaging and contents materials (fissile, moderating, neutron absorbing) within the package. The experiments recommended in this paper are needed to address certain areas related to transportation of unirradiated fissile materials in drum-type containers (packagings) for which current data are inadequate or are lacking

  12. Detection of tiny amounts of fissile materials in large-sized containers with radioactive waste

    Directory of Open Access Journals (Sweden)

    Batyaev V.F.

    2018-01-01

    Full Text Available The paper is devoted to non-destructive control of tiny amounts of fissile materials in large-sized containers filled with radioactive waste (RAW. The aim of this work is to model an active neutron interrogation facility for detection of fissile ma-terials inside NZK type containers with RAW and determine the minimal detectable mass of U-235 as a function of various param-eters: matrix type, nonuniformity of container filling, neutron gen-erator parameters (flux, pulse frequency, pulse duration, meas-urement time. As a result the dependence of minimal detectable mass on fissile materials location inside container is shown. Nonu-niformity of the thermal neutron flux inside a container is the main reason of the space-heterogeneity of minimal detectable mass in-side a large-sized container. Our experiments with tiny amounts of uranium-235 (<1 g confirm the detection of fissile materials in NZK containers by using active neutron interrogation technique.

  13. 10 CFR 71.59 - Standards for arrays of fissile material packages.

    Science.gov (United States)

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 false Standards for arrays of fissile material packages. 71.59 Section 71.59 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PACKAGING AND TRANSPORTATION OF RADIOACTIVE.... The value of the CSI may be zero provided that an unlimited number of packages are subcritical, such...

  14. Safeguard and security issues for the U.S. Fissile Materials Disposition Program

    International Nuclear Information System (INIS)

    Jaeger, C.D.; Moya, R.W.; Duggan, R.A.

    1995-01-01

    The Department of Energy's Office of Materials Disposition (MD) is analyzing long-term storage and disposition options for fissile materials, preparing a Programmatic Environmental Impact Statement (PEIS), preparing for a Record of Decision (ROD) regarding this material, and conducting other related activities. A primary objective of this program is to support U.S. nonproliferation policy by reducing major security risks. Particular areas of concern are the acquisition of this material by unauthorized persons and preventing the reintroduction of the material for use in weapons. This paper presents some of the issues, definitions, and assumptions addressed by the Safeguards and Security Project Team in support of the Fissile Materials Disposition Program (FMDP). The discussion also includes some preliminary ideas regarding safeguards and security criteria that are applicable to the screening of disposition options

  15. Safeguards and security issues for the U.S. Fissile Materials Disposition Program

    International Nuclear Information System (INIS)

    Jaeger, C.D.; Moya, R.W.; Duggan, R.A.

    1995-01-01

    The Department of Energy's Office of Materials Disposition (MD) is analyzing long-term storage and disposition options for fissile materials, preparing a Programmatic Environmental Impact Statement (PEIS), preparing for a Record of Decision (ROD) regarding this material, and conducting other related activities. A primary objective of this program is to support US nonproliferation policy by reducing major security risks. Particular areas of concern are the acquisition of this material by unauthorized persons and preventing the reintroduction of the material for use in weapons. This paper presents some of the issues, definitions, and assumptions addressed by the Safeguards and Security Project Team in support of the Fissile Materials Disposition Program (FMDP). The discussion also includes some preliminary ideas regarding safeguards and security criteria that are applicable to the screening of disposition options

  16. Proliferation resistance criteria for fissile material disposition issues

    International Nuclear Information System (INIS)

    Rutherford, D.A.; Fearey, B.L.; Markin, J.T.; Close, D.A.; Tolk, K.M.; Mangan, D.L.; Moore, L.

    1995-01-01

    The 1994 National Acdaemy of Sciences study ''Management and Disposition of Excess Weapons Plutonium'' defined options for reducing the national and international proliferation risks of materials declared excess to the nuclear weapons program. This paper proposes criteria for assessing the proliferation resistance of these options as well defining the ''Standards'' from the report. The criteria are general, encompassing all stages of the disposition process from storage through intermediate processing to final disposition including the facilities, processing technologies and materials, the level of safeguards for these materials, and the national/subnational threat to the materials

  17. Fissile materials and international security in the post-Cold War world

    International Nuclear Information System (INIS)

    Anon.

    1996-01-01

    It is essential that members of industry, government and international organizations be able to come together to discuss the latest developments in this vital field at events such as this. Given the number of years this organization has devoted to the issue, the INMM must find it interesting that the control of fissile materials has become such a high-profile issue in the policy and political communities. But, this evolution in policy is a natural outgrowth of the changing world situation. While just 10 years ago the US and Soviet Union were churning out the fissile materials needed for weapons, today these former rivals are working together, hand in hand, to corral the danger posed by these materials. And, while it is clear that the world no longer lives on the edge of nuclear war, the nuclear danger still exists, though in a less obvious and perhaps more insidious form. It is a great challenge in this post-Cold War world to contain this nuclear threat. It is prudent and necessary for the US to be in the forefront of efforts to address and tame this problem. The fundamental threat posed by the proliferation of nuclear weapons and materials is a direct challenge to US and world security. President Clinton has clearly recognized the changed nature of the nuclear danger. To meet this challenge, he has labored to put in place a comprehensive and integrated plan for addressing this threat. The US Department of Energy has a unique role in this effort because, as an institution with many decades of experience in fissile material matters, it is able to provide expertise and technical analyses that are essential in defining and implementing policy prescriptions. The president's comprehensive plan to prevent nuclear proliferation and reduce the danger posed by weapons-usable nuclear materials has four essential elements: secure existing nuclear material stockpiles; limit fissile material production and use, eliminate warheads, and strengthen the nonproliferation regime

  18. Glass material oxidation and dissolution system: Converting miscellaneous fissile materials to glass

    International Nuclear Information System (INIS)

    Forsberg, C.W.; Ferrada, J.J.

    1996-01-01

    The cold war and the development of nuclear energy have resulted in significant inventories of miscellaneous fissile materials (MFMs). MFMs include (1) plutonium scrap and residue, (2) miscellaneous spent nuclear fuel (SNF), (3) certain hot cell wastes, and (4) many one-of-a-kind materials. Major concerns associated with the long-term management of these materials include: safeguards and nonproliferation issues; health, environment, and safety concerns. waste management requirements; and high storage costs. These issues can be addressed by converting the MFMs to glass for secure, long-term storage or repository disposal; however, conventional glass-making processes require oxide-like feed materials. Converting MFMs to oxide-like materials with subsequent vitrification is a complex and expensive process. A new vitrification process has been invented, the Glass Material Oxidation and Dissolution System (GMODS), which directly converts metals, ceramics, and amorphous solids to glass; oxidizes organics with the residue converted to glass; and converts chlorides to borosilicate glass and a secondary sodium chloride (NaCl) stream. Laboratory work has demonstrated the conversion of cerium (a plutonium surrogate), uranium, Zircaloy, stainless steel, multiple oxides, and other materials to glass. However, significant work is required to develop GMODS further for applications at an industrial scale. If implemented, GMODS will provide a new approach to manage these materials

  19. IAEA verification of weapon-origin fissile material in the Russian Federation and the United States

    International Nuclear Information System (INIS)

    2001-01-01

    The Secretary of Energy of the United States, Spencer Abraham, Minister of the Russian Federation on Atomic Energy, Alexander Rumyantsev, and Director General of the International Atomic Energy Agency (IAEA), Mohamed ElBaradei, met in Vienna on 18 September 2001 to review progress on the Trilateral Initiative. The Initiative was launched in 1996 to develop a new IAEA verification system for weapon-origin material designated by the United States and the Russian Federation as released from their defence programmes. The removal of weapon-origin fissile material from the defence programmes of the Russian Federation and the United States is in furtherance of the commitment to disarmament undertaken by the two States pursuant to Article VI of the Treaty on the Non-Proliferation of Nuclear Weapons (NPT). IAEA verification under this Initiative is intended to promote international confidence that fissile material made subject by either of the two States to Agency verification remains irreversibly removed from nuclear weapon programmes

  20. Processing fissile material mixtures containing zirconium and/or carbon

    Science.gov (United States)

    Johnson, Michael Ernest; Maloney, Martin David

    2013-07-02

    A method of processing spent TRIZO-coated nuclear fuel may include adding fluoride to complex zirconium present in a dissolved TRIZO-coated fuel. Complexing the zirconium with fluoride may reduce or eliminate the potential for zirconium to interfere with the extraction of uranium and/or transuranics from fission materials in the spent nuclear fuel.

  1. Assessment and recommendations for fissile-material packaging exemptions and general licenses within 10 CFR Part 71

    International Nuclear Information System (INIS)

    Parks, C.V.; Hopper, C.M.; Lichtenwalter, J.L.

    1998-07-01

    This report provides a technical and regulatory assessment of the fissile material general licenses and fissile material exemptions within Title 10 of the Code of Federal Regulations Part 71. The assessment included literature studies and calculational analyses to evaluate the technical criteria; review of current industry practice and concerns; and a detailed evaluation of the regulatory text for clarity, consistency and relevance. Recommendations for potential consideration by the Nuclear Regulatory Commission staff are provided. The recommendations call for a simplification and consolidation of the general licenses and a change in the technical criteria for the first fissile material exemptions

  2. India and the fissile material cut-off treaty: policy options

    International Nuclear Information System (INIS)

    Nayan, Rajiv

    2011-01-01

    The international community inside and outside the Conference of Disarmament is underscoring the need for concluding a fissile material cut-off treaty (FMCT). The Indian government, for a long period, has been sponsoring the idea. Notwithstanding the international stagnation, the issue has been instigating periodic debate in India on the Indian approach. The periodic revival of the issue requires that India revisit its policy on fissile material production as well as its approach towards a possible EVICT. This article examines the question: should India's approach to conclude an FMCT be within the UN institutional framework? The new international reality is pushing for a new context, new realignments and a fresh outlook for an FMCT. India should take its own time to support conclusion of an FMCT so that its national interests and security are not adversely affected. (author)

  3. Update on Monitoring Technologies for International Safeguards and Fissile Material Verification

    International Nuclear Information System (INIS)

    Croessmann, C. Dennis; Glidewell, Don D.; Mangan, Dennis L.; Smathers, Douglas C.

    1999-01-01

    Monitoring technologies are playing an increasingly important part in international safeguards and fissile material verification. The developments reduce the time an inspector must spend at a site while assuring continuity of knowledge. Monitoring technologies' continued development has produced new seal systems and integrated video surveillance advances under consideration for Trilateral Initiative use. This paper will present recent developments for monitoring systems at Embalse, Argentina, VNHEF, Sarov, Russian, and Savannah River Site, Aiken, South Carolina

  4. Applications of the ANSI/ANS standard on the storage of fissile materials

    International Nuclear Information System (INIS)

    Thomas, J.T.

    1985-01-01

    The American National Standard ''Guide for Nuclear Criticality Safety in the Storage of Fissile Materials,'' ANSI/N16.5-1975 is the subject of this paper. The 'Guide' was reaffirmed in 1982. The technical bases for the conditions and requirements are discussed. Suggestions for applications and several general problems addressed by the Guide are presented. The development of information needed for future extensions of the area of applicability is given

  5. A review of the prospects for fusion breeding of fissile material

    International Nuclear Information System (INIS)

    Geiger, J.S.; Bartholomew, G.A.

    1981-10-01

    This report is the result of an eight month study by the AECL Fusion Status Study Group. The objectives of this study were to review the current status of fusion research, to evaluate the neutronic performance of various fusion-breeder systems, and to assess the economic and technological outlook for the fusion breeder as a source of fissile material to support CANDU reactors operating on the thorium fuel cycle

  6. Unified instrumentation for determining fissile and radioactive materials

    International Nuclear Information System (INIS)

    Voronov, V.L.; Gorokhov, V.A.; Drozdov, V.Yu.; Morozov, O.S.; Novikov, V.M.

    1999-01-01

    The instrumentation is aimed to equip various facilities: nuclear facilities (including radioactive plant and nuclear material storages), border check stations at the customs, transport junctions, administrative buildings and other facilities. The monitor under design are based on the gamma-spectrometric method of radiation monitoring which consists in recording and analyzing characteristics of X-ray and gamma-sources power spectra within the range of 40-3000 keV at the background level whose value is measured and taken into account during the signal analysis. The designed universal set of instrumentation based on common technical solutions and metrological support plus its small dimensions allows to install it actually in any check point without any significant changes in the room lay-out to facilitate its maintenance [ru

  7. Actualization of physical-chemical properties and criticality data of specific fissile materials

    International Nuclear Information System (INIS)

    Strauch, V.; Deutsch, K.H.

    1991-09-01

    The purpose of this project is to update the criticality curves contained in DIN 25 403, Parts 2-8. This report contains criticality data for aqueous uranium and plutonium systems of various concentrations for spherical, cylindrical and layer geometries. The critical dimensions were calculated with the single dimensional transport code XSDRNPM-S and the 27 group-library from Scale 3.1. A 30 cm thick water reflector was taken into account. The critical masses were obtained by multiplying the volume of a critical sphere with the fissile material concentration. The moderator/fissile material relationship for each of the investigated concentration ranges were described. Checks were made using experiments with comparable fissile material systems. Due to the complex geometry of some of the chosen experiments some calculation checks were carried out using the Monte-Carlo-Codes KENO IV-S and Va. The calculation results compared very well with the experiments. Comparison of the results with the currently valid DIN curves does not show any serious differences. The new values lie however slightly below the current values and therefore represent conservative values, so that the criticality curves of DIN 25 403, Parts 2-6 and 8 should be replaced. (orig./HP) [de

  8. Harmonisation of criticality assessments of packages for the transport of fissile nuclear fuel cycle materials

    International Nuclear Information System (INIS)

    Farrington, L.

    2004-01-01

    The transport of fissile nuclear fuel cycle materials is an international business, and for international shipments the regulations require a package to be certified by each country through or into which the consignment is to be transported. This raises a number of harmonisation issues, which have an important bearing on transport activities. National authorities carry out independent reviews of the criticality safety of packages containing fissile materials but the underlying assumptions used in the calculations can differ, and the outcome is that implementation of the regulations is not uniform. A single design may require multiple criticality analyses to obtain base approval and foreign validations. When several competent authorities are involved, the approval and validation process of package design can often become a time-consuming, expensive and unpredictably lengthy process that can have a significant detrimental effect upon the businesses involved. The characteristics of the fissile nuclear fuel cycle materials transported by the various countries have much in common and so have the designs of the packages to contain them. A greater degree of standardisation should allow criticality safety to be assessed consistently and efficiently with benefits for the nuclear transport industry and the regulatory bodies. (author)

  9. Harmonisation of criticality assessments of packages for the transport of fissile nuclear fuel cycle materials

    International Nuclear Information System (INIS)

    Farrington, L.

    2004-01-01

    The transport of fissile nuclear fuel cycle materials is an international business and for international shipments the regulations require a package to be certified by each country through or into which the consignment is to be transported. This raises a number of harmonisation issues, which have an important bearing on transport activities. National authorities carry out independent reviews of criticality safety of packages containing fissile materials but the underlying assumptions used in the calculations can differ, and the outcome is that implementation of the regulations is not uniform. A single design may require multiple criticality analyses to obtain base approval and foreign validations. When several Competent Authorities are involved, the approval and validation process of package design can often become time consuming, expensive and an unpredictably lengthy process that can have a significant detrimental effect upon the businesses involved. The characteristics of the fissile nuclear fuel cycle materials transported by the various countries have much in common and so have the designs of the packages to contain them. A greater degree of standardisation should allow criticality safety to be assessed consistently and efficiently with benefits for the nuclear transport industry and the regulatory bodies

  10. Verification of a Fissile Material Cut-off Treaty (FMCT): The Potential Role of the IAEA

    International Nuclear Information System (INIS)

    Chung, Jin Ho

    2016-01-01

    The objective of a future verification of a FMCT(Fissile Material Cut-off Treaty) is to deter and detect non-compliance with treaty obligations in a timely and non-discriminatory manner with regard to banning the production of fissile material for nuclear weapons or other nuclear devices. Since the International Atomic Energy Agency (IAEA) has already established the IAEA safeguards as a verification system mainly for Non -Nuclear Weapon States (NNWSs), it is expected that the IAEA's experience and expertise in this field will make a significant contribution to setting up a future treaty's verification regime. This paper is designed to explore the potential role of the IAEA in verifying the future treaty by analyzing verification abilities of the Agency in terms of treaty verification and expected challenges. Furthermore, the concept of multilateral verification that could be facilitated by the IAEA will be examined as a measure of providing a credible assurance of compliance with a future treaty. In this circumstance, it is necessary for the IAEA to be prepared for playing a leading role in FMCT verifications as a form of multilateral verification by taking advantage of its existing verification concepts, methods, and tools. Also, several challenges that the Agency faces today need to be overcome, including dealing with sensitive and proliferative information, attribution of fissile materials, lack of verification experience in military fuel cycle facilities, and different attitude and culture towards verification between NWSs and NNWSs

  11. Verification of a Fissile Material Cut-off Treaty (FMCT): The Potential Role of the IAEA

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Jin Ho [Korea Institute of Nuclear Nonproliferation and Control, Daejeon (Korea, Republic of)

    2016-05-15

    The objective of a future verification of a FMCT(Fissile Material Cut-off Treaty) is to deter and detect non-compliance with treaty obligations in a timely and non-discriminatory manner with regard to banning the production of fissile material for nuclear weapons or other nuclear devices. Since the International Atomic Energy Agency (IAEA) has already established the IAEA safeguards as a verification system mainly for Non -Nuclear Weapon States (NNWSs), it is expected that the IAEA's experience and expertise in this field will make a significant contribution to setting up a future treaty's verification regime. This paper is designed to explore the potential role of the IAEA in verifying the future treaty by analyzing verification abilities of the Agency in terms of treaty verification and expected challenges. Furthermore, the concept of multilateral verification that could be facilitated by the IAEA will be examined as a measure of providing a credible assurance of compliance with a future treaty. In this circumstance, it is necessary for the IAEA to be prepared for playing a leading role in FMCT verifications as a form of multilateral verification by taking advantage of its existing verification concepts, methods, and tools. Also, several challenges that the Agency faces today need to be overcome, including dealing with sensitive and proliferative information, attribution of fissile materials, lack of verification experience in military fuel cycle facilities, and different attitude and culture towards verification between NWSs and NNWSs.

  12. Requirements for the transport of surplus fissile materials in the United States

    International Nuclear Information System (INIS)

    Wilson, R.K.

    1995-01-01

    This paper discusses the requirements and issues associated with the transportation of surplus fissile materials in the United States. The paper describes the materials that will be transported, the permissible modes of transport for these materials, and the safety and security requirements for each mode of transport. The paper also identifies transportation issues associated with these requirements, including the differences in requirements corresponding to who owns the material and whether the transport is on-site or off-site. Finally, the paper provides a discussion that suggests that by adopting the spent fuel standard and stored weapon standard proposed by the National Academy of Sciences, the requirements for transportation become straightforward

  13. Disposition of excess fissile materials in deep boreholes

    International Nuclear Information System (INIS)

    Halsey, W.G.; Danker, W.; Morley, R.

    1995-09-01

    As a result of recent changes throughout the world, a substantial inventory of excess separated plutonium is expected to result from dismantlement of US nuclear weapons. The safe and secure management and eventual disposition of this plutonium, and of a similar inventory in Russia, is a high priority. A variety of options (both interim and permanent) are under consideration to manage this material. The permanent solutions can be categorized into two broad groups: direct disposal and utilization. Plutonium utilization options have in common the generation of high-level radioactive waste which will be disposed of in a mined geologic disposal system to be developed for spent reactor fuel and defense high level waste. Other final disposition forms, such as plutonium metal, plutonium oxide and plutonium immobilized without high-level radiation sources may be better suited to placement in a custom facility. This paper discusses a leading candidate for such a facility; deep (several kilometer) borehole disposition. The deep borehole disposition concept involves placing excess plutonium deep into old stable rock formations with little free water present. The safety argument centers around ancient groundwater indicating lack of migration, and thus no expected communication with the accessible environment until the plutonium has decayed

  14. Experimental verification of neutron emission method for measuring of fissile material content in spent fuel

    International Nuclear Information System (INIS)

    Abou-Zaid, A.A.; Pytel, K.

    1999-01-01

    A non-destructive method of measurement of fissile nuclides content remained in spent fuel from research reactor is presented. The method, called the neutron emission one, is based on counting of fission neutrons emitted from fissile isotopes: 235 U, 239 Pu, 241 Pu. Fissions are induced mainly by neutrons supplied by the external neutron source. Another effects contribute also to the measured neutron population, e. g. source neutrons from penetrating the fuel without being captured and scattered, neutrons (α,n) reactions and from spontaneous fissions of actinides. Complexity of phenomena occurring within the measurement facility required the detailed numerical simulation and experimental studies prior design of ultimate measurement stand. In the previous paper, the results of Monte Carlo simulation on optimisation of measuring stand for neutron emission method were presented. On the basis of those results, the experimental stand for Maria reactor fuel investigation has been designed and manufactured. The present paper, being the continuation of previous one, contains the description of experimental facility and the results of measurements for the fresh fuel (without burnup) and the fuel mock-up (without fissile materials). Although some discrepancies were found between Monte Carlo and experimental results, the main conclusions concerning the optimal geometry of measuring facility have been confirmed. (author)

  15. Fissile material and international security in the post-Cold War world

    International Nuclear Information System (INIS)

    Luongo, K.N.

    1995-01-01

    Given the number of years this organization has devoted to the issue, the INMM must find it quite interesting that the control of fissile materials has become such a high profile issue in the policy and political communities. But, this evolution in policy is a natural outgrowth of the changing world situation. While just ten years ago the United States and the Soviet Union were churning out the fissile materials needed for weapons, today these former rivals are working together, hand in hand, to corral the danger posed by these materials. And, while it is clear that the world no longer lives on the edge of nuclear war, the nuclear danger still exists, though in a less obvious and perhaps more insidious form. It is a great challenge in this post Cold War-world to contain this nuclear threat. It is prudent and necessary for the United States to be in the forefront of efforts to address and tame this problem. The fundamental threat posed by the proliferation of nuclear weapons and materials is a direct challenge to US and world security. President Clinton has clearly recognized the changed nature of the nuclear danger. To meet this challenge, he also labored to put in place a comprehensive and integrated plan for addressing this threat. The Department of Energy has a unique role in this effort because, as an institution with man decades of experience in fissile material matters, it is able to provide expertise and technical analyses which are essential in defining and implementing policy prescriptions. The President's comprehensive plan to prevent nuclear proliferation and reduce the danger posed by weapons-usable nuclear materials has four essential elements: (1) secure existing stockpiles; (2) limit production and use; (3) eliminate warheads; and (4) strengthen the nonproliferation regime

  16. Prospects for a fissile material cut-off: Achieving a successful NPT review process

    International Nuclear Information System (INIS)

    Kalinowski, M.

    1999-01-01

    Finding new and creative ways to overcome the current deadlock in progress in nuclear arms control became the most important question in the past year. For a long time it had been expected that after the conclusion of the Comprehensive Test Ban Treaty, the next step would be to ban production of fissile materials for weapon purposes. Three strategies are proposed for reaching relevant cut-off agreements. First suggests possible fore for achievement of relevant agreements, second is the proposal to begin with international register of inventories and production capabilities for all relevant nuclear materials, and the third one is ti identify equivalent steps obligatory for all the parties involved

  17. Comparative analysis of non-destructive methods to control fissile materials in large-size containers

    Directory of Open Access Journals (Sweden)

    Batyaev V.F.

    2017-01-01

    Full Text Available The analysis of various non-destructive methods to control fissile materials (FM in large-size containers filled with radioactive waste (RAW has been carried out. The difficulty of applying passive gamma-neutron monitoring FM in large containers filled with concreted RAW is shown. Selection of an active non-destructive assay technique depends on the container contents; and in case of a concrete or iron matrix with very low activity and low activity RAW the neutron radiation method appears to be more preferable as compared with the photonuclear one.

  18. Improvements of neutron activation techniques for the determination of fissile material concentrations

    International Nuclear Information System (INIS)

    Papadopoulos, N.N.

    1987-01-01

    Certain experimental improvements, as variable sample size and irradiation position, automation and flexibility in radiation detection, broaden the measurable concentration range, increase the possible rate and accuracy of analysis and enlarge the application range of home-made nuclear analyzer for fissile material analysis by delayed fission neutron counting and for short-lived multielement analysis by neutron activation gamma-ray spectrometry. Intercomparisons of results by various methods and laboratories show the need for regular checks of techniques to ensure reliable measurements. (author)

  19. The preliminary design of real-time neutron fissile material monitoring system

    International Nuclear Information System (INIS)

    Shi Jun; Ren Zhongguo; Zhang Ming; Zhao Zhiping; Chen Qi

    2013-01-01

    In this paper we present the preliminary design to carry out real-time neutron fissile material monitoring system, The system includes hardware and data acquisition software. For the hardware, it is employed with He3 proportional tubes as neutron detectors, polyethylene as moderator, and, to achieve the remote counting, RM4036 counting modules are connected to the remote computer through the 485 ports. The software with real-time data display and storage, alarm and other functions are developed using Visual Basic 6.0. (authors)

  20. IAEA technical meeting on fissile material strategies for sustainable nuclear energy

    International Nuclear Information System (INIS)

    Ganguly, Chaitanyamoy; Koyama, Kazutoshi

    2005-01-01

    A Technical Meeting (TM) on 'Fissile Material Management Strategies for Sustainable Nuclear Energy' was organized by the International Atomic Energy Agency (IAEA) in Vienna from 12 to 15 September 2005. Prior to the TM, three Working Groups (WG) composed of experts from 10 countries prepared Key Issues papers on: 1) Uranium Demand and Supply through 2050; 2) Back-end Fuel Cycle Options; and 3) Sustainable Nuclear Energy beyond 2050: Cross-cutting Issues. Some 36 papers, including 3 key issue papers, were presented during the TM in 3 different sessions. The present paper summarizes the deliberations of the TM. (author)

  1. Comparative analysis of non-destructive methods to control fissile materials in large-size containers

    Science.gov (United States)

    Batyaev, V. F.; Sklyarov, S. V.

    2017-09-01

    The analysis of various non-destructive methods to control fissile materials (FM) in large-size containers filled with radioactive waste (RAW) has been carried out. The difficulty of applying passive gamma-neutron monitoring FM in large containers filled with concreted RAW is shown. Selection of an active non-destructive assay technique depends on the container contents; and in case of a concrete or iron matrix with very low activity and low activity RAW the neutron radiation method appears to be more preferable as compared with the photonuclear one. Note to the reader: the pdf file has been changed on September 22, 2017.

  2. A method for managing the storage of fissile materials using criticality indices

    International Nuclear Information System (INIS)

    Philbin, J.S.; Harms, G.A.

    1995-01-01

    This paper describes a method for criticality control at fissile material storage facilities. The method involves the use criticiality indices for storage canisters. The logic, methodology, and results for selected canisters are presented. A concept for an interactive computer program using the method is also introduced. The computer program can be used in real time (using precalulated data) to select a Criticality Index (CI) for a container when it is delivered to or packaged at a site. Criticality safety is assured by controlling the sum of the CIs at each storage location below a defined Emit value when containers are moved

  3. Fissile materials in solution concentration measured by active neutron interrogation; Mesure de concentration en matiere fissile dans les liquides par interrogation neutronique active

    Energy Technology Data Exchange (ETDEWEB)

    Romeyer Dherbey, J.; Passard, Ch.; Cloue, J.; Bignan, G.

    1993-12-31

    The use of the active neutron interrogation to measure the concentration of plutonium contained in flow solutions is particularly interesting for fuel reprocessing plants. Indeed, this method gives a signal which is in a direct relation with the fissile materials concentration. Moreover, it is less sensitive to the gamma dose rate than the other nondestructive methods. Two measure methods have been evolved in CEA. Their principles are given into details in this work. The first one consists to detect fission delayed neutrons induced by a {sup 252} Cf source. In the second one fission prompt neutrons induced by a neutron generator of 14 MeV are detected. (O.M.). 6 refs.

  4. Development of a Fissile Materials Irradiation Capability for Advanced Fuel Testing at the MIT Research Reactor

    International Nuclear Information System (INIS)

    Hu Linwen; Bernard, John A.; Hejzlar, Pavel; Kohse, Gordon

    2005-01-01

    A fissile materials irradiation capability has been developed at the Massachusetts Institute of Technology (MIT) Research Reactor (MITR) to support nuclear engineering studies in the area of advanced fuels. The focus of the expected research is to investigate the basic properties of advanced nuclear fuels using small aggregates of fissile material. As such, this program is intended to complement the ongoing fuel evaluation programs at test reactors. Candidates for study at the MITR include vibration-packed annular fuel for light water reactors and microparticle fuels for high-temperature gas reactors. Technical considerations that pertain to the design of the MITR facility are enumerated including those specified by 10 CFR 50 concerning the definition of a research reactor and those contained in a separate license amendment that was issued by the U.S. Nuclear Regulatory Commission to MIT for these types of experiments. The former includes limits on the cross-sectional area of the experiment, the physical form of the irradiated material, and the removal of heat. The latter addresses experiment reactivity worth, thermal-hydraulic considerations, avoidance of fission product release, and experiment specific temperature scrams

  5. Proceedings from the Fissile Material Cut-off seminar in Stockholm

    International Nuclear Information System (INIS)

    Arbman, G.

    1998-01-01

    The Swedish Defence Research Establishment hosted an international expert seminar on the subject of verifying a prohibition of the production of fissile material for nuclear weapons purpose (cut-off) in Stockholm, June 3-5 1998. The objective of the seminar was to provide an opportunity for informal discussions among scientific and technical experts on various technical matters relating to the verification of a future Fissile Material Cut-off Treaty (FMCT). A stated aim of the seminar was to keep issues of scope to a minimum. Invited speakers and commentators were given an opportunity to present their views as written contributions. The present seminar proceedings are essentially the result of these views. In addition, short summaries of the discussions following each session are included. Although an attempt was made to be as complete and accurate as possible in reproducing these discussions, the editors apologise if some important points or statements have been omitted. If so, the main reason is that the documentation of the discussions were based on written notes, not taped recordings. Eight longer contributions have been separately indexed

  6. Proceedings from the Fissile Material Cut-off seminar in Stockholm

    Energy Technology Data Exchange (ETDEWEB)

    Arbman, G. [ed.

    1998-07-01

    The Swedish Defence Research Establishment hosted an international expert seminar on the subject of verifying a prohibition of the production of fissile material for nuclear weapons purpose (cut-off) in Stockholm, June 3-5 1998. The objective of the seminar was to provide an opportunity for informal discussions among scientific and technical experts on various technical matters relating to the verification of a future Fissile Material Cut-off Treaty (FMCT). A stated aim of the seminar was to keep issues of scope to a minimum. Invited speakers and commentators were given an opportunity to present their views as written contributions. The present seminar proceedings are essentially the result of these views. In addition, short summaries of the discussions following each session are included. Although an attempt was made to be as complete and accurate as possible in reproducing these discussions, the editors apologise if some important points or statements have been omitted. If so, the main reason is that the documentation of the discussions were based on written notes, not taped recordings. Eight longer contributions have been separately indexed.

  7. Non-proliferation issues for the disposition of fissile materials using reactor alternatives

    International Nuclear Information System (INIS)

    Jaeger, C.D.; Duggan, R.A.; Tolk, K.M.

    1996-01-01

    The Department of Energy (DOE) is analyzing long-term storage on options for excess weapons-usable fissile materials. A number of the disposition alternatives are being considered which involve the use of reactors. The various reactor alternatives are all very similar and include front-end processes that could convert plutonium to a usable form for fuel fabrication, a MOX fuel fab facility, reactors to bum the MOX fuel and ultimate disposal of spent fuel in some geologic repository. They include existing, partially completed, advanced or evolutionary light water reactors and Canadian deuterium uranium (CANDU) reactors. In addition to the differences in the type of reactors, other variants on these alternatives are being evaluated to include the location and number of the reactors, the location of the mixed oxide (MOX) fabrication facility, the ownership of the facilities (private or government) and the colocation and/or separation of these facilities. All of these alternatives and their variants must be evaluated with respect to non-proliferation resistance. Both domestic and international safeguards support are being provided to DOE's Fissile Materials Disposition Program (FMDP) and includes such areas as physical protection, nuclear materials accountability and material containment and surveillance. This paper will focus on how the non-proliferation objective of reducing security risks and strengthening arms reduction will be accomplished and what some of the nonproliferation issues are for the reactor alternatives. Proliferation risk has been defined in terms of material form, physical environment, and the level of security and safeguards that is applied to the material. Metrics have been developed for each of these factors. The reactor alternatives will be evaluated with respect to these proliferation risk factors at each of the unit process locations in the alternative

  8. Non-proliferation issues for the disposition of fissile materials using reactor alternatives

    International Nuclear Information System (INIS)

    Jaeger, C.D.; Duggan, R.A.; Tolk, K.M.

    1996-01-01

    The Department of Energy (DOE) is analyzing long-term storage imposition options for excess weapons-usable fissile materials. A number of the disposition alternatives are being considered which involve the use of reactors. The various reactor alternatives are all very similar and include front-end processes that could convert plutonium to a usable form for fuel fabrication, a MOX fuel fab facility, reactors to burn the MOX fuel and ultimate disposal of spent fuel in some geologic repository. They include existing, partially completed, advanced or evolutionary light water reactors and Canadian deuterium uranium (CANDU) reactors. In addition to the differences in the type of reactors, other variants on these alternatives are being evaluated to include the location and number of the reactors, the location of the mixed oxide (MOX) fabrication facility, the ownership of the facilities (private or government) and the colocation and/or separation of these facilities. All of these alternatives and their variants must be evaluated with respect to non-proliferation resistance. Both domestic and international safeguards support are being provided to DOE's Fissile Materials Disposition Program (FMDP) and includes such areas as physical protection, nuclear materials accountability and material containment and surveillance. This paper will focus on how the non-proliferation objective of reducing security risks and strengthening arms reduction will be accomplished and what some of the non-proliferation issues are for the reactor alternatives. Proliferation risk has been defined in terms of material form, physical environment, and the level of security and safeguards that is applied to the material. Metrics have been developed for each of these factors. The reactor alternatives will be evaluated with respect to these proliferation risk factors at each of the unit process locations in the alternative

  9. Device for characterization of fissile materials comprising at least a neutron detector embedded inside a scintillator for gamma radiation detection

    International Nuclear Information System (INIS)

    Bernard, P.; Dherbey, J.R.; Bosser, R.; Berne, R.

    1989-01-01

    Fissile materials, for instance in radioactive wastes, are characterized by measurement of prompt and delayed neutrons and gamma radiation from induced fission by a neutron source. Gamma radiation is detected with a scintillation detector associated to a photomultiplier, the scintillation material is at the same time a moderator for thermalization of fast neutrons emitted by the neutron source and also of neutrons from spontaneous fission, (α, n) reactions and neutrons from induced fission in the fissile material. Preferentially the moderator is made of Altustipe (Plexiglas with anthracene as additive) [fr

  10. General principles of the nuclear criticality safety for handling, processing and transportation fissile materials in the USSR

    International Nuclear Information System (INIS)

    Vnukov, V.S.; Rjazanov, B.G.; Sviridov, V.I.; Frolov, V.V.; Zubkov, Y.N.

    1991-01-01

    The paper describes the general principles of nuclear criticality safety for handling, processing, transportation and fissile materials storing. Measures to limit the consequences of critical accidents are discussed for the fuel processing plants and fissile materials storage. The system of scientific and technical measures on nuclear criticality safety as well as the system of control and state supervision based on the rules, limits and requirements are described. The criticality safety aspects for various stages of handling nuclear materials are considered. The paper gives descriptions of the methods and approaches for critical risk assessments for the processing facilities, plants and storages. (Author)

  11. EXAFS and XANES analysis of plutonium and cerium edges from titanate ceramics for fissile materials disposal

    International Nuclear Information System (INIS)

    Fortner, J. A.; Kropf, A. J.; Bakel, A. J.; Hash, M. C.; Aase, S. B.; Buck, E. C.; Chamerlain, D. B.

    1999-01-01

    We report x-ray absorption near edge structure (XANES) and extended x-ray absorption fine structure (EXAFS) spectra from the plutonium L III edge and XANES from the cerium L II edge in prototype titanate ceramic hosts. The titanate ceramics studied are based upon the hafnium-pyrochlore and zirconolite mineral structures and will serve as an immobilization host for surplus fissile materials, containing as much as 10 weight % fissile plutonium and 20 weight % (natural or depleted) uranium. Three ceramic formulations were studied: one employed cerium as a ''surrogate'' element, replacing both plutonium and uranium in the ceramic matrix, another formulation contained plutonium in a ''baseline'' ceramic formulation, and a third contained plutonium in a formulation representing a high-impurity plutonium stream. The cerium XANES from the surrogate ceramic clearly indicates a mixed III-IV oxidation state for the cerium. In contrast, XANES analysis of the two plutonium-bearing ceramics shows that the plutonium is present almost entirely as Pu(IV) and occupies the calcium site in the zirconolite and pyrochlore phases. The plutonium EXAFS real-space structure shows a strong second-shell peak, clearly distinct from that of PuO 2 , with remarkably little difference in the plutonium crystal chemistry indicated between the baseline and high-impurity formulations

  12. IAEA verification of weapon-origin fissile material in the Russian Federation and the United States

    International Nuclear Information System (INIS)

    2002-01-01

    Full text: Russian Federation Minister of Atomic Energy Alexander Rumyantsev, United States Secretary of Energy Spencer Abraham and Director General of the International Atomic Energy Agency (IAEA) Mohamed ElBaradei met in Vienna on 16 September 2002 to review the status of the Trilateral Initiative and agree on its future direction. The parties concluded that the task entrusted to the Trilateral Initiative Working Group in 1996 has been fulfilled. The work completed has demonstrated practical approaches for IAEA verification of weapon-origin fissile material designated as released from defence programmes in classified forms or at certain sensitive facilities. The work included the examination of technical, legal and financial issues associated with such verification. The removal of weapon-origin fissile material from defence programmes of the Russian Federation and the United States is in furtherance of the commitment to disarmament steps undertaken by the two States pursuant to Article VI of the Treaty on the Non-Proliferation of Nuclear Weapons (NPT). IAEA verification of the materials declared excess to nuclear weapons programmes and made subject to this Initiative would build international confidence that this material will never again be used in nuclear weapons. Minister Rumyantsev, Secretary Abraham and Director General ElBaradei recognized the value of the groundbreaking work completed over the last six years. Building on the work completed, they directed the technical experts to begin without delay discussions on future possible cooperation within the trilateral format. Minister Rumyantsev, Secretary Abraham and Director General ElBaradei agreed that the Principals would meet again in September 2003 to review progress within the trilateral format. (IAEA)

  13. Calculation of multiplication factors regarding criticality aiming at the storage of fissile material

    International Nuclear Information System (INIS)

    Lima Barros, M. de.

    1982-04-01

    The multiplication factors of several systems with low enrichment, 3,5% and 3,2% in the isotope 235 U, aiming at the storage of fuel of ANGRA-I and ANGRA II, through the method of Monte Carlo, by the computacional code KENO-IV and the library of section of cross Hansen - Roach with 16 groups of energy. The method of Monte Carlo is specially suitable to the calculation of the factor of multiplication, because it is one of the most acurate models of solution and allows the description of complex tridimensional systems. Various tests of sensibility of this method have been done in order to present the most convenient way of working with KENO-IV code. The safety on criticality of stores of fissile material of the 'Fabrica de Elementos Combustiveis ', has been analyzed through the method of Monte Carlo. (Author) [pt

  14. Non-destructive assay of fissile materials by detection and multiplicity analysis of spontaneous neutrons

    International Nuclear Information System (INIS)

    Prosdocimi, A.

    1979-01-01

    A method for determining the absolute reaction rate of nuclear events giving rise to neutron emission, according to their neutron multiplicity, is proposed. A typical application is the measurement of the (α, n) and spontaneous fission rates in a fissile material sample, particularly of Pu oxide composition. An analysis of random and correlated neutron pulses is carried out on the basis of sequential order without requiring any time interval analysis, then the primary nuclear events are sorted versus their neutron multiplicity. Suitable theoretical relationships enable to derive the absolute (α, n) and SF reaction rates when the physical parameters of the neutron detector and the multiplicity spectrumm of pulses are known. A typical device is described and the results of experiments leading to Pu-239 and Pu-240 assay are given

  15. Safety analysis report: packages 238Pu oxide shipping cask (packaging of fissile and other radioactive materials). Final report

    International Nuclear Information System (INIS)

    Evans, J.E.; Gates, A.A.

    1975-06-01

    Plutonium-238 (as PuO 2 powder) is shipped in triple-container stainless steel shipping casks in compliance with ERDA Manual Chapter 0529 (ERDAM 0529), Safety Standards for the Packaging of Fissile and Other Radioactive Materials. (U.S.)

  16. Safety analysis report for packages: packaging of fissile and other radioactive materials. Final report

    International Nuclear Information System (INIS)

    Chalfant, G.G.

    1984-01-01

    The 9965, 9966, 9967, and 9968 packages are designed for surface shipment of fissile and other radioactive materials where a high degree of containment (either single or double) is required. Provisions are made to add shielding material to the packaging as required. The package was physically tested to demonstrate that it meets the criteria specified in USDOE Order No. 5480.1, chapter III, dated 5/1/81, which invokes Title 10, Code of Federal Regulations, Part 71 (10 CFR 71), Packing and Transportation of Radioactive Material, and Title 49, Code of Federal Regulations, Part 100-179, Transportation. By restricting the maximum normal operating pressure of the packages to less than 7 kg/cm 2 (gauge) (99 to 54 psig), the packages will comply with Type B(U) regulations of the International Atomic Energy Agency (IAEA) in its Regulations for the Safe Transport of Radioactive Materials, Safety Series No. 6, 1973 Revised Edition, and may be used for export and import shipments. These packages have been assessed for transport of up to 14.5 kilograms of uranium, excluding uranium-233, or 4.4 kilograms of plutonium metal, oxides, or scrap having a maximum radioactive decay energy of 30 watts. Specific maximum package contents are given. This quantity and the configuration of uranium or plutonium metal cannot be made critical by any combination of hydrogeneous reflection and moderation regardless of the condition of the package. For a uranium-233 shipment, a separate criticality evaluation for the specific package is required

  17. A treaty on the cutoff of fissile material for nuclear weapons - What to cover? How to verify?

    International Nuclear Information System (INIS)

    Schaper, A.

    1998-01-01

    Since 1946, a cutoff has been proposed. In 1993, the topic was placed on the agenda of the CD. The establishment of an Ad Hoc Committee in the CD with a mandate to negotiate a fissile material cutoff treaty struggled with difficulties for more than a year. The central dispute was whether the mandate should refer to existing un-safeguarded stockpiles. The underlying conflict of the CTBT negotiations can be summarized as nuclear disarmament versus nuclear nonproliferation The same conflict is now blocking progress with FMCT negotiations in the CD. At the center of technical proliferation concerns is direct use material that can be used for nuclear warheads without any further enrichment or reprocessing. Those materials are plutonium and highly enriched uranium (HEU). A broader category of materials is defined as all those containing any fissile isotopes, called special fissionable materials. In order ta verify that no direct use materials are abused for military purposes, also special fissionable materials must be controlled. An even broader category is simply called nuclear materials. Pu and HEU can be distinguished into the following categories of utilisation: 1. military direct use material in operational nuclear weapons and their logistics pipeline, 2. military direct use material held in reserve for military purposes, in assembled weapons or in other forms, 3. military direct use material withdrawn from dismantled weapons, 4. military direct use material considered excess and designated for transfer into civilian use, 5. military direct use material considered excess and declared for transfer into civilian use, 6. direct use material currently in reactors or their logistics pipelines and storages, and 7. irradiated Pu and HEU in spent fuel from reactors, or in vitrified form for final disposal. Large quantities of materials are neither inside weapons nor declared excess. So far, there are no legal obligations for NWS for limitations, declarations, or

  18. In-Drift Accumulation of Fissile Material From Waste Packages Containing Plutonium Disposition Waste Forms

    International Nuclear Information System (INIS)

    H.W. Stockman; S. LeStrange

    2000-01-01

    The objective of this calculation is to provide estimates of the amount of fissile material flowing out of the waste package (source term) and the accumulation of fissile elements (U and Pu) in a crushed-tuff invert. These calculations provide input for the analysis of repository impacts of the Pu-ceramic waste forms. In particular, the source term results are used as input to the far-field accumulation calculation reported in Ref. 51, and the in-drift accumulation results are used as inputs for the criticality calculations reported in Ref. 2. The results are also summarized and interpreted in Ref. 52. The scope of this calculation is the waste package (WP) Viability Assessment (VA) design, which consists of an outer corrosion-allowance material (CAM) and an inner corrosion-resistant material (CRM). This design is used in this calculation in order to be consistent with earlier Pu-ceramic degradation calculations (Ref. 15). The impact of the new Enhanced Design Alternative-I1 (EDA-11) design on the results will be addressed in a subsequent report. The design of the invert (a leveling foundation, which creates a level surface of the drift floor and supports the WP mounting structure) is consistent with the EDA-I1 design. The invert will be composed of crushed stone and a steel support structure (Ref. 17). The scope of this calculation is also defined by the nominal degradation scenario, which involves the breach of the WP (Section 10.5.1.2, Ref. 48), followed by the influx of water. Water in the WP may, in time, gradually leach the fissile components and neutron absorbers out of the ceramic waste forms. Thus, the water in the WP may become laden with dissolved actinides (e.g., Pu and U), and may eventually overflow or leak from the WP. Once the water leaves the WP, it may encounter the invert, in which the actinides may reprecipitate. Several factors could induce reprecipitation; these factors include: the high surface area of the crushed stone, and the presence of

  19. In-Drift Accumulation of Fissile Material From Waste Packages Containing Plutonium Disposition Waste Form

    Energy Technology Data Exchange (ETDEWEB)

    H.W> Stockman; S. LeStrange

    2000-09-28

    The objective of this calculation is to provide estimates of the amount of fissile material flowing out of the waste package (source term) and the accumulation of fissile elements (U and Pu) in a crushed-tuff invert. These calculations provide input for the analysis of repository impacts of the Pu-ceramic waste forms. In particular, the source term results are used as input to the far-field accumulation calculation reported in Ref. 51, and the in-drift accumulation results are used as inputs for the criticality calculations reported in Ref. 2. The results are also summarized and interpreted in Ref. 52. The scope of this calculation is the waste package (WP) Viability Assessment (VA) design, which consists of an outer corrosion-allowance material (CAM) and an inner corrosion-resistant material (CRM). This design is used in this calculation in order to be consistent with earlier Pu-ceramic degradation calculations (Ref. 15). The impact of the new Enhanced Design Alternative-I1 (EDA-11) design on the results will be addressed in a subsequent report. The design of the invert (a leveling foundation, which creates a level surface of the drift floor and supports the WP mounting structure) is consistent with the EDA-I1 design. The invert will be composed of crushed stone and a steel support structure (Ref. 17). The scope of this calculation is also defined by the nominal degradation scenario, which involves the breach of the WP (Section 10.5.1.2, Ref. 48), followed by the influx of water. Water in the WP may, in time, gradually leach the fissile components and neutron absorbers out of the ceramic waste forms. Thus, the water in the WP may become laden with dissolved actinides (e.g., Pu and U), and may eventually overflow or leak from the WP. Once the water leaves the WP, it may encounter the invert, in which the actinides may reprecipitate. Several factors could induce reprecipitation; these factors include: the high surface area of the crushed stone, and the presence of

  20. The swelling behavior of Ti-stabilized austenitic steels used as structural materials of fissile subassemblies in Phenix

    International Nuclear Information System (INIS)

    Seran, J.L.; Touron, H.; Maillard, A.; Dubuisson, P.; Hugot, J.P.; Blanchard, P.; Pelletier, M.

    1988-06-01

    In this paper we analyse the main results obained on pressurized tubes, fissile pins and hexagonal cans, allowing us to characterize the swelling and irradiation creep resistance of Ti-Mod. austenitic steels, used as reference materials for the fast breeder subassembly. After having compared the global behavior of 316Ti and 15-15Ti steels irradiated as fissile pins we examine in more detail the leading variables acting on swelling and irradiation creep resistance of CW 316Ti clads and wrappers. The irradiation creep associated to the principal mechanical stresses (sodium pressure for the wrapper, fission gas pressure for the clad) explain the plastic deformation observed on the wrappers not on the clads. Fissile pins swell more and the scatter of the results is larger than for wrappers or samples. It does not seem possible to invoque flux or primary stress differences to explain this fact. On the opposite the thermal gradient in the thickness of the components appears to be a significant parameter. In fissile pins it gives rise to a swelling gradient observed by electron microscopy that must be taken into account when comparing to the wrapper. As compared to CW 316Ti, CW 15-15Ti is an important improvement since its incubation dose for swelling is far beyond 100 dpa. Further more since it swelling temperature dependence does not seem to be as important as for 316Ti, it should be less sensitive to the effect of thermal gradients

  1. Quantification of Fissile Materials by Photon Activation Method in a Highly Shielded Enclosure

    International Nuclear Information System (INIS)

    Dighe, P.M.; Pithawa, C.K.; Goswami, A.; Dixit, K.P.; Mittal, K.C.; Sunil, C.; Sarkar, P.K.; Mukhopadhyay, P.K.; Patil, R.K.; Srivastava, G.P.; Ganesan, S.; Venugopal, V.

    2010-01-01

    For active and non-destructive quantitative identification of heavily shielded fissile materials, photo fission is one of the most often used techniques. High energy photon beams can be conveniently generated with the help of electron LINACs. 10MeV energy electron LINACs are extensively used for various industrial applications such as food irradiation, X-ray radiography, etc. The radiological safety consideration favours the use of electron beam of upto 10 MeV energy. The photonuclear data available on 10 MeV end point energy is very scarce. The present paper gives the results of our initial experiments carried out using natural uranium samples at 10 MeV LINAC facility. Water cooled tantalum target converter was used to produce intense Bremsstrahlung to induce photofission in the samples. Neutron detection system consists of six numbers of high sensitivity Helium-3 proportional counters and gamma detection system consists of two numbers of 76 mm diameter BGO scintillators. Delayed neutron and delayed gamma radiations were measured and analyzed. The mass to count rate relationship has been established for both delayed neutron and gamma radiations. Delayed gamma decay constants of natural uranium have been derived for the 10 MeV end point energy. (author)

  2. Current status and recommended future studies of underground supercriticality of fissile material

    International Nuclear Information System (INIS)

    Bowman, C.D.

    1996-06-01

    More than a year has passed since we released our original report pointing out the possibility of natural or induced rearrangement of fissile material underground into a critical mass, the possibility of positive feedback in underground configurations, the confinement of the rock to produce significant yield, and the possibility of venting or explosion. The nuclear weapons and repository storage groups at both Los Alamos and Livermore have been critical of our work while others have defended our calculations on wet and dry criticality. The conditions we identified for positive and negative feedback are no longer contested. The role of confinement of the rock in enhancing the yield from the explosion is still unsettled, and that is addressed later in this paper. The likelihood of confinement, venting, or explosive dispersion also remains unsettled and that is addressed here as well. Some critics of our work have tried to show that the probability of reconfiguration by natural processes is very small. They argue further that emplacement can be done in such a way as to make the probability even smaller. Of course these additional efforts will raise the cost of waste emplacement and the question arises as to how much is enough. The answer to this question seems to not be an easy one

  3. Analysis of triso packing fraction and fissile material to DB-MHR using LWR reprocessed fuel

    International Nuclear Information System (INIS)

    Silva, Clarysson A.M. da; Pereira, Claubia; Costa, Antonella L.; Veloso, Maria Auxiliadora F.; Gual, Maritza R.

    2013-01-01

    Gas-cooled and graphite-moderated reactor is being considered the next generation of nuclear power plants because of its characteristic to operate with reprocessed fuel. The typical fuel element consists of a hexagonal block with coolant and fuel channels. The fuel pin is manufactured into compacted ceramic-coated particles (TRISO) which are used to achieve both a high burnup and a high degree of passive safety. This work uses the MCNPX 2.6.0 to simulate the active core of Deep Burn Modular Helium Reactor (DB-MHR) employing PWR (Pressurized Water Reactor) reprocessed fuel. However, before a complete study of DB-MHR fuel cycle and recharge, it is necessary to evaluate the neutronic parameters to some values of TRISO Packing Fractions (PF) and Fissile Material (FM). Each PF and FM combination would generate the best behaviour of neutronic parameters. Therefore, this study configures several PF and FM combinations considering the heterogeneity of TRISO layers and lattice. The results present the best combination of PF and FM values according with the more appropriated behaviour of the neutronic parameters during the burnup. In this way, the optimized combination can be used to future works of MHR fuel cycle and recharge. (author)

  4. Multicounter neutron detector for examination of content and spatial distribution of fissile materials in bulk samples

    International Nuclear Information System (INIS)

    Swiderska-Kowalczyk, M.; Starosta, W.; Zoltowski, T.

    1999-01-01

    A new neutron coincidence well-counter is presented. This experimental device can be applied for passive assay of fissile and, in particular, for plutonium bearing materials. It contains of a set of the 3 He tubes placed inside a polyethylene moderator. Outputs from the tubes, first processed by preamplifier/amplifier/discriminator circuits, are then analysed using a correlator connected with PC, and correlation techniques implemented in software. Such a neutron counter enables determination of the 240 Pu effective mass in samples of a small Pu content (i.e., where the multiplication effects can be neglected) having a fairly big volume (up to 0.17 m 3 ), if only the isotopic composition is known. For determination of neutron sources distribution inside a sample, a heuristic method based on hierarchical cluster analysis was applied. As input parameters, amplitudes and phases of two-dimensional Fourier transformation of the count profiles matrices for known point sources distributions and for the examined samples were taken. Such matrices of profiles counts are collected using the sample scanning with detection head. In the clustering processes, process, counts profiles of unknown samples are fitted into dendrograms employing the 'proximity' criterion of the examined sample profile to standard samples profiles. Distribution of neutron sources in the examined sample is then evaluated on the basis of a comparison with standard sources distributions. (author)

  5. Safety analysis report: packages. Argonne National Laboratory SLSF test train shipping container, P-1 shipment. Fissile material. Final report

    International Nuclear Information System (INIS)

    Meyer, C.A.

    1975-06-01

    The package is used to ship an instrumented test fuel bundle (test train) containing fissile material. The package assembly is Argonne National Laboratory (ANL) Model R1010-0032. The shipment is fissile class III. The packaging consists of an outer carbon steel container into which an inner container is placed; the inner container is separated from the outer container by urethane foam cushioning material. The test train is supported in the inner container by a series of transverse supports spaced along the length of the test train. Both the inner and outer containers are closed with bolted covers. The covers do not seal the containers in a leaktight manner. The gross weight of the shipment is about 8350 lb. The unirradiated fissile material content is less than 3 kg of UO 2 of up to 93.2 percent enrichment. This is a Type A quantity (transport group III and less than 3 curies) of radioactive material which does not require shielding, cooling or heating, or neutron absorption or moderation functions in its packaging. The maximum exterior dimensions of the container are 37 ft 11 in. long, 24 1 / 2 in. wide, and 19 3 / 4 in. high

  6. Storage capacity for fissile material as a function of facility shape (room length-to-width ratio)

    International Nuclear Information System (INIS)

    Altschuler, S.J.

    1975-01-01

    The results of a previous study for applying surface density methods to square room of varying size are shown to be conservative for rectangular rooms as well. The surface density required to produce criticality has been calculated as a function of the facility length-to-width ratio for a variety of room widths and unit sizes, shapes, and fissile material compositions. For a length to width ratio greater than or equal to 6, the critical surface density is essentially constant. This allows further economies since more fissile material can be stored at a given subcritical value of k/ sub eff/(0.90) in a rectangular vault of given usable area than in a square one. (U.S.)

  7. Safety analysis report, packages. Drath and Schrader Double Lidded Drum (packaging of fissile and other radioactive materials). Final report

    International Nuclear Information System (INIS)

    Chalfant, G.G.

    1985-07-01

    The preceding Safety Analysis Report - Packages qualifies the Drath and Schrader Double Lidded Drum (see appendix E) as a Department of Transportation DOT 7A Type A packaging and/or ''Type A'' foreign made packaging. The allowable contents shall be: in solid form; non-fissile or exempt fissile material (as defined by 49 CFR 173.453); less than 700 pounds (318 kg) in weight; equal to or less than the A 1 or A 2 quantities of radioactive material as appropriate (see 49 CFR 173.435 for tables of A 1 /A 2 values); and hydrogen gas generation in radioactive waste shall be limited to a maximum of 2-1/2% and total gas pressure limited to 5 psig. Package marking shall be as specified in 49 CFR 178.350-3 or as specified by the foreign country of origin

  8. In field application of differential Die-Away time technique for detecting gram quantities of fissile materials

    Science.gov (United States)

    Remetti, Romolo; Gandolfo, Giada; Lepore, Luigi; Cherubini, Nadia

    2017-10-01

    In the frame of Chemical, Biological, Radiological, and Nuclear defense European activities, the ENEA, the Italian National Agency for New Technologies, Energy and Sustainable Economic Development, is proposing the Neutron Active Interrogation system (NAI), a device designed to find transuranic-based Radioactive Dispersal Devices hidden inside suspected packages. It is based on Differential Die-Away time Analysis, an active neutron technique targeted in revealing the presence of fissile material through detection of induced fission neutrons. Several Monte Carlo simulations, carried out by MCNPX code, and the development of ad-hoc design methods, have led to the realization of a first prototype based on a 14 MeV d-t neutron generator coupled with a tailored moderating structure, and an array of helium-3 neutron detectors. The complete system is characterized by easy transportability, light weight, and real-time response. First results have shown device's capability to detect gram quantities of fissile materials.

  9. High-power, photofission-inducing bremsstrahlung source for intense pulsed active detection of fissile material

    Directory of Open Access Journals (Sweden)

    J. C. Zier

    2014-06-01

    Full Text Available Intense pulsed active detection (IPAD is a promising technique for detecting fissile material to prevent the proliferation of special nuclear materials. With IPAD, fissions are induced in a brief, intense radiation burst and the resulting gamma ray or neutron signals are acquired during a short period of elevated signal-to-noise ratio. The 8 MV, 200 kA Mercury pulsed-power generator at the Naval Research Laboratory coupled to a high-power vacuum diode produces an intense 30 ns bremsstrahlung beam to study this approach. The work presented here reports on Mercury experiments designed to maximize the photofission yield in a depleted-uranium (DU object in the bremsstrahlung far field by varying the anode-cathode (AK diode gap spacing and by adding an inner-diameter-reducing insert in the outer conductor wall. An extensive suite of diagnostics was fielded to measure the bremsstrahlung beam and DU fission yield as functions of diode geometry. Delayed fission neutrons from the DU proved to be a valuable diagnostic for measuring bremsstrahlung photons above 5 MeV. The measurements are in broad agreement with particle-in-cell and Monte Carlo simulations of electron dynamics and radiation transport. These show that with increasing AK gap, electron losses to the insert and outer conductor wall increase and that the electron angles impacting the bremsstrahlung converter approach normal incidence. The diode conditions for maximum fission yield occur when the gap is large enough to produce electron angles close to normal, yet small enough to limit electron losses.

  10. High-power, photofission-inducing bremsstrahlung source for intense pulsed active detection of fissile material

    Science.gov (United States)

    Zier, J. C.; Mosher, D.; Allen, R. J.; Commisso, R. J.; Cooperstein, G.; Hinshelwood, D. D.; Jackson, S. L.; Murphy, D. P.; Ottinger, P. F.; Richardson, A. S.; Schumer, J. W.; Swanekamp, S. B.; Weber, B. V.

    2014-06-01

    Intense pulsed active detection (IPAD) is a promising technique for detecting fissile material to prevent the proliferation of special nuclear materials. With IPAD, fissions are induced in a brief, intense radiation burst and the resulting gamma ray or neutron signals are acquired during a short period of elevated signal-to-noise ratio. The 8 MV, 200 kA Mercury pulsed-power generator at the Naval Research Laboratory coupled to a high-power vacuum diode produces an intense 30 ns bremsstrahlung beam to study this approach. The work presented here reports on Mercury experiments designed to maximize the photofission yield in a depleted-uranium (DU) object in the bremsstrahlung far field by varying the anode-cathode (AK) diode gap spacing and by adding an inner-diameter-reducing insert in the outer conductor wall. An extensive suite of diagnostics was fielded to measure the bremsstrahlung beam and DU fission yield as functions of diode geometry. Delayed fission neutrons from the DU proved to be a valuable diagnostic for measuring bremsstrahlung photons above 5 MeV. The measurements are in broad agreement with particle-in-cell and Monte Carlo simulations of electron dynamics and radiation transport. These show that with increasing AK gap, electron losses to the insert and outer conductor wall increase and that the electron angles impacting the bremsstrahlung converter approach normal incidence. The diode conditions for maximum fission yield occur when the gap is large enough to produce electron angles close to normal, yet small enough to limit electron losses.

  11. Far-Field Accumulation of Fissile Material From Waste Packages Containing Plutonium Disposition Waste Form

    International Nuclear Information System (INIS)

    J.P. Nicot

    2000-01-01

    The objective of this calculation is to estimate the quantity of fissile material that could accumulate in fractures in the rock beneath plutonium-ceramic (Pu-ceramic) and Mixed-Oxide (MOX) waste packages (WPs) as they degrade in the potential monitored geologic repository at Yucca Mountain. This calculation is to feed another calculation (Ref. 31) computing the probability of criticality in the systems described in Section 6 and then ultimately to a more general report on the impact of plutonium on the performance of the proposed repository (Ref. 32), both developed concurrently to this work. This calculation is done in accordance with the development plan TDP-DDC-MD-000001 (Ref. 9), item 5. The original document described in item 5 has been split into two documents: this calculation and Ref. 4. The scope of the calculation is limited to only very low flow rates because they lead to the most conservative cases for Pu accumulation and more generally are consistent with the way the effluent from the WP (called source term in this calculation) was calculated (Ref. 4). Ref. 4 (''In-Drift Accumulation of Fissile Material from WPs Containing Plutonium Disposition Waste Forms'') details the evolution through time (breach time is initial time) of the chemical composition of the solution inside the WP as degradation of the fuel and other materials proceed. It is the chemical solution used as a source term in this calculation. Ref. 4 takes that same source term and reacts it with the invert; this calculation reacts it with the rock. In addition to reactions with the rock minerals (that release Si and Ca), the basic mechanisms for actinide precipitation are dilution and mixing with resident water as explained in Section 2.1.4. No other potential mechanism such as flow through a reducing zone is investigated in this calculation. No attempt was made to use the effluent water from the bottom of the invert instead of using directly the effluent water from the WP. This

  12. Far-Field Accumulation of Fissile Material From Waste Packages Containing Plutonium Disposition Waste Form

    Energy Technology Data Exchange (ETDEWEB)

    J.P. Nicot

    2000-09-29

    The objective of this calculation is to estimate the quantity of fissile material that could accumulate in fractures in the rock beneath plutonium-ceramic (Pu-ceramic) and Mixed-Oxide (MOX) waste packages (WPs) as they degrade in the potential monitored geologic repository at Yucca Mountain. This calculation is to feed another calculation (Ref. 31) computing the probability of criticality in the systems described in Section 6 and then ultimately to a more general report on the impact of plutonium on the performance of the proposed repository (Ref. 32), both developed concurrently to this work. This calculation is done in accordance with the development plan TDP-DDC-MD-000001 (Ref. 9), item 5. The original document described in item 5 has been split into two documents: this calculation and Ref. 4. The scope of the calculation is limited to only very low flow rates because they lead to the most conservative cases for Pu accumulation and more generally are consistent with the way the effluent from the WP (called source term in this calculation) was calculated (Ref. 4). Ref. 4 (''In-Drift Accumulation of Fissile Material from WPs Containing Plutonium Disposition Waste Forms'') details the evolution through time (breach time is initial time) of the chemical composition of the solution inside the WP as degradation of the fuel and other materials proceed. It is the chemical solution used as a source term in this calculation. Ref. 4 takes that same source term and reacts it with the invert; this calculation reacts it with the rock. In addition to reactions with the rock minerals (that release Si and Ca), the basic mechanisms for actinide precipitation are dilution and mixing with resident water as explained in Section 2.1.4. No other potential mechanism such as flow through a reducing zone is investigated in this calculation. No attempt was made to use the effluent water from the bottom of the invert instead of using directly the effluent water from the

  13. Decree of 4 November 1982 on conditions for notification of possession of special fissile materials and source materials and for keeping accounts thereof

    International Nuclear Information System (INIS)

    1982-01-01

    This Decree lays down a detailed procedure for notification of the possession and accounting of special fissile materials and source materials. The Decree was made in pursuance of Decree No. 185 of 13 February 1964 of the President of the Republic concerning radiation protection and licensing procedures. (NEA) [fr

  14. Open literature review of threats including sabotage and theft of fissile material transport in Japan

    International Nuclear Information System (INIS)

    Cochran, John Russell; Furaus, James Phillip; Marincel, Michelle K.

    2005-01-01

    This report is a review of open literature concerning threats including sabotage and theft related to fissile material transport in Japan. It is intended to aid Japanese officials in the development of a design basis threat. This threat includes the external threats of the terrorist, criminal, and extremist, and the insider threats of the disgruntled employee, the employee forced into cooperation via coercion, the psychotic employee, and the criminal employee. Examination of the external terrorist threat considers Japanese demographics, known terrorist groups in Japan, and the international relations of Japan. Demographically, Japan has a relatively homogenous population, both ethnically and religiously. Japan is a relatively peaceful nation, but its history illustrates that it is not immune to terrorism. It has a history of domestic terrorism and the open literature points to the Red Army, Aum Shinrikyo, Chukaku-Ha, and Seikijuku. Japan supports the United States in its war on terrorism and in Iraq, which may make Japan a target for both international and domestic terrorists. Crime appears to remain low in Japan; however sources note that the foreign crime rate is increasing as the number of foreign nationals in the country increases. Antinuclear groups' recent foci have been nuclear reprocessing technology, transportation of MOX fuel, and possible related nuclear proliferation issues. The insider threat is first defined by the threat of the disgruntled employee. This threat can be determined by studying the history of Japan's employment system, where Keiretsu have provided company stability and lifetime employment. Recent economic difficulties and an increase of corporate crime, due to sole reliability on the honor code, have begun to erode employee loyalty

  15. Safety analysis report: packages. Pu oxide and Am oxide shipping cask (Packaging of fissile and other radioactive materials). Final report

    International Nuclear Information System (INIS)

    Chalfant, G.G.

    1980-05-01

    The PuO 2 cask or SP 5320-2 and 3 cask is designed for surface shipment of americium or plutonium. The cask design was physically tested to demonstrate that it met the criteria specified in US ERDA Manual Chapter 0529, and Chapter I, Interstate Commerce Commission. The package has been assessed for transport of up to 357 grams of plutonium (403 grams PuO 2 powder) and up to 176 grams of americium (200 grams AmO 2 powder), having a maximum decay heat of 203 watts. Criticality evaluation alone would allow the shipment as Fissile Class II but the radiation level of the cask, measured at the time of shipment, may exceed 50 mrem/h at the surface and require shipment as Fissile Class III. Sample calculations address only the more restrictive of the two materials, which in most cases is 238 PuO 2

  16. Contribution of civilian industry to the management of military fissile materials

    International Nuclear Information System (INIS)

    Montalembert de, J.A.

    2001-01-01

    The situation about using of highly enriched uranium (HEU) and weapon grade plutonium (WgPu) for nuclear fuel preparation in U.S.A. and Russian Federation is reviewed. A few remarks were concluded: (1) We stand at the onset of a process that will be lengthy and which is unlikely to stop with the elimination of the 700 t of HEU and 2 x 34.5 t of WgPu concerned so far. If the announced negotiation of the third START treaty concludes favorably, additional tonnages will have to be recycled, particularly on the Russian side whose estimated inventory is larger. (2) The time scales necessitated by the management of these materials should be no surprise. On the one hand, the aim is to reduce an arsenal built up during 45 years of a Cold War. And this return to civilian life of materials of military origin must be achieved in conditions of safety and bilateral or international safeguards (IAEA), which obviously did not constitute the primary concern of the powers who produced them. Besides, insofar as it enlists the services of civilian industry, this return must be carried out with due respect for the equilibrium of markets that are severely mauled today, in other words, in an orderly and progressive manner. (3) Finally, it is important to recognize that without the contribution of the nuclear power industry, the elimination of military fissile materials would raise problems at another scale and would inevitably lead to regrettable waste. It is to be hoped that this will jog the minds of those who urge a rapid end to nuclear energy, when all the evidence demonstrates that the best way to eliminate surplus weapon grade materials is to recycle them in a reactor, in other words, to destroy them or to denature them while generating electricity. (4) The civilian nuclear industry is happy to contribute concretely and significantly to the solution of a problem of surplus nuclear weaponry, while at the same time utilizing technologies successfully developed for power generation

  17. Increasing transparency of nuclear-warhead and fissile-material stocks as a step toward disarmament -- Proposals for the NPT PrepCom, Geneva

    International Nuclear Information System (INIS)

    2013-04-01

    These proposals made by the International Panel on Fissile Materials IPFM at a conference in Geneva, Switzerland, in April 2013 discuss how increasing transparency can help disarmament efforts. After a short introduction to IPFM and its mission, the action plan on nuclear disarmament is looked at and the various nations involved are listed. A set of baseline declarations proposed are discussed. These include warhead stocks, potential new declarations and fissile material stocks. Monitoring by the International Atomic Energy Authority IAEA is also reviewed. Preparations for future declarations concerning warhead and delivery systems locations, stockpile histories and fissile material production and disposal aspects are reported on. Finally, co-operative verification projects, warhead dismantlement and past fissile material production are examined

  18. Fissile and fertile nuclear material measurements using a new differential die-away self-interrogation technique

    International Nuclear Information System (INIS)

    Menlove, H.O.; Menlove, S.H.; Tobin, S.J.

    2009-01-01

    This paper presents a new technique for the measurement of fissile and fertile nuclear materials in spent fuel and plutonium-laden materials such as mixed oxide (MOX) fuel. The technique, called differential die-away self-interrogation, is similar to traditional differential die-away analysis, but it does not require a pulsed neutron generator or pulsed beam accelerator, and it can measure the fertile mass in addition to the fissile mass. The new method uses the spontaneous fission neutrons from 244 Cm in spent fuel and 240 Pu effective neutrons in MOX as the 'pulsed' neutron source, with an average of ∼2.7 neutrons per pulse. The time-correlated neutrons from the spontaneous fission and the subsequent induced fissions are analyzed as a function of time to determine the spontaneous fission rate, the induced fast-neutron fissions, and the induced thermal-neutron fissions. The fissile mass is determined from the induced thermal-neutron fissions that are produced by reflected thermal neutrons that originated from the spontaneous fission reaction. The sensitivity of the fissile mass measurement is enhanced by the use of two measurements, with and without a cadmium liner between the sample and a hydrogenous moderator that surrounds the sample. The fertile mass is determined from the multiplicity analysis of the neutrons detected soon after the initial triggering neutron is detected. The method obtains good sensitivity by the optimal design of two different neutron die-away regions: a short die-away for the neutron detector region and a longer die-away for the sample interrogation region.

  19. Direct conversion of surplus fissile materials, spent nuclear fuel, and other materials to high-level-waste glass

    International Nuclear Information System (INIS)

    Forsberg, C.W.; Elam, K.R.

    1995-01-01

    With the end of the cold war the United States, Russia, and other countries have excess plutonium and other materials from the reductions in inventories of nuclear weapons. The United States Academy of Sciences (NAS) has recommended that these surplus fissile materials (SFMs) be processed so they are no more accessible than plutonium in spent nuclear fuel (SNF). This spent fuel standard, if adopted worldwide, would prevent rapid recovery of SFMs for the manufacture of nuclear weapons. The NAS recommended investigation of three sets of options for disposition of SFMs while meeting the spent fuel standard: (1) incorporate SFMs with highly radioactive materials and dispose of as waste, (2) partly burn the SFMs in reactors with conversion of the SFMs to SNF for disposal, and (3) dispose of the SFMs in deep boreholes. The US Government is investigating these options for SFM disposition. A new method for the disposition of SFMs is described herein: the simultaneous conversion of SFMs, SNF, and other highly radioactive materials into high-level-waste (HLW) glass. The SFMs include plutonium, neptinium, americium, and 233 U. The primary SFM is plutonium. The preferred SNF is degraded SNF, which may require processing before it can be accepted by a geological repository for disposal

  20. Plutonium-bearing materials feed report for the DOE Fissile Materials Disposition Program alternatives

    International Nuclear Information System (INIS)

    Brough, W.G.; Boerigter, S.T.

    1995-01-01

    This report has identified all plutonium currently excess to DOE Defense Programs under current planning assumptions. A number of material categories win clearly fan within the scope of the MD (Materials Disposition) program, but the fate of the other categories are unknown at the present time. MD planning requires that estimates be made of those materials likely to be considered for disposition actions so that bounding cases for the PEIS (Programmatic Environmental Impact Statement) can be determined and so that processing which may be required can be identified in considering the various alternatives. A systematic analysis of the various alternatives in reachmg the preferred alternative requires an understanding of the possible range of values which may be taken by the various categories of feed materials. One table identifies the current total inventories excess to Defense Program planning needs and represents the bounding total of Pu which may become part of the MD disposition effort for all materials, except site return weapons. The other categories, principally irradiated fuel, rich scrap, and lean scrap, are discussed. Another table summarizes the ranges and expected quantities of Pu which could become the responsibility of the MD program. These values are to be used for assessing the impact of the various alternatives and for scaling operations to assess PEIS impact. Determination of the actual materials to be included in the disposition program will be done later

  1. The Molten Salt Reactor option for beneficial use of fissile material from dismantled weapons

    International Nuclear Information System (INIS)

    Gat, U.; Engel, J.R.; Dodds, H.L.

    1991-01-01

    The Molten Salt Reactor (MSR) option for burning fissile fuel from dismantled weapons is examined. It is concluded that MSRs are very suitable for beneficial utilization of the dismantled fuel. The MSRs can utilize any fissile fuel in continuous operation with no special modifications, as demonstrated in the Molten Salt Reactor Experiment. Thus MSRs are flexible while maintaining their economy. MSRs further require a minimum of special fuel preparation and can tolerate denaturing and dilution of the fuel. Fuel shipments can be arbitrarily small, all of which supports nonproliferation and averts diversion. MSRs have inherent safety features which make them acceptable and attractive. They can burn a fuel type completely and convert it to other fuels. MSRs also have the potential for burning the actinides and delivering the waste in an optimal form, thus contributing to the solution of one of the major remaining problems for deployment of nuclear power. 19 refs

  2. Operational experience in the non-destructive assay of fissile material in General Electric's nuclear fuel fabrication facility

    International Nuclear Information System (INIS)

    Stewart, J.P.

    1976-01-01

    Operational experience in the non-destructive assay of fissile material in a variety of forms and containers and incorporation of the assay devices into the accountability measurement system for General Electric's Wilmington Fuel Fabrication Facility measurement control programme is detailed. Description of the purpose and related operational requirements of each non-destructive assay system is also included. In addition, the accountability data acquisition and processing system is described in relation to its interaction with the various non-destructive assay devices and scales used for accountability purposes within the facility. (author)

  3. Nonproliferation and arms control assessment of weapons-usable fissile material storage and excess plutonium disposition alternatives

    International Nuclear Information System (INIS)

    1997-01-01

    This report has been prepared by the Department of Energy's Office of Arms Control and Nonproliferation (DOE-NN) with support from the Office of Fissile Materials Disposition (DOE-MD). Its purpose is to analyze the nonproliferation and arms reduction implications of the alternatives for storage of plutonium and HEU, and disposition of excess plutonium, to aid policymakers and the public in making final decisions. While this assessment describes the benefits and risks associated with each option, it does not attempt to rank order the options or choose which ones are best. It does, however, identify steps which could maximize the benefits and mitigate any vulnerabilities of the various alternatives under consideration

  4. Electric breeding of fissile materials with low Q, non-mainline fusion drivers

    International Nuclear Information System (INIS)

    Benford, J.; Bailey, V.; Oliver, D.; DiCapua, M.; Cooper, R.; Lopez, O.; Lindsey, H.

    1977-10-01

    The application of two novel fusion reactor concepts to the production of fissile fuel for existing and planned fission reactors has been shown to be technically feasible and potentially economically competitive. The performance required of fusion based breeders has been derived in terms of the fusion gain, blanket neutron and energy multiplication, and the performance and economic parameters of the fission reactors. Electron beam heated, linear solenoid confined plasmas were one concept which showed the most promise. A shock heated, wall confined reactor also appeared attractive for breeding

  5. Problems in future negotiations for a treaty on the cut-off of fissile material for nuclear weapons

    International Nuclear Information System (INIS)

    Schaper, A.

    1999-01-01

    A treaty to end the production of fissile material for nuclear weapons, the so-called cutoff, is one of the most important next steps on the disarmament agenda.' But meanwhile, the Conference on Disarmament (CD) is deadlocked, and confidence in negotiations taking place in the near future is replaced by bewilderment at the inaction. The underlying conflict of the Comprehensive Test Ban Treaty (CTBT) negotiations can be summarized as nuclear disarmament versus nuclear nonproliferation. The same conflict is now blocking progress with negotiations in the CD on the Fissile Material Cut-off Treaty (FMCT). Nevertheless, the cut-off would be the major policy driver to insert transparency and irreversibility into the disarmament process,' and we need to harness all our efforts to overcome the current difficulties. The CTBT can be regarded as a tool to cap the qualitative nuclear arms race, for example to hinder the future development of qualitatively new nuclear explosives, and an FMCT can be seen as its quantitative counterpart, capping the amount of material available for new nuclear weapons. The complex questions involve political, technical, legal, and economic aspects and constitute a challenge for diplomats and decision makers

  6. Partitioning of fissile and radio-toxic materials from spent nuclear fuel

    International Nuclear Information System (INIS)

    Bychkov, A.V.; Skiba, O.V.; Kormilitsyn, M.V.

    2007-01-01

    these elements as fuel components, they could be involved in the recycling together with the main actinides, and they could be jointly extracted in the partitioning processes. It is also possible to design some special reactor systems for energy generation. For instance, Np, Am and Cm could be considered as fuel components for fast reactors. It would be possible to apply similar approaches even to the burning of uranium isotopes ( 232,234,236 U), which should be produced in a concentrated form during the re-enrichment. So the future development of innovative technologies should be directed from a complete reprocessing towards partitioning of fissile and radio-toxic materials from the spent nuclear fuel. The objectives of technology optimisation can be stated as follows: (1) reprocessing/partitioning with the view of non-proliferation, (2) partitioning with a minimal effect on the environment (3) partitioning using advanced economical methods. The criteria for the partitioning in future (after the year 2050) can be taken from the INPRO methodology. (authors)

  7. Applications of Monte Carlo technique in the detection of explosives, narcotics and fissile material using neutron sources

    International Nuclear Information System (INIS)

    Sinha, Amar; Kashyap, Yogesh; Roy, Tushar; Agrawal, Ashish; Sarkar, P.S.; Shukla, Mayank

    2009-01-01

    The problem of illicit trafficking of explosives, narcotics or fissile materials represents a real challenge to civil security. Neutron based detection systems are being actively explored worldwide as a confirmatory tool for applications in the detection of explosives either hidden inside a vehicle or a cargo container or buried inside soil. The development of a system and its experimental testing is a tedious process and to develop such a system each experimental condition needs to be theoretically simulated. Monte Carlo based methods are used to find an optimized design for such detection system. In order to design such systems, it is necessary to optimize source and detector system for each specific application. The present paper deals with such optimization studies using Monte Carlo technique for tagged neutron based system for explosives and narcotics detection hidden in a cargo and landmine detection using backscatter neutrons. We will also discuss some simulation studies on detection of fissile material and photo-neutron source design for applications on cargo scanning. (author)

  8. Summary report of the screening process to determine reasonable alternatives for long-term storage and disposition of weapons-usable fissile materials

    International Nuclear Information System (INIS)

    1995-01-01

    Significant quantities of weapons-usable fissile materials (primarily plutonium and highly enriched uranium) have become surplus to national defense needs both in the US and Russia. These stocks of fissile materials pose significant dangers to national and international security. The dangers exist not only in the potential proliferation of nuclear weapons but also in the potential for environmental, safety and health consequences if surplus fissile materials are not properly managed. As announced in the Notice of Intent (NOI) to prepare a Programmatic Environmental Impact Statement (PEIS), the Department of Energy is currently conducting an evaluation process for disposition of surplus weapons-usable fissile materials determined surplus to National Security needs, and long-term storage of national security and programmatic inventories, and surplus weapons-usable fissile materials that are not able to go directly from interim storage to disposition. An extensive set of long-term storage and disposition options was compiled. Five broad long-term storage options were identified; thirty-seven options were considered for plutonium disposition; nine options were considered for HEU disposition; and eight options were identified for Uranium-233 disposition. Section 2 discusses the criteria used in the screening process. Section 3 describes the options considered, and Section 4 provides a detailed summary discussions of the screening results

  9. Summary report of the screening process to determine reasonable alternatives for long-term storage and disposition of weapons-usable fissile materials

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-03-29

    Significant quantities of weapons-usable fissile materials (primarily plutonium and highly enriched uranium) have become surplus to national defense needs both in the US and Russia. These stocks of fissile materials pose significant dangers to national and international security. The dangers exist not only in the potential proliferation of nuclear weapons but also in the potential for environmental, safety and health consequences if surplus fissile materials are not properly managed. As announced in the Notice of Intent (NOI) to prepare a Programmatic Environmental Impact Statement (PEIS), the Department of Energy is currently conducting an evaluation process for disposition of surplus weapons-usable fissile materials determined surplus to National Security needs, and long-term storage of national security and programmatic inventories, and surplus weapons-usable fissile materials that are not able to go directly from interim storage to disposition. An extensive set of long-term storage and disposition options was compiled. Five broad long-term storage options were identified; thirty-seven options were considered for plutonium disposition; nine options were considered for HEU disposition; and eight options were identified for Uranium-233 disposition. Section 2 discusses the criteria used in the screening process. Section 3 describes the options considered, and Section 4 provides a detailed summary discussions of the screening results.

  10. Royal Order of 30 March 1981 determining the duties and conditions of operation of the public body responsible for radioactive waste and fissile materials management

    International Nuclear Information System (INIS)

    1981-01-01

    The purpose of this Royal Order is to set up a public body to be responsible for management of the storage of conditioned radioactive waste, waste disposal, its transport as well as that of plutonium-bearing or enriched fissile materials, and plutonium storage. It must become operational as soon as possible, in particular in the perspective of the Eurochemic Company's technical operations ceasing as from 31 December 1981. This body will be named the National Body for Radioactive Waste and Fissile Materials (ONDRAF). As respects plutonium-bearing or enriched fissile materials, ONDRAF will deal with the transport of materials which, in accordance with the IAEA recommendations [INFCIRC/225/Rev. 1], require physical protection measures (NEA) [fr

  11. Analyse of the potential of the high temperature reactor with respect to the use of fissile materials

    International Nuclear Information System (INIS)

    Damian, F.

    2001-01-01

    The high temperature reactors fuel is made of micro-particles dispersed in a graphite matrix. This configuration makes it possible to reach high burnup, higher than 700 GWj/t. Thanks to the decoupling between the thermal and the neutronic behaviors in the core many types of fuels can be used. These characteristics give to HTR reactor very good capacities to burn fissile materials. This work was done in the frame of the evaluation of HTR capacities to enhance the value of the plutonium stocks. These stocks are currently composed of the irradiated fuels discharged from classical PWR or the dismantling of the nuclear weapons and represent a significant energy potential. These studies concluded that high cycles length can be reached whatever the plutonium quality is (from 50 % to 94 % of fissile plutonium). In addition, it was demonstrated that the moderator temperature coefficient becomes locally positive for highly burn fuel while the core global moderator temperature coefficient remained negative in the operation range of the reactor. A significant share of this work was first devoted to the setting of a modeling of the fuel element but also of the reactor's core with the codes of system SAPHYR. The whole of modeling was validated by reference calculations. This work of code assessment is justified by a preliminary work that showed that the classical calculation scheme used for PWR could not be transposed directly to HTR core. (author)

  12. Fate Of Fissile Material Bound To Monosodium Titanate During Cooper Catalyzed Peroxide Oxidation Of Tank 48H Waste

    International Nuclear Information System (INIS)

    Taylor-Pashow, K.

    2012-01-01

    At the Savannah River Site (SRS), Tank 48H currently holds approximately 240,000 gallons of slurry which contains potassium and cesium tetraphenylborate (TPB). A copper catalyzed peroxide oxidation (CCPO) reaction is currently being examined as a method for destroying the TPB present in Tank 48H. Part of the development of that process includes an examination of the fate of the Tank 48H fissile material which is adsorbed onto monosodium titanate (MST) particles. This report details results from experiments designed to examine the potential degradation of MST during CCPO processing and the subsequent fate of the adsorbed fissile material. Experiments were conducted to simulate the CCPO process on MST solids loaded with sorbates in a simplified Tank 48H simulant. Loaded MST solids were placed into the Tank 48H simplified simulant without TPB, and the experiments were then carried through acid addition (pH adjustment to 11), peroxide addition, holding at temperature (50 C) for one week, and finally NaOH addition to bring the free hydroxide concentration to a target concentration of 1 M. Testing was conducted without TPB to show the maximum possible impact on MST since the competing oxidation of TPB with peroxide was absent. In addition, the Cu catalyst was also omitted, which will maximize the interaction of H 2 O 2 with the MST; however, the results may be non-conservative assuming the Cu-peroxide active intermediate is more reactive than the peroxide radical itself. The study found that both U and Pu desorb from the MST when the peroxide addition begins, although to different extents. Virtually all of the U goes into solution at the beginning of the peroxide addition, whereas Pu reaches a maximum of ∼34% leached during the peroxide addition. Ti from the MST was also found to come into solution during the peroxide addition. Therefore, Ti is present with the fissile in solution. After the peroxide addition is complete, the Pu and Ti are found to precipitate from

  13. Nonproliferation and arms control assessment of weapons-usable fissile material storage and excess plutonium disposition alternatives

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-01-01

    This report has been prepared by the Department of Energy`s Office of Arms Control and Nonproliferation (DOE-NN) with support from the Office of Fissile Materials Disposition (DOE-MD). Its purpose is to analyze the nonproliferation and arms reduction implications of the alternatives for storage of plutonium and HEU, and disposition of excess plutonium, to aid policymakers and the public in making final decisions. While this assessment describes the benefits and risks associated with each option, it does not attempt to rank order the options or choose which ones are best. It does, however, identify steps which could maximize the benefits and mitigate any vulnerabilities of the various alternatives under consideration.

  14. Trilateral Initiative: IAEA authentication and national certification of verification equipment for facilities with classified forms of fissile material

    International Nuclear Information System (INIS)

    Haas, Eckard; Sukhanov, Alexander; Murphy, John

    2001-01-01

    Full text: Within the framework of the Trilateral Initiative, technical challenges have arisen due to the potential of the International Atomic Energy Agency (IAEA) monitoring fissile material with classified characteristics, as well as the IAEA using facility or host country supplied monitoring equipment. In monitoring material with classified characteristics, it is recognized that the host country needs to assure that classified information is not made available to the IAEA inspectors. Thus, any monitoring equipment used to monitor material with classified characteristics has to contain information security capabilities, such as information barriers. But likewise in using host-country-supplied monitoring equipment, regarding the material being monitored the IAEA has to have confidence that the information provided by the equipment is genuine and can be used by the IAEA in fulfilling its obligation to derive conclusions based on independent verification measures. Thus the IAEA needs to go through the process of authenticating the monitoring equipment. In the same way the host country needs to go through the process to assure itself that the monitoring equipment integrated with an information barrier will not divulge any classified information about an inspected sensitive item. Both processes require on large extent identical measures, but partially also may conflict with each other. The fact that monitoring equipment needs to exhibit information security throughout its lifecycle while at the same time be capable of being authenticated necessitates the need for creative technical approaches to be pursued. (author)

  15. Material correlations and models for the irradiation behavior of fissile and fertile material in SNR-300, Mark-II and KNK II, third core

    International Nuclear Information System (INIS)

    Fenneker; Steinmetz; Toebbe

    1986-07-01

    The report contains the material correlations and models used in the fuel pin design code IAMBUS for the irradiation behavior of PuO 2 -UO 2 fissile materials and UO 2 fertile materials of the SNR-300 Mark-II reload and the KNK II third core. They are applicable for pellet densities of more than 90 % of the theoretical density. The presented models of the fuel behavior and the applied material correlations have been derived either from single experiments or from the comparison of theoretically predicted integral fuel behavior with the results of fuel pin irradiation experiments. The material correlations have been examined and extended in the frame of the collaborations INTERATOM/KWU and INTERATOM/KfK. French and British results were included, when available from the European fast reactor knowledge exchange [de

  16. Review of the bases for regulations governing the transport of fissile and other radioactive material

    International Nuclear Information System (INIS)

    Smith, D.R.; Thomas, J.T.

    1978-01-01

    The outstanding record of transport of radioactive materials prompted this brief review of the history of the regulations. IAEA as well as DOT regulations are discussed, as are all classes of shipments and materials (Class I, II, III)

  17. Analyse of the potential of the high temperature reactor with respect to the use of fissile materials; Analyse des capacites des reacteurs a haute temperature sous l'aspect de l'utilisation des matieres fissiles

    Energy Technology Data Exchange (ETDEWEB)

    Damian, F

    2001-07-01

    The high temperature reactors fuel is made of micro-particles dispersed in a graphite matrix. This configuration makes it possible to reach high burnup, higher than 700 GWj/t. Thanks to the decoupling between the thermal and the neutronic behaviors in the core many types of fuels can be used. These characteristics give to HTR reactor very good capacities to burn fissile materials. This work was done in the frame of the evaluation of HTR capacities to enhance the value of the plutonium stocks. These stocks are currently composed of the irradiated fuels discharged from classical PWR or the dismantling of the nuclear weapons and represent a significant energy potential. These studies concluded that high cycles length can be reached whatever the plutonium quality is (from 50 % to 94 % of fissile plutonium). In addition, it was demonstrated that the moderator temperature coefficient becomes locally positive for highly burn fuel while the core global moderator temperature coefficient remained negative in the operation range of the reactor. A significant share of this work was first devoted to the setting of a modeling of the fuel element but also of the reactor's core with the codes of system SAPHYR. The whole of modeling was validated by reference calculations. This work of code assessment is justified by a preliminary work that showed that the classical calculation scheme used for PWR could not be transposed directly to HTR core. (author)

  18. Analyse of the potential of the high temperature reactor with respect to the use of fissile materials; Analyse des capacites des reacteurs a haute temperature sous l'aspect de l'utilisation des matieres fissiles

    Energy Technology Data Exchange (ETDEWEB)

    Damian, F

    2001-07-01

    The high temperature reactors fuel is made of micro-particles dispersed in a graphite matrix. This configuration makes it possible to reach high burnup, higher than 700 GWj/t. Thanks to the decoupling between the thermal and the neutronic behaviors in the core many types of fuels can be used. These characteristics give to HTR reactor very good capacities to burn fissile materials. This work was done in the frame of the evaluation of HTR capacities to enhance the value of the plutonium stocks. These stocks are currently composed of the irradiated fuels discharged from classical PWR or the dismantling of the nuclear weapons and represent a significant energy potential. These studies concluded that high cycles length can be reached whatever the plutonium quality is (from 50 % to 94 % of fissile plutonium). In addition, it was demonstrated that the moderator temperature coefficient becomes locally positive for highly burn fuel while the core global moderator temperature coefficient remained negative in the operation range of the reactor. A significant share of this work was first devoted to the setting of a modeling of the fuel element but also of the reactor's core with the codes of system SAPHYR. The whole of modeling was validated by reference calculations. This work of code assessment is justified by a preliminary work that showed that the classical calculation scheme used for PWR could not be transposed directly to HTR core. (author)

  19. Device for the determination of concentrations of fissile and/or fertile materials by means of x-ray fluorescence spectrometry

    International Nuclear Information System (INIS)

    Von Baeckmann, A.; Neuber, J.

    1975-01-01

    In analyzing fissile and/or fertile materials in the thorium, uranium, neptunium, plutonium, americium and curium group, time and accuracy are significant factors. An automated system for rapidly analyzing these materials includes: sample preparation device in which aliquots of sample are weighed and mixed with known amounts of solution; x-ray fluorescence spectrometer; and, a central control system for controlling the operation and analyzing the data. (auth)

  20. Destructive and non-destructive methods of measuring the quantity and isotopic composition of fissile materials for purposes of national safeguards in the German Democratic Republic

    International Nuclear Information System (INIS)

    Villun, K.; Gruner, V.; Siebert, Kh.U.; Hoffmann, D.

    1979-01-01

    The authors give a brief description of the destructive and non-destructive methods of measuring the quantity and isotopic composition of fissile materials used in the nuclear materials accounting and control system of the German Democratic Republic. They cite examples of the use of gamma-spectrometry, X-ray fluorescence analysis, neutron activation, radiochemical techniques, mass-spectrometry and alpha-spectrometry. (author)

  1. To the question of definition of fissile material mass and neutron multiplication in deep sub-critical systems

    International Nuclear Information System (INIS)

    Dulin, V.V.

    2006-01-01

    A method of determination neutrons multiplication in deep sub-critical multiplying media has been developed. It is based on a modified of Rossi - alpha method. It will consist in use of integral on time (a method of the areas) from correlated parts of distribution and integral in area, independent of time a part of distribution (area of a constant background). It allows to spend the calculated analysis, using the integrated equation on time for a neutrons flux and to not use representation of point kinetic model. A calculation spatially-correlation factor the adjoint (relative the detector count rate) inhomogeneous equation is used. Its calculation takes into account fission both in multiplying media and in a spontaneous neutron source. Measurements with plutonium-steel and uranium-steel blocks, and blocks from uranium and plutonium dioxide of different enrichment are have been carried out. The measured values of neutrons multiplication in a range 1.03-1.82 will be well coordinated to results of calculations. The question on an opportunity of definition of weight of the measured blocks of fissile material is considered [ru

  2. The determination by irradiation with a pulsed neutron generator and delayed neutron counting of the amount of fissile material present in a sample; Determination de la quantite de matiere fissile presente dans un echantillon par irradiation au moyen d'une source pulsee de neutrons et comptage des neutrons retardes

    Energy Technology Data Exchange (ETDEWEB)

    Beliard, L; Janot, P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1967-07-01

    A preliminary study was conducted to determine the amount of fissile material present in a sample. The method used consisted in irradiating the sample by means of a pulsed neutron generator and delayed neutron counting. Results show the validity of this method provided some experimental precautions are taken. Checking on the residual proportion of fissile material in leached hulls seems possible. (authors) [French] Ce rapport rend compte d'une etude preliminaire effectuee en vue de determiner la quantite de matiere fissile presente dans un echantillon. La methode utilisee consiste a irradier l'echantillon considere au moyen d'une source puisee de neutrons et a compter les neutrons retardes produits. Les resultats obtenus permettent de conclure a la validite de la methode moyennant certaines precautions. Un controle de la teneur residuelle en matiere fissile des gaines apres traitement semble possible. (auteurs)

  3. Safety analysis report: packages cobalt-60 shipping cask (packaging of radioactive and fissile materials)

    International Nuclear Information System (INIS)

    Evans, J.E.; Langhaar, J.W.

    1973-07-01

    Safety Analysis Report DPSPU-73-124-1 replaces DPSPU-69-124-1 and Supplement 1 to permit shipment of 350,000 curies of 60 Co (maximum) in cobalt-60 shipping casks in compliance with 10 CFR Part 71, Packaging of Radioactive Materials for Transport

  4. IAEA verification of weapon-origin fissile material in the Russian Federation and the United States

    International Nuclear Information System (INIS)

    2000-01-01

    The document informs about the meeting of the Minister of the Russian Federation on Atomic Energy, the Administrator of the National Nuclear Security Administration of the United States, and the Director General of the IAEA, on 18 September 2000 in Vienna, to review progress on the Trilateral Initiative which was launched in 1996 to develop a new IAEA verification system for weapon-origin material designated as released from defense programs by the United States or the Russian Federation

  5. Determining fissile content of nuclear fuel elements

    International Nuclear Information System (INIS)

    Arya, S.P.; Grossman, L.N.; Schoenig, F.C.

    1980-01-01

    This invention relates to the determination of the fissile fuel content of fuel for nuclear reactors. A nondestructive method is described for determining rapidly, accurately and simultaneously the fissile content, enrichment and location of fuel material which may also contain amounts of burnable poison, by detecting the γ-rays emitted from the fuel material due to natural radioactive decay. (U.K.)

  6. User's guide for shipping Type B quantities of radioactive and fissile material, including plutonium, in DOT-6M specification packaging configurations

    International Nuclear Information System (INIS)

    Kelly, D.L.

    1994-09-01

    The need for developing a user's guide for shipping Type B quantities of radioactive and fissile material, including plutonium, in a US Department of Transportation Specification 6M (DOT-6M) packaging was identified by the US Department of Energy (DOE)-Headquarters, Transportation Management Division (EM-261) because the DOT-6M packaging is widely used by DOE site contractors and the DOE receives many questions about approved packaging configuration. Currently, EM-261 has the authority to approve new DOT-6M packaging configurations for use by the DOE Operations Offices. This user's guide identifies the DOE-approved DOT-6M packaging configurations and explains how to have new configurations approved by the DOE. The packaging configurations described in this guide are approved by the DOE, and satisfy the applicable DOT requirements and the identified DOE restrictions. These packaging configurations are acceptable for transport of Type B quantities of radioactive and fissile material, including plutonium

  7. Criticality safety analysis of the fissile material storage arrays in the east end of building 6592

    International Nuclear Information System (INIS)

    McKeon, D.C.; Philbin, J.S.

    1981-03-01

    A criticality safety analysis of nine concrete storage holes that have been formed in the floor of the Materials Balance Area (MBA) in Building 6592 is reported. Unit cell dimensions and unit mass limits are defined for the most likely plutonium and uranium fuel types that will be stored there. Two tables of mass limits are derived. The first table is to be used for short units that can be stacked with fixed separation in the same hole. The second table will permit units greater than one foot in length providing that the appropriate linear mass density limit (in kg/ft) is not exceeded

  8. International report to validate criticality safety calculations for fissile material transport

    International Nuclear Information System (INIS)

    Whitesides, G.E.

    1984-01-01

    During the past three years a Working Group established by the Organization for Economic Co-operation and Development's Nuclear Energy Agency (OECD-NEA) in Paris, France, has been studying the validity and applicability of a variety of criticality safety computer programs and their associated nuclear data for the computation of the neutron multiplication factor, k/sub eff/, for various transport packages used in the fuel cycle. The principal objective of this work has been to provide an internationally acceptable basis for the licensing authorities in a country to honor licensing approvals granted by other participating countries. Eleven countries participated in the initial study which consisted of examining criticality safety calculations for packages designed for spent light water reactor fuel transport. This paper presents a summary of this study which has been completed and reported in an OECD-NEA Report No. CSNI-71. The basic goal of this study was to outline a satisfactory validation procedure for this particular application. First, a set of actual critical experiments were chosen which contained the various material and geometric properties present in typical LWR transport containers. Secondly, calculations were made by each of the methods in order to determine how accurately each method reproduced the experimental values. This successful effort in developing a benchmark procedure for validating criticality calculations for spent LWR transport packages along with the successful intercomparison of a number of methods should provide increased confidence by licensing authorities in the use of these methods for this area of application. 4 references, 2 figures

  9. Method of storing fissile mateiral

    International Nuclear Information System (INIS)

    Onoshita, Toshio; Ishitobi, Masuhiro

    1989-01-01

    Upon storing nuclear fissile materials in a storing building, vessels packed with fissile materials are inserted into a containing chamber divided with partition walls comprising neutron absorbers and neutron moderators. Thus, released neutrons permeating the vessel are moderated by the neutron moderators and then absorbed by the neutron absorbers. Accordingly, the neutron absorbing effect by the neutron absorbers is improved, and irradiation of neutrons released from one of vessels to the other of vessels can be suppressed. Accordingly, it is possible to shorten the distance between the vessels in a contained state as much as possible, while securing the critical safety, to improve the containing density during storage. (T.M.)

  10. Thermal energy of nuclear origin produced in non-fissile materials (1962); Energie calorifique d'origine nucleaire degagee dans les materiaux non fissiles (1962)

    Energy Technology Data Exchange (ETDEWEB)

    Naudet, G [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Millies, P; Berger, J [Commissariat a l' Energie Atomique, Grenoble (France). Centre d' Etudes Nucleaires

    1962-07-01

    A first part is devoted to the description of the interaction phenomena between elementary particles and material that may be observed during the irradiation process in a nuclear reactor: nuclear reactions due to neutrons, production of gamma rays and absorption of those gamma rays through various processes. In a second part the phenomena producing calorific energy in irradiated material are quantitatively examined. In the third part results are summed up in a formulary. The fourth part presents tables and figures giving to the reader all the numerical values necessary for practical calculations. (authors) [French] Une premiere partie est consacree a l'examen des principaux phenomenes d'interaction des particules avec la matiere qui interviennent lors d'une irradiation dans un reacteur: reactions nucleaires dues aux neutrons, production des rayons gamma et absorption de ces derniers par les divers processus. Une deuxieme partie etudie quantitativement les phenomenes qui conduisent a l'apparition d'energie calorifique dans le materiau irradie. En troisieme partie, un formulaire resume les resultats etablis. Dans une quatrieme partie, des tableaux et des courbes fournissent a l'experimentateur toutes les valeurs numeriques necessaires aux calculs pratiques. (auteurs)

  11. Using the sampling method to propagate uncertainties of physical parameters in systems with fissile material

    International Nuclear Information System (INIS)

    Campolina, Daniel de Almeida Magalhães

    2015-01-01

    There is an uncertainty for all the components that comprise the model of a nuclear system. Assessing the impact of uncertainties in the simulation of fissionable material systems is essential for a realistic calculation that has been replacing conservative model calculations as the computational power increases. The propagation of uncertainty in a simulation using a Monte Carlo code by sampling the input parameters is recent because of the huge computational effort required. By analyzing the propagated uncertainty to the effective neutron multiplication factor (k eff ), the effects of the sample size, computational uncertainty and efficiency of a random number generator to represent the distributions that characterize physical uncertainty in a light water reactor was investigated. A program entitled GB s ample was implemented to enable the application of the random sampling method, which requires an automated process and robust statistical tools. The program was based on the black box model and the MCNPX code was used in and parallel processing for the calculation of particle transport. The uncertainties considered were taken from a benchmark experiment in which the effects in k eff due to physical uncertainties is done through a conservative method. In this work a script called GB s ample was implemented to automate the sampling based method, use multiprocessing and assure the necessary robustness. It has been found the possibility of improving the efficiency of the random sampling method by selecting distributions obtained from a random number generator in order to obtain a better representation of uncertainty figures. After the convergence of the method is achieved, in order to reduce the variance of the uncertainty propagated without increase in computational time, it was found the best number o components to be sampled. It was also observed that if the sampling method is used to calculate the effect on k eff due to physical uncertainties reported by

  12. Fissile fingerprints

    International Nuclear Information System (INIS)

    Edwards, R.

    1995-01-01

    This article looks at recent research which may allow police and customs officers to detect smuggled weapons-grade plutonium and uranium. Contrary to popular opinion, nuclear materials do not have a nuclear ''fingerprint'' but enough information can be gleaned from sources to confirm what has been learnt from other data. Indeed, two leading nuclear laboratories can look at the same analytical results and draw different conclusions. The case of a lead cylinder seized from a German garage is examined to illustrate the confusion. (UK)

  13. 49 CFR 173.420 - Uranium hexafluoride (fissile, fissile excepted and non-fissile).

    Science.gov (United States)

    2010-10-01

    ... 49 Transportation 2 2010-10-01 2010-10-01 false Uranium hexafluoride (fissile, fissile excepted....420 Uranium hexafluoride (fissile, fissile excepted and non-fissile). (a) In addition to any other... non-fissile uranium hexafluoride must be offered for transportation as follows: (1) Before initial...

  14. LSDS Development for Isotopic Fissile Assay in Spent Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Yong Deok; Park, Chang Je; Park, Geun Il; Lee, Jung Won; Song, Kee Chan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2011-07-01

    As an option to reduce a spent fuel and reuse an existing fissile material in spent fuel, sodium fast reactor SFR program linked with pyro-processing is under development in KAERI. A uranium-TRU mixture through a pyro-process is used to fabricate SFR fuel. An assay of isotopic fissile content plays an important role in an optimum design of storage site and reuse of fissile materials of spent fuel. Lead slowing down spectrometer LSDS is being developed in KAERI to analyze isotopic fissile material content. LSDS has several features: direct fissile assay, near real time fissile assay, no influence from radiation background, fissile isotopic assay and applicable to spent fuel and recycled fuel. Based on the designed geometry, neutron energy resolution was investigated. The neutron energy spectrum was analyzed as well. Spent fuel emits large number of neutrons by spontaneous fission. Neutron generator must overcome the neutron background to get the pure fission signals from fissile materials. Neutron generator is planned to have compact system with one section electron linac which is easy maintenance, less cost and high neutron yield. The LSD has the power to resolve the fission characteristics from each fissile material. This feature can analyze the content of isotopic fissile. From 1keV to 0.1eV energy range, the energy resolution is enough to get the individual fissile fission signatures. The dominant fission signature is shown below 1eV for each fissile isotope. The neutron generation system with target was designed to get fission signals by fissile materials. The system was decided to overcome neutron backgrounds and to get good counting statistics. Finally, an accurate fissile material content will contribute to safety of spent fuel reuse in future nuclear energy system and optimum design of spent fuel storage site. Additionally, an accurate fissile material content will increase international transparence and credibility for the reuse of PWR spent fuel.

  15. LSDS Development for Isotopic Fissile Assay in Spent Fuel

    International Nuclear Information System (INIS)

    Lee, Yong Deok; Park, Chang Je; Park, Geun Il; Lee, Jung Won; Song, Kee Chan

    2011-01-01

    As an option to reduce a spent fuel and reuse an existing fissile material in spent fuel, sodium fast reactor SFR program linked with pyro-processing is under development in KAERI. A uranium-TRU mixture through a pyro-process is used to fabricate SFR fuel. An assay of isotopic fissile content plays an important role in an optimum design of storage site and reuse of fissile materials of spent fuel. Lead slowing down spectrometer LSDS is being developed in KAERI to analyze isotopic fissile material content. LSDS has several features: direct fissile assay, near real time fissile assay, no influence from radiation background, fissile isotopic assay and applicable to spent fuel and recycled fuel. Based on the designed geometry, neutron energy resolution was investigated. The neutron energy spectrum was analyzed as well. Spent fuel emits large number of neutrons by spontaneous fission. Neutron generator must overcome the neutron background to get the pure fission signals from fissile materials. Neutron generator is planned to have compact system with one section electron linac which is easy maintenance, less cost and high neutron yield. The LSD has the power to resolve the fission characteristics from each fissile material. This feature can analyze the content of isotopic fissile. From 1keV to 0.1eV energy range, the energy resolution is enough to get the individual fissile fission signatures. The dominant fission signature is shown below 1eV for each fissile isotope. The neutron generation system with target was designed to get fission signals by fissile materials. The system was decided to overcome neutron backgrounds and to get good counting statistics. Finally, an accurate fissile material content will contribute to safety of spent fuel reuse in future nuclear energy system and optimum design of spent fuel storage site. Additionally, an accurate fissile material content will increase international transparence and credibility for the reuse of PWR spent fuel

  16. Staatsblad 343 - Order of 4 June 1987 amending the Order concerning transport of fissile materials, ores and radioactive materials

    International Nuclear Information System (INIS)

    1987-01-01

    This Decree amends the 1969 Decree to take account of developments in international transport regulations, already taken into account in the national regulations for all modes of transport of dangerous materials or goods. Further amendments concern physical protection requirements in compliance which the Convention on the Physical Protection of Nuclear Material which the Netherlands signed as a Member State of the European Communities. In essence, the modifications relate to licensing requirements in particular packaging and transport conditions for the different levels of activity of the materials carried, certificates of approval etc., and surveillance during transport. The Decree entered into force on 23 August 1987 [fr

  17. Ceramics for Molten Materials Transfer

    Science.gov (United States)

    Standish, Evan; Stefanescu, Doru M.; Curreri, Peter A.

    2009-01-01

    The paper reviews the main issues associated with molten materials transfer and handling on the lunar surface during the operation of a hig h temperature electrowinning cell used to produce oxygen, with molten iron and silicon as byproducts. A combination of existing technolog ies and purposely designed technologies show promise for lunar exploi tation. An important limitation that requires extensive investigation is the performance of refractory currently used for the purpose of m olten metal containment and transfer in the lunar environment associa ted with electrolytic cells. The principles of a laboratory scale uni t at a scale equivalent to the production of 1 metric ton of oxygen p er year are introduced. This implies a mass of molten materials to be transferred consistent with the equivalent of 1kg regolithlhr proces sed.

  18. New Technology For Fissile Assay In Spent Fuel Using LSDS

    International Nuclear Information System (INIS)

    Lee, Yongdeok; Park, Changje; Park, Geunil; Lee, Jungwon; Song, Keechan

    2012-01-01

    The principle of LSDS is very simple. The interrogated neutron induces energy dependent characteristic fission from fissile materials in spent fuel. The fission threshold detector screens the prompt fast fission neutrons from background and fissionable materials. However, intense source neutron is necessary to overcome radiation background. The detected signals have a direct relationship to the content of each fissile material. The isotopic fissile assay using LSDS is applicable for optimum design of spent fuel storage and management, quality assurance of recycled nuclear material, maximization of burnup credit. Another important application is verity burnup code and provide correction factor for improving the fissile material content, fission product correction factor for improving the fissile material content, fission product content and theoretical burnup. Additionally, the isotopic fissile content assay will increase the transparence and credibility for spent fuel storage and its re-utilization, as internationally demanded

  19. Use of borosilicate-glass raschig rings as a neutron absorber in solutions of fissile material-ANSI/ANS-8.5-1996

    International Nuclear Information System (INIS)

    Rothe, R.E.; Ketzlach, N.; Finch, D.R.

    1996-01-01

    American National Standards Institute/American Nuclear Society (ANSI/ANS)-8.5 is one of several standards prepared by the ANS Standards Committee to provide guidance to enhance criticality safety in the handling, storage, and processing of fissionable materials. American National Standard ANSI/ANS-8.5-1996 provides this guidance for one type of boron-loaded glass in one type of geometry (cylindrical rings) for use with fissile solutions. Recorded use of such fixed neutron absorbers for criticality control of fissile solutions dates back to 1958, but some less-well-documented applications were recorded as early as the mid-1940's. The first solid efforts to collect recommendations derived from experience and technology were begun in 1965. Over the next 6 yr additional experiments were performed, and supporting data for the proposed standard were gathered. The first standard on this safety matter was issued in 1971. It was reaffirmed in 1979 with only minor changes and a slight expansion of the coverage. The standard was last revised in 1986

  20. Fissile Material Disposition Program: Deep borehole disposal Facility PEIS date input report for immobilized disposal. Immobilized disposal of plutonium in coated ceramic pellets in grout with canisters. Version 3.0

    International Nuclear Information System (INIS)

    Wijesinghe, A.M.; Shaffer, R.J.

    1996-01-01

    Following President Clinton's Non-Proliferation Initiative, launched in September, 1993, an Interagency Working Group (IWG) was established to conduct a comprehensive review of the options for the disposition of weapons-usable fissile materials from nuclear weapons dismantlement activities in the United States and the former Soviet Union. The IWG review process will consider technical, nonproliferation, environmental budgetary, and economic considerations in the disposal of plutonium. The IWG is co-chaired by the White House Office of Science and Technology Policy and the National Security Council. The Department of Energy (DOE) is directly responsible for the management, storage, and disposition of all weapons-usable fissile material. The Department of Energy has been directed to prepare a comprehensive review of long-term options for Surplus Fissile Material (SFM) disposition, taking into account technical, nonproliferation, environmental, budgetary, and economic considerations

  1. Fissile Material Disposition Program: Deep borehole disposal Facility PEIS date input report for immobilized disposal. Immobilized disposal of plutonium in coated ceramic pellets in grout with canisters. Version 3.0

    Energy Technology Data Exchange (ETDEWEB)

    Wijesinghe, A.M.; Shaffer, R.J.

    1996-01-15

    Following President Clinton`s Non-Proliferation Initiative, launched in September, 1993, an Interagency Working Group (IWG) was established to conduct a comprehensive review of the options for the disposition of weapons-usable fissile materials from nuclear weapons dismantlement activities in the United States and the former Soviet Union. The IWG review process will consider technical, nonproliferation, environmental budgetary, and economic considerations in the disposal of plutonium. The IWG is co-chaired by the White House Office of Science and Technology Policy and the National Security Council. The Department of Energy (DOE) is directly responsible for the management, storage, and disposition of all weapons-usable fissile material. The Department of Energy has been directed to prepare a comprehensive review of long-term options for Surplus Fissile Material (SFM) disposition, taking into account technical, nonproliferation, environmental, budgetary, and economic considerations.

  2. Fissile Content Assay of Spent Fuel Using LSDS System

    International Nuclear Information System (INIS)

    Jeon, Ju Young; Lee, Yong Deok; Park, Chang Je

    2016-01-01

    About 1.5 % fissile materials still exist in the spent fuel. Therefore, for reutilization of fissile materials in spent fuel at SFR, resource material is produced through the pyro process. Fissile material contents in the resource material must be analyzed before fabricating SFR fuel for reactor safety and economics. The new technology for an isotopic fissile material content assay is under development at KAERI using a lead slowing down spectrometer (LSDS). LSDS is very sensitive to distinguish fission signals from each fissile isotope in spent and recycled fuel. In an assay of fissile content of spent fuel and recycled fuel, an intense radiation background gives limits the direct analysis of fissile materials. However, LSDS is not influenced by such a radiation background in a fissile assay. Based on the decided LSDS geometry set up, a self shielding parameter was calculated at the fuel assay zone by introducing spent fuel or pyro produced nuclear material. When nuclear material is inserted into the assay area, the spent fuel assembly or pyro recycled fuel material perturbs the spatial distribution of slowing down neutrons in lead and the prompt fast fission neutrons produced by fissile materials are also perturbed. The self shielding factor is interpreted as how much of the absorption is created inside the fuel area when it is in the lead. The self shielding effect provides a non-linear property in the isotopic fissile assay. When the self shielding is severe, the assay system becomes more complex and needs a special parameter to treat this non linear effect. Additionally, an assay of isotopic fissile content will contribute to an accuracy improvement of the burn-up code and increase the transparency and credibility for spent fuel storage and usage, as internationally increasing demand. The fissile contents result came out almost exactly with relative error ∼ 2% in case of Pu239, Pu241 for two different plutonium contents. In this study, meaningful results were

  3. Self shielding in cylindrical fissile sources in the APNea system

    International Nuclear Information System (INIS)

    Hensley, D.

    1997-01-01

    In order for a source of fissile material to be useful as a calibration instrument, it is necessary to know not only how much fissile material is in the source but also what the effective fissile content is. Because uranium and plutonium absorb thermal neutrons so Efficiently, material in the center of a sample is shielded from the external thermal flux by the surface layers of the material. Differential dieaway measurements in the APNea System of five different sets of cylindrical fissile sources show the various self shielding effects that are routinely encountered. A method for calculating the self shielding effect is presented and its predictions are compared with the experimental results

  4. Epithermal interrogation of fissile waste

    International Nuclear Information System (INIS)

    Coop, K.L.; Hollas, C.L.

    1996-01-01

    Self-shielding of interrogating thermal neutrons in lumps of fissile material can be a major source of error in transuranic waste assay using the widely employed differential dieaway technique. We are developing a new instrument, the combined thermal/epithermal neutron (CTEN) interrogation instrument to detect the occurrence of self- shielding and mitigate its effects. Neutrons are moderated in the graphite walls of the CTEN instrument to provide an interrogating flux of epithermal and thermal neutrons. The induced prompt fission neutrons are detected in proportional counters. We report the results of measurements made with the CTEN instrument, using minimal and highly self-shielding plutonium and uranium sources in 55 gallon drums containing a variety of mock waste matrices. Fissile isotopes and waste forms for which the method is most applicable, and limitations associated with the hydrogen content of the waste package/matrix are described

  5. International conference on military conversion and science. Utilization/disposal of the excess fissile weapon materials: scientific, technological and socio-economic aspects

    International Nuclear Information System (INIS)

    Kouzminov, V.; Martellini, M.

    1996-01-01

    The Proceedings of the Conference includes the papers presented by the eminent specialists in the field of utilisation and/or disposal of excess fissile materials, each with a separate abstract, as well as the Conference opening and introduction speeches. According to the concerned subjects presentations were divided into following five sessions: perspectives of nuclear research and development; Technical problems and possibilities of civilian utilization of Highly enriched uranium (HEU) and plutonium including alternate strategies (application of MOX fuel) and operational and safety problems; Comparison of different options for weapon-grade Pu utilization connected to present programme for recycling of civilian Pu; Socio-economic aspects including cost of Pu conversion and fabrication of MOX fuel; Effects of different strategies of waste disposal including environmental and safety related issues

  6. Proceedings of the workshop on a comparative analysis of approaches to the protection of fissile materials, Stanford University, July 28-30, 1997

    International Nuclear Information System (INIS)

    Goodby, J.E.; Lehman, R. III; Potter, W.C.

    1998-01-01

    Events in recent years have caused heightened concern about the security of weapons-usable nuclear material. The possibility of illicit trafficking in, or seizure of, such material, leading to nuclear terrorism, is a worry for all states and their citizens. And given the relatively small quantities required, material obtained in one part of the world could be made into a weapon in another and threaten lives in a third. It is truly a global problem. Since the beginning of the nuclear era, the physical protection of fissile material has been a responsibility of the individual states possessing the material. These states have different organizational approaches for providing physical protection; and while cognizant of recommended general standards, they tend to follow their own practices, shaped by custom, costs, and threat perception. Moreover, the existence of military as well as civil programs in some states adds another dimension to the physical protection issue. Because physical protection is a sovereign matter and not part of an international regime (except for transit of civil material across borders), there has been less attention in much of the world community to the issues of physical protection than to the other elements of nuclear safeguards and controls. (An important exception to this situation is the effort being made to assist the states of the former Soviet Union in the disposition of their weapons-usable nuclear materials.) The lack of a general dialog about a problem of growing concern motivated us to hold a three-day workshop at Stanford University to develop a better understanding of some of the important underlying questions and issues, and to undertake a comparative examination of states' approaches to physical protection. We were pleased to have knowledgeable participants from a number of the countries and regions where physical protection of fissile materials is, or will become, a day-to-day matter. The results of the workshop are reported in

  7. Source modulation-correlation measurement for fissile mass flow in gas or liquid fissile streams

    International Nuclear Information System (INIS)

    Mihalczo, J.T.; March-Leuba, J.A.; Valentine, T.E.; Abston, R.A.; Mattingly, J.K.; Mullens, J.A.

    1996-01-01

    The method of monitoring fissile mass flow on all three legs of a blending point, where the input is high-enriched uranium (HEU) and low-enriched uranium (LEU) and the product is PEU, can yield the fissile stream velocity and, with calibration, the [sup235]U content. The product of velocity and content integrated over the pipe gives the fissile mass flow in each leg. Also, the ratio of fissile contents in each pipe: HEU/LEU, HEU/PEU, and PEU/LEU, are obtained. By modulating the source on the input HEU pipe differently from that on the output pipe, the HEU gas can be tracked through the blend point. This method can be useful for monitoring flow velocity, fissile content, and fissile mass flow in HEU blenddown of UF[sub 6] if the pressures are high enough to contain some of the induced fission products. This method can also be used to monitor transfer of fissile liquids and other gases and liquids that emit radiation delayed from particle capture. These preliminary experiments with the Oak Ridge apparatus show that the method will work and the modeling is adequate

  8. Variants of Regenerated Fissile Materials Usage in Thermal Reactors as the First Stage of Fuel Cycle Closing

    Science.gov (United States)

    Andrianova, E. A.; Tsibul'skiy, V. F.

    2017-12-01

    At present, 240 000 t of spent nuclear fuel (SF) has been accumulated in the world. Its long-term storage should meet safety conditions and requires noticeable finances, which grow every year. Obviously, this situation cannot exist for a long time; in the end, it will require a final decision. At present, several variants of solution of the problem of SF management are considered. Since most of the operating reactors and those under construction are thermal reactors, it is reasonable to assume that the structure of the nuclear power industry in the near and medium-term future will be unchanged, and it will be necessary to utilize plutonium in thermal reactors. In this study, different strategies of SF management are compared: open fuel cycle with long-term SF storage, closed fuel cycle with MOX fuel usage in thermal reactors and subsequent long-term storage of SF from MOX fuel, and closed fuel cycle in thermal reactors with heterogeneous fuel arrangement. The concept of heterogeneous fuel arrangement is considered in detail. While in the case of traditional fuel it is necessary to reprocess the whole amount of spent fuel, in the case of heterogeneous arrangement, it is possible to separate plutonium and 238U in different fuel rods. In this case, it is possible to achieve nearly complete burning of fissile isotopes of plutonium in fuel rods loaded with plutonium. These fuel rods with burned plutonium can be buried after cooling without reprocessing. They would contain just several percent of initially loaded plutonium, mainly even isotopes. Fuel rods with 238U alone should be reprocessed in the usual way.

  9. New glass material oxidation and dissolution system facility: Direct conversion of surplus fissile materials, spent nuclear fuel, and other material to high-level-waste glass. Storage and disposition of weapons-usable fissile materials programmatic environmental impact statement data report: Predecisional draft

    International Nuclear Information System (INIS)

    Forsberg, C.W.; Elam, K.R.; Reich, W.J.

    1995-01-01

    With the end of the Cold War, countries have excess plutonium and other materials from the reductions in inventories of nuclear weapons. It has been recommended that these surplus fissile materials (SFMs) be processed so that they are no more accessible than plutonium in spent nuclear fuel (SNF). This SNF standard, if adopted worldwide, would prevent rapid recovery of SFMs for the manufacture of nuclear weapons. This report provides for the PEIS the necessary input data on a new method for the disposition of SFMs: the simultaneous conversion of SFMs, SNF, and other highly radioactive materials into high-level-waste (HLW) glass. The SFMs include plutonium, neptunium, americium, and 233 U. The primary SFM is plutonium. The preferred SNF is degraded SNF, which may require processing before it can be accepted by a geological repository for disposal. The primary form of this SNF is Hanford-N SNF with preirradiation uranium enrichments between 0.95 and 1.08%. The final product is a plutonium, low-enriched-uranium, HLW, borosilicate glass for disposition in a geological repository. The proposed conversion process is the Glass Material Oxidation and Dissolution System (GMODS), which is a new process. The initial analysis of the GMODS process indicates that a MODS facility for this application would be similar in size and environmental impact to the Defense Waste Processing Facility (DWPF) at the Savannah River Site. Because of this, the detailed information available on DWPF was used as the basis for much of the GMODS input into the SFMs PEIS

  10. Material Transfer Agreement (MTA) | FNLCR Staging

    Science.gov (United States)

    Material Transfer Agreements are appropriate for exchange of materials into or out of the Frederick National Labfor research or testing purposes, with no collaborative research by parties involving the materials.

  11. 16 October 1991-Royal Order amending the Royal Order of 30 March 1981 determining the duties and fixing the operating conditions of the Public Body for the Management of Radioactive Waste and Fissile Materials

    International Nuclear Information System (INIS)

    1991-01-01

    The 1991 Royal Order amends and supplements the provisions of the 1981 0rder dealing with the duties and resources of ONDRAF, the National Body for the Management of Radioactive Waste and Fissile Materials. Its duties include, inter alia, treatment and conditioning of waste on behalf of producers without the necessary facilities, training of specialists for such work for the producers with such facilities, transport, storage and disposal of radioactive waste, transport, and storage of certain enriched fissile materials and plutonium-bearing materials. As regards decommissioned nuclear installations, ONDRAF must establish management programmes for the resulting waste and must also decommission a nuclear installation at the operator's request or if he defaults. (NEA)

  12. Safety Analysis Report: Packages, Pu oxide and Am oxide shipping cask: Packaging of fissile and other radioactive materials: Final report

    International Nuclear Information System (INIS)

    Chalfant, G.G.

    1984-12-01

    The PuO 2 cask or 5320-3 cask is designed for shipment of americium or plutonium by surface transportation modes. The cask design was physically tested to demonstrate that it met the criteria specified in US ERDA Manual Chapter 0529, dated 12/21/76, which invokes Title 10 Code of Federal Regulations, Part 71 (10 CFR 71) ''Packaging of Radioactive Materials for Transport,'' and Title 49 CFR Parts 171.179 ''Hazardous Materials Regulations.'' (US DOE Order 4580.1A, Chapter III, superseded manual chapter 0529 effective May 1981, but it retained the same 10 CFR 71 and 49 CFR 171-179 references

  13. A setup for active neutron analysis of the fissile material content in fuel assemblies of nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bushuev, A. V.; Kozhin, A. F., E-mail: alexfkozhin@yandex.ru; Aleeva, T. B.; Zubarev, V. N.; Petrova, E. V.; Smirnov, V. E. [National Research Nuclear University MEPhI (Russian Federation)

    2016-12-15

    An active neutron method for measuring the residual mass of {sup 235}U in spent fuel assemblies (FAs) of the IRT MEPhI research reactor is presented. The special measuring stand design and uniform irradiation of the fuel with neutrons along the entire length of the active part of the FA provide high accuracy of determination of the residual {sup 235}U content. AmLi neutron sources yield a higher effect/background ratio than other types of sources and do not induce the fission of {sup 238}U. The proposed method of transfer of the isotope source in accordance with a given algorithm may be used in experiments where the studied object needs to be irradiated with a uniform fluence.

  14. Max-von-Laue-lecture: Unmaking the bomb: A fissile material approach to nuclear disarmament and nonproliferation

    Energy Technology Data Exchange (ETDEWEB)

    Von Hippel, Frank N. [Princeton University, Princeton, NJ (United States)

    2015-07-01

    The number of operational nuclear weapons in the world has dropped from about 65,000 at the end of the Cold war to about 10,000 and can be driven much lower. But we have a huge amount of highly enriched uranium and separated plutonium from these dismantled Cold War nuclear weapons and from failed civilian plutonium breeder reactor commercialization programs. To make nuclear disarmament irreversible and prevent nuclear terrorism, all this material must be secured and disposed of. We also must abandon the idea of using a nuclear-weapon-usable material as a fuel * that is plutonium in power reactors and highly enriched uranium in naval-propulsion and research reactors. Fortunately, using plutonium as a fuel is uneconomic and research and naval reactors can be designed to use low-enriched uranium. Finally, we must move away from ambiguous national enrichment programs like Iran*s to multinational enrichment programs such as Urenco.

  15. Some aspects of in-pile swelling of fissile materials, 1. part: non-alloyed α uranium

    International Nuclear Information System (INIS)

    Mikailoff, H.

    1964-01-01

    An examination has been carried out of non-alloyed uranium samples, having various structural states, cold-worked and recrystallized, as-cast and β-treated, and irradiated at temperatures of between 450 and 600 C and with burn-ups from 1300 to 5500 MW days/metric ton. These samples swelled because of precipitation of the fission gases the porosity thus produced has a morphology depending mainly on the type of deformation to which the metal has been subjected and which is due to in-pile growth. The most homogeneous distribution of pores, and thus that leading to the minimum swelling, is only observed in the material having a marked [010] texture in which the growth and perhaps the thermal cycling introduce little or no strain. For other materials the deformation /swelling association causes a more rapid destruction of the samples either by cracking when the deformation is due to twinning, or by pronounced swelling localized in the bands when deformation is due to slipping. Finally the fission-gas precipitation considerably facilitates, above 500 C, the germination and growth of the intergranular cracks which can then develop at low stresses. (author) [fr

  16. Characterization of a facility for the measurement of fission fragment transport effects: experimental determination of the fission rates for fissile and fissionable isotopes

    International Nuclear Information System (INIS)

    Benetti, P.; Raselli, G.L.; Tigliole, A. Borio di; Cagnazzo, M.; Cesana, A.; Mongelli, S.; Terrani, M.

    2002-01-01

    The transfer facility of the LENA laboratory allows the direct neutron irradiation of fissionable material in the D channel of the TRIGA reactor. A test measurement carried out with a ionization chamber and a 239 Pu sample shows the possibility to use this tool for the study of the transport effects of the fission fragment emerging from thin layers of fissile materials. (author)

  17. Los Alamos National Laboratory summary plan to fabricate mixed oxide lead assemblies for the fissile material disposition program

    Energy Technology Data Exchange (ETDEWEB)

    Buksa, J.J.; Eaton, S.L.; Trellue, H.R.; Chidester, K.; Bowidowicz, M.; Morley, R.A.; Barr, M.

    1997-12-01

    This report summarizes an approach for using existing Los Alamos National Laboratory (Laboratory) mixed oxide (MOX) fuel-fabrication and plutonium processing capabilities to expedite and assure progress in the MOX/Reactor Plutonium Disposition Program. Lead Assembly MOX fabrication is required to provide prototypic fuel for testing in support of fuel qualification and licensing requirements. It is also required to provide a bridge for the full utilization of the European fabrication experience. In part, this bridge helps establish, for the first time since the early 1980s, a US experience base for meeting the safety, licensing, safeguards, security, and materials control and accountability requirements of the Department of Energy and Nuclear Regulatory Commission. In addition, a link is needed between the current research and development program and the production of disposition mission fuel. This link would also help provide a knowledge base for US regulators. Early MOX fabrication and irradiation testing in commercial nuclear reactors would provide a positive demonstration to Russia (and to potential vendors, designers, fabricators, and utilities) that the US has serious intent to proceed with plutonium disposition. This report summarizes an approach to fabricating lead assembly MOX fuel using the existing MOX fuel-fabrication infrastructure at the Laboratory.

  18. Los Alamos National Laboratory summary plan to fabricate mixed oxide lead assemblies for the fissile material disposition program

    International Nuclear Information System (INIS)

    Buksa, J.J.; Eaton, S.L.; Trellue, H.R.; Chidester, K.; Bowidowicz, M.; Morley, R.A.; Barr, M.

    1997-12-01

    This report summarizes an approach for using existing Los Alamos National Laboratory (Laboratory) mixed oxide (MOX) fuel-fabrication and plutonium processing capabilities to expedite and assure progress in the MOX/Reactor Plutonium Disposition Program. Lead Assembly MOX fabrication is required to provide prototypic fuel for testing in support of fuel qualification and licensing requirements. It is also required to provide a bridge for the full utilization of the European fabrication experience. In part, this bridge helps establish, for the first time since the early 1980s, a US experience base for meeting the safety, licensing, safeguards, security, and materials control and accountability requirements of the Department of Energy and Nuclear Regulatory Commission. In addition, a link is needed between the current research and development program and the production of disposition mission fuel. This link would also help provide a knowledge base for US regulators. Early MOX fabrication and irradiation testing in commercial nuclear reactors would provide a positive demonstration to Russia (and to potential vendors, designers, fabricators, and utilities) that the US has serious intent to proceed with plutonium disposition. This report summarizes an approach to fabricating lead assembly MOX fuel using the existing MOX fuel-fabrication infrastructure at the Laboratory

  19. Self Shielding in Nuclear Fissile Assay Using LSDS

    International Nuclear Information System (INIS)

    Lee, Yong Deok; Park, Chang Je; Park, Geun Il; Song, Kee Chan

    2012-01-01

    The new technology for isotopic fissile material contents assay is under development at KAERI using lead slowing down spectrometer(LSDS). LSDS is very sensitive to distinguish fission signals from each fissile isotope in spent and recycled fuel. The accumulation of spent fuel is current big issue. The amount of spent fuels will reach the maximum storage capacity of the pools soon. Therefore, an interim storage must be searched and it should be optimized in design by applying accurate fissile content. When the storage has taken effect, all the nuclear materials must be also specified and verified for safety, economics and management. Generally, the spent fuel from PWR has unburned ∼1 % U235, produced ∼0.5 % plutonium from decay chain, ∼3 % fission products, ∼ 0.1 % minor actinides (MA) and uranium remainder. About 1.5 % fissile materials still exist in the spent fuel. Therefore, for reutilization of fissile materials in spent fuel at SFR, resource material is produced through pyro process. Fissile material contents in resource material must be analyzed before fabricating SFR fuel for reactor safety and economics. In assay of fissile content of spent fuel and recycled fuel, intense radiation background gives limitation on the direct analysis of fissile materials. However, LSDS is not influenced by such a radiation background in fissile assay. Based on the decided geometry setup, self shielding parameter was calculated at the fuel assay zone by introducing spent fuel or pyro produced nuclear material. When nuclear material is inserted into the assay area, the spent fuel assembly or pyro recycled fuel material perturbs the spatial distribution of the slowing down neutrons in lead and the prompt fast fission neutrons produced by fissile materials are also perturbed. The self shielding factor is interpreted as that how much of absorption is created inside the fuel area when it is in the lead. Self shielding effect provides a non-linear property in the isotopic

  20. LSDS Development for Isotopic Fissile Content Assay

    International Nuclear Information System (INIS)

    Lee, Yong Deok; Park, Chang Je; Park, Geun Il; Lee, Jung Won; Song, Kee Chan

    2010-01-01

    Concerning the sustainable energy supply and green house effect, nuclear energy became the most feasible option to meet the energy demand in Korea. However, the production of the spent nuclear fuel is the inevitable situation. Since the first nuclear power plant started to produce the electricity in Korea, the accumulated amount of spent fuels exceeded 10k tomes recently. The accumulation of the spent fuels is the big issue in the society. Therefore, as an option which strengthens the nuclear proliferation resistance and reduces the amount of spent fuels, sodium fast reactor (SFR) program linked with pyro-processing is under development to re-use the PWR spent fuel and produce the energy. In the process, the produced metallic material involves uranium and TRU (transuranic; neptunium, plutonium, and americium). The uranium-TRU is used to fabricate SFR fuel. The burning the recycled fuel in the reactor is to solve the current spent fuel storage problem and to minimize the actinides accumulation having long half-life. Generally, the spent fuel from PWR has unburned ∼1 % U235, produced ∼0.5 % plutonium from decay chain, ∼3 % fission products, ∼ 0.1 % minor actinides (MA) and uranium remainder. About 1.5 % fissile materials still exist in the spent fuel. Therefore, spent fuel is not only waste but energy resource. The direct and isotopic fissile content assay is the crucial technology for the spent fuel reuse. Additionally, the fissile content analysis will contribute to the optimum storage design and safe spent fuel management. Several nondestructive technologies have been developed for the spent fuel assay; gamma ray measurement, passive and active neutron measurements. Spent fuel emits intense gamma rays and neutrons by (a, n) and spontaneous fission. This intense background has the limitation on the direct analysis of fissile materials. Recently, to analyze the individual fissile content, leadslowing down spectrometer (LSDS) has been being developed in Korea

  1. Quantitative Fissile Assay In Used Fuel Using LSDS System

    Science.gov (United States)

    Lee, YongDeok; Jeon, Ju Young; Park, Chang-Je

    2017-09-01

    A quantitative assay of isotopic fissile materials (U235, Pu239, Pu241) was done at Korea Atomic Energy Research Institute (KAERI), using lead slowing down spectrometer (LSDS). The optimum design of LSDS was performed based on economics, easy maintenance and assay effectiveness. LSDS system consists of spectrometer, neutron source, detection and control. LSDS system induces fissile fission and fast neutrons are collected at fission chamber. The detected signal has a direct relation to the mass of existing fissile isotopes. Many current commercial assay technologies have a limitation in direct application on isotopic fissile assay of spent fuel, except chemical analysis. In the designed system, the fissile assay model was setup and the correction factor for self-shield was obtained. The isotopic fissile content assay was performed by changing the content of Pu239. Based on the fuel rod, the isotopic content was consistent with 2% uncertainty for Pu239. By applying the covering (neutron absorber), the effective shielding was obtained and the activation was calculated on the target. From the assay evaluation, LSDS technique is very powerful and direct to analyze the isotopic fissile content. LSDS is applicable for nuclear fuel cycle and spent fuel management for safety and economics. Additionally, an accurate fissile content will contribute to the international transparency and credibility on spent fuel.

  2. Thermal transfer in multilayer materials

    Energy Technology Data Exchange (ETDEWEB)

    Bouayad, H.; Mokhtari, A.; Martin, C.; Fauchais, P. [Laboratoire de Materiaux Ceramiques et Traitements de Surface, 87 - Limoges (France)

    1993-12-31

    It is easier to measure the thermal diffusivity (a) of material rather than its thermal conductivity (k), a simple relationship (k=a cp) allowing to calculate k provided and cp are measured. However this relationship applies only if the considered material is homogenous. For composite materials, especially for multilayers ones, we have developed an analytical model and a numerical one. The first one allows to determine the thermal diffusivity and conductivity of a two-layer material. The second one allows to determine the thermal diffusivity of one of the layers provided the values of (a) are known for the two other layers (for a two or three-layer material). The use of the two models to calculate the apparent diffusivity of a two layer material results in values in fairly good agreement. (Authors). 4 refs., 3 figs., 3 tabs.

  3. Heat transfer in multi-phase materials

    CERN Document Server

    Öchsner, Andreas

    2011-01-01

    This book provides a profound understanding, which physical processes and mechanisms cause the heat transfer in composite and cellular materials. It shows models for all important classes of composite materials and introduces into the latest advances. In three parts, the book covers Composite Materials (Part A), Porous and Cellular Materials (Part B) and the appearance of a conjoint solid phase and fluid aggregate (Part C).

  4. Development of lead slowing down spectrometer for isotopic fissile assay

    International Nuclear Information System (INIS)

    Lee, Yong Deok; Park, Chang Je; Ahn, Sang Joon; Kim, Ho Dong

    2014-01-01

    A lead slowing down spectrometer (LSDS) is under development for analysis of isotopic fissile material contents in pyro-processed material, or spent fuel. Many current commercial fissile assay technologies have a limitation in accurate and direct assay of fissile content. However, LSDS is very sensitive in distinguishing fissile fission signals from each isotope. A neutron spectrum analysis was conducted in the spectrometer and the energy resolution was investigated from 0.1eV to 100keV. The spectrum was well shaped in the slowing down energy. The resolution was enough to obtain each fissile from 0.2eV to 1keV. The detector existence in the lead will disturb the source neutron spectrum. It causes a change in resolution and peak amplitude. The intense source neutron production was designed for ∼E12 n's/sec to overcome spent fuel background. The detection sensitivity of U238 and Th232 fission chamber was investigated. The first and second layer detectors increase detection efficiency. Thorium also has a threshold property to detect the fast fission neutrons from fissile fission. However, the detection of Th232 is about 76% of that of U238. A linear detection model was set up over the slowing down neutron energy to obtain each fissile material content. The isotopic fissile assay using LSDS is applicable for the optimum design of spent fuel storage to maximize burnup credit and quality assurance of the recycled nuclear material for safety and economics. LSDS technology will contribute to the transparency and credibility of pyro-process using spent fuel, as internationally demanded.

  5. 49 CFR 172.441 - FISSILE label.

    Science.gov (United States)

    2010-10-01

    ... 49 Transportation 2 2010-10-01 2010-10-01 false FISSILE label. 172.441 Section 172.441... SECURITY PLANS Labeling § 172.441 FISSILE label. (a) Except for size and color, the FISSILE label must be... FISSILE label must be white. [69 FR 3669, Jan. 26, 2004] ...

  6. Some aspects of in-pile swelling of fissile materials, 1. part: non-alloyed {alpha} uranium; Quelques aspects du gonflement en pile des materiaux fissiles. 1. partie: uranium {alpha} non allie

    Energy Technology Data Exchange (ETDEWEB)

    Mikailoff, H [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

    1964-07-01

    An examination has been carried out of non-alloyed uranium samples, having various structural states, cold-worked and recrystallized, as-cast and {beta}-treated, and irradiated at temperatures of between 450 and 600 C and with burn-ups from 1300 to 5500 MW days/metric ton. These samples swelled because of precipitation of the fission gases the porosity thus produced has a morphology depending mainly on the type of deformation to which the metal has been subjected and which is due to in-pile growth. The most homogeneous distribution of pores, and thus that leading to the minimum swelling, is only observed in the material having a marked [010] texture in which the growth and perhaps the thermal cycling introduce little or no strain. For other materials the deformation /swelling association causes a more rapid destruction of the samples either by cracking when the deformation is due to twinning, or by pronounced swelling localized in the bands when deformation is due to slipping. Finally the fission-gas precipitation considerably facilitates, above 500 C, the germination and growth of the intergranular cracks which can then develop at low stresses. (author) [French] On a examine des echantillons d'uranium non allie, de divers etats structuraux, marteles et recristallises, bruts de coulee et traites {beta}, irradies a des temperatures comprises entre 450 et 600 C, et a des taux de combustion allant de 1300 a 5500 MWj/t. Ces echantillons ont gonfle par suite de la precipitation de gaz de fission: la porosite ainsi fournie a une morphologie qui depend principalement des modes de deformation subie par le metal et due a la croissance en pile. La repartition la plus homogene des pores, donc celle qui donnera le gonflement minimum, est observee seulement dans le materiau a forte texture [010] dans lequel la croissance et eventuellement le cyclage thermique introduisent peu ou pas de contraintes. Dans les autres materiaux l'association deformation/gonflement rend plus rapide

  7. Design of LSDS for Isotopic Fissile Assay in Spent Fuel

    International Nuclear Information System (INIS)

    Lee, Yongdeok; Park, Changje; Kim, Hodong; Song, Kee Chan

    2013-01-01

    A future nuclear energy system is being developed at Korea Atomic Energy Research Institute (KAERI), the system involves a Sodium Fast Reactor (SFR) linked with the pyro-process. The pyro-process produces a source material to fabricate a SFR fuel rod. Therefore, an isotopic fissile content assay is very important for fuel rod safety and SFR economics. A new technology for an analysis of isotopic fissile content has been proposed using a lead slowing down spectrometer (LSDS). The new technology has several features for a fissile analysis from spent fuel: direct isotopic fissile assay, no background interference, and no requirement from burnup history information. Several calculations were done on the designed spectrometer geometry: detection sensitivity, neutron energy spectrum analysis, neutron fission characteristics, self shielding analysis, and neutron production mechanism. The spectrum was well organized even at low neutron energy and the threshold fission chamber was a proper choice to get prompt fast fission neutrons. The characteristic fission signature was obtained in slowing down neutron energy from each fissile isotope. Another application of LSDS is for an optimum design of the spent fuel storage, maximization of the burnup credit and provision of the burnup code correction factor. Additionally, an isotopic fissile content assay will contribute to an increase in transparency and credibility for the utilization of spent fuel nuclear material, as internationally demanded

  8. DESIGN OF LSDS FOR ISOTOPIC FISSILE ASSAY IN SPENT FUEL

    Directory of Open Access Journals (Sweden)

    YONGDEOK LEE

    2013-12-01

    Full Text Available A future nuclear energy system is being developed at Korea Atomic Energy Research Institute (KAERI, the system involves a Sodium Fast Reactor (SFR linked with the pyro-process. The pyro-process produces a source material to fabricate a SFR fuel rod. Therefore, an isotopic fissile content assay is very important for fuel rod safety and SFR economics. A new technology for an analysis of isotopic fissile content has been proposed using a lead slowing down spectrometer (LSDS. The new technology has several features for a fissile analysis from spent fuel: direct isotopic fissile assay, no background interference, and no requirement from burnup history information. Several calculations were done on the designed spectrometer geometry: detection sensitivity, neutron energy spectrum analysis, neutron fission characteristics, self shielding analysis, and neutron production mechanism. The spectrum was well organized even at low neutron energy and the threshold fission chamber was a proper choice to get prompt fast fission neutrons. The characteristic fission signature was obtained in slowing down neutron energy from each fissile isotope. Another application of LSDS is for an optimum design of the spent fuel storage, maximization of the burnup credit and provision of the burnup code correction factor. Additionally, an isotopic fissile content assay will contribute to an increase in transparency and credibility for the utilization of spent fuel nuclear material, as internationally demanded.

  9. Methodology for interpretation of fissile mass flow measurements

    International Nuclear Information System (INIS)

    March-Leuba, J.; Mattingly, J.K.; Mullens, J.A.

    1997-01-01

    This paper describes a non-intrusive measurement technique to monitor the mass flow rate of fissile material in gaseous or liquid streams. This fissile mass flow monitoring system determines the fissile mass flow rate by relying on two independent measurements: (1) a time delay along a given length of pipe, which is inversely proportional to the fissile material flow velocity, and (2) an amplitude measurement, which is proportional to the fissile concentration (e.g., grams of 235 U per length of pipe). The development of this flow monitor was first funded by DOE/NE in September 95, and initial experimental demonstration by ORNL was described in the 37th INMM meeting held in July 1996. This methodology was chosen by DOE/NE for implementation in November 1996; it has been implemented in hardware/software and is ready for installation. This paper describes the methodology used to interpret the data measured by the fissile mass flow monitoring system and the models used to simulate the transport of fission fragments from the source location to the detectors

  10. Fissile fuel assembly for a sub-moderated nuclear reactor

    International Nuclear Information System (INIS)

    Millot, J.P.; Dejeux, Pol.; Alibran, Patrice.

    1983-01-01

    Each of the core assemblies is composed of a prismatic case made of a neutron absorbing material, inside which very long rods containing the fissile material are arranged parallel to the height of the case and according to a regular network in the straight sections of the case. At least one piece in a fertile material exposed to the neutrons emitted by the fissile material of the assembly is arranged on each one of the side faces of the case. The invention applies in particular to sub-moderated reactors, cooled and moderated by pressurized water [fr

  11. Criteria for onsite transfers of radioactive material

    International Nuclear Information System (INIS)

    Opperman, E.K.; Jackson, E.J.; Eggers, A.G.

    1992-01-01

    A general description of the requirements for making onsite transfers of radioactive material is provided in Chapter 2, along with the required sequencey of activities. Various criteria for package use are identified in Chapters 3-13. These criteria provide protection against undue radiation exposure. Package shielding, containment, and surface contamination requirements are established. Criteria for providing criticality safety are enumerated in Chapter 6. Criteria for providing hazards information are established in Chapter 13. A glossary is provided

  12. Accounting Systems for Heavy Water and Fissionable Materials; Comptabilite de l'Eau Lourde et des Matieres Fissiles; Sistema ucheta tyazheloj vody i delyashchikhsya materialov; Sistemas de Contabilidad para el Agua Pesada y los Materiales Fisionables

    Energy Technology Data Exchange (ETDEWEB)

    Fletcher, G. W.; Reid, H. B.; Jenkinson, W. G. [Atomic Energy of Canada Ltd., Chalk River, ON (Canada)

    1966-02-15

    Detailed accounting and reporting procedures used by Atomic Energy of Canada Limited (AECL) for maintaining adequate records and control of heavy water supplies and stocks of fissionable materials are described, along with the duties and responsibilities of those administering the system. An appraisal is made of these procedures with respect to their adaptability for use in rapidly expanding research and power programmes. In particular the use of electronic data processing equipment is evaluated. A senior management committee is responsible for ensuring that there is a proper system for recording, reporting and controlling fissionable materials. The Production Planning and Control Branch (Pp and C B) of the Operations Division at the Chalk River Nuclear Laboratories (CRNL) is responsible to the committee for keeping the over-all records and for the general administration of the system. The duties involved are detailed in the report. The system for fissionable materials is segregated into several accountability units 15 of which are allocated to AECL departments and the others to Canadian industries and research organizations. A control ledger is kept by PP and CB for each of the units; however, the units are responsible for preparing detailed accounts of all material under their jurisdiction. The basic recording procedures covering the movement Of materials between units, the changing of forms within units, the handling of gains and losses, and disposals, are outlined in the report. The transfer of this data to IBM cards, the ultimate processing through an IBM 1401 computer and the preparation of reports for management approval are described. The heavy-water accounting system based on the same principles as used for the fissionable materials is explained. In this case the control ledger lists the pounds of D{sub 2}O allocated to each of the 15 accountability units. Again the basic recording methods and the use of a computer system are outlined. (author) [French

  13. Laser induced forward transfer of soft materials

    International Nuclear Information System (INIS)

    Palla-Papavlu, A; Dinca, V; Luculescu, C; Dinescu, M; Shaw-Stewart, J; Lippert, T; Nagel, M

    2010-01-01

    A strong research effort is presently aimed at patterning methodologies for obtaining controlled defined micrometric polymeric structures for a wide range of applications, including electronics, optoelectronics, sensors, medicine etc. Lasers have been identified as appropriate tools for processing of different materials, such as ceramics and metals, but also for soft, easily damageable materials (biological compounds and polymers). In this work we study the dynamics of laser induced forward transfer (LIFT) with a gap between the donor and the receiver substrates, which is the basis for possible applications that require multilayer depositions with high spatial resolution

  14. Status of LSDS Development for Isotopic Fissile Assay in Used Fuel

    International Nuclear Information System (INIS)

    Lee, Y.D.; Ahn, S.; Kim, H.-D.; Song, K.C.; Park, C.J.

    2015-01-01

    Because of the large amount accumulation of spent fuel, a research to solve the spent fuel problem is actively performed in Korea. One option is to develop the SFR linked with the pyro process to reuse the existing fissile materials in spent fuel. Therefore, an accurate isotopic fissile content assay becomes a key factor in the reuse of fissile material for safety and safeguards purpose. There are several commercial non-destructive technologies for nuclear material assay. However, technology for direct isotopic fissile content assay in spent fuel is not developed yet. Internationally, a verification of special nuclear material in spent fuel, mainly U-235, Pu239, Pu241, is very important for the safeguards objective. These fissile materials can be misused for nuclear weapon purpose, not for peaceful use. As a future nuclear system is developed,, improved safeguards technology must be developed for an approval of fissile materials. A direct measurement of fissile materials is very important to provide a continuous of knowledge on nuclear materials. LSDS (Lead Slowing Down Spectrometer) has an advantage to assay an isotopic fissile content directly, without any help of burnup code and history. LSDS system is under development in KAERI (Korea Atomic Energy Research Institute) for spent fuel and recycled fuel. A linear assay model was setup for U235, Pu239 and Pu241. The dominant individual fission characteristic is appeared between 0.1 eV and 1 keV range. An electron linear accelerator for compact and low cost is under development to produce high source neutron effectively and efficiently. The LSDS is also applicable for optimum design of spent fuel storage and management. The advanced fissile assay technology will contribute to increase the transparency and credibility internationally on a reuse of fissile materials in future nuclear energy system development. (author)

  15. CarbAl Heat Transfer Material

    Science.gov (United States)

    Fink, Richard

    2015-01-01

    The increasing use of power electronics, such as high-current semiconductor devices and modules, within space vehicles is driving the need to develop specialty thermal management materials in both the packaging of these discrete devices and the packaging of modules consisting of these device arrays. Developed by Applied Nanotech, Inc. (ANI), CarbAl heat transfer material is uniquely characterized by its low density, high thermal diffusivity, and high thermal conductivity. Its coefficient of thermal expansion (CTE) is similar to most power electronic materials, making it an effective base plate substrate for state-of-the-art silicon carbide (SiC) super junction transistors. The material currently is being used to optimize hybrid vehicle inverter packaging. Adapting CarbAl-based substrates to space applications was a major focus of the SBIR project work. In Phase I, ANI completed modeling and experimentation to validate its deployment in a space environment. Key parameters related to cryogenic temperature scaling of CTE, thermal conductivity, and mechanical strength. In Phase II, the company concentrated on improving heat sinks and thermally conductive circuit boards for power electronic applications.

  16. Isotopic fissile assay of spent fuel in a lead slowing-down spectrometer system

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Yong Deok; Jeon, Ju Young [Dept. of Fuel Cycle Technology, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Park, Chang Je [Dept. of Nuclear Engineering, Sejong University, Seoul (Korea, Republic of)

    2017-04-15

    A lead slowing-down spectrometer (LSDS) system is under development to analyze isotopic fissile content that is applicable to spent fuel and recycled material. The source neutron mechanism for efficient and effective generation was also determined. The source neutron interacts with a lead medium and produces continuous neutron energy, and this energy generates dominant fission at each fissile, below the unresolved resonance region. From the relationship between the induced fissile fission and the fast fission neutron detection, a mathematical assay model for an isotopic fissile material was set up. The assay model can be expanded for all fissile materials. The correction factor for self-shielding was defined in the fuel assay area. The corrected fission signature provides well-defined fission properties with an increase in the fissile content. The assay procedure was also established. The assay energy range is very important to take into account the prominent fission structure of each fissile material. Fission detection occurred according to the change of the Pu239 weight percent (wt%), but the content of U235 and Pu241 was fixed at 1 wt%. The assay result was obtained with 2∼3% uncertainty for Pu239, depending on the amount of Pu239 in the fuel. The results show that LSDS is a very powerful technique to assay the isotopic fissile content in spent fuel and recycled materials for the reuse of fissile materials. Additionally, a LSDS is applicable during the optimum design of spent fuel storage facilities and their management. The isotopic fissile content assay will increase the transparency and credibility of spent fuel storage.

  17. Fissile mass estimation by pulsed neutron source interrogation

    Energy Technology Data Exchange (ETDEWEB)

    Israelashvili, I., E-mail: israelashvili@gmail.com [Nuclear Research Center of the Negev, P.O.B 9001, Beer Sheva 84190 (Israel); Dubi, C.; Ettedgui, H.; Ocherashvili, A. [Nuclear Research Center of the Negev, P.O.B 9001, Beer Sheva 84190 (Israel); Pedersen, B. [Nuclear Security Unit, Institute for Transuranium Elements, Joint Research Centre, Via E. Fermi, 2749, 21027 Ispra (Italy); Beck, A. [Nuclear Research Center of the Negev, P.O.B 9001, Beer Sheva 84190 (Israel); Roesgen, E.; Crochmore, J.M. [Nuclear Security Unit, Institute for Transuranium Elements, Joint Research Centre, Via E. Fermi, 2749, 21027 Ispra (Italy); Ridnik, T.; Yaar, I. [Nuclear Research Center of the Negev, P.O.B 9001, Beer Sheva 84190 (Israel)

    2015-06-11

    Passive methods for detecting correlated neutrons from spontaneous fissions (e.g. multiplicity and SVM) are widely used for fissile mass estimations. These methods can be used for fissile materials that emit a significant amount of fission neutrons (like plutonium). Active interrogation, in which fissions are induced in the tested material by an external continuous source or by a pulsed neutron source, has the potential advantages of fast measurement, alongside independence of the spontaneous fissions of the tested fissile material, thus enabling uranium measurement. Until recently, using the multiplicity method, for uranium mass estimation, was possible only for active interrogation made with continues neutron source. Pulsed active neutron interrogation measurements were analyzed with techniques, e.g. differential die away analysis (DDA), which ignore or implicitly include the multiplicity effect (self-induced fission chains). Recently, both, the multiplicity and the SVM techniques, were theoretically extended for analyzing active fissile mass measurements, made by a pulsed neutron source. In this study the SVM technique for pulsed neutron source is experimentally examined, for the first time. The measurements were conducted at the PUNITA facility of the Joint Research Centre in Ispra, Italy. First promising results, of mass estimation by the SVM technique using a pulsed neutron source, are presented.

  18. Administrative Co-ordination of Fissile Material Management and Accounting in the U.K.A.E.A; Coordination Administrative de la Gestion et de la Comptabilite des Matieres Fissiles dans les Etabussements de l'Autorite de l'Energie Atomique du Royaume-Uni; Administrativnaya koordinatsiya kontrolya i ucheta delyashchikhsya materialov v upravlenii po atomnoj ehnergii soedinennogo korolevstva; Coordinacion Administrativa de la Gestion y la Contabilidad de Materiales Fisionables en la Comision de Energia Atomica del Reino Unido

    Energy Technology Data Exchange (ETDEWEB)

    Hood, St. C.C. [United Kingdom Atomic Energy Authority, London (United Kingdom)

    1966-02-15

    The Authority are engaged as suppliers in fissile material production, distribution, recycle and reprocessing. As consumers, the Authority require fissile material for power reactors, a variety of prototypes, MTRs, zero-energy facilities and fuel development projects; and for other experimental and research purposes in laboratory quantities. Executive responsibility for these activities lies with the four Groups through which the Authority discharge these functions. It has been found useful to keep these activities under review in specialized inter-Group Committees, with a common secretariat. These Committees: (a) study all projects all proposals or work involving significant quantities of fissile material (plutonium and enriched uranium, other than natural U or U depleted in {sup 235}U) in the light of expected supplies over a number of years from all sources, including new production, scrap recovery and imports; and all uses including burn-up, losses and exports; (b) recommend the optimum allocation of specific amounts for approved purposes in relation to other calls upon available supplies, and having regard to the economic issues involved; (c) record and progress all approved allocations, and examine the nature, amount and purpose of all existing stockholdings in relation to current policies and objectives; (d) record and study all losses of fissile material during fabrication or other processing and the measures taken to reduce them; (e) assist in developing procedures and incentives to ensure that material is used economically and returned promptly. Each Group has considerable autonomy in its day-to-day use of fissile material. The administrative machinery described above provides a means by which the Authority's scientists, engineers, accountants and administrators concerned with fissile material problems can operate collectively in a common frame of reference with a minimum of paperwork. The paper is illustrated with a simplified flowsheet of the main flows

  19. Material Transfer Agreement (MTA) | Frederick National Laboratory for Cancer Research

    Science.gov (United States)

    Material Transfer Agreements are appropriate for exchange of materials into or out of the Frederick National Laboratory for research or testing purposes, with no collaborative research by parties involving the materials.

  20. Fissile solution dynamics: Student research

    Energy Technology Data Exchange (ETDEWEB)

    Hetrick, D.L.

    1994-09-01

    There are two research projects in criticality safety at the University of Arizona: one in dynamic simulation of hypothetical criticality accidents in fissile solutions, and one in criticality benchmarks using transport theory. We have used the data from nuclear excursions in KEWB, CRAC, and SILENE to help in building models for solution excursions. An equation of state for liquids containing gas bubbles has been developed and coupled to point-reactor dynamics in an attempt to predict fission rate, yield, pressure, and kinetic energy. It appears that radiolytic gas is unimportant until after the first peak, but that it does strongly affect the shape of the subsequent power decrease and also the dynamic pressure.

  1. Historic transfer of forest reproductive material in the Nordic region

    DEFF Research Database (Denmark)

    Myking, Tor; Rusanen, Mari; Steffenrem, Arne

    2016-01-01

    Large-scale transfer of reproductive material is a common phenomenon in forestry and is not only limited to recent history. Here we review the historical transfer of forest reproductive material (FRM) in Fennoscandia, the directions, their drivers, and the reported consequences for adaptation...

  2. What should ''damaged'' mean in air transport of fissile packages

    International Nuclear Information System (INIS)

    Luna, R.E.; Falci, F.P.; Blackman, D.

    1995-01-01

    It is likely that the ongoing process to produce the 1996 version of the IAEA Regulation for the Safe Transport of Radioactive Materials, IAEA Safety Series 6(SS 6) will result in a more stringent package qualification standard for air transport of large quantities of radioactive materials (RAM) than is included in the 1990 version. During the process to define the scope of the new requirements there was extensive discussion of their impact on, and application to, fissile material package qualification criteria. Since fissile materials are shipped in a variety of packagings ranging from exempt to Type B, each packaging of each type must be evaluated for its ability to maintain subcriticality both alone and in arrays and in both damaged and undamaged condition. In the 1990 version of SS 6 ''damaged'' means the condition of a package after it had undergone the ''tests for demonstrating the ability to withstand accident conditions in transport,'' i.e., Type B qualification tests. These tests conditions are typical of severe accidents in surface modes, but are less severe than air mode qualification test environments to be applied to Type C packages. As a result, questions arose about the need for a corresponding change in the 1996 SS 6 to define ''damaged'' to include the Type C test regime for criticality evaluations of fissile packages in air transport

  3. Heat transfer in Rockwool modelling and method of measurement. Modelling radiative heat transfer in fibrous materials

    Energy Technology Data Exchange (ETDEWEB)

    Dyrboel, Susanne

    1998-05-01

    Fibrous materials are some of the most widely used materials for thermal insulation. In this project the focus of interest has been on fibrous materials for building application. Interest in improving the thermal properties of insulation materials is increasing as legislation is being tightened to reduce the overall energy consumption. A knowledge of the individual heat transfer mechanisms - whereby heat is transferred within a particular material is an essential tool to improve continuously the thermal properties of the material. Heat is transferred in fibrous materials by four different transfer mechanisms: conduction through air, conduction through fibres, thermal radiation and convection. In a particular temperature range the conduction through air can be regarded as a constant, and conduction through fibres is an insignificant part of the total heat transfer. Radiation, however, constitutes 25-40% of the total heat transfer in light fibrous materials. In Denmark and a number of other countries convection in fibrous materials is considered as non-existent when calculating heat transmission as well as when designing building structures. Two heat transfer mechanisms have been the focus of the current project: radiation heat transfer and convection. The radiation analysis serves to develop a model that can be used in further work to gain a wider knowledge of the way in which the morphology of the fibrous material, i.e. fibre diameter distribution, fibre orientation distribution etc., influences the radiation heat transfer under different conditions. The convection investigation serves to examine whether considering convection as non-existent is a fair assumption to use in present and future building structures. The assumption applied in practically is that convection makes a notable difference only in very thick insulation, at external temperatures below -20 deg. C, and at very low densities. For lager thickness dimensions the resulting heat transfer through the

  4. Irradiation performance of HTGR recycle fissile fuel

    International Nuclear Information System (INIS)

    Homan, F.J.; Long, E.L. Jr.

    1976-08-01

    The irradiation performance of candidate HTGR recycle fissile fuel under accelerated testing conditions is reviewed. Failure modes for coated-particle fuels are described, and the performance of candidate recycle fissile fuels is discussed in terms of these failure modes. The bases on which UO 2 and (Th,U)O 2 were rejected as candidate recycle fissile fuels are outlined, along with the bases on which the weak-acid resin (WAR)-derived fissile fuel was selected as the reference recycle kernel. Comparisons are made relative to the irradiation behavior of WAR-derived fuels of varying stoichiometry and conclusions are drawn about the optimum stoichiometry and the range of acceptable values. Plans for future testing in support of specification development, confirmation of the results of accelerated testing by real-time experiments, and improvement in fuel performance and reliability are described

  5. Sensing Fissile Materials at Long Range

    Science.gov (United States)

    2016-04-01

    and then with it, accounting for its energy  consumption by  eddy   current  heating and subsequent thermal conduction into the  coil (quench back). The...slightly modify the field in mature‐design  machines , if needed.     The  current  in the cyclotron coils can be high, to provide protection through...internal energy dump for protection, either by using  eddy   current  quench  or by imbedded heaters, allows for low  current  operation.  Low  current  is

  6. Fourier analysis of conductive heat transfer for glazed roofing materials

    Energy Technology Data Exchange (ETDEWEB)

    Roslan, Nurhana Lyana; Bahaman, Nurfaradila; Almanan, Raja Noorliyana Raja; Ismail, Razidah [Faculty of Computer and Mathematical Sciences, Universiti Teknologi MARA, 40450 Shah Alam, Selangor (Malaysia); Zakaria, Nor Zaini [Faculty of Applied Sciences, Universiti Teknologi MARA, 40450 Shah Alam, Selangor (Malaysia)

    2014-07-10

    For low-rise buildings, roof is the most exposed surface to solar radiation. The main mode of heat transfer from outdoor via the roof is conduction. The rate of heat transfer and the thermal impact is dependent on the thermophysical properties of roofing materials. Thus, it is important to analyze the heat distribution for the various types of roofing materials. The objectives of this paper are to obtain the Fourier series for the conductive heat transfer for two types of glazed roofing materials, namely polycarbonate and polyfilled, and also to determine the relationship between the ambient temperature and the conductive heat transfer for these materials. Ambient and surface temperature data were collected from an empirical field investigation in the campus of Universiti Teknologi MARA Shah Alam. The roofing materials were installed on free-standing structures in natural ventilation. Since the temperature data are generally periodic, Fourier series and numerical harmonic analysis are applied. Based on the 24-point harmonic analysis, the eleventh order harmonics is found to generate an adequate Fourier series expansion for both glazed roofing materials. In addition, there exists a linear relationship between the ambient temperature and the conductive heat transfer for both glazed roofing materials. Based on the gradient of the graphs, lower heat transfer is indicated through polyfilled. Thus polyfilled would have a lower thermal impact compared to polycarbonate.

  7. Theory of light transfer in food and biological materials

    Science.gov (United States)

    In this chapter, we first define the basic radiometric quantities that are needed for describing light propagation in food and biological materials. Radiative transfer theory is then derived, according to the principle of the conservation of energy. Because the radiative transfer theory equation is ...

  8. Issues related to the inter-utility transfer of material

    International Nuclear Information System (INIS)

    1993-08-01

    An option that utilities have for obtaining material is to procure the desired item(s) from another utility. There are several reasons utilities choose another utility as the procurement source including item obsolescence, prohibitive cost on the commercial market, and excessive lead time. This document provides information on the technical, quality, and commercial issues which utilities may need to address when selling material to or procuring material from other utilities. This report provides suggested approaches for each of the following technical and quality issues: Design considerations; item acceptability considerations; original supplier considerations; commercial grade item dedication considerations; reportability considerations; packaging, shipping, and storage considerations; documentation considerations; receipt inspection considerations. The information is provided primarily for the inter-utility transfer of safety-related material. Several of the topics, however, may also apply to the transfer of non-safety-related material. The report also provides considerations on commercial issues which may be addressed during the inter-utility transfer of materials

  9. Research Tools and Materials | NCI Technology Transfer Center | TTC

    Science.gov (United States)

    Research Tools can be found in TTC's Available Technologies and in scientific publications. They are freely available to non-profits and universities through a Material Transfer Agreement (or other appropriate mechanism), and available via licensing to companies.

  10. Hardware implementation of the ORNL fissile mass flow monitor

    International Nuclear Information System (INIS)

    McEvers, J.; Sumner, J.; Jones, R.; Ferrell, R.; Martin, C.; Uckan, T.; March-Leuba, J.

    1998-01-01

    This paper provides an overall description of the implementation of the Oak Ridge National Laboratory (ORNL) Fissile Mass Flow Monitor, which is part of a Blend Down Monitoring System (BDMS) developed by the US Department of Energy (DOE). The Fissile Mass Flow Monitor is designed to measure the mass flow of fissile material through a gaseous or liquid process stream. It consists of a source-modulator assembly, a detector assembly, and a cabinet that houses all control, data acquisition, and supporting electronics equipment. The development of this flow monitor was first funded by DOE/NE in September 95, and an initial demonstration by ORNL was described in previous INMM meetings. This methodology was chosen by DOE/NE for implementation in November 1996, and the hardware/software development is complete. Successful BDMS installation and operation of the complete BDMS has been demonstrated in the Paducah Gaseous Diffusion Plant (PGDP), which is operated by Lockheed Martin Utility Services, Inc. for the US Enrichment Corporation and regulated by the Nuclear Regulatory Commission. Equipment for two BDMS units has been shipped to the Russian Federation

  11. Radiative heat transfer in 2D Dirac materials

    International Nuclear Information System (INIS)

    Rodriguez-López, Pablo; Tse, Wang-Kong; Dalvit, Diego A R

    2015-01-01

    We compute the radiative heat transfer between two sheets of 2D Dirac materials, including topological Chern insulators and graphene, within the framework of the local approximation for the optical response of these materials. In this approximation, which neglects spatial dispersion, we derive both numerically and analytically the short-distance asymptotic of the near-field heat transfer in these systems, and show that it scales as the inverse of the distance between the two sheets. Finally, we discuss the limitations to the validity of this scaling law imposed by spatial dispersion in 2D Dirac materials. (paper)

  12. Internal transfers of special nuclear material - March 1975

    International Nuclear Information System (INIS)

    Anon.

    1976-01-01

    Paragraph 70.51(e) of 10 CFR Part 70 requires, with certain exceptions stated in the rule, that each licensee authorized to possess more than one effective kilogram of special nuclear material (SNM) maintain certain procedures. These procedures are to include: (1) records of the quantities of SNM added to or removed from the process; (2) documentation of all transfers of SNM between material-balance areas to show the identity and quantity of SNM transferred; (3) requirements for authorized signatures on each document used to record the transfer of SNM between material-balance areas; and (4) means for control of and accounting for internal transfer documents. Paragraph 70.58(e) requires licensees to establish, maintain, and follow a system for measuring the SNM transferred between material-balance areas and item-control areas. Paragraph 70.58(f) requires that licensees have a program that evaluates and controls the quality of their measurement system. Additionally, all licensees authorized to possess SNM must comply with paragraph 70.51(b) of 10 CFR Part 70. That rule requires licensees to keep records showing, among other things, the inventory of all SNM in their possession and its location. This guide sets forth acceptable methods for controlling and documenting transfers of SNM within a plant site in order to meet the requirements listed above

  13. Effect of phase change material on the heat transfer rate of different building materials

    Science.gov (United States)

    Hasan, Mushfiq; Alam, Shahnur; Ahmed, Dewan Hasan

    2017-12-01

    Phase change material (PCM) is widely known as latent heat storage. A comprehensive study is carried out to investigate the effect of PCM on heat transfer rate of building materials. Paraffin is used as PCM along with different conventional building materials to investigate the heat transfer rate from the heated region to the cold region. PCM is placed along with the three different types of building materials like plaster which is well know building material in urban areas and wood and straw which are commonly used in rural areas for roofing as well as wall panel material and investigated the heat transfer rate. An experimental setup was constructed with number of rectangular shape aluminum detachable casing (as cavity) and placed side by side. Series of rectangular cavity filled with convent ional building materials and PCM and these were placed in between two chambers filled with water at different temperature. Building materials and PCM were placed in different cavities with different combinations and investigated the heat transfer rate. The results show that using the PCM along with other building materials can be used to maintain lower temperature at the inner wall and chamber of the cold region. Moreover, the placement or orientation of the building materials and PCM make significant contribution to heat transfer rate from the heated zone to the cold zone.

  14. 36 CFR 1275.48 - Transfer of materials.

    Science.gov (United States)

    2010-07-01

    ... be neither related to abuses of governmental power nor otherwise of general historical significance... identified as private or personal. (b) Materials determined to be neither related to abuses of governmental power nor otherwise of general historical significance, and transferred pursuant to paragraph (a) of...

  15. Neutronic studies of fissile and fusile breeding blankets

    International Nuclear Information System (INIS)

    Taczanowski, S.

    1984-08-01

    In light of the need of convincing motivation substantiating expensive and inherently applied research (nuclear energy), first a simple comparative study of fissile breeding economics of fusion fission hybrids, spallators and also fast breeder reactors has been carried out. As a result, the necessity of maximization of fissile production (in the first two ones, in fast breeders rather the reprocessing costs should be reduced) has been shown, thus indicating the design strategy (high support ratio) for these systems. In spite of the uncertainty of present projections onto further future and discrepancies in available data even quite conservative assumptions indicate that hybrids and perhaps even earlier - spallators can become economic at realistic uranium price increase and successfully compete against fast breeders. Then on the basis of the concept of the neutron flux shaping aimed at the correlation of the selected cross-sections with the neutron flux, the indications for the maximization of respective reaction rates has been formulated. In turn, these considerations serve as the starting point for the guidelines of breeding blanket nuclear design, which are as follows: 1) The source neutrons must face the multiplying layer (of proper thickness) of possibly low concentration of nuclides attenuating the neutron multiplication (i.e. structure materials, nongaseous coolants). 2) For the most effective trapping of neutrons within the breeding zone (leakage and void streaming reduction) it must contain an efficient moderator (not valid for fissile breeding blankets). 3) All regions of significant slow flux should contain 6 Li in order to reduce parasite neutron captures in there. (orig./HP)

  16. Interregional technology transfer on advanced materials and renewable energy systems

    International Nuclear Information System (INIS)

    Agrianidis, P.; David, C.; Anthymidis, K.; Ekhrawat, M.

    2008-01-01

    Advanced materials are used in most industrial sectors and human activities and all developing and developed countries as well as international organizations eg. United Nations have established work groups, which survey the national and global state and developments in the area of advanced materials trying to establish strategies on that crucial technology sector. These strategies are focused on research and technology activities including education and vocation training, as well as stimulus for the starting up of new industrial applications. To introduce such a concept in Greece and especially in Northern Greece, the Technological Education Institute of Serres has initiated an Interregional technology transfer project in this scientific field. This project includes mod topics of advanced materials technology with emphasison specific industrial applications (renewable energy systems). The project demonstrates the development of a prototype photovoltaic thermal system in terms of a new industrial product. The product development procedure consists of steps such as initial product design, materials selection and processing, prototype design and manufacturing, quality control, performance optimization, but also control of materials ecocompatibility according to the national trends of life cycle design and recycling techniques. Keywords: Interregional technology transfer, materials, renewable energy systems

  17. Interregional technology transfer on advanced materials and renewable energy systems

    Energy Technology Data Exchange (ETDEWEB)

    Agrianidis, P.; David, C.; Anthymidis, K.; Ekhrawat, M. [Department of Mechanical Engineering, Technological Educational Institute of Serres, Serres (Greece)

    2008-07-01

    Advanced materials are used in most industrial sectors and human activities and all developing and developed countries as well as international organizations eg. United Nations have established work groups, which survey the national and global state and developments in the area of advanced materials trying to establish strategies on that crucial technology sector. These strategies are focused on research and technology activities including education and vocation training, as well as stimulus for the starting up of new industrial applications. To introduce such a concept in Greece and especially in Northern Greece, the Technological Education Institute of Serres has initiated an Interregional technology transfer project in this scientific field. This project includes mod topics of advanced materials technology with emphasison specific industrial applications (renewable energy systems). The project demonstrates the development of a prototype photovoltaic thermal system in terms of a new industrial product. The product development procedure consists of steps such as initial product design, materials selection and processing, prototype design and manufacturing, quality control, performance optimization, but also control of materials ecocompatibility according to the national trends of life cycle design and recycling techniques. Keywords: Interregional technology transfer, materials, renewable energy systems.

  18. Spectrum analysis in lead spectrometer for isotopic fissile assay in used fuel

    International Nuclear Information System (INIS)

    Lee, Y.D.; Park, C.J.; Kim, H.D.; Song, K.C.

    2014-01-01

    The LSDS system is under development for analyzing isotopic fissile content applicable in a hot cell for the pyro process. The fuel assay area and nuclear material composition were selected for simulation. The source mechanism for efficient neutron generation was also determined. A neutron is produced at the Ta target by hitting it from accelerated electron. The parameters for an electron accelerator are being researched for cost effectiveness, easy maintenance, and compact size. The basic principle of LSDS is that isotopic fissile has its own fission structure below the unresolved resonance region. The source neutron interacts with a lead medium and produces continuous neutron energy, which generates dominant fission at each fissile. Therefore, a spectrum analysis is very important at a lead medium and fuel area for system working. The energy spectrum with respect to slowing down energy and the energy resolution were investigated in lead. A spectrum analysis was done by the existence of surrounding detectors. In particular, high resonance energy was considered. The spectrum was well organized at each slowing down energy and the energy resolution was acceptable to distinguish isotopic fissile fissions. Additionally, LSDS is applicable for the optimum design of spent fuel storage and management.The isotopic fissile content assay will increase the transparency and credibility for spent fuel storage and its re-utilization, as demanded internationally. (author)

  19. Development and production of Zenith fissile elements

    Energy Technology Data Exchange (ETDEWEB)

    George, D; Wheatley, C C.H.; Lloyd, H

    1959-06-15

    The development of a new glass-bonded alumina-uranium oxide composition forming the fissile component of the Zenith fuel elements is described, together with the production of the initial charge containing 15 Kg. of U{sub 235]; the composition is capable of retaining fission product gases at high temperatures. The description includes criticality considerations, details of manufacture, and production statistics of the 11,000 discs produced.

  20. Local tissue distribution of fissile nuclides

    International Nuclear Information System (INIS)

    Smith, J.M.

    1981-01-01

    Conventional tissue-section autoradiography of alpha-emitting actinide elements may require prohibitively long exposure times. Neutron-induced or fission-track autoradiography can be used for fissile nuclides such as 233 U, 235 U, and 239 Pu to circumvent this difficulty. The detection limit for these nuclides is about 4 x 10 -13 (weight fraction). This paper describes a specific technique for determining their microdistribution with histologically stained tissue sections

  1. Material transfer mechanisms between aluminum and fluorinated carbon interfaces

    Energy Technology Data Exchange (ETDEWEB)

    Sen, F.G. [NSERC/General Motors of Canada Industrial Research Chair, Department of Mechanical, Automotive and Materials Engineering, University of Windsor, 401 Sunset Avenue, Windsor, Ontario, N9B 3P4 (Canada); Qi, Y. [Chemical Sciences and Materials Systems Laboratory, General Motors R and D Center, 30500 Mound Road, Warren, MI 48090-9055 (United States); Alpas, A.T., E-mail: aalpas@uwindsor.ca [NSERC/General Motors of Canada Industrial Research Chair, Department of Mechanical, Automotive and Materials Engineering, University of Windsor, 401 Sunset Avenue, Windsor, Ontario, N9B 3P4 (Canada)

    2011-04-15

    First-principles calculations and sliding contact experiments were conducted to elucidate material transfer mechanisms between aluminum and fluorinated carbon (diamond, diamond-like carbon (DLC)) surfaces. An interface model that examined interactions between Al (1 1 1) and F-terminated diamond (1 1 1) surfaces revealed that F atoms would transfer to the Al surface in increasing quantities with an increase in the contact pressure, and this F transfer would lead to the formation of a stable AlF{sub 3} compound at the Al surface. The presence of AlF{sub 3} on the transfer layers formed at the Al counterface placed in sliding contact against DLC containing 3 at.% F was confirmed by both X-ray photoelectron spectroscopy and cross-sectional focussed-ion beam transmission electron microscopy analyses. The coefficient of friction (COF) of the DLC coating was high initially due to deformation and wear of Al counterface, but formation of -OH and -H passivated C-rich transfer layers on Al reduced the COF to a low steady-state value of 0.20. The repulsive forces generated between the two F-passivated surfaces further decreased the COF to 0.14.

  2. Material transfer mechanisms between aluminum and fluorinated carbon interfaces

    International Nuclear Information System (INIS)

    Sen, F.G.; Qi, Y.; Alpas, A.T.

    2011-01-01

    First-principles calculations and sliding contact experiments were conducted to elucidate material transfer mechanisms between aluminum and fluorinated carbon (diamond, diamond-like carbon (DLC)) surfaces. An interface model that examined interactions between Al (1 1 1) and F-terminated diamond (1 1 1) surfaces revealed that F atoms would transfer to the Al surface in increasing quantities with an increase in the contact pressure, and this F transfer would lead to the formation of a stable AlF 3 compound at the Al surface. The presence of AlF 3 on the transfer layers formed at the Al counterface placed in sliding contact against DLC containing 3 at.% F was confirmed by both X-ray photoelectron spectroscopy and cross-sectional focussed-ion beam transmission electron microscopy analyses. The coefficient of friction (COF) of the DLC coating was high initially due to deformation and wear of Al counterface, but formation of -OH and -H passivated C-rich transfer layers on Al reduced the COF to a low steady-state value of 0.20. The repulsive forces generated between the two F-passivated surfaces further decreased the COF to 0.14.

  3. Corrosion and material transfer in a sodium loop

    International Nuclear Information System (INIS)

    Garcia, A.M.; Espigares, M.M.; Arroyo, J.; Borgstedt, H.U.; Kernforschungszentrum Karlsruhe G.m.b.H.

    1984-01-01

    The corrosion and material transfer behaviour of the martensitic steel X18 CrMoVNb 12 1 as a function of the temperature and the position is studied in the ML-1 sodium loop. Up to 600 C the material has the same good compatibility with liquid sodium as austenitic stainless steels, as well in the corrosion region of the loop as in the deposition zone in the cooled leg. The steel is not sensitive to carburization or decarburization under the conditions in the sodium rig. (author)

  4. Material Selection for Microchannel Heatsink: Conjugate Heat Transfer Simulation

    Science.gov (United States)

    Uday Kumar, A.; Javed, Arshad; Dubey, Satish K.

    2018-04-01

    Heat dissipation during the operation of electronic devices causes rise in temperature, which demands an effective thermal management for their performance, life and reliability. Single phase liquid cooling in microchannels is an effective and proven technology for electronics cooling. However, due to the ongoing trends of miniaturization and developments in the microelectronics technology, the future needs of heat flux dissipation rate are expected to rise to 1 kW/cm2. Air cooled systems are unable to meet this demand. Hence, liquid cooled heatsinks are preferred. This paper presents conjugate heat transfer simulation of single phase flow in microchannels with application to electronic cooling. The numerical model is simulated for different materials: copper, aluminium and silicon as solid and water as liquid coolant. The performances of microchannel heatsink are analysed for mass flow rate range of 20-40 ml/min. The investigation has been carried out on same size of electronic chip and heat flux in order to have comparative study of different materials. This paper is divided into two sections: fabrication techniques and numerical simulation for different materials. In the first part, a brief discussion of fabrication techniques of microchannel heatsink have been presented. The second section presents conjugate heat transfer simulation and parametric investigation for different material microchannel heatsink. The presented study and findings are useful for selection of materials for microchannel heatsink.

  5. Underground processing method for radiation-contaminated material and transferring method for buffer molding material

    International Nuclear Information System (INIS)

    Akasaka, Hidenari; Shimura, Satoshi; Asano, Eiichi; Yamagata, Junji; Ninomiya, Nobuo; Kawakami, Susumu.

    1995-01-01

    A bottomed molding material (buffer molding material) is formed into a bottomed cylindrical shape by solidifying, under pressure, powders such as of bentonite into a highly dense state by a cold isotropic pressing or the like, having a hole for accepting and containing a vessel for radiation-contaminated materials. The bottomed cylindrical molding material is loaded on a transferring vessel, and transferred to a position near the site for underground disposal. The bottomed cylindrical molding material having a upwarded containing hole is buried in the cave for disposal. The container for radiation-contaminated material is loaded and contained in the containing hole of the bottomed cylindrical molding material. A next container for radiation-contaminated materials is juxtaposed thereover. Then, a bottomed cylindrical molding material having a downwarded containing hole is covered to the container for the radiation-contaminated material in a state being protruded upwardly. The radiation-contaminated material is thus closed by a buffer material of the same material at the circumference thereof. (I.N.)

  6. Determination of fissile fraction in MOX (mixed U + Pu oxides) fuels for different burnup values

    International Nuclear Information System (INIS)

    Ozdemir, Levent; Acar, Banu Bulut; Zabunoglu, Okan H.

    2011-01-01

    When spent Light Water Reactor fuels are processed by the standard Purex method of reprocessing, plutonium (Pu) and uranium (U) in spent fuel are obtained as pure and separate streams. The recovered Pu has a fissile content (consisting of 239 Pu and 241 Pu) greater than 60% typically (although it mainly depends on discharge burnup of spent fuel). The recovered Pu can be recycled as mixed-oxide (MOX) fuel after being blended with a fertile U makeup in a MOX fabrication plant. The burnup that can be obtained from MOX fuel depends on: (1) isotopic composition of Pu, which is closely related to the discharge burnup of spent fuel from which Pu is recovered; (2) the type of fertile U makeup material used (depleted U, natural U, or recovered U); and (3) fraction of makeup material in the mix (blending ratio), which in turn determines the total fissile fraction of MOX. Using the Non-linear Reactivity Model and the code MONTEBURNS, a step-by-step procedure for computing the total fissile content of MOX is introduced. As was intended, the resulting expression is simple enough for quick/hand calculations of total fissile content of MOX required to reach a desired burnup for a given discharge burnup of spent fuel and for a specified fertile U makeup. In any case, due to non-fissile (parasitic) content of recovered Pu, a greater fissile fraction in MOX than that in fresh U is required to obtain the same burnup as can be obtained by the fresh U fuel.

  7. Heat transfer characteristics of building walls using phase change material

    Science.gov (United States)

    Irsyad, M.; Pasek, A. D.; Indartono, Y. S.; Pratomo, A. W.

    2017-03-01

    Minimizing energy consumption in air conditioning system can be done with reducing the cooling load in a room. Heat from solar radiation which passes through the wall increases the cooling load. Utilization of phase change material on walls is expected to decrease the heat rate by storing energy when the phase change process takes place. The stored energy is released when the ambient temperature is low. Temperature differences at noon and evening can be utilized as discharging and charging cycles. This study examines the characteristics of heat transfer in walls using phase change material (PCM) in the form of encapsulation and using the sleeve as well. Heat transfer of bricks containing encapsulated PCM, tested the storage and released the heat on the walls of the building models were evaluated in this study. Experiments of heat transfer on brick consist of time that is needed for heat transfer and thermal conductivity test as well. Experiments were conducted on a wall coated by PCM which was exposed on a day and night cycle to analyze the heat storage and heat release. PCM used in these experiments was coconut oil. The measured parameter is the temperature at some points in the brick, walls and ambient temperature as well. The results showed that the use of encapsulation on an empty brick can increase the time for thermal heat transfer. Thermal conductivity values of a brick containing encapsulated PCM was lower than hollow bricks, where each value was 1.3 W/m.K and 1.6 W/m.K. While the process of heat absorption takes place from 7:00 am to 06:00 pm, and the release of heat runs from 10:00 pm to 7:00 am. The use of this PCM layer can reduce the surface temperature of the walls of an average of 2°C and slows the heat into the room.

  8. Fuel conditioning facility material accountancy

    International Nuclear Information System (INIS)

    Yacout, A.M.; Bucher, R.G.; Orechwa, Y.

    1995-01-01

    The operation of the Fuel conditioning Facility (FCF) is based on the electrometallurgical processing of spent metallic reactor fuel. It differs significantly, therefore, from traditional PUREX process facilities in both processing technology and safeguards implications. For example, the fissile material is processed in FCF only in batches and is transferred within the facility only as solid, well-characterized items; there are no liquid steams containing fissile material within the facility, nor entering or leaving the facility. The analysis of a single batch lends itself also to an analytical relationship between the safeguards criteria, such as alarm limit, detection probability, and maximum significant amount of fissile material, and the accounting system's performance, as it is reflected in the variance associated with the estimate of the inventory difference. This relation, together with the sensitivity of the inventory difference to the uncertainties in the measurements, allows a thorough evaluation of the power of the accounting system. The system for the accountancy of the fissile material in the FCF has two main components: a system to gather and store information during the operation of the facility, and a system to interpret this information with regard to meeting safeguards criteria. These are described and the precision of the inventory closure over one batch evaluated

  9. Computational simulation of heat transfer in laser melted material flow

    International Nuclear Information System (INIS)

    Shankar, V.; Gnanamuthu, D.

    1986-01-01

    A computational procedure has been developed to study the heat transfer process in laser-melted material flow associated with surface heat treatment of metallic alloys to improve wear-and-tear and corrosion resistance. The time-dependent incompressible Navier-Stokes equations are solved, accounting for both convective and conductive heat transfer processes. The convection, induced by surface tension and high surface temperature gradients, sets up a counterrotating vortex flow within the molten pool. This recirculating material flow is responsible for determining the molten pool shape and the associated cooling rates which affect the solidifying material composition. The numerical method involves an implicit triple-approximate factorization scheme for the energy equation, and an explicit treatment for the momentum and the continuity equations. An experimental setup, using a continuous wave CO 2 laser beam as a heat source, has been carried out to generate data for validation of the computational model. Results in terms of the depth, width, and shape of the molten pool and the heat-affected zone for various power settings and shapes of the laser, and for various travel speeds of the workpiece, compare very well with experimental data. The presence of the surface tension-induced vortex flow is demonstrated

  10. Conduit for high temperature transfer of molten semiconductor crystalline material

    Science.gov (United States)

    Fiegl, George (Inventor); Torbet, Walter (Inventor)

    1983-01-01

    A conduit for high temperature transfer of molten semiconductor crystalline material consists of a composite structure incorporating a quartz transfer tube as the innermost member, with an outer thermally insulating layer designed to serve the dual purposes of minimizing heat losses from the quartz tube and maintaining mechanical strength and rigidity of the conduit at the elevated temperatures encountered. The composite structure ensures that the molten semiconductor material only comes in contact with a material (quartz) with which it is compatible, while the outer layer structure reinforces the quartz tube, which becomes somewhat soft at molten semiconductor temperatures. To further aid in preventing cooling of the molten semiconductor, a distributed, electric resistance heater is in contact with the surface of the quartz tube over most of its length. The quartz tube has short end portions which extend through the surface of the semiconductor melt and which are lef bare of the thermal insulation. The heater is designed to provide an increased heat input per unit area in the region adjacent these end portions.

  11. Fusion-Fission Hybrid for Fissile Fuel Production without Processing

    Energy Technology Data Exchange (ETDEWEB)

    Fratoni, M; Moir, R W; Kramer, K J; Latkowski, J F; Meier, W R; Powers, J J

    2012-01-02

    Two scenarios are typically envisioned for thorium fuel cycles: 'open' cycles based on irradiation of {sup 232}Th and fission of {sup 233}U in situ without reprocessing or 'closed' cycles based on irradiation of {sup 232}Th followed by reprocessing, and recycling of {sup 233}U either in situ or in critical fission reactors. This study evaluates a third option based on the possibility of breeding fissile material in a fusion-fission hybrid reactor and burning the same fuel in a critical reactor without any reprocessing or reconditioning. This fuel cycle requires the hybrid and the critical reactor to use the same fuel form. TRISO particles embedded in carbon pebbles were selected as the preferred form of fuel and an inertial laser fusion system featuring a subcritical blanket was combined with critical pebble bed reactors, either gas-cooled or liquid-salt-cooled. The hybrid reactor was modeled based on the earlier, hybrid version of the LLNL Laser Inertial Fusion Energy (LIFE1) system, whereas the critical reactors were modeled according to the Pebble Bed Modular Reactor (PBMR) and the Pebble Bed Advanced High Temperature Reactor (PB-AHTR) design. An extensive neutronic analysis was carried out for both the hybrid and the fission reactors in order to track the fuel composition at each stage of the fuel cycle and ultimately determine the plant support ratio, which has been defined as the ratio between the thermal power generated in fission reactors and the fusion power required to breed the fissile fuel burnt in these fission reactors. It was found that the maximum attainable plant support ratio for a thorium fuel cycle that employs neither enrichment nor reprocessing is about 2. This requires tuning the neutron energy towards high energy for breeding and towards thermal energy for burning. A high fuel loading in the pebbles allows a faster spectrum in the hybrid blanket; mixing dummy carbon pebbles with fuel pebbles enables a softer spectrum in

  12. Modeling of heat transfer within porous multi-constituent materials

    International Nuclear Information System (INIS)

    Niezgoda, M.

    2012-01-01

    The CEA works a great deal with porous materials - carbon composites, ceramics - and aims to optimize their properties for specific uses. These materials can be composed of several constituents and generally has a complex structure with pore size of several tens of micrometers. It is used in large-scale systems that are bigger than its own characteristic scale in which they are considered as equivalent to a homogeneous medium for the simulation of its behavior in its using environment without taking into account its local morphology. We are especially interested in the effective thermal diffusivity of heterogeneous materials that we estimate as a function of temperature with the help of an inverse method by considering they are homogeneous. The identification of the diffusivity of porous and/or semi-transparent materials is made difficult because of the strong conducto-radiative coupling can quickly occur when the temperature increases. We have thus modeled the coupled conductive and radiative heat transfer as a function of the temperature within porous multi-constituent materials from their morphology discretized into a set of homogeneous voxels. We have developed a methodology that consists in starting from a 3D-microstructure of the studied materials obtained by tomography. The microstructures constitute the numerical support to this modeling that renders it possible, on the one hand, to simulate any kind of numerical thermal experiments, especially the flash method whose the results render it possible to estimate the thermal diffusivity, and on the other hand, to reproduce the thermal behavior of our materials in their using conditions. (author) [fr

  13. Counterstreaming-ion-tokamak fissile breeder

    International Nuclear Information System (INIS)

    Jassby, D.L.; Lee, J.D.

    1976-08-01

    Tokamak plasmas fueled and heated by energetic neutral-atom beams are characterized by total ion energy greatly exceeding the electron energy. For smaller devices the largest fusion reactivity of energetic-ion plasmas is obtained when oppositely injected D 0 and T 0 beams sustain counterstreaming velocity distributions of deuterons and tritons. This scoping study investigates the net fissile and power productions of a tokamak fusion-fission reactor with a counterstreaming-ion fusion driver and a fertile blanket optimized for fissile breeding. The fusion driver has parameters R/sub o/ = 4.7 m, a = 1.0 m, B/sub t/ = 5.6 T, W/sub b/ = 100 keV (D 0 ), n tau/sub E/ = 1.4 x 10 13 cm -3 s, Q = 1.5, 14-MeV neutron production = 175 MW. The blanket contains a fast-fission zone of natural U plus Mo (7 percent), followed by a Li-bearing zone for T breeding. The reactor produces a net power of 480 MWe and supplies sufficient Pu to support a system of LWR's producing 3800 MWe, with an estimated electrical energy cost for the entire system of 27 mills/kWh

  14. Heat transfer in Rockwool modelling and method of measurement. The effect of natural convection on heat transfer in fibrous materials

    Energy Technology Data Exchange (ETDEWEB)

    Dyrboel, Susanne

    1998-05-01

    Fibrous materials are some of the most widely used materials for thermal insulation. In this project the focus of interest has been on fibrous materials for building application. Interest in improving the thermal properties of insulation materials is increasing as legislation is being tightened to reduce the overall energy consumption. A knowledge of the individual heat transfer mechanisms - whereby heat is transferred within a particular material is an essential tool to improve continuously the thermal properties of the material. Heat is transferred in fibrous materials by four different transfer mechanisms: conduction through air, conduction through fibres, thermal radiation and convection. In a particular temperature range the conduction through air can be regarded as a constant, and conduction through fibres is an insignificant part of the total heat transfer. Radiation, however, constitutes 25-40% of the total heat transfer in light fibrous materials. In Denmark and a number of other countries convection in fibrous materials is considered as non-existent when calculating heat transmission as well as when designing building structures. Two heat transfer mechanisms have been the focus of the current project: radiation heat transfer and convection. The radiation analysis serves to develop a model that can be used in further work to gain a wider knowledge of the way in which the morphology of the fibrous material, i.e. fibre diameter distribution, fibre orientation distribution etc., influences the radiation heat transfer under different conditions. The convection investigation serves to examine whether considering convection as non-existent is a fair assumption to use in present and future building structures. The assumption applied in practically is that convection makes a notable difference only in very thick insulation, at external temperatures below -20 deg. C, and at very low densities. For large thickness dimensions the resulting heat transfer through the

  15. Vibration damping and heat transfer using material phase changes

    Science.gov (United States)

    Kloucek, Petr (Inventor); Reynolds, Daniel R. (Inventor)

    2009-01-01

    A method and apparatus wherein phase changes in a material can dampen vibrational energy, dampen noise and facilitate heat transfer. One embodiment includes a method for damping vibrational energy in a body. The method comprises attaching a material to the body, wherein the material comprises a substrate, a shape memory alloy layer, and a plurality of temperature change elements. The method further comprises sensing vibrations in the body. In addition, the method comprises indicating to at least a portion of the temperature change elements to provide a temperature change in the shape memory alloy layer, wherein the temperature change is sufficient to provide a phase change in at least a portion of the shape memory alloy layer, and further wherein the phase change consumes a sufficient amount of kinetic energy to dampen at least a portion of the vibrational energy in the body. In other embodiments, the shape memory alloy layer is a thin film. Additional embodiments include a sensor connected to the material.

  16. Vibration damping and heat transfer using material phase changes

    Science.gov (United States)

    Kloucek, Petr [Houston, TX; Reynolds, Daniel R [Oakland, CA

    2009-03-24

    A method and apparatus wherein phase changes in a material can dampen vibrational energy, dampen noise and facilitate heat transfer. One embodiment includes a method for damping vibrational energy in a body. The method comprises attaching a material to the body, wherein the material comprises a substrate, a shape memory alloy layer, and a plurality of temperature change elements. The method further comprises sensing vibrations in the body. In addition, the method comprises indicating to at least a portion of the temperature change elements to provide a temperature change in the shape memory alloy layer, wherein the temperature change is sufficient to provide a phase change in at least a portion of the shape memory alloy layer, and further wherein the phase change consumes a sufficient amount of kinetic energy to dampen at least a portion of the vibrational energy in the body. In other embodiments, the shape memory alloy layer is a thin film. Additional embodiments include a sensor connected to the material.

  17. 49 CFR 173.477 - Approval of packagings containing greater than 0.1 kg of non-fissile or fissile-excepted uranium...

    Science.gov (United States)

    2010-10-01

    ... kg of non-fissile or fissile-excepted uranium hexafluoride. 173.477 Section 173.477 Transportation... non-fissile or fissile-excepted uranium hexafluoride. (a) Each offeror of a package containing more than 0.1 kg of uranium hexafluoride must maintain on file for at least one year after the latest...

  18. Heat transfer between relocated materials and the RPV lower head

    Energy Technology Data Exchange (ETDEWEB)

    Rempe, J.L.; Knudson, D.L. [Idaho National Engineering and Environmental Lab., Idaho Falls, ID (United States); Kohriyama, T. [INSS, Fukui (Japan)

    2001-07-01

    Questions about the coolability of a continuous mass of relocated corium were raised during the Three Mile Island Unit 2 (TMI-2) Vessel Investigation Project (VIP) Post-accident examinations indicate that nearly half of the material that relocated to the vessel lower head during the TMI-2 accident formed a cohesive or ''continuous'' layer. TMI-2 VIP results and other evidence suggest that conduction through this continuous layer of solidified corium materials was assisted by other cooling mechanisms. Because increased knowledge about in-vessel coolability of corium materials may assist reactor designers in demonstrating that their concepts are passively safe, there is international interest in this topic. However, data are needed to identify what cooling mechanism(s) occurred and to develop a validated model for predicting this cooling. Corium cooling models significantly impact predictions for subsequent accident progression, such as the estimated time and mode of vessel failure. Hence, improved cooling models will provide a much needed, missing component of severe accident analyses. This paper provides a critical review of research investigating the coolability of corium relocating to a water-filled lower head. Where possible, existing models and data for predicting cooling are quantitatively compared; and governing relationships are identified. Key phenomena that should be incorporated into models for predicting this heat transfer are discussed, and deficiencies in current models and available data for predicting cooling are noted. Recommendations for improving these models and for obtaining data to validate these models are also provided. (author)

  19. Heat transfer between relocated materials and the RPV lower head

    International Nuclear Information System (INIS)

    Rempe, J.L.; Knudson, D.L.; Kohriyama, T.

    2001-01-01

    Questions about the coolability of a continuous mass of relocated corium were raised during the Three Mile Island Unit 2 (TMI-2) Vessel Investigation Project (VIP) Post-accident examinations indicate that nearly half of the material that relocated to the vessel lower head during the TMI-2 accident formed a cohesive or ''continuous'' layer. TMI-2 VIP results and other evidence suggest that conduction through this continuous layer of solidified corium materials was assisted by other cooling mechanisms. Because increased knowledge about in-vessel coolability of corium materials may assist reactor designers in demonstrating that their concepts are passively safe, there is international interest in this topic. However, data are needed to identify what cooling mechanism(s) occurred and to develop a validated model for predicting this cooling. Corium cooling models significantly impact predictions for subsequent accident progression, such as the estimated time and mode of vessel failure. Hence, improved cooling models will provide a much needed, missing component of severe accident analyses. This paper provides a critical review of research investigating the coolability of corium relocating to a water-filled lower head. Where possible, existing models and data for predicting cooling are quantitatively compared; and governing relationships are identified. Key phenomena that should be incorporated into models for predicting this heat transfer are discussed, and deficiencies in current models and available data for predicting cooling are noted. Recommendations for improving these models and for obtaining data to validate these models are also provided. (author)

  20. Detector and front-end electronics of a fissile mass flow monitoring system

    International Nuclear Information System (INIS)

    Paulus, M.J.; Uckan, T.; Lenarduzzi, R.; Mullens, J.A.; Castleberry, K.N.; McMillan, D.E.; Mihalczo, J.T.

    1997-01-01

    A detector and front-end electronics unit with secure data transmission has been designed and implemented for a fissile mass flow monitoring system for fissile mass flow of gases and liquids in a pipe. The unit consists of 4 bismuth germanate (BGO) scintillation detectors, pulse-shaping and counting electronics, local temperature sensors, and on-board local area network nodes which locally acquire data and report to the master computer via a secure network link. The signal gain of the pulse-shaping circuitry and energy windows of the pulse-counting circuitry are periodicially self calibrated and self adjusted in situ using a characteristic line in the fissile material pulse height spectrum as a reference point to compensate for drift such as in the detector gain due to PM tube aging. The temperature- dependent signal amplitude variations due to the intrinsic temperature coefficients of the PM tube gain and BGO scintillation efficiency have been characterized and real-time gain corrections introduced. The detector and electronics design, measured intrinsic performance of the detectors and electronics, and the performance of the detector and electronics within the fissile mass flow monitoring system are described

  1. Reactor physics ideas to design novel reactors with faster fissile growth

    International Nuclear Information System (INIS)

    Jagannathan, V.; Pal, U.; Karthikeyan, R.; Raj, D.; Srivastava, A.; Khan, S. A.

    2007-01-01

    There are several types of fission reactors operating in the world adopting generally the open fuel cycle which considers the naturally available fissile nuclide, viz., 2 35U. The accumulated discharged fuel is considered as waste in some countries. However the discharged fuel contains the precious man-made fissile plutonium which would provide the sole means of harnessing the nuclear energy from either depleted uranium or the natural thorium in future. It must be emphasized that the present day power reactors use just about 0.5% of the mined uranium and it would be imprudent to discard the rest of the mass as waste. It is therefore necessary to explore ways and means of exploiting the fertile mass which has the potential of providing the energy without the green house effects for millennia to come. This has to be done by innovating means of large scale fertile to fissile conversion and then using the man-made fissile material for sustenance as well as growth of fission nuclear power. This paper attempts to give a broad picture of the available options and the challenges in realizing the theoretical possibilities

  2. PETIs as High-Temperature Resin-Transfer-Molding Materials

    Science.gov (United States)

    Connell, John N.; Smith, Joseph G., Jr.; Hergenrother, Paul M.

    2005-01-01

    Compositions of, and processes for fabricating, high-temperature composite materials from phenylethynyl-terminated imide (PETI) oligomers by resin-transfer molding (RTM) and resin infusion have been developed. Composites having a combination of excellent mechanical properties and long-term high-temperature stability have been readily fabricated. These materials are particularly useful for the fabrication of high-temperature structures for jet-engine components, structural components on highspeed aircraft, spacecraft, and missiles. Phenylethynyl-terminated amide acid oligomers that are precursors of PETI oligomers are easily made through the reaction of a mixture of aromatic diamines with aromatic dianhydrides at high stoichiometric offsets and 4-phenylethynylphthalic anhydride (PEPA) as an end-capper in a polar solvent such as N-methylpyrrolidinone (NMP). These oligomers are subsequently cyclodehydrated -- for example, by heating the solution in the presence of toluene to remove the water by azeotropic distillation to form low-molecular-weight imide oligomers. More precisely, what is obtained is a mixture of PETI oligomeric species, spanning a range of molecular weights, that exhibits a stable melt viscosity of less than approximately 60 poise (and generally less than 10 poise) at a temperature below 300 deg C. After curing of the oligomers at a temperature of 371 deg C, the resulting polymer can have a glass-transition temperature (Tg) as high as 375 C, the exact value depending on the compositions.

  3. Fissile fuel dynamics of breeder/converter reactors

    International Nuclear Information System (INIS)

    Harms, A.A.

    1978-01-01

    The long-term fissile fuel dynamics for a hierarchy of fission reactors covering the range from pure-burners to super-breeders is examined. It is found that the breeding gains of the core and blanket can be used to identify several distinct fissile fuel histories and elucidate the importance of fuel cycle characteristics such as the time dependence of the fissile fuel doubling time. On this basis, a self-sufficient fission reactor is introduced and its determining characteristics are identified. (author)

  4. 10 CFR 73.28 - Security background checks for secure transfer of nuclear materials.

    Science.gov (United States)

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 false Security background checks for secure transfer of nuclear... PLANTS AND MATERIALS Physical Protection of Special Nuclear Material in Transit § 73.28 Security background checks for secure transfer of nuclear materials. Licensees are excepted from the security...

  5. The optimisation of transfer chutes in the bulk materials industry / M.N. van Aarde

    OpenAIRE

    Van Aarde, Michiel Nicolaas

    2009-01-01

    Bulk materials handling is a rapidly growing global industry. Immense challenges exist to improve the efficiency and cost effectiveness of transporting and handling bulk materials continuously. The nature and scale of bulk materials handling varies from country to country. This study specifically focuses on the handling of bulk materials in the mining sector. Within this industry, transfer chutes are a key component used for transferring bulk material from one conveyor to another. Among o...

  6. Storage and disposition of weapons usable fissile materials (FMD) PEIS: Blending of U-233 to <12% or <5% enrichment at the Idaho National Engineering Laboratory. Data report, Draft: Version 1

    International Nuclear Information System (INIS)

    Shaber, E.L.

    1995-08-01

    Uranium-233 (U-233), a uranium isotope, is a fissionable material capable of fueling nuclear reactors or being utilized in the manufacturing of nuclear weapons. As such, it is controlled as a special nuclear material. The Idaho National Engineering Laboratory (INEL) and Oak Ridge National Laboratory (ORNL) currently store the Department of Energy's (DOE's) supply of unirradiated U-233 fuel materials. Irradiated U-233 is covered by the national spent nuclear fuel (SNF) program and is not in the scope of this report. The U-233 stored at ORNL is relatively pure uranium oxide in the form of powder or monolithic solids. This material is currently stored in stainless steel canisters of variable lengths measuring about 3 inches in diameter. The ORNL material enrichment varies with some material containing considerable amounts of U-235. The INEL material is fuel from the Light Water Breeder Reactor (LWBR) Program and consists of enriched uranium and thorium oxides in zircaloy cladding. The DOE inventory of U-233 contains trace quantities of U-232, and daughter products from the decay of U-232 and U-233, resulting in increased radioactivity over time. These increased levels of radioactivity generally result in the need for special handling considerations

  7. Storage and disposition of weapons usable fissile materials (FMD) PEIS: Blending of U-233 to {lt}12% or {lt}5% enrichment at the Idaho National Engineering Laboratory. Data report, Draft: Version 1

    Energy Technology Data Exchange (ETDEWEB)

    Shaber, E.L.

    1995-08-01

    Uranium-233 (U-233), a uranium isotope, is a fissionable material capable of fueling nuclear reactors or being utilized in the manufacturing of nuclear weapons. As such, it is controlled as a special nuclear material. The Idaho National Engineering Laboratory (INEL) and Oak Ridge National Laboratory (ORNL) currently store the Department of Energy`s (DOE`s) supply of unirradiated U-233 fuel materials. Irradiated U-233 is covered by the national spent nuclear fuel (SNF) program and is not in the scope of this report. The U-233 stored at ORNL is relatively pure uranium oxide in the form of powder or monolithic solids. This material is currently stored in stainless steel canisters of variable lengths measuring about 3 inches in diameter. The ORNL material enrichment varies with some material containing considerable amounts of U-235. The INEL material is fuel from the Light Water Breeder Reactor (LWBR) Program and consists of enriched uranium and thorium oxides in zircaloy cladding. The DOE inventory of U-233 contains trace quantities of U-232, and daughter products from the decay of U-232 and U-233, resulting in increased radioactivity over time. These increased levels of radioactivity generally result in the need for special handling considerations.

  8. The differential dieaway technique applied to the measurement of the fissile content of drums of cement encapsulated waste

    International Nuclear Information System (INIS)

    Swinhoe, M.T.

    1986-01-01

    This report describes calculations of the differential dieaway technique as applied to cement encapsulated waste. The main difference from previous applications of the technique are that only one detector position is used (diametrically opposite the neutron source) and the chamber walls are made of concrete. The results show that by rotating the drum the response to fissile material across the central plane of the drum can be made relatively uniform. The absolute size of the response is about 0.4. counts per minute per gram fissile for a neutron source of 10 8 neutrons per second. Problems of neutron and gamma background and water content are considered. (author)

  9. 10 CFR 32.18 - Manufacture, distribution and transfer of exempt quantities of byproduct material: Requirements...

    Science.gov (United States)

    2010-01-01

    ... 10 Energy 1 2010-01-01 2010-01-01 false Manufacture, distribution and transfer of exempt... COMMISSION SPECIFIC DOMESTIC LICENSES TO MANUFACTURE OR TRANSFER CERTAIN ITEMS CONTAINING BYPRODUCT MATERIAL Exempt Concentrations and Items § 32.18 Manufacture, distribution and transfer of exempt quantities of...

  10. Operational Characteristics of an Accelerator Driven Fissile Solution System

    International Nuclear Information System (INIS)

    Kimpland, Robert Herbert

    2016-01-01

    Operational characteristics represent the set of responses that a nuclear system exhibits during normal operation. Operators rely on this behavior to assess the status of the system and to predict the consequences of off-normal events. These characteristics largely refer to the relationship between power and system operating conditions. The static and dynamic behavior of a chain-reacting system, operating at sufficient power, is primarily governed by reactivity effects. The science of reactor physics has identified and evaluated a number of such effects, including Doppler broadening and shifts in the thermal neutron spectrum. Often these reactivity effects are quantified in the form of feedback coefficients that serve as coupling coefficients relating the neutron population and the physical mechanisms that drive reactivity effects, such as fissile material temperature and density changes. The operational characteristics of such nuclear systems usually manifest themselves when perturbations between system power (neutron population) and system operating conditions arise. Successful operation of such systems requires the establishment of steady equilibrium conditions. However, prior to obtaining the desired equilibrium (steady-state) conditions, an approach from zero-power (startup) must occur. This operational regime may possess certain limiting system conditions that must be maintained to achieve effective startup. Once steady-state is achieved, a key characteristic of this operational regime is the level of stability that the system possesses. Finally, a third operational regime, shutdown, may also possess limiting conditions of operation that must be maintained. This report documents the operational characteristics of a ''generic'' Accelerator Driven Fissile Solution (ADFS) system during the various operational regimes of startup, steady-state operation, and shutdown. Typical time-dependent behavior for each operational regime will be illustrated, and key system

  11. Material Balance Assessment for Double-Shell Tank Waste Pipeline Transfer

    International Nuclear Information System (INIS)

    Onishi, Yasuo; Wells, Beric E; Hartley, Stacey A; Enderlin, Carl W

    2001-01-01

    PNNL developed a material balance assessment methodology based on conservation of mass for detecting leaks and mis-routings in pipeline transfer of double-shell tank waste at Hanford. The main factors causing uncertainty in these transfers are variable property and tank conditions of density, existence of crust, and surface disturbance due to mixer pump operation during the waste transfer. The methodology was applied to three waste transfers from Tanks AN-105 and AZ-102

  12. Implementation of the Fissile Mass Flow Monitor Source Verification and Confirmation

    Energy Technology Data Exchange (ETDEWEB)

    Uckan, Taner [ORNL; March-Leuba, Jose A [ORNL; Powell, Danny H [ORNL; Nelson, Dennis [Sandia National Laboratories (SNL); Radev, Radoslav [Lawrence Livermore National Laboratory (LLNL)

    2007-12-01

    This report presents the verification procedure for neutron sources installed in U.S. Department of Energy equipment used to measure fissile material flow. The Fissile Mass Flow Monitor (FMFM) equipment determines the {sup 235}U fissile mass flow of UF{sub 6} gas streams by using {sup 252}Cf neutron sources for fission activation of the UF{sub 6} gas and by measuring the fission products in the flow. The {sup 252}Cf sources in each FMFM are typically replaced every 2 to 3 years due to their relatively short half-life ({approx} 2.65 years). During installation of the new FMFM sources, the source identity and neutronic characteristics provided by the manufacturer are verified with the following equipment: (1) a remote-control video television (RCTV) camera monitoring system is used to confirm the source identity, and (2) a neutron detection system (NDS) is used for source-strength confirmation. Use of the RCTV and NDS permits remote monitoring of the source replacement process and eliminates unnecessary radiation exposure. The RCTV, NDS, and the confirmation process are described in detail in this report.

  13. Development for fissile assay in recycled fuel using lead slowing down spectrometer

    International Nuclear Information System (INIS)

    Lee, Yong Deok; Je Park, C.; Kim, Ho-Dong; Song, Kee Chan

    2013-01-01

    A future nuclear energy system is under development to turn spent fuels produced by PWRs into fuels for a SFR (Sodium Fast Reactor) through the pyrochemical process. The knowledge of the isotopic fissile content of the new fuel is very important for fuel safety. A lead slowing down spectrometer (LSDS) is under development to analyze the fissile material content (Pu 239 , Pu 241 and U 235 ) of the fuel. The LSDS requires a neutron source, the neutrons will be slowed down through their passage in a lead medium and will finally enter the fuel and will induce fission reactions that will be analysed and the isotopic content of the fuel will be then determined. The issue is that the spent fuel emits intense gamma rays and neutrons by spontaneous fission. The threshold fission detector screens the prompt fast fission neutrons and as a result the LSDS is not influenced by the high level radiation background. The energy resolution of LSDS is good in the range 0.1 eV to 1 keV. It is also the range in which the fission reaction is the most discriminating for the considered fissile isotopes. An electron accelerator has been chosen to produce neutrons with an adequate target through (e - ,γ)(γ,n) reactions

  14. Implementation of the Fissile Mass Flow Monitor Source Verification and Confirmation

    International Nuclear Information System (INIS)

    Uckan, Taner; March-Leuba, Jose A.; Powell, Danny H.; Nelson, Dennis; Radev, Radoslav

    2007-01-01

    This report presents the verification procedure for neutron sources installed in U.S. Department of Energy equipment used to measure fissile material flow. The Fissile Mass Flow Monitor (FMFM) equipment determines the 235 U fissile mass flow of UF 6 gas streams by using 252 Cf neutron sources for fission activation of the UF 6 gas and by measuring the fission products in the flow. The 252 Cf sources in each FMFM are typically replaced every 2 to 3 years due to their relatively short half-life (∼ 2.65 years). During installation of the new FMFM sources, the source identity and neutronic characteristics provided by the manufacturer are verified with the following equipment: (1) a remote-control video television (RCTV) camera monitoring system is used to confirm the source identity, and (2) a neutron detection system (NDS) is used for source-strength confirmation. Use of the RCTV and NDS permits remote monitoring of the source replacement process and eliminates unnecessary radiation exposure. The RCTV, NDS, and the confirmation process are described in detail in this report.

  15. Derivation of plutonium-239 materials disposition categories

    International Nuclear Information System (INIS)

    Brough, W.G.

    1995-01-01

    At this time, the Office of Fissile Materials Disposition within the DOE, is assessing alternatives for the disposition of excess fissile materials. To facilitate the assessment, the Plutonium-Bearing Materials Feed Report for the DOE Fissile Materials Disposition Program Alternatives report was written. The development of the material categories and the derivation of the inventory quantities associated with those categories is documented in this report

  16. Fissile Material Disposition Program: Deep Borehole Disposal Facility PEIS data input report for direct disposal. Direct disposal of plutonium metal/plutonium dioxide in compound metal canisters. Version 3.0

    Energy Technology Data Exchange (ETDEWEB)

    Wijesinghe, A.M.; Shaffer, R.J.

    1996-01-15

    The US Department of Energy (DOE) is examining options for disposing of excess weapons-usable nuclear materials [principally plutonium (Pu) and highly enriched uranium (HEU)] in a form or condition that is substantially and inherently more difficult to recover and reuse in weapons production. This report is the data input report for the Programmatic Environmental Impact Statement (PEIS). The PEIS examines the environmental, safety, and health impacts of implementing each disposition alternative on land use, facility operations, and site infrastructure; air quality and noise; water, geology, and soils; biotic, cultural, and paleontological resources; socioeconomics; human health; normal operations and facility accidents; waste management; and transportation. This data report is prepared to assist in estimating the environmental effects associated with the construction and operation of a Deep Borehole Disposal Facility, an alternative currently included in the PEIS. The facility projects under consideration are, not site specific. This report therefore concentrates on environmental, safety, and health impacts at a generic site appropriate for siting a Deep Borehole Disposal Facility.

  17. Fissile Material Disposition Program: Deep Borehole Disposal Facility PEIS data input report for direct disposal. Direct disposal of plutonium metal/plutonium dioxide in compound metal canisters. Version 3.0

    International Nuclear Information System (INIS)

    Wijesinghe, A.M.; Shaffer, R.J.

    1996-01-01

    The US Department of Energy (DOE) is examining options for disposing of excess weapons-usable nuclear materials [principally plutonium (Pu) and highly enriched uranium (HEU)] in a form or condition that is substantially and inherently more difficult to recover and reuse in weapons production. This report is the data input report for the Programmatic Environmental Impact Statement (PEIS). The PEIS examines the environmental, safety, and health impacts of implementing each disposition alternative on land use, facility operations, and site infrastructure; air quality and noise; water, geology, and soils; biotic, cultural, and paleontological resources; socioeconomics; human health; normal operations and facility accidents; waste management; and transportation. This data report is prepared to assist in estimating the environmental effects associated with the construction and operation of a Deep Borehole Disposal Facility, an alternative currently included in the PEIS. The facility projects under consideration are, not site specific. This report therefore concentrates on environmental, safety, and health impacts at a generic site appropriate for siting a Deep Borehole Disposal Facility

  18. Capture and transfer of pions in hydrogenous materials

    International Nuclear Information System (INIS)

    Armstrong, D.S.

    1990-05-01

    Pionic hydrogen is a short-lived exotic hydrogen isotope in which a negative pion replaces the atomic electron. The formation and subsequent interactions of pionic hydrogen are discussed, with emphasis on the process of pion transfer. Recent results using the pion charge-exchange reaction (π - , π 0 ) obtained at TRIUMF are reviewed. (Author) (35 refs., 3 tabs., 9 figs.)

  19. Potential for fissile breeding with the fusion-fission hybrid reactor

    International Nuclear Information System (INIS)

    Bender, D.J.; Lee, J.D.

    1976-01-01

    The general features of the mirror reactor design are discussed. Details of the blanket-coil geometry are shown. The inside face of the blanket segments are divided into individual pressure vessels. These submodules contain fissile breeding material located directly behind the first wall, a fusile breeding material behind the fertile breeder, and then coolant inlet and outlet plena. Two blankets are examined and compared in this study. One contains natural uranium plus 7 wt. percent Mo, the second contains thorium metal. The performance of these blankets is discussed

  20. Development of a fissile particle for HTGR fuel recycle

    International Nuclear Information System (INIS)

    Homan, F.J.; Long, E.L. Jr.; Lindemer, T.B.; Beatty, R.L.; Tiegs, T.N.

    1976-12-01

    Recycle fissile fuel particles for high-temperature gas-cooled reactors (HTGRs) have been under development since the mid-1960s. Irradiation performance on early UO 2 and Th 0 . 8 U 0 . 2 O 2 kernels is described in this report, and the performance limitations associated with the dense oxide kernels are presented. The development of the new reference fuel kernel, the weak-acid-resin-derived (WAR) UO 2 --UC 2 , is discussed in detail, including an extensive section on the irradiation performance of this fuel in HFIR removable beryllium capsules HRB-7 through -10. The conclusion is reached that the irradiation performance of the WAR fissile fuel kernel is better than that of any coated particle fuel yet tested. Further, the present fissile kernel is adequate for steam cycle HTGRs as well as for many advanced applications such as gas turbine and process heat HTGRs

  1. Thick film laser induced forward transfer for deposition of thermally and mechanically sensitive materials

    International Nuclear Information System (INIS)

    Kattamis, Nicholas T.; Purnick, Priscilla E.; Weiss, Ron; Arnold, Craig B.

    2007-01-01

    Laser forward transfer processes incorporating thin absorbing films can be used to deposit robust organic and inorganic materials but the deposition of more delicate materials has remained elusive due to contamination and stress induced during the transfer process. Here, we present the approach to high resolution patterning of sensitive materials by incorporating a thick film polymer absorbing layer that is able to dissipate shock energy through mechanical deformation. Multiple mechanisms for transfer as a function of incident laser energy are observed and we show viable and contamination-free deposition of living mammalian embryonic stem cells

  2. Fires involving radioactive materials : transference model; operative recommendations

    International Nuclear Information System (INIS)

    Rodriguez, C.E.; Puntarulo, L.J.; Canibano, J.A.

    1988-01-01

    In all aspects related to the nuclear activity, the occurrence of an explosion, fire or burst type accident, with or without victims, is directly related to the characteristics of the site. The present work analyses the different parameters involved, describing a transference model and recommendations for evaluation and control of the radiological risk for firemen. Special emphasis is placed on the measurement of the variables existing in this kind of operations

  3. Heat transfer in composite materials disintegrating under high-rate one-sided heating

    Science.gov (United States)

    Isaev, K. B.

    1993-12-01

    A mathematical model of heat transfer in heat-protective materials is suggested with the proviso of a squarelaw temperature depence of the material density in the zone of thermal destruction of its binder. The influence of certain factors on the experimental temperature field and thermal conductivity of a glass-reinforced epoxy plastic material is shown.

  4. Experiment of forced convection heat transfer using microencapsulated phase-change-material slurries

    International Nuclear Information System (INIS)

    Kubo, Shinji; Akino, Norio; Tanaka, Amane; Nagashima, Akira.

    1997-01-01

    The present study describes an experiment on forced convective heat transfer using a water slurry of Microencapsulated Phase-change-material. A normal paraffin hydrocarbon is microencapsulated by melamine resin, melting point of 28.1degC. The heat transfer coefficient and pressure drop in a circular tube were evaluated. The heat transfer coefficient using the slurry in case with and without phase change were compared to in case of using pure water. (author)

  5. Fissile interrogation using gamma rays from oxygen

    Science.gov (United States)

    Smith, Donald; Micklich, Bradley J.; Fessler, Andreas

    2004-04-20

    The subject apparatus provides a means to identify the presence of fissionable material or other nuclear material contained within an item to be tested. The system employs a portable accelerator to accelerate and direct protons to a fluorine-compound target. The interaction of the protons with the fluorine-compound target produces gamma rays which are directed at the item to be tested. If the item to be tested contains either a fissionable material or other nuclear material the interaction of the gamma rays with the material contained within the test item with result in the production of neutrons. A system of neutron detectors is positioned to intercept any neutrons generated by the test item. The results from the neutron detectors are analyzed to determine the presence of a fissionable material or other nuclear material.

  6. Mechanics/heat-transfer relation for particulate materials

    Energy Technology Data Exchange (ETDEWEB)

    Campbell, C.S.; Wang, D.G.; Rahman, K.

    1991-11-01

    The original goal of this study was to try and understand the relationship between the thermal and mechanical properties of particulate flows. Two situations were examined. The first is a study of the effects of simple shear flows, as a embryonic flow type on the apparent thermal conductivity and apparent viscosity of a dry granular flow. The second study involved fluidized beds. The original idea was to try and relate the heat transfer behavior of a fluidized bed to the particle pressure,'' the forces by only the particle phase of the two-phase mixture. (VC)

  7. Mechanics/heat-transfer relation for particulate materials. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Campbell, C.S.; Wang, D.G.; Rahman, K.

    1991-11-01

    The original goal of this study was to try and understand the relationship between the thermal and mechanical properties of particulate flows. Two situations were examined. The first is a study of the effects of simple shear flows, as a embryonic flow type on the apparent thermal conductivity and apparent viscosity of a dry granular flow. The second study involved fluidized beds. The original idea was to try and relate the heat transfer behavior of a fluidized bed to the ``particle pressure,`` the forces by only the particle phase of the two-phase mixture. (VC)

  8. Multiscale network model for simulating liquid water and water vapour transfer properties of porous materials

    NARCIS (Netherlands)

    Carmeliet, J.; Descamps, F.; Houvenaghel, G.

    1999-01-01

    A multiscale network model is presented to model unsaturated moisture transfer in hygroscopic capillary-porous materials showing a broad pore-size distribution. Both capillary effects and water sorption phenomena, water vapour and liquid water transfer are considered. The multiscale approach is

  9. Using a fully automatic mass spectrometer for fissile material control

    International Nuclear Information System (INIS)

    Wilhelmi, M.

    1978-08-01

    The demand for higher accuracy and a shorter delay in the analysis together with better objectifiability and data security needed in safeguards, lead to the automation of a mass spectrometer. Starting with a continuous feeding of samples via a high vacuum lock and including the subsequent heating, focussing and scanning of the samples as well as the final evaluation of the source data (taking alpha spectrometry and the weights required for the isotope dilution technique into account), the mass spectrometric procedure was completely automated. For this purpose, a serial CH-5 instrument of varian mat was modified to be operated by a varian 620/I computer. A newly developed three chamber high vacuum lock was attached to this system and the final evaluation is made with an IBM 370. The system has been used in operation for the isotope analysis of U, Pu and Nd for one year. Major breakdowns of the hardware did not occur, however, the computer programmes had to be steadily improved according to the changing characteristics of the samples. Compared to manual operation, the automat is superior in its throughput and speed of analysing series of similar samples. The automatic procedure objectifies the analysis and the complete evaluation ensures a better data security. (Orig./HP). (author)

  10. Fissile materials from nuclear arms reductions: A question of disposition

    International Nuclear Information System (INIS)

    Sutcliffe, W.G.

    1991-01-01

    This Session, 35T-2, of the Annual Meeting of the American Association for the Advancement of Science (AAAS) was held on February 18, 1991. The papers presented during this session covered a variety of issues and technologies concerning the disposition of the highly enriched uranium and plutonium salvaged from retired nuclear warheads. However, circumstances, including the amount of time available for the session, imposed limitations on the number and breadth of these papers. A comprehensive study of this topic should include a broader range of papers. This session included a paper on molten salt reactors designed to use highly enriched uranium or plutonium as fuel. Other options for the disposal of plutonium, such as transmutation using accelerators and underground vitrification using nuclear explosions, were not discussed during this session, but need to be considered. Individual papers are indexed separately

  11. Conversion ratio and consumption of fissile material in PWR reactors

    International Nuclear Information System (INIS)

    Tiba, C.

    1977-01-01

    It has been shown that the uranium resources will be insufficient for future projected demand. The many solutions to this problem are considered and, in particular, the effect of enrichment on the conversion ratio and hence total uranium comsumption is studied. The developed computacional method employs the one-group neutron diffusion theory. The model is verified by calculating typical burn-up, conversion ratio, U-235 comsumption and plutonium production values in PWR's, and comparing results with those in the published literature. The associated costs of U and U-Pu fuel cycles are also studied for various enrichment values [pt

  12. 49 CFR 173.417 - Authorized fissile materials packages.

    Science.gov (United States)

    2010-10-01

    ... for export and import shipments. (2) A residual “heel” of enriched solid uranium hexafluoride may be... made in accordance with Table 2, as follows: Table 2—Allowable Content of Uranium Hexafluoride (UF6... Liters Cubic feet Maximum Uranium 235-enrichment (weight)percent Maximum “Heel” weight per cylinder UF6...

  13. Creep of fissile ceramic materials under neutron irradiation

    International Nuclear Information System (INIS)

    Brucklacher, D.

    1975-01-01

    Theoretical estimation of the irradiation-induced creep rate of U0 2 by a modification of the Nabarro-Herring model for diffusional creep resulted in a creep rate range between about 6 x 10 -6 to 8 x 10 -5 h -1 for a fission rate of 1 x 10 14 f/cm 3 s and a stress of 2 kgf/mm 2 . Accordingly, the creep rate is enhanced by irradiation at temperatures below 1000 0 to 1200 0 C. It is essentially due to the 'thermal rods' along the fission fragment tracks. Therefore, irradiation-induced creep rates should depend only slightly on temperature and must be markedly lower for carbide and nitride fuel. In-reactor creep experiments on UO 2 were performed at fuel temperatures between 250 0 to 850 0 C. At burnups between 0.3 to 3% the steady-state compressive creep rates are proportional to stress (0 to 4 kgf/mm 2 ) and to fission rate (1 x 10 13 to 2 x 10 14 f/cm 3 s), and are in the range estimated before. The increase in the creep rate with increasing temperature is low and corresponds to an apparent activation energy of only 5200 cal/mol. At burnups above 3 to 4% the stress exponent of the irradiation-induced creep rate increased from n = 1 to n = 1.5. Creep measurements on UO 2 to 15 wt-%Pu0 2 (mechanically mixed, sintered density 86% TD) showed the same temperature dependence as UO 2 below 700 0 C. However, the creep rates were higher by a factor of about 20 compared to fully dense UO 2 . This difference may be explained by assuming a high 'effective' porosity. In-pile creep tests on some UN samples resulted in creep rates that were lower by an order of magnitude than for UO 2 under comparable conditions. (author)

  14. Transfer tunnel transporter system for the Fuels and Materials Examination Facility

    International Nuclear Information System (INIS)

    Petty, J.A.; Miller, S.C.; Richards, J.T.

    1981-01-01

    The detail design is complete and fabrication is approximately 75% complete on the Transfer Tunnel Transporter System. This system provides material handling capability for large, bulky equipment between two hot cells in a new Breeder Reactor Program support facility, the Fuels and Materials Examination Facility. One hot cell has an air atmosphere, the other a high purity inert gas atmosphere which must be maintained during transfer operations. System design features, operational capabilities and remote recovery provisions are described

  15. Transfer of toxic and radioactive materials to and from a work enclosure

    International Nuclear Information System (INIS)

    Hackney, S.

    1980-01-01

    Apparatus for transfer of toxic and radioactive materials between a work enclosure, e.g. a glove box, and a container for storing and transporting the materials comprises a 'double-cover' through which materials are moved. A port in the enclosure is closed by a first cover and the container is closed by a second cover. During transfer, the covers are connected together and the space between the covers is swept by an air stream supplied by a pipe to prevent ingress and deposition of toxic or radioactive material on the facing surfaces of the cover which are subsequently exposed to the environment on separation of the covers. (author)

  16. Safety assessment requirements for onsite transfers of radioactive material

    International Nuclear Information System (INIS)

    Opperman, E.K.; Jackson, E.J.; Eggers, A.G.

    1992-05-01

    This document contains the requirements for developing a safety assessment document for an onsite package containing radioactive material. It also provides format and content guidance to establish uniformity in the safety assessment documentation and to ensure completeness of the information provided

  17. Ternary fission of spontaneously fissile uranium isomers excited by neutrons

    International Nuclear Information System (INIS)

    Makarenko, V.E.; Molchanov, Y.D.; Otroshchenko, G.A.; Yan'kov, G.B.

    1989-01-01

    Spontaneously fissile isomers (SFI) of uranium were excited in the reactions 236,238 U(n,n') at an average neutron energy 4.5 MeV. A pulsed electrostatic accelerator and time analysis of the fission events were used. Fission fragments were detected by the scintillation method, and long-range particles from fission were detected by an ionization method. The relative probability of fission of nuclei through a spontaneously fissile isomeric state was measured: (1.30±0.01)·10 -4 ( 236 U) and (1.48±0.02)·10 -4 ( 238 U). Half-lives of the isomers were determined: 121±2 nsec (the SFI 236 U) and 267±13 nsec (the SFI 238 U). In study of the ternary fission of spontaneously fissile isotopes of uranium it was established that the probability of the process amounts to one ternary fission per 163±44 binary fissions of the SFI 236 U and one ternary fission per 49±14 binary fissions of the SFI 238 U. The substantial increase of the probability of ternary fission of SFI of uranium in comparison with the case of ternary fission of nuclei which are not in an isomeric state may be related to a special nucleon configuration of the fissile isomers of uranium

  18. Multilevel parametrization of fissile nuclei resonance cross sections

    International Nuclear Information System (INIS)

    Lukyanov, A.A.; Kolesov, V.V.; Janeva, N.

    1987-01-01

    Because the resonance interference has an important influence on the resonance structure of neutron cross sections energy dependence at lowest energies, multilevel scheme of the cross section parametrization which take into account the resonance interference is used for the description with the same provisions in the regions of the interferential maximum and minimum of the resonance cross sections of the fissile nuclei

  19. Apparatus and method for quantitatively evaluating total fissile and total fertile nuclide content in samples

    International Nuclear Information System (INIS)

    Caldwell, J.T.; Cates, M.R.; Franks, L.A.; Kunz, W.E.

    1985-01-01

    Simultaneous photon and neutron interrogation of samples for the quantitative determination of total fissile nuclide and total fertile nuclide material present is made possible by the use of an electron accelerator. Prompt and delayed neutrons produced from resulting induced fissions are counted using a single detection system and allow the resolution of the contributions from each interrogating flux leading in turn to the quantitative determination sought. Detection limits for 239 Pu are estimated to be about 3 mg using prompt fission neutrons and about 6 mg using delayed delayed neutrons

  20. Raman spectral indicators of catalyst decoupling for transfer of CVD grown 2D materials

    DEFF Research Database (Denmark)

    Whelan, Patrick Rebsdorf; Jessen, Bjarke Sørensen; Wang, Ruizhi

    2017-01-01

    .7% monolayer graphene coverage, for up to 300 mm diameter wafers.We find a strong correlation between the transfer coverage obtained for graphene and the emergence of a lower wavenumber 2D peak component, with the concurrent disappearance of the higher wavenumber 2Dþ peak component during oxidation......Through a combination of monitoring the Raman spectral characteristics of 2D materials grown on copper catalyst layers, and wafer scale automated detection of the fraction of transferred material, we reproducibly achieve transfers with over 97.5% monolayer hexagonal boron nitride and 99...... of the catalyst surface. The 2D peak characteristics can therefore act as an unambiguous predictor of the success of the transfer. The combined monitoring and transfer process presented here is highly scalable and amenable for roll-to-roll processing....

  1. Experimental study on method for heat transfer enhancement using a porous material

    International Nuclear Information System (INIS)

    Shimura, Takuya; Takeda, Tetsuaki

    2011-01-01

    There are several methods for enhancement of heat transfer; for example, there are attaching various fins on the heat transfer surface, processing the surface roughly, and so on. When cooling high temperature circular or rectangular channels by forced convection of gas, there are several methods for enhancement of heat transfer such as attaching radial or spiral fins on the channel surface or inserting twisted tape in the channel. In the case of the gas heating type steam reformer, disk type fins are attached on the outside surface of the reformer tube, and the tube is inserted into the guide tube to increase an amount of heat transferred from the high temperature gas. However, it has to take into consideration the deterioration of the structure strength by attaching the fins on the tube surface with the design of the steam reformer. The objective of this study is to clarify performances of a method for heat transfer enhancement using porous material with high porosity. The experiment has been performed using an apparatus which simulated the passage structure of the steam reformer to obtain characteristics of heat transfer and pressure drop. From the results obtained in this experiment, the heat transfer rate by this method showed a good performance in the laminar flow region. It was also found that the method for heat transfer enhancement using porous material with high porosity is further improved under the high temperature condition as compared with the other methods for heat transfer enhancement. (author)

  2. Synchronized fusion development considering physics, materials and heat transfer

    Science.gov (United States)

    Wong, C. P. C.; Liu, Y.; Duan, X. R.; Xu, M.; Li, Q.; Feng, K. M.; Zheng, G. Y.; Li, Z. X.; Wang, X. Y.; Li, B.; Zhang, G. S.

    2017-12-01

    Significant achievements have been made in the last 60 years in the development of fusion energy with the tokamak configuration. Based on the accumulated knowledge, the world is embarking on the construction and operation of ITER (International Thermonuclear Experimental Reactor) with a production of 500 MWf fusion power and the demonstration of physics Q  =  10. ITER will demonstrate D-T burn physics for a duration of a few hundred seconds to prepare for the next long-burn or steady state nuclear testing tokamak operating at much higher neutron fluence. With the evolution into a steady state nuclear device, such as the China Fusion Engineering Test Reactor (CFETR), it is necessary to examine the boundary conditions imposed by the combined development of tokamak physics, fusion materials and fusion technology for a reactor. The development of ferritic steel alloys as the structural material suitable for use at high neutron fluence leads to the use of helium as the most likely reactor coolant. This points to the fundamental technology limitation on the removal of chamber wall maximum heat flux at around 1 MW m-2 and an average heat flux of 0.1 MW m-2 for the next test reactor. Future reactor performance will then depend on the control of spatial and temporal edge heat flux peaking in order to increase the average heat flux to the chamber wall. With these severe material and technological limitations, system studies were used to scope out a few robust steady state synchronized fusion reactor (SFR) designs. As an example, a low fusion power design at 131.6 MWf, which can satisfy steady state design requirements, would have a major radius of 5.5 m and minor radius of 1.6 m. Such a design with even more advanced structural materials like W f/W composite could allow higher performance and provide a net electrical production of 62 MWe. These can be incorporated into the CFETR program.

  3. Technology transfer and international development: Materials and manufacturing technology

    Science.gov (United States)

    1982-01-01

    Policy oriented studies on technological development in several relatively advanced developing countries were conducted. Priority sectors defined in terms of technological sophistication, capital intensity, value added, and export potential were studied in Brazil, Venezuela, Israel, and Korea. The development of technological policy alternatives for the sponsoring country is assessed. Much emphasis is placed on understanding the dynamics of the sectors through structured interviews with a large sample of firms in the leading manufacturing and materials processing sectors.

  4. Corrosion of heat exchanger materials under heat transfer conditions

    International Nuclear Information System (INIS)

    Tapping, R.L.; Lavoie, P.A.; Disney, D.J.

    1987-01-01

    Severe pitting has occurred in moderator heat exchangers tubed with Incoloy-800 in Pickering Nuclear Generating Station. The pitting originated on the cooling side (outside) of the tubes and perforation occurred in less than two years. It was known from corrosion testing at CRNL that Incoloy-800 was not susceptible to pitting in Lake Ontario water under isothermal conditions. Corrosion testing with heat transfer across the tube wall was carried out, and it was noted that severe pitting could occur under deposits formed on the tubes in silty Lake Ontario water. Subsequent testing, carried out in co-operation with Ontario Hydro Research Division, investigated the pitting resistance of other candidate tubing alloys: Incoloy-825, 904 L stainless steel, AL-6X, Inconel-625, 70:30 Cu:Ni, titanium, Sanicro-30 and Sanicro-28 1 . Of these, only titanium and Sanicro-28 have not suffered some degree of pitting attack in silt-containing Lake Ontario Water. In the absence of silt, and hence deposits, no pitting took place on any of the alloys tested

  5. Charge transfer and redistribution at interfaces between metals and 2D materials

    NARCIS (Netherlands)

    Bokdam, Menno

    2013-01-01

    Large potential steps are observed at the interfaces between metals and novel 2D materials. They can lower the work function by more than 1 eV, even when the two parts are only weakly interacting. In this thesis the transfer and redistribution of electrons in metal|2D material heterostructures are

  6. The influence of surface treatment on mass transfer between air and building material

    DEFF Research Database (Denmark)

    Kwiatkowski, Jerzy; Rode, Carsten; Hansen, Kurt Kielsgaard

    2008-01-01

    for the experiments: gypsum board and calcium silicate. The wallpaper and paint were used as finishing materials. Impact of the following parameters for changes of RH was studied: coating, temperature and air movement. The measurements showed that acryl paint (diffusion open) can significantly decrease mass uptake......The processes of mass transfer between air and building structure and in the material influence not only the conditions within the material but also inside the connected air spaces. The material which absorbs and desorbs water vapour can be used to moderate the amplitude of indoor relative humidity...... and therefore to participate in the improvement of the indoor air quality and energy saving. Many parameters influence water vapour exchange between indoor air and building material. The aim of this work is to present the change of mass transfer under different climatic and material conditions. The measurements...

  7. Characterization of the heat transfer properties of thermal interface materials

    Science.gov (United States)

    Fullem, Travis Z.

    Physicists have studied the thermal conductivity of solids for decades. As a result of these efforts, thermal conduction in crystalline solids is well understood; there are detailed theories describing thermal conduction due to electrons and phonons. Phonon scattering and transmission at solid/solid interfaces, particularly above cryogenic temperatures, is not well understood and more work is needed in this area. The desire to solve engineering problems which require good thermal contact between mating surfaces has provided enhanced motivation for furthering the state of the art on this topic. Effective thermal management is an important design consideration in microelectronic systems. A common technique for removing excess heat from an electronic device is to attach a heatsink to the device; it is desirable to minimize the thermal resistance between the device and the heatsink. This can be accomplished by placing a thermal interface material (TIM) between the two surfaces. Due to the ever-increasing power densities found in electronic components, there is a desire to design better TIMs, which necessitates the ability to characterize TIM bondlines and to better understand the physics of heat conduction through TIM bondlines. A micro Fourier apparatus which employs Pt thin film thermometers of our design has been built and is capable of precisely quantifying the thermal resistance of thermal interface materials. In the present work several types of commercially available TIMs have been studied using this apparatus, including: greases, filled epoxies, and thermally conductive pads. In the case of filled epoxies, bondlines of various thicknesses, ranging from thirty microns to several hundred microns, have been measured. The microstructure of these bondlines has been investigated using optical microscopy and acoustic microscopy. Measured values of thermal conductivity are considered in terms of microstructural features such as percolation networks and filler particle

  8. Knowledge and Technology Transfer in Materials Science and Engineering in Europe

    OpenAIRE

    Bressler, Patrick; Dürig, Urs; González-Elipe, Agustin; Quandt, Eckhard; Ritschkoff, Anne-Christine; Vahlas, Constantin

    2015-01-01

    Advanced Materials is one of the Key Enabling 3 Technologies identified by the European Commission1. Together with Advanced Manufacturing it underpins almost all other Key Enabling and Industrial Technologies. The basic science and engineering research that results in the development of Advanced Materials lies within the field of Materials Science and Engineering (MSE). The transfer of knowledge from basic research into final products and applications in the field of MSE involves certain MSE-...

  9. Investigation of material transfer in sliding friction-topography or surface chemistry?

    OpenAIRE

    Westlund, V.; Heinrichs, J.; Olsson, M.; Jacobson, S.

    2016-01-01

    To differentiate between the roles of surface topography and chemical composition on influencing friction and transfer in sliding contact, a series of tests were performed in situ in an SEM. The initial sliding during metal forming was investigated, using an aluminum tip representing the work material, put into sliding contact with a polished flat tool material. Both DLC-coated and uncoated tool steel was used. By varying the final polishing step of the tool material, different surface topogr...

  10. Non-local spatial frequency response of photopolymer materials containing chain transfer agents: I. Theoretical modelling

    International Nuclear Information System (INIS)

    Guo, Jinxin; Gleeson, Michael R; Liu, Shui; Sheridan, John T

    2011-01-01

    The non-local photopolymerization driven diffusion (NPDD) model predicts that a reduction in the non-local response length within a photopolymer material will improve its high spatial frequency response. The introduction of a chain transfer agent reduces the average molecular weight of polymer chains formed during free radical polymerization. Therefore a chain transfer agent (CTA) provides a practical method to reduce the non-local response length. An extended NPDD model is presented, which includes the chain transfer reaction and most major photochemical processes. The addition of a chain transfer agent into an acrylamide/polyvinyl alcohol photopolymer material is simulated and the predictions of the model are examined. The predictions of the model are experimentally examined in part II of this paper

  11. Simultaneous Heat and Mass Transfer Model for Convective Drying of Building Material

    Science.gov (United States)

    Upadhyay, Ashwani; Chandramohan, V. P.

    2018-04-01

    A mathematical model of simultaneous heat and moisture transfer is developed for convective drying of building material. A rectangular brick is considered for sample object. Finite-difference method with semi-implicit scheme is used for solving the transient governing heat and mass transfer equation. Convective boundary condition is used, as the product is exposed in hot air. The heat and mass transfer equations are coupled through diffusion coefficient which is assumed as the function of temperature of the product. Set of algebraic equations are generated through space and time discretization. The discretized algebraic equations are solved by Gauss-Siedel method via iteration. Grid and time independent studies are performed for finding the optimum number of nodal points and time steps respectively. A MATLAB computer code is developed to solve the heat and mass transfer equations simultaneously. Transient heat and mass transfer simulations are performed to find the temperature and moisture distribution inside the brick.

  12. Feasibility of fissile mass assay of spent nuclear fuel using 252Cf-source-driven frequency-analysis

    International Nuclear Information System (INIS)

    Mattingly, J.K.; Valentine, T.E.; Mihalczo, J.T.

    1996-01-01

    The feasibility was evaluated using MCNP-DSP, an analog Monte Carlo transport cod to simulate source-driven measurements. Models of an isolated Westinghouse 17x17 PWR fuel assembly in a 1500-ppM borated water storage pool were used. In the models, the fuel burnup profile was represented using seven axial burnup zones, each with isotopics estimated by the PDQ code. Four different fuel assemblies with average burnups from fresh to 32 GWd/MTU were modeled and analyzed. Analysis of the fuel assemblies was simulated by inducing fission in the fuel using a 252 Cf source adjacent to the assembly and correlating source fissions with the response of a bank of 3 He detectors adjacent to the assembly opposite the source. This analysis was performed at 7 different axial positions on each of the 4 assemblies, and the source-detector cross-spectrum signature was calculated for each of these 28 simulated measurements. The magnitude of the cross-spectrum signature follows a smooth upward trend with increasing fissile material ( 235 U and 239 Pu) content, and the signature is independent of the concentration of spontaneously fissioning isotopes (e.g., 244 Cm) and (α,n) sources. Furthermore, the cross-spectrum signature is highly sensitive to changes in fissile material content. This feasibility study indicated that the signature would increase ∼100% in response to an increase of only 0.1 g/cm 3 of fissile material

  13. Development of Novel Electrode Materials for the Electrocatalysis of Oxygen-Transfer and Hydrogen-Transfer Reactions

    Energy Technology Data Exchange (ETDEWEB)

    Simpson, Brett Kimball [Iowa State Univ., Ames, IA (United States)

    2002-01-01

    Throughout this thesis, the fundamental aspects involved in the electrocatalysis of anodic O-transfer reactions and cathodic H-transfer reactions have been studied. The investigation into anodic O-transfer reactions at undoped and Fe(III)[doped MnO2 films] revealed that MnO2 film electrodes prepared by a cycling voltammetry deposition show improved response for DMSO oxidation at the film electrodes vs. the Au substrate. Doping of the MnO2 films with Fe(III) further enhanced electrode activity. Reasons for this increase are believed to involve the adsorption of DMSO by the Fe(III) sites. The investigation into anodic O-transfer reactions at undoped and Fe(III)-doped RuO2 films showed that the Fe(III)-doped RuO2-film electrodes are applicable for anodic detection of sulfur compounds. The Fe(III) sites in the Fe-RuO2 films are speculated to act as adsorption sites for the sulfur species while the Ru(IV) sites function for anodic discharge of H2O to generate the adsorbed OH species. The investigation into cathodic H-transfer reactions, specifically nitrate reduction, at various pure metals and their alloys demonstrated that the incorporation of metals into alloy materials can create a material that exhibits bifunctional properties for the various steps involved in the overall nitrate reduction reaction. The Sb10Sn20Ti70, Cu63Ni37 and Cu25Ni75 alloy electrodes exhibited improved activity for nitrate reduction as compared to their pure component metals. The Cu63Ni37 alloy displayed the highest activity for nitrate reduction. The final investigation was a detailed study of the electrocatalytic activity of cathodic H-transfer reactions (nitrate reduction) at various compositions of Cu-Ni alloy electrodes. Voltammetric response for NO3- at the Cu-Ni alloy electrode is superior to

  14. Experimental study of dynamic effects in moisture transfer in building materials

    DEFF Research Database (Denmark)

    Janssen, Hans; Scheffler, Gregor Albrecht; Plagge, Rudolf

    2016-01-01

    transfer in building materials, similar to moisture transfer in soils, is not free of dynamic effects. The findings imply that the widely accepted static theory for moisture storage in porous media is not generally valid and should be corrected for the occurrences of dynamic effects. Considering......In relation to moisture storage in porous materials, it is often assumed that the process dynamics do not affect the moisture retention. There is mounting evidence though that this notion is incorrect: various studies demonstrate that the moisture retention is influenced by the (de)saturation rates...... of the moisture transfer processes involved. The available evidence primarily stems from imbibition and drainage experiments on soils however, and compared to many other porous media, these tests consider rather permeable materials with relatively dominant liquid transport at comparatively large (de...

  15. Deterministic transfer of two-dimensional materials by all-dry viscoelastic stamping

    International Nuclear Information System (INIS)

    Castellanos-Gomez, Andres; Buscema, Michele; Molenaar, Rianda; Singh, Vibhor; Janssen, Laurens; Van der Zant, Herre S J; Steele, Gary A

    2014-01-01

    The deterministic transfer of two-dimensional crystals constitutes a crucial step towards the fabrication of heterostructures based on the artificial stacking of two-dimensional materials. Moreover, controlling the positioning of two-dimensional crystals facilitates their integration in complex devices, which enables the exploration of novel applications and the discovery of new phenomena in these materials. To date, deterministic transfer methods rely on the use of sacrificial polymer layers and wet chemistry to some extent. Here, we develop an all-dry transfer method that relies on viscoelastic stamps and does not employ any wet chemistry step. This is found to be very advantageous to freely suspend these materials as there are no capillary forces involved in the process. Moreover, the whole fabrication process is quick, efficient, clean and it can be performed with high yield. (letter)

  16. Polyoxometalate active charge-transfer material for mediated redox flow battery

    Energy Technology Data Exchange (ETDEWEB)

    Anderson, Travis Mark; Hudak, Nicholas; Staiger, Chad; Pratt, Harry

    2017-01-17

    Redox flow batteries including a half-cell electrode chamber coupled to a current collecting electrode are disclosed herein. In a general embodiment, a separator is coupled to the half-cell electrode chamber. The half-cell electrode chamber comprises a first redox-active mediator and a second redox-active mediator. The first redox-active mediator and the second redox-active mediator are circulated through the half-cell electrode chamber into an external container. The container includes an active charge-transfer material. The active charge-transfer material has a redox potential between a redox potential of the first redox-active mediator and a redox potential of the second redox-active mediator. The active charge-transfer material is a polyoxometalate or derivative thereof. The redox flow battery may be particularly useful in energy storage solutions for renewable energy sources and for providing sustained power to an electrical grid.

  17. Erosion and mass transfer of Mo, W and Nb under neutron irradiation of high temperature materials

    International Nuclear Information System (INIS)

    Berzhatyj, V.I.; Luk'yanov, A.N.; Zavalishin, A.A.; Tkach, V.N.; Fedorenko, A.I.

    1980-01-01

    Studies have been made of the medium composition in thermionic fuel elements of two types during reactor tests; erosion and mass transfer of electrode materials have been investigated in the after-reactor analysis of the tested fuel elements. The studies of electrode material evaporation at the conditions approaching (in environment temperature and composition) those of reactor tests of thermionic fuel elements have shown that the process proceeds in the form of metal oxides. Evaporation rates are determined, the mechanism of evaporation is discussed, and the analytical dependences are obtained for calculating the evaporation rates of Mo and W at certain temperature and gaseous medium composition. It is found that the main contribution to the material transfer off the Mo and Nb surfaces under a high-temperature reactor irradiation comes through the thermal evaporation; in the case of tungsten at the same experimental conditions the rates of mass transfer due to thermal evaporation and neutron sputtering are nearly the same [ru

  18. Calibration measurements using the ORNL fissile mass flow monitor

    International Nuclear Information System (INIS)

    March-Leuba, J.; Uckan, T.; Sumner, J.; Mattingly, J.; Mihalczo, J.

    1998-01-01

    This paper presents a demonstration of fissile-mass-flow measurements using the Oak Ridge National Laboratory (ORNL) Fissile Mass Flow Monitor in the Paducah Gaseous Diffusion Plant (PGDP). This Flow Monitor is part of a Blend Down Monitoring System (BDMS) that will be installed in at least two Russian Federation (R.F.) blending facilities. The key objectives of the demonstration of the ORNL Flow Monitor are two: (a) demonstrate that the ORNL Flow Monitor equipment is capable of reliably monitoring the mass flow rate of 235 UF 6 gas, and (b) provide a demonstration of ORNL Flow Monitor system in operation with UF 6 flow for a visiting R.F. delegation. These two objectives have been met by the PGDP demonstration, as presented in this paper

  19. The burnable poisons utilization for fissile enriched CANDU fuel bundle

    Energy Technology Data Exchange (ETDEWEB)

    Serghiuta, D; Nainer, O [Team 3 Solutions, Don Mills, ON (Canada)

    1996-12-31

    Utilization of burnable poison for the fissile enriched fueled CANDU 6 Mk1 core is investigated. The main incentives for this analysis are the reduction of void reactivity effects, the maximization of the fissile content of fresh fuel bundles, and the achievement of better power shape control, in order to preserve the power envelope of the standard 37 rod fuel bundle. The latter allows also the preservation of construction parameters of the standard core (for example: number and location of reactivity devices). It also permits the use of regular shift fueling schemes. The paper makes analyses of MOX weapons-grade plutonium and 1.2% SEU fueled CANDU 6 Mk 1 cores. (author). 6 refs., 4 tabs., 10 figs.

  20. Reduction of the uncertainty due to fissile clusters in radioactive waste characterization with the Differential Die-away Technique

    Science.gov (United States)

    Antoni, R.; Passard, C.; Perot, B.; Guillaumin, F.; Mazy, C.; Batifol, M.; Grassi, G.

    2018-07-01

    AREVA NC is preparing to process, characterize and compact old used fuel metallic waste stored at La Hague reprocessing plant in view of their future storage ("Haute Activité Oxyde" HAO project). For a large part of these historical wastes, the packaging is planned in CSD-C canisters ("Colis Standard de Déchets Compacté s") in the ACC hulls and nozzles compaction facility ("Atelier de Compactage des Coques et embouts"). . This paper presents a new method to take into account the possible presence of fissile material clusters, which may have a significant impact in the active neutron interrogation (Differential Die-away Technique) measurement of the CSD-C canisters, in the industrial neutron measurement station "P2-2". A matrix effect correction has already been investigated to predict the prompt fission neutron calibration coefficient (which provides the fissile mass) from an internal "drum flux monitor" signal provided during the active measurement by a boron-coated proportional counter located in the measurement cavity, and from a "drum transmission signal" recorded in passive mode by the detection blocks, in presence of an AmBe point source in the measurement cell. Up to now, the relationship between the calibration coefficient and these signals was obtained from a factorial design that did not consider the potential for occurrence of fissile material clusters. The interrogative neutron self-shielding in these clusters was treated separately and resulted in a penalty coefficient larger than 20% to prevent an underestimation of the fissile mass within the drum. In this work, we have shown that the incorporation of a new parameter in the factorial design, representing the fissile mass fraction in these clusters, provides an alternative to the penalty coefficient. This new approach finally does not degrade the uncertainty of the original prediction, which was calculated without taking into consideration the possible presence of clusters. Consequently, the

  1. Minimizing residues and strain in 2D materials transferred from PDMS

    Science.gov (United States)

    Jain, Achint; Bharadwaj, Palash; Heeg, Sebastian; Parzefall, Markus; Taniguchi, Takashi; Watanabe, Kenji; Novotny, Lukas

    2018-06-01

    Integrating layered two-dimensional (2D) materials into 3D heterostructures offers opportunities for novel material functionalities and applications in electronics and photonics. In order to build the highest quality heterostructures, it is crucial to preserve the cleanliness and morphology of 2D material surfaces that come in contact with polymers such as PDMS during transfer. Here we report that substantial residues and up to ∼0.22% compressive strain can be present in monolayer MoS2 transferred using PDMS. We show that a UV-ozone pre-cleaning of the PDMS surface before exfoliation significantly reduces organic residues on transferred MoS2 flakes. An additional 200 ◦C vacuum anneal after transfer efficiently removes interfacial bubbles and wrinkles as well as accumulated strain, thereby restoring the surface morphology of transferred flakes to their native state. Our recipe is important for building clean heterostructures of 2D materials and increasing the reproducibility and reliability of devices based on them.

  2. Alternative repository criticality-control strategies for fissile uranium wastes

    International Nuclear Information System (INIS)

    Forsberg, C.W.

    1998-01-01

    Methods to prevent long term, disposal site nuclear criticality from fissile uranium isotopes in wastes were investigated. Long term refers to the time period after waste package (WP) failure and the subsequent loss of geometry and chemistry control within the WP. The preferred method of control was found to be the addition of sufficient depleted uranium to each WP so that the uranium enrichment is reduced to 235 U and 233 U in 238 U

  3. Development of AGNES, a kinetics code for fissile solutions, 1

    International Nuclear Information System (INIS)

    Nakajima, Ken; Ohnishi, Nobuaki

    1986-01-01

    A kinetics code for fissile solutions, AGNES (Accidentally Generated Nuclear Excursion Simulation code), has been developed. This code calculates the radiolytic gas void effect as a reactivity feedback. Physical and calculative models of the radiolytic gas void are summarized and the usage of AGNES is described. In addition, some benchmark calculations were performed and results of calculations show good agreement with those of experiments. (author)

  4. A role for research ethics committees in exchanges of human biospecimens through material transfer agreements.

    Science.gov (United States)

    Chalmers, Donald; Nicol, Dianne; Nicolás, Pilar; Zeps, Nikolajs

    2014-09-01

    International transfers of human biological material (biospecimens) and data are increasing, and commentators are starting to raise concerns about how donor wishes are protected in such circumstances. These exchanges are generally made under contractual material transfer agreements (MTAs). This paper asks what role, if any, should research ethics committees (RECs) play in ensuring legal and ethical conduct in such exchanges. It is recommended that RECs should play a more active role in the future development of best practice MTAs involving exchange of biospecimens and data and in monitoring compliance.

  5. The simultaneous neutron and photon interrogation method for fissile and non-fissile element separation in radioactive waste drums

    International Nuclear Information System (INIS)

    Jallu, F.; Lyoussi, A.; Passard, C.; Payan, E.; Recroix, H.; Nurdin, G.; Buisson, A.; Allano, J.

    2000-01-01

    Measuring α-emitters such as ( 234,235,236,238 U, 238,239,240,242,244 Pu, 237 Np, 241,243 Am, ...), in solid radioactive waste allows us to quantify the α-activity in a drum and then to classify it. The simultaneous photon and neutron interrogation experiment (SIMPHONIE) method dealt with in this paper, combines both active neutron interrogation and induced photofission interrogation techniques simultaneously. Its purpose is to quantify fissile ( 235 U, 239,241 Pu, ...) and non-fissile ( 236,238 U, 238,240 Pu, ...) elements separately in only one measurement. This paper presents the principle of the method, the experimental setup, and the first experimental results obtained using the DGA/ETCA Linac and MiniLinatron pulsed linear electron accelerators located at Arcueil, France. First studies were carried out with U and Pu bare samples

  6. Transferable tight-binding model for strained group IV and III-V materials and heterostructures

    Science.gov (United States)

    Tan, Yaohua; Povolotskyi, Michael; Kubis, Tillmann; Boykin, Timothy B.; Klimeck, Gerhard

    2016-07-01

    It is critical to capture the effect due to strain and material interface for device level transistor modeling. We introduce a transferable s p3d5s* tight-binding model with nearest-neighbor interactions for arbitrarily strained group IV and III-V materials. The tight-binding model is parametrized with respect to hybrid functional (HSE06) calculations for varieties of strained systems. The tight-binding calculations of ultrasmall superlattices formed by group IV and group III-V materials show good agreement with the corresponding HSE06 calculations. The application of the tight-binding model to superlattices demonstrates that the transferable tight-binding model with nearest-neighbor interactions can be obtained for group IV and III-V materials.

  7. The SVM Method for Fissile Mass Estimation through Passive Neutron Interrogation: Advances and Developments

    International Nuclear Information System (INIS)

    Dubi, C.; Shvili, Israel I.

    2014-01-01

    Fissile mass estimation through passive neutron interrogation is now one of the main techniques for NDT of fissile mass estimation, due to the relative transparency of neutron radiation to structural materials- making it extremely effective in poorly characterized or dirty samples . Passive neutron interrogation relies on the fact that the number of neutrons emitted (per time unit) due to spontaneous fissions from the sample is proportional to the mass of the detected sample. However, since the measurement is effected by additional neutron sources- mainly (D±n) reactions and induced fission chain in the tested sample, a naive estimation, assuming a linear correspondence between the mass of the detected sample and the average number of detections, is bound to give an over estimation of the mass. Since most passive interrogation facilities are based on 3He detectors, the origin of the neutron cannot be determined by analyzing the energy spectrum (as all neutrons arrive at the detector in more or less the same energy), and a mathematical 'filter' is used to evaluate the noise to source ratio in the detection signal. The basic idea behind the mathematical filter is to utilize the fact that the different neutron sources have different statistical attributes- in particular, both the source event rate and the distribution of the number of neutrons released in each event differs between the different sources. There for, by studying the higher moments of the neutron population, new information about the source to noise ration may be obtained

  8. Evaluation of criticality criteria for fissile class II packages in transportation

    International Nuclear Information System (INIS)

    Thomas, J.T.

    1976-01-01

    The nuclear criticality safety of packages in transportation is explored systematically by a surface density representation of reflected array criticality of air-spaced units. Typical perturbations to arrays are shown to be related analytically to the corresponding reactivity changes they produce. The reactivity change associated with the removal of three reflecting surfaces from a totally water reflected array is shown to depend upon the fissile material loading of the packages. For U(93.2) metal, the expected reactivity loss can range from 2 to 21%. Replacement of a three-sided reflector of water on a critical array by one of concrete results in a reactivity increase ranging from 0 to 6%. Mass limits established by criticality data for reflected arrays of air-spaced units can provide a minimum, uniform margin of safety, expressible in terms of reactivity, to more reliably specify subcriticality in transport. Mass limits less than those defined by air-spaced units in water-reflected arrays are unnecessary for Fissile Class II packages. (author)

  9. Heat transfer characteristics of coconut oil as phase change material to room cooling application

    Science.gov (United States)

    Irsyad, M.; Harmen

    2017-03-01

    Thermal comfort in a room is one of human needs in the workplace and dwellings, so that the use of air conditioning system in tropical countries is inevitable. This equipment has an impact on the increase of energy consumption. One method of minimizing the energy use is by using the phase change material (PCM) as thermal energy storage. This material utilizes the temperature difference between day and night for the storage and release of thermal energy. PCM development on application as a material for air cooling inlet, partitioning and interior needs to be supported by the study of heat transfer characteristics when PCM absorbs heat from ambient temperature. This study was conducted to determine the heat transfer characteristics on coconut oil as a phase change material. There are three models of experiments performed in this research. Firstly, an experiment was conducted to analyze the time that was needed by material to phase change by varying the temperature. The second experiment analyzed the heat transfer characteristics of air to PCM naturally convection. The third experiment analyzed the forced convection heat transfer on the surface of the PCM container by varying the air velocity. The data of experimental showed that, increasing ambient air temperature resulted in shorter time for phase change. At temperatures of 30°C, the time for phase change of PCM with the thickness of 8 cm was 1700 min, and it was stable at temperatures of 27°C. Increasing air temperature accelerated the phase change in the material. While for the forced convection heat transfer, PCM could reduce the air temperature in the range of 30 to 35°C at about 1 to 2°C, with a velocity of 1-3 m/s.

  10. Use and Misuse of Material Transfer Agreements: Lessons in Proportionality from Research, Repositories, and Litigation

    OpenAIRE

    Bubela, Tania; Guebert, Jenilee; Mishra, Amrita

    2015-01-01

    Material transfer agreements exist to facilitate the exchange of materials and associated data between researchers as well as to protect the interests of the researchers and their institutions. But this dual mandate can be a source of frustration for researchers, creating administrative burdens and slowing down collaborations. We argue here that in most cases in pre-competitive research, a simple agreement would suffice; the more complex agreements and mechanisms for their negotiation should ...

  11. Simultaneous Contact Sensing and Characterizing of Mechanical and Dynamic Heat Transfer Properties of Porous Polymeric Materials

    Directory of Open Access Journals (Sweden)

    Bao-guo Yao

    2017-10-01

    Full Text Available Porous polymeric materials, such as textile fabrics, are elastic and widely used in our daily life for garment and household products. The mechanical and dynamic heat transfer properties of porous polymeric materials, which describe the sensations during the contact process between porous polymeric materials and parts of the human body, such as the hand, primarily influence comfort sensations and aesthetic qualities of clothing. A multi-sensory measurement system and a new method were proposed to simultaneously sense the contact and characterize the mechanical and dynamic heat transfer properties of porous polymeric materials, such as textile fabrics in one instrument, with consideration of the interactions between different aspects of contact feels. The multi-sensory measurement system was developed for simulating the dynamic contact and psychological judgment processes during human hand contact with porous polymeric materials, and measuring the surface smoothness, compression resilience, bending and twisting, and dynamic heat transfer signals simultaneously. The contact sensing principle and the evaluation methods were presented. Twelve typical sample materials with different structural parameters were measured. The results of the experiments and the interpretation of the test results were described. An analysis of the variance and a capacity study were investigated to determine the significance of differences among the test materials and to assess the gage repeatability and reproducibility. A correlation analysis was conducted by comparing the test results of this measurement system with the results of Kawabata Evaluation System (KES in separate instruments. This multi-sensory measurement system provides a new method for simultaneous contact sensing and characterizing of mechanical and dynamic heat transfer properties of porous polymeric materials.

  12. An artificial compressibility CBS method for modelling heat transfer and fluid flow in heterogeneous porous materials

    CSIR Research Space (South Africa)

    Malan, AG

    2011-08-01

    Full Text Available to modelling both forced convection as well as heat transfer and fluid flow through heterogeneous saturated porous materials via an edge-based finite volume discretization scheme. A volume-averaged set of local thermal disequilibrium governing equations...

  13. Nanoscale heat transfer in carbon nanotube - sugar alcohol composites as heat storage materials

    NARCIS (Netherlands)

    Zhang, H.; Rindt, C.C.M.; Smeulders, D.M.J.; Gaastra - Nedea, S.V.

    2016-01-01

    Nanoscale carbon structures such as graphene and carbon nanotubes (CNTs) can greatly improve the effective thermal conductivity of thermally sluggish heat storage materials, such as sugar alcohols (SAs). The specific improvement depends on the heat transfer rate across the carbon structure. Besides,

  14. Rebuilding conveyor transfer points to cut fugitive material and improve operations

    Energy Technology Data Exchange (ETDEWEB)

    Stahura, R P [Martin Engineering Company, Neponset, MA (USA)

    1992-10-01

    The article describes a three-part programme to control spillage and prevent fugitive material at conveyor transfer points. The three parts are: adequate belt support; the installation of a wear line inside the chute to preserve the rubber seal system; and the maintenance of an effective edge seal. The article also discusses designing for ease of maintenance. 9 figs.

  15. Electron Transfer in Donor-Bridge-Acceptor Systems and Derived Materials

    NARCIS (Netherlands)

    Oosterbaan, W.D.

    2002-01-01

    Some aspects of photoinduced electron transfer (ET) in (electron donor)-bridge-(electron acceptor) compounds (D-B-A) and derived materials are investigated. Aim I is to determine how and to which extent non-conjugated double bonds in an otherwise saturated hydrocarbon bridge affect the rate of

  16. The Effect of Anisotropy of Building Materials on the Moisture Transfer

    Directory of Open Access Journals (Sweden)

    J. Drchalová

    2000-01-01

    Full Text Available The effect of anisotropy of building materials on the moisture transfer in the design of envelope parts of building structures is studied. Two typical fibre containing plate building materials produced in the Czech Republic, Dekalux and Dekalit P, are chosen for the demonstration of this effect. Experimental results show that while for lighter Dekalit P, an order of magnitude difference in the moisture diffusivities k for the two basic orientations, i.e. along and across the plate, is observed, for the heavier Dekalux the differences in k are within the errorbar of the experimental method. As follows from the experimental results, compacting of surface layers of the plates of light fibred materials is very favorable from the point of view of moisture penetration but one should keep in mind that any local damage of the surface layer can result in a considerably faster moisture transfer in the direction along the plate.

  17. Applications of free-electron lasers to measurements of energy transfer in biopolymers and materials

    Science.gov (United States)

    Edwards, Glenn S.; Johnson, J. B.; Kozub, John A.; Tribble, Jerri A.; Wagner, Katrina

    1992-08-01

    Free-electron lasers (FELs) provide tunable, pulsed radiation in the infrared. Using the FEL as a pump beam, we are investigating the mechanisms for energy transfer between localized vibrational modes and between vibrational modes and lattice or phonon modes. Either a laser-Raman system or a Fourier transform infrared (FTIR) spectrometer will serve as the probe beam, with the attribute of placing the burden of detection on two conventional spectroscopic techniques that circumvent the limited response of infrared detectors. More specifically, the Raman effect inelastically shifts an exciting laser line, typically a visible frequency, by the energy of the vibrational mode; however, the shifted Raman lines also lie in the visible, allowing for detection with highly efficient visible detectors. With regards to FTIR spectroscopy, the multiplex advantage yields a distinct benefit for infrared detector response. Our group is investigating intramolecular and intermolecular energy transfer processes in both biopolymers and more traditional materials. For example, alkali halides contain a number of defect types that effectively transfer energy in an intermolecular process. Similarly, the functioning of biopolymers depends on efficient intramolecular energy transfer. Understanding these mechanisms will enhance our ability to modify biopolymers and materials with applications to biology, medecine, and materials science.

  18. Experimental and analytical investigations of granular materials: Shear flow and convective heat transfer

    Science.gov (United States)

    Ahn, Hojin

    1989-12-01

    Granular materials flowing down an inclined chute were studied experimentally and analytically. Characteristics of convective heat transfer to granular flows were also investigated experimentally and numerically. Experiments on continuous, steady flows of granular materials in an inclined chute were conducted with the objectives of understanding the characteristics of chute flows and of acquiring information on the rheological behavior of granular material flow. Existing constitutive equations and governing equations were used to solve for fully developed chute flows of granular materials, and thus the boundary value problem was formulated with two parameters (the coefficient of restitution between particles, and the chute inclination) and three boundary values at the chute base wall (the values of solid fraction, granular temperature, and mean velocity at the wall). The boundary value problem was numerically solved by the shooting method. These analytical results were also compared with the present experimental values and with the computer simulations by other investigators in their literature. Experiments on heat transfer to granular flows over a flat heating plate were conducted with three sizes of glass beads, polystyrene beads, and mustard seeds. A modification on the existing model for the convective heat transfer was made using the effective Nusselt number and the effective Peclet number, which include the effects of solid fraction variations. The slightly modified model could describe the heat transfer characteristics of both fast and slow flows (supercritical and subcritical). A numerical analysis of the transfer to granular flows was also performed. The results were compared with the present experimental data, and reasonable agreement was found in the comparison.

  19. Fissile fuel doubling time characteristics for reactor lifetime fuel logistics

    International Nuclear Information System (INIS)

    Heindler, M.; Harms, A.A.

    1978-01-01

    The establishment of nuclear fuel requirements and their efficient utilization requires a detailed knowledge of some aspects of fuel dynamics and processing during the reactor lifetime. It is shown here that the use of the fuel stockpile inventory concept can serve effectively for this fuel management purpose. The temporal variation of the fissile fuel doubling time as well as nonequilibrium core conditions are among the characteristics which thus become more evident. These characteristics - rather than a single figure-of-merit - clearly provide an improved description of the expansion capacity and/or fuel requirements of a nuclear reactor energy system

  20. Cyclododecane as support material for clean and facile transfer of large-area few-layer graphene

    International Nuclear Information System (INIS)

    Capasso, A.; Leoni, E.; Dikonimos, T.; Buonocore, F.; Lisi, N.; De Francesco, M.; Lancellotti, L.; Bobeico, E.; Sarto, M. S.; Tamburrano, A.; De Bellis, G.

    2014-01-01

    The transfer of chemical vapor deposited graphene is a crucial process, which can affect the quality of the transferred films and compromise their application in devices. Finding a robust and intrinsically clean material capable of easing the transfer of graphene without interfering with its properties remains a challenge. We here propose the use of an organic compound, cyclododecane, as a transfer material. This material can be easily spin coated on graphene and assist the transfer, leaving no residues and requiring no further removal processes. The effectiveness of this transfer method for few-layer graphene on a large area was evaluated and confirmed by microscopy, Raman spectroscopy, x-ray photoemission spectroscopy, and four-point probe measurements. Schottky-barrier solar cells with few-layer graphene were fabricated on silicon wafers by using the cyclododecane transfer method and outperformed reference cells made by standard methods.

  1. The simulation calculation of acoustics energy transfer through the material structure

    Directory of Open Access Journals (Sweden)

    Zvolenský Peter

    2016-01-01

    Full Text Available The paper deals with the modification of the rail passenger coach floor design aimed at improvement of sound reduction index. Refurbishing was performed by using a new acoustic material with a filamentary microstructure. The materials proposed in research were compared by simulation calculation of acoustic energy transfer trough porous microstructure of filamentary material, and the effect of material porosity on sound reduction index and sound absorption coefficient were observed. This proposed filamentary material can be used in the railway bed structure, too. High degree of noise absorbing, resistance to climate conditions, low specific mass, enable to choose a system of low anti-noise barriers having similar properties as standard high anti-noise walls..

  2. A comparative study between transport and criticality safety indexes for fissile uranium nuclearly pure

    Energy Technology Data Exchange (ETDEWEB)

    Moraes da Silva, T. de; Sordi, G.M.A.A. [Instituto de Pesquisas Energeticas e Nucleares, IPEN/CNEN (Brazil)]. e-mail: tmsilva@ipen.br

    2006-07-01

    The international and national standards determine that during the transport of radioactive materials the package to be sent should be identified by labels of risks specifying content, activity and the transport index. The result of the monitoring of the package to 1 meter identifies the transport index, TI, which represents the dose rate to 1 meter of this. The transport index is, by definition, a number that represents a gamma radiation that crosses the superficial layer the radioactive material of the package to 1 meter of distance. For the fissile radioactive material that is the one in which a neutron causes the division of the atom, the international standards specify criticality safety index CSI, which is related with the safe mass of the fissile element. In this work it was determined the respective safe mass for each considered enrichment for the compounds of uranium oxides UO{sub 2}, U{sub 3}O{sub 8} and U{sub 3}Si{sub 2}. In the study of CSI it was observed that the value 50 of the expression 50/N being N the number of packages be transported in subcriticality conditions it represents a fifth part of the safe mass of the element uranium or 9% of the smallest mass critical for a transport not under exclusive use. As conclusion of the accomplished study was observed that the transport index starting from 7% of enrichment doesn't present contribution and that criticality safety index is always greater than the transport index. Therefore what the standards demand to specify, the largest value between both indexes, was clearly identified in this study as being the criticality safety index. (Author)

  3. Spin-transfer phenomena in layered magnetic structures: Physical phenomena and materials aspects

    International Nuclear Information System (INIS)

    Gruenberg, P.; Buergler, D.E.; Dassow, H.; Rata, A.D.; Schneider, C.M.

    2007-01-01

    During the past 20 years, layered structures consisting of ferromagnetic layers and spacers of various material classes with a thickness of only a few nanometers have revealed a variety of exciting and potentially very useful phenomena not present in bulk material. Representing distinct manifestations of spin-transfer processes, these phenomena may be categorized into interlayer exchange coupling (IEC), giant magnetoresistance (GMR), tunneling magnetoresistance (TMR), and the more recently discovered spin-transfer torque effect leading to current-induced magnetization switching (CIMS) and current-driven magnetization dynamics. These phenomena clearly confer novel material properties on magnetic layered structures with respect to the (magneto-)transport and the magnetostatic as well as magnetodynamic behavior. Here, we will first concentrate on the less well understood aspects of IEC across insulating and semiconducting interlayers and relate the observations to TMR in the corresponding structures. In this context, we will also discuss more recent advances in TMR due to the use of electrodes made from Heusler alloys and the realization of coherent tunneling in epitaxial magnetic tunneling junctions. Finally, we will review our results on CIMS in epitaxial magnetic nanostructures showing that normal and inverse CIMS can occur simultaneously in a single nanopillar device. In all cases discussed, material issues play a major role in the detailed understanding of the spin-transfer effects, in particular in those systems that yield the largest effects and are thus of utmost interest for applications

  4. Magnetic field concentration using ferromagnetic material to propel a wireless power transfer based micro-robot

    Directory of Open Access Journals (Sweden)

    Dongwook Kim

    2018-05-01

    Full Text Available In this paper, we propose a novel coil structure, using a ferromagnetic material which concentrates the magnetic field, as the propulsion system of a wireless power transfer (WPT based micro-robot. This structure uses an incident magnetic field to induce current during wireless power transfer, to generate a Lorentz force. To prevent net cancelation of the Lorentz force in the load coil, ferrite films were applied to one side of the coil segment. The demonstrated simplicity and effectiveness of the proposed micro-robot showed its suitability for applications. Simulation and experimental results confirmed a velocity of 1.02 mm/s with 6 mW power transfer capacity for the 3 mm sized micro-robot.

  5. Natural convection heat transfer from a heated horizontal cylinder with Microencapsulated Phase-Change-Material slurries

    International Nuclear Information System (INIS)

    Kubo, Shinji; Akino, Norio; Tanaka, Amane; Nagashima, Akira

    1998-01-01

    The present study investigates natural convection heat transfer from a heated cylinder cooled by a water slurry of Microencapsulated Phase Change Material (MCPCM). A normal paraffin hydrocarbon with carbon number of 18 and melting point of 27.9degC, is microencapsulated by Melamine resin into particles of which average diameter is 9.5 μm and specific weight is same as water. The slurry of the MCPCM and water is put into a rectangular enclosure with a heated horizontal cylinder. The heat transfer coefficients of the cylinder were evaluated. Changing the concentrations of PCM and temperature difference between cylinder surface and working fluid. Addition of MCPCM into water, the heat transfer is enhanced significantly comparison with pure water in cases with phase change and is reduced slightly in cases without phase change. (author)

  6. Magnetic field concentration using ferromagnetic material to propel a wireless power transfer based micro-robot

    Science.gov (United States)

    Kim, Dongwook; Park, Bumjin; Park, Jaehyoung; Park, Hyun Ho; Ahn, Seungyoung

    2018-05-01

    In this paper, we propose a novel coil structure, using a ferromagnetic material which concentrates the magnetic field, as the propulsion system of a wireless power transfer (WPT) based micro-robot. This structure uses an incident magnetic field to induce current during wireless power transfer, to generate a Lorentz force. To prevent net cancelation of the Lorentz force in the load coil, ferrite films were applied to one side of the coil segment. The demonstrated simplicity and effectiveness of the proposed micro-robot showed its suitability for applications. Simulation and experimental results confirmed a velocity of 1.02 mm/s with 6 mW power transfer capacity for the 3 mm sized micro-robot.

  7. Heat transfer characteristics of liquid-gas Taylor flows incorporating microencapsulated phase change materials

    International Nuclear Information System (INIS)

    Howard, J A; Walsh, P A

    2014-01-01

    This paper presents an investigation on the heat transfer characteristics associated with liquid-gas Taylor flows in mini channels incorporating microencapsulated phase change materials (MPCM). Taylor flows have been shown to result in heat transfer enhancements due to the fluid recirculation experienced within liquid slugs which is attributable to the alternating liquid slug and gas bubble flow structure. Microencapsulated phase change materials (MPCM) also offer significant potential with increased thermal capacity due to the latent heat required to cause phase change. The primary aim of this work was to examine the overall heat transfer potential associated with combining these two novel liquid cooling technologies. By investigating the local heat transfer characteristics, the augmentation/degradation over single phase liquid cooling was quantified while examining the effects of dimensionless variables, including Reynolds number, liquid slug length and gas void fraction. An experimental test facility was developed which had a heated test section and allowed MPCM-air Taylor flows to be subjected to a constant heat flux boundary condition. Infrared thermography was used to record high resolution experimental wall temperature measurements and determine local heat transfer coefficients from the thermal entrance point. 30.2% mass particle concentration of the MPCM suspension fluid was examined as it provided the maximum latent heat for absorption. Results demonstrate a significant reduction in experimental wall temperatures associated with MPCM-air Taylor flows when compared with the Graetz solution for conventional single phase coolants. Total enhancement in the thermally developed region is observed to be a combination of the individual contributions due to recirculation within the liquid slugs and also absorption of latent heat. Overall, the study highlights the potential heat transfer enhancements that are attainable within heat exchange devices employing MPCM

  8. Charge-transfer channel in quantum dot-graphene hybrid materials

    Science.gov (United States)

    Cao, Shuo; Wang, Jingang; Ma, Fengcai; Sun, Mengtao

    2018-04-01

    The energy band theory of a classical semiconductor can qualitatively explain the charge-transfer process in low-dimensional hybrid colloidal quantum dot (QD)-graphene (GR) materials; however, the definite charge-transfer channels are not clear. Using density functional theory (DFT) and time-dependent DFT, we simulate the hybrid QD-GR nanostructure, and by constructing its orbital interaction diagram, we show the quantitative coupling characteristics of the molecular orbitals (MOs) of the hybrid structure. The main MOs are derived from the fragment MOs (FOs) of GR, and the Cd13Se13 QD FOs merge with the GR FOs in a certain proportion to afford the hybrid system. Upon photoexcitation, electrons in the GR FOs jump to the QD FOs, leaving holes in the GR FOs, and the definite charge-transfer channels can be found by analyzing the complex MOs coupling. The excited electrons and remaining holes can also be localized in the GR or the QD or transfer between the QD and GR with different absorption energies. The charge-transfer process for the selected excited states of the hybrid QD-GR structure are testified by the charge difference density isosurface. The natural transition orbitals, charge-transfer length analysis and 2D site representation of the transition density matrix also verify the electron-hole delocalization, localization, or coherence chacracteristics of the selected excited states. Therefore, our research enhances understanding of the coupling mechanism of low-dimensional hybrid materials and will aid in the design and manipulation of hybrid photoelectric devices for practical application in many fields.

  9. Transfer

    DEFF Research Database (Denmark)

    Wahlgren, Bjarne; Aarkrog, Vibe

    Bogen er den første samlede indføring i transfer på dansk. Transfer kan anvendes som praksis-filosofikum. Den giver en systematisk indsigt til den studerende, der spørger: Hvordan kan teoretisk viden bruges til at reflektere over handlinger i situationer, der passer til min fremtidige arbejdsplads?...

  10. Coupled heat transfer in high temperature transporting system with semitransparent/opaque material

    International Nuclear Information System (INIS)

    Du Shenghua; Xia Xinjin

    2010-01-01

    The heat transfer model of the aerodynamic heating coupled with radiative cooling was developed. The thermal protect system includes the higher heat flux region with high temperature semitransparent material, the heat transporting channel and the lower heat flux region with metal. The control volume method was combined with the Monte Carlo method to calculate the coupled heat transfer of the transporting system, and the thermal equilibrium equation for the transporting channel was solved simultaneously. The effect of the aeroheating flux radio, the area ratio of radiative surfaces, the convective heat transfer coefficient of the heat transporting channel on the radiative surface temperature and the fluid temperature in the heat transporting channel were analyzed. The effect of radiation and conduction in the semitransparent material was discussed. The result shows that to increase the convective heat transfer coefficient in heat flux channel can enhance the heat transporting ability of the system, but the main parameter to effect on the temperature of the heat transporting system is the area ratio of radiative surfaces. (authors)

  11. Calculation of the minimum critical mass of fissile nuclides

    International Nuclear Information System (INIS)

    Wright, R.Q.; Hopper, Calvin Mitchell

    2008-01-01

    The OB-1 method for the calculation of the minimum critical mass of fissile actinides in metal/water systems was described in a previous paper. A fit to the calculated minimum critical mass data using the extended criticality parameter is the basis of the revised method. The solution density (grams/liter) for the minimum critical mass is also obtained by a fit to calculated values. Input to the calculation consists of the Maxwellian averaged fission and absorption cross sections and the thermal values of nubar. The revised method gives more accurate values than the original method does for both the minimum critical mass and the solution densities. The OB-1 method has been extended to calculate the uncertainties in the minimum critical mass for 12 different fissile nuclides. The uncertainties for the fission and capture cross sections and the estimated nubar uncertainties are used to determine the uncertainties in the minimum critical mass, either in percent or grams. Results have been obtained for U-233, U-235, Pu-236, Pu-239, Pu-241, Am-242m, Cm-243, Cm-245, Cf-249, Cf-251, Cf-253, and Es-254. Eight of these 12 nuclides are included in the ANS-8.15 standard.

  12. 1987 target values for uncertainty components in fissile isotope and element assay

    International Nuclear Information System (INIS)

    De Bievre, P.; Baumann, S.; Gorgenyi, T.; Kuhn, E.; Deron, S.; Dalton, J.; Perrin, R.E.; Pietri, C.; De Regge, P.

    1987-01-01

    The Working Group on Techniques and Standards for Destructive Analysis (WGDA) of the European Safeguards Research and Development Association (ESARDA), which at present includes the representation of 37 nuclear analytical laboratories, has long been concerned with defining realistic performance characteristics of destructive analysis techniques. One of the terms of reference of the working groups is: ''to evaluate and recommend criteria for destructive analysis of nuclear materials for use by plant operators and safeguarding authorities''. Some of the most important and most badly needed criteria are those to be used for judging results of quantitative determinations of fissile isotope and element amounts. The working group has recognized and discussed this problem at several meetings and decided that it was appropriate to fix reasonable levels of performance as ''goals'' for nuclear analytical laboratories

  13. A Polarizable and Transferable PHAST CO 2 Potential for Materials Simulation

    KAUST Repository

    Mullen, Ashley L.

    2013-12-10

    Reliable PHAST (Potentials with High Accuracy Speed and Transferability) intermolecular potential energy functions for CO2 have been developed from first principles for use in heterogeneous systems, including one with explicit polarization. The intermolecular potentials have been expressed in a transferable form and parametrized from nearly exact electronic structure calculations. Models with and without explicit many-body polarization effects, known to be important in simulation of interfacial processes, are constructed. The models have been validated on pressure-density isotherms of bulk CO 2 and adsorption in three metal-organic framework (MOF) materials. The present models appear to offer advantages over high quality fluid/liquid state potentials in describing CO2 interactions in interfacial environments where sorbates adopt orientations not commonly explored in bulk fluids. Thus, the nonpolar CO2-PHAST and polarizable CO 2-PHAST* potentials are recommended for materials/interfacial simulations. © 2013 American Chemical Society.

  14. Use and misuse of material transfer agreements: lessons in proportionality from research, repositories, and litigation.

    Directory of Open Access Journals (Sweden)

    Tania Bubela

    2015-02-01

    Full Text Available Material transfer agreements exist to facilitate the exchange of materials and associated data between researchers as well as to protect the interests of the researchers and their institutions. But this dual mandate can be a source of frustration for researchers, creating administrative burdens and slowing down collaborations. We argue here that in most cases in pre-competitive research, a simple agreement would suffice; the more complex agreements and mechanisms for their negotiation should be reserved for cases where the risks posed to the institution and the potential commercial value of the research reagents is high.

  15. Use and misuse of material transfer agreements: lessons in proportionality from research, repositories, and litigation.

    Science.gov (United States)

    Bubela, Tania; Guebert, Jenilee; Mishra, Amrita

    2015-02-01

    Material transfer agreements exist to facilitate the exchange of materials and associated data between researchers as well as to protect the interests of the researchers and their institutions. But this dual mandate can be a source of frustration for researchers, creating administrative burdens and slowing down collaborations. We argue here that in most cases in pre-competitive research, a simple agreement would suffice; the more complex agreements and mechanisms for their negotiation should be reserved for cases where the risks posed to the institution and the potential commercial value of the research reagents is high.

  16. Material Flow and Stakeholder Analysis for a Transfer & Recycling Station in Gaborone, Botswana

    OpenAIRE

    Andersson, Emil

    2014-01-01

    Landfilling waste material is still one of the most common methods to take care of waste in a big part of the world. Gaborone, the capital of Botswana located in the southern part of Africa is no different in this way. The major part of all waste is landfilled in Gaborone and there is only a minor part of all collected material that is recycled. One solution that earlier studies suggest is to build a transfer and recycling station in the city of Gaborone that can contribute to a more sustaina...

  17. Experimental and numerical investigations of a hydrogen-assisted laser-induced materials transfer procedure

    International Nuclear Information System (INIS)

    Toet, D.; Smith, P. M.; Sigmon, T. W.; Thompson, M. O.

    2000-01-01

    We present investigations of the mechanisms of a laser-induced transfer technique, which can be used for the spatially selective deposition of materials such as Si. This transfer is effected by irradiating the backside of a hydrogenated amorphous silicon film, deposited on a transparent substrate with an excimer laser pulse. The resulting release and accumulation of hydrogen at the film/substrate interface propels the silicon onto an adjacent receptor wafer. Time-resolved infrared transmission measurements indicate that the amorphous film is melted by the laser pulse and breaks into droplets during ejection. These droplets travel towards the receptor substrate and coalesce upon arrival. The transfer velocity increases as a function of fluence, the rate of increase dropping noticeably around the full melt threshold of the film. At this fluence, the transfer velocity reaches values of around 1000 m/s for typical films. Atomic force microscopy reveals that films transferred below the full melt threshold only partially cover the receptor substrate, while uniform, well-adhering films, which can be smoothed by subsequent laser irradiation, are obtained above it. Transfer of hydrogen-free Si films, on the other hand, does not occur until much higher fluences. The dynamics of the process have been simulated using a semiquantitative numerical model. In this model, hydrogen released from the melt front is instantaneously accumulated at the interface with an initial kinetic energy given by the melting temperature of Si and the enthalpy of solution. The resulting pressure accelerates the Si film, the dynamics of which are modeled using Newtonian mechanics, and the gas cools adiabatically as its kinetic energy is converted to the film's momentum. The results of the calculations are in good agreement with the experimental data. (c) 2000 American Institute of Physics

  18. A general model for the transfer of radioactive materials in terrestrial food chains

    International Nuclear Information System (INIS)

    Simmonds, J.R.; Linsley, G.S.; Jones, J.A.

    1979-09-01

    A general methodology for modelling the transfer of radionuclides in the food chains to man is described. The models are dynamic in nature so that the long-term time dependence of processes in environmental materials can be represented, for example, the build-up of activity concentrations in soils during continuous deposition from atmosphere. Modules for radionuclide migration are described in well-mixed (cultivated) soil and undisturbed soil (pasture). The methods by which the transfer coefficients used in plant and animal modules are derived are also given. The foodstuffs considered are those derived from green vegetables, grain, and root vegetables together with meat and liver products from the cow and sheep and cow dairy products. The dynamic model permits the time dependence of food chain transfer processes to be represented for different land contamination scenarios; in particular, the model can be adapted to represent behaviour following a single deposit. Using the sensitivity of results to the variation of transfer parameters the model can be used to determine the parts of the food chain where improved data would be most effective in increasing the reliability of radiological assessments; a worked example is given. (author)

  19. Hole-Transfer Dependence on Blend Morphology and Energy Level Alignment in Polymer: ITIC Photovoltaic Materials.

    Science.gov (United States)

    Eastham, Nicholas D; Logsdon, Jenna L; Manley, Eric F; Aldrich, Thomas J; Leonardi, Matthew J; Wang, Gang; Powers-Riggs, Natalia E; Young, Ryan M; Chen, Lin X; Wasielewski, Michael R; Melkonyan, Ferdinand S; Chang, Robert P H; Marks, Tobin J

    2018-01-01

    Bulk-heterojunction organic photovoltaic materials containing nonfullerene acceptors (NFAs) have seen remarkable advances in the past year, finally surpassing fullerenes in performance. Indeed, acceptors based on indacenodithiophene (IDT) have become synonymous with high power conversion efficiencies (PCEs). Nevertheless, NFAs have yet to achieve fill factors (FFs) comparable to those of the highest-performing fullerene-based materials. To address this seeming anomaly, this study examines a high efficiency IDT-based acceptor, ITIC, paired with three donor polymers known to achieve high FFs with fullerenes, PTPD3T, PBTI3T, and PBTSA3T. Excellent PCEs up to 8.43% are achieved from PTPD3T:ITIC blends, reflecting good charge transport, optimal morphology, and efficient ITIC to PTPD3T hole-transfer, as observed by femtosecond transient absorption spectroscopy. Hole-transfer is observed from ITIC to PBTI3T and PBTSA3T, but less efficiently, reflecting measurably inferior morphology and nonoptimal energy level alignment, resulting in PCEs of 5.34% and 4.65%, respectively. This work demonstrates the importance of proper morphology and kinetics of ITIC → donor polymer hole-transfer in boosting the performance of polymer:ITIC photovoltaic bulk heterojunction blends. © 2017 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  20. Remotely operated facility for in situ solidification of fissile uranium

    International Nuclear Information System (INIS)

    McGinnis, C.P.; Collins, E.D.; Patton, B.D.

    1986-01-01

    A heavily shielded, remotely operated facility, located within the Radiochemical processing Plant at Oak Ridge National Laboratory (ORNL), has been designed and is being operated to convert approx.1000 kg of fissile uranium (containing approx.75% 235 U, approx.10% 233 U, and approx.140 ppM 232 U) from a nitrate solution (130 g of uranium per L) to a solid oxide form. This project, the Consolidated Edison Uranium Solidification Program (CEUSP), is being carried out in order to prepare a stable uranium form for longterm storage. This paper describes the solidification process selected, the equipment and facilities required, the experimental work performed to ensure successful operation, some problems that were solved, and the initial operations

  1. Fundamental mass transfer modeling of emission of volatile organic compounds from building materials

    Science.gov (United States)

    Bodalal, Awad Saad

    In this study, a mass transfer theory based model is presented for characterizing the VOC emissions from building materials. A 3-D diffusion model is developed to describe the emissions of volatile organic compounds (VOCs) from individual sources. Then the formulation is extended to include the emissions from composite sources (system comprising an assemblage of individual sources). The key parameters for the model (The diffusion coefficient of the VOC in the source material D, and the equilibrium partition coefficient k e) were determined independently (model parameters are determined without the use of chamber emission data). This procedure eliminated to a large extent the need for emission testing using environmental chambers, which is costly, time consuming, and may be subject to confounding sink effects. An experimental method is developed and implemented to measure directly the internal diffusion (D) and partition coefficients ( ke). The use of the method is illustrated for three types of VOC's: (i) Aliphatic Hydrocarbons, (ii) Aromatic Hydrocarbons and ( iii) Aldehydes, through typical dry building materials (carpet, plywood, particleboard, vinyl floor tile, gypsum board, sub-floor tile and OSB). Then correlations for predicting D and ke based solely on commonly available properties such as molecular weight and vapour pressure were proposed for each product and type of VOC. These correlations can be used to estimate the D and ke when direct measurement data are not available, and thus facilitate the prediction of VOC emissions from the building materials using mass transfer theory. The VOC emissions from a sub-floor material (made of the recycled automobile tires), and a particleboard are measured and predicted. Finally, a mathematical model to predict the diffusion coefficient through complex sources (floor adhesive) as a function of time was developed. Then this model (for diffusion coefficient in complex sources) was used to predict the emission rate from

  2. Transfer printing techniques for materials assembly and micro/nanodevice fabrication.

    Science.gov (United States)

    Carlson, Andrew; Bowen, Audrey M; Huang, Yonggang; Nuzzo, Ralph G; Rogers, John A

    2012-10-09

    Transfer printing represents a set of techniques for deterministic assembly of micro-and nanomaterials into spatially organized, functional arrangements with two and three-dimensional layouts. Such processes provide versatile routes not only to test structures and vehicles for scientific studies but also to high-performance, heterogeneously integrated functional systems, including those in flexible electronics, three-dimensional and/or curvilinear optoelectronics, and bio-integrated sensing and therapeutic devices. This article summarizes recent advances in a variety of transfer printing techniques, ranging from the mechanics and materials aspects that govern their operation to engineering features of their use in systems with varying levels of complexity. A concluding section presents perspectives on opportunities for basic and applied research, and on emerging use of these methods in high throughput, industrial-scale manufacturing. Copyright © 2012 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  3. Graphene materials as 2D non-viral gene transfer vector platforms.

    Science.gov (United States)

    Vincent, M; de Lázaro, I; Kostarelos, K

    2017-03-01

    Advances in genomics and gene therapy could offer solutions to many diseases that remain incurable today, however, one of the critical reasons halting clinical progress is due to the difficulty in designing efficient and safe delivery vectors for the appropriate genetic cargo. Safety and large-scale production concerns counter-balance the high gene transfer efficiency achieved with viral vectors, while non-viral strategies have yet to become sufficiently efficient. The extraordinary physicochemical, optical and photothermal properties of graphene-based materials (GBMs) could offer two-dimensional components for the design of nucleic acid carrier systems. We discuss here such properties and their implications for the optimization of gene delivery. While the design of such vectors is still in its infancy, we provide here an exhaustive and up-to-date analysis of the studies that have explored GBMs as gene transfer vectors, focusing on the functionalization strategies followed to improve vector performance and on the biological effects attained.

  4. Enhancing heat capacity of colloidal suspension using nanoscale encapsulated phase-change materials for heat transfer.

    Science.gov (United States)

    Hong, Yan; Ding, Shujiang; Wu, Wei; Hu, Jianjun; Voevodin, Andrey A; Gschwender, Lois; Snyder, Ed; Chow, Louis; Su, Ming

    2010-06-01

    This paper describes a new method to enhance the heat-transfer property of a single-phase liquid by adding encapsulated phase-change nanoparticles (nano-PCMs), which absorb thermal energy during solid-liquid phase changes. Silica-encapsulated indium nanoparticles and polymer-encapsulated paraffin (wax) nanoparticles have been made using colloid method, and suspended into poly-alpha-olefin (PAO) and water for potential high- and low-temperature applications, respectively. The shells prevent leakage and agglomeration of molten phase-change materials, and enhance the dielectric properties of indium nanoparticles. The heat-transfer coefficients of PAO containing indium nanoparticles (30% by mass) and water containing paraffin nanoparticles (10% by mass) are 1.6 and 1.75 times higher than those of corresponding single-phase fluids. The structural integrity of encapsulation allows repeated use of such nanoparticles for many cycles in high heat generating devices.

  5. Natural convection heat transfer enhancement using Microencapsulated Phase-Change-Material slurries

    International Nuclear Information System (INIS)

    Kubo, Shinji; Akino, Norio; Tanaka, Amane; Nakano, Fumihiko; Nagashima, Akira.

    1997-01-01

    The present study investigates natural convection heat transfer from a heated cylinder cooled by a water slurry of Microencapsulated Phase Change Material (MCPCM). A normal paraffin hydrocarbon with carbon number of 18 and melting point of 27.9degC, is microencapsulated by Melamine resin into particles of which average diameter is 9.5μm and specific weight is same as water. The slurry of the MCPCM and water is put into a test apparatus, which is a rectangular enclosure with a heated horizontal cylinder. As the concentrations of PCM in the slurry are changed in 1,3 and 5%, the heat transfer coefficients of the cylinder are larger than that of water as working fluid, by 3,20 and 35% enhancements respectively. (author)

  6. Communications Received from Certain Member States Regarding Guidelines for the Export of Nuclear Material, Equipment and Technology. Nuclear Transfers and Nuclear-Related Dual-Use Transfers

    International Nuclear Information System (INIS)

    1993-04-01

    The Director General has received a Note Verbale dated 5 March 1993 from the Ministry of Foreign Affairs of the Slovak Republic. The purpose of the Note Verbale is to provide information on that Governments' guidelines for Nuclear Transfers and for Transfers of of Nuclear-related Dual-use Equipment, Material and Related Technology. In the light of the wish expressed at the end of each Note Verbale, the text of the Note Verbale is annexed hereto [fr

  7. Communications Received from Certain Member States Regarding Guidelines for the Export of Nuclear Material, Equipment and Technology. Nuclear Transfers and Nuclear-Related Dual-Use Transfers

    International Nuclear Information System (INIS)

    1993-04-01

    The Director General has received a Note Ver bale dated 5 March 1993 from the Ministry of Foreign Affairs of the Slovak Republic. The purpose of the Note Ver bale is to provide information on that Governments' guidelines for Nuclear Transfers and for Transfers of of Nuclear-related Dual-use Equipment, Material and Related Technology. In the light of the wish expressed at the end of each Note Ver bale, the text of the Note Ver bale is annexed hereto

  8. Communications Received from Certain Member States Regarding Guidelines for the Export of Nuclear Material, Equipment and Technology. Nuclear Transfers and Nuclear-Related Dual-Use Transfers

    International Nuclear Information System (INIS)

    1993-04-01

    The Director General has received a Note Verbale dated 5 March 1993 from the Ministry of Foreign Affairs of the Slovak Republic. The purpose of the Note Verbale is to provide information on that Governments' guidelines for Nuclear Transfers and for Transfers of of Nuclear-related Dual-use Equipment, Material and Related Technology. In the light of the wish expressed at the end of each Note Verbale, the text of the Note Verbale is annexed hereto [es

  9. Grey radiative transfer in binary statistical media with material temperature coupling: asymptotic limits

    International Nuclear Information System (INIS)

    Prinja, A.K.; Olson, G.L.

    2005-01-01

    Simplified models for the unconditional ensemble-averaged radiation intensity and material energy are developed for radiative transfer in binary statistical media. Asymptotic analysis is used to construct an effective transport model with homogenized opacities in two limits. In the first, the material properties are assumed to have low contrast on average, and is shown to correctly reproduce the well-known atomic mix model in both time-dependent and equilibrium situations. Our analysis successfully resolves an inconsistency previously noted in the literature with the application of the standard definition of the atomic mix limit to radiative transfer in participating random media. In the second limit considered, the materials are assumed to have highly contrasting opacities, yielding a reduced transport model with effective scattering. The existence of these limits requires the mean chunk sizes to be independent of the photon direction and this creates an ambiguity in the interpretation of the models when the underlying stochastic geometry is comprised of alternating one-dimensional slabs. A consistent one-dimensional setting is defined and the asymptotic models are numerically validated over a broad range of physical parameter values

  10. Capillary-Force-Assisted Clean-Stamp Transfer of Two-Dimensional Materials.

    Science.gov (United States)

    Ma, Xuezhi; Liu, Qiushi; Xu, Da; Zhu, Yangzhi; Kim, Sanggon; Cui, Yongtao; Zhong, Lanlan; Liu, Ming

    2017-11-08

    A simple and clean method of transferring two-dimensional (2D) materials plays a critical role in the fabrication of 2D electronics, particularly the heterostructure devices based on the artificial vertical stacking of various 2D crystals. Currently, clean transfer techniques rely on sacrificial layers or bulky crystal flakes (e.g., hexagonal boron nitride) to pick up the 2D materials. Here, we develop a capillary-force-assisted clean-stamp technique that uses a thin layer of evaporative liquid (e.g., water) as an instant glue to increase the adhesion energy between 2D crystals and polydimethylsiloxane (PDMS) for the pick-up step. After the liquid evaporates, the adhesion energy decreases, and the 2D crystal can be released. The thin liquid layer is condensed to the PDMS surface from its vapor phase, which ensures the low contamination level on the 2D materials and largely remains their chemical and electrical properties. Using this method, we prepared graphene-based transistors with low charge-neutral concentration (3 × 10 10 cm -2 ) and high carrier mobility (up to 48 820 cm 2 V -1 s -1 at room temperature) and heterostructure optoelectronics with high operation speed. Finally, a capillary-force model is developed to explain the experiment.

  11. TRANSFER

    African Journals Online (AJOL)

    This paper reports on further studies on long range energy transfer between curcumine as donor and another thiazine dye, thionine, which is closely related to methylene blue as energy harvester (Figure 1). Since thionine is known to have a higher quantum yield of singlet oxygen sensitization than methylene blue [8], it is ...

  12. Recent advances in energy transfer in bulk and nanoscale luminescent materials: from spectroscopy to applications.

    Science.gov (United States)

    Liu, Xiaofeng; Qiu, Jianrong

    2015-12-07

    Transfer of energy occurs endlessly in our universe by means of radiation. Compared to energy transfer (ET) in free space, in solid state materials the transfer of energy occurs in a rather confined manner, which is usually mediated by real or virtual particles, including not only photons, but also electrons, phonons, and excitons. In the present review, we discuss the recent advances in optical ET by resonance mediated with photons in solid materials as well as their nanoscale counterparts, with focus on the photoluminescence behavior pertaining to ET between optically active centers, such as rare earth (RE) ions. This review begins with a brief discussion on the classification of optical ET together with an overview of the theoretical formulations and experimental method for the examination of ET. We will then present a comprehensive discussion on the ET in practical systems in which normal photoluminescence, upconversion and quantum cutting resulted from ET involving metal ions, QDs, organic species, 2D materials and plasmonic nanostructures. Diverse ET systems are therefore simply categorized into cases of ion-ion interactions and non-ion interactions. Special attention has been paid to the progress in the manipulation of spatially confined ET in nanostructured systems including core-shell structures, as well as the ET in multiple exciton generation found in QDs and organic molecules, which behave quite similarly to resonance ET between metal ion centers. Afterwards, we will discuss the broad spectrum of applications of ET in the aforementioned systems, including solid state lighting, solar energy utilization, bio-imaging and diagnosis, and sensing. In the closing part, along with a short summary, we discuss further research focus regarding the problems and possible future directions of optical ET in solids.

  13. The phenomenon of microscale flow and mass transfer in medicinal herb materials

    Energy Technology Data Exchange (ETDEWEB)

    Yang, J.H.; Di, Q.Q.; Sun, M.D. [Tianjin Univ., Tianjin (China). School of Mechanical Engineering; Zhang, T.J.; Gong, S.X. [Tianjin Inst. of Pharmaceutical Research, Tianjin (China)

    2008-07-01

    Microwave assisted extraction (MAE) is a combination of a microwave technique and conventional solvent extraction used in the modernization of traditional Chinese medicine. The effective component of medicinal herbs is mostly cellular material which can be released via solvent extraction. The material is diffused to solvents via the porous membrane wall. The structure of herb morphology and characteristics of the solute's molecular weight play an important role in the extraction process of target compounds. Astragalus pieces were chosen for this study in which an ultra-filtration membrane method was used to determine the molecular weight distribution characteristics of Astragalus water extraction liquid in the process of MAE. The fine structure of matrix materials was also characterized by scanning election microscopy (SEM). The phenomenon of mass flow and mass transfer in the plant porous media was discussed along with the enhancement mechanism of microwave field on medicinal plant solvent extraction. The results showed that the water-soluble components in the parenchyma cells of Astragalus pieces pass through the plasmodesma with a diameter of 10 nm to adjacent cell, then through an aperture with a diameter of 0.1 {mu}m to 1 {mu}m into a trachea with a diameter of about 10 {mu}m. The water-soluble components then come onto the surface of matrix material and the main solution via the trachea. The main mass transfer occurs by the trachea and its aperture. It was concluded that in order to promote the dissolution of effective components in medicinal herb in the extraction process, a suitable extraction technology is needed to maintain the permeability of transportation tissue and parenchyma in materials. 11 refs., 1 tab., 3 figs.

  14. Materials and Physics Challenges for Spin Transfer Torque Magnetic Random Access Memories

    Energy Technology Data Exchange (ETDEWEB)

    Heinonen, O.

    2014-10-05

    Magnetic random access memories utilizing the spin transfer torque effect for writing information are a strong contender for non-volatile memories scalable to the 20 nm node, and perhaps beyond. I will here examine how these devices behave as the device size is scaled down from 70 nm size to 20 nm. As device sizes go below ~50 nm, the size becomes comparable to intrinsic magnetic length scales and the device behavior does not simply scale with size. This has implications for the device design and puts additional constraints on the materials in the device.

  15. Impacts on health and safety from transfer/consolidation of nuclear materials and hazardous chemicals

    International Nuclear Information System (INIS)

    Gallucci, R.H.V.

    1994-11-01

    Environmental restoration plans at the US Department of Energy (USDOE) Hanford Site calls for transfer/consolidation of ''targets/threats,'' namely nuclear materials and hazardous chemicals. Reductions in the health and safety hazards will depend on the plans implemented. Pacific Northwest Laboratory (PNL) estimated these potential impacts, assuming implementation of the current reference plan and employing ongoing risk and safety analyses. The results indicated the potential for ''significant'' reductions in health and safety hazards in the long term (> 25 years) and a potentially ''noteworthy'' reduction in health hazard in the short term (≤ 25 years)

  16. Coupled heat transfer model and experiment study of semitransparent barrier materials in aerothermal environment

    Science.gov (United States)

    Wang, Da-Lin; Qi, Hong

    Semi-transparent materials (such as IR optical windows) are widely used for heat protection or transfer, temperature and image measurement, and safety in energy , space, military, and information technology applications. They are used, for instance, ceramic coatings for thermal barriers of spacecrafts or gas turbine blades, and thermal image observation under extreme or some dangerous environments. In this paper, the coupled conduction and radiation heat transfer model is established to describe temperature distribution of semitransparent thermal barrier medium within the aerothermal environment. In order to investigate this numerical model, one semi-transparent sample with black coating was considered, and photothermal properties were measured. At last, Finite Volume Method (FVM) was used to solve the coupled model, and the temperature responses from the sample surfaces were obtained. In addition, experiment study was also taken into account. In the present experiment, aerodynamic heat flux was simulated by one electrical heater, and two experiment cases were designed in terms of the duration of aerodynamic heating. One case is that the heater irradiates one surface of the sample continually until the other surface temperature up to constant, and the other case is that the heater works only 130 s. The surface temperature responses of these two cases were recorded. Finally, FVM model of the coupling conduction-radiation heat transfer was validated based on the experiment study with relative error less than 5%.

  17. Coolant material effect on the heat transfer rates of the molten metal pool with solidification

    International Nuclear Information System (INIS)

    Cho, Jae Seon; Suh, Kune Y.; Chung, Chang Hyun; Park, Rae Joon; Kim, Sang Baik

    1998-01-01

    Experimental studies on heat transfer and solidification of the molten metal pool with overlying coolant with boiling were performed. The simulant molten pool material is tin (Sn) with the melting temperature of 232 degree C. Demineralized water and R113 are used as the working coolant. This work examines the crust formation and the heat transfer characteristics of the molten metal pool immersed in the boiling coolant. The Nusselt number and the Rayleigh number in the molten metal pool region of this study are compared between the water coolant case and the R113 coolant case. The experimental results for the water coolant are higher than those for R113. Also, the empirical relationship of the Nusselt number and the Rayleigh number is compared with the literature correlations measured from mercury. The present experimental results are higher than the literature correlations. It is believed that this discrepancy is caused by the effect of the heat loss to the environment on the natural convection heat transfer in the molten pool

  18. Transfer of microorganisms, including Listeria monocytogenes, from various materials to beef.

    Science.gov (United States)

    Midelet, Graziella; Carpentier, Brigitte

    2002-08-01

    The quantity of microorganisms that may be transferred to a food that comes into contact with a contaminated surface depends on the density of microorganisms on the surface and on the attachment strengths of the microorganisms on the materials. We made repeated contacts between pieces of meat and various surfaces (stainless steel and conveyor belt materials [polyvinyl chloride and polyurethane]), which were conditioned with meat exudate and then were contaminated with Listeria monocytogenes, Staphylococcus sciuri, Pseudomonas putida, or Comamonas sp. Attachment strengths were assessed by the slopes of the two-phase curves obtained by plotting the logarithm of the number of microorganisms transferred against the order number of the contact. These curves were also used to estimate the microbial population on the surface by using the equation of A. Veulemans, E. Jacqmain, and D. Jacqmain (Rev. Ferment. Ind. Aliment. 25:58-65, 1970). The biofilms were characterized according to their physicochemical surface properties and structures. Their exopolysaccharide-producing capacities were assessed from biofilms grown on polystyrene. The L. monocytogenes biofilms attached more strongly to polymers than did the other strains, and attachment strength proved to be weaker on stainless steel than on the two polymers. However, in most cases, it was the population of the biofilms that had the strongest influence on the total number of CFU detached. Although attachment strengths were weaker on stainless steel, this material, carrying a smaller population of bacteria, had a weaker contaminating capacity. In most cases the equation of Veulemans et al. revealed more bacteria than did swabbing the biofilms, and it provided a better assessment of the contaminating potential of the polymeric materials studied here.

  19. Requirements for materials of dispersion fuel elements

    International Nuclear Information System (INIS)

    Samojlov, A.G.; Kashtanov, A.I.; Volkov, V.S.

    1982-01-01

    Requirements for materials of dispersion fuel elements are considered. The necessity of structural and fissile materials compatibility at maximum permissible operation temperatures and temperatures arising in a fuel element during manufacture is pointed out. The fuel element structural material must be ductile, possess high mechanical strength minimum neutron absorption cross section, sufficient heat conductivity, good corrosion resistance in a coolant and radiation resistance. The fissile material must have high fissile isotope concentration, radiation resistance, high thermal conductivity, certain porosity high melting temperature must not change the composition under irradiation

  20. Stoichiometric transfer of material in the infrared pulsed laser deposition of yttrium doped Bi-2212 films

    International Nuclear Information System (INIS)

    De Vero, Jeffrey C.; Blanca, Glaiza Rose S.; Vitug, Jaziel R.; Garcia, Wilson O.; Sarmago, Roland V.

    2011-01-01

    Highlights: → This work describes the stoichiometric transfer of Y-doped Bi-2212 during IR-PLD. → As-deposited films show spheroidal morphology with similar composition as the target. Relatively flat and highly c-axis oriented films were obtained after heat treatment. → IR-PLD can be a viable technique in growing other high Tc superconducting materials. - Abstract: Films of Y-doped Bi-2212 were successfully grown on MgO (1 0 0) substrates by infrared pulsed laser deposition (IR-PLD). With post-heat treatments, smooth and highly c-axis oriented films were obtained. The average compositions of the films have the same stoichiometry as the target. Y content is also preserved on the grown films at all doping levels. The electrical properties of the grown Y-doped Bi-2212 films exhibit the expected electrical properties of the bulk Y-doped Bi-2212. This is attributed to the stoichiometric transfer of material by IR-PLD.

  1. Correlation between charge transfer and exchange coupling in carbon-based magnetic materials

    Energy Technology Data Exchange (ETDEWEB)

    Nguyen, Anh Tuan, E-mail: tuanna@hus.edu.vn [Faculty of Physics, VNU University of Science, 334 Nguyen Trai, Thanh Xuan, Ha Noi (Viet Nam); Science and Technology Department, Vietnam National University, Hanoi, 144 Xuan Thuy, Cau Giay, Hanoi (Viet Nam); Japan Advanced Institute of Science and Technology, 1-1, Asahidai, Nomi, Ishikawa, 923-1292 Japan (Japan); Nguyen, Van Thanh; Nguyen, Huy Sinh [Faculty of Physics, VNU University of Science, 334 Nguyen Trai, Thanh Xuan, Ha Noi (Viet Nam); Pham, Thi Tuan Anh [Faculty of Physics, VNU University of Science, 334 Nguyen Trai, Thanh Xuan, Ha Noi (Viet Nam); Faculty of Science, College of Hai Duong, Nguyen Thi Due, Hai Duong (Viet Nam); Do, Viet Thang [Faculty of Physics, VNU University of Science, 334 Nguyen Trai, Thanh Xuan, Ha Noi (Viet Nam); Faculty of Science, Haiphong University, 171 Phan Dang Luu, Kien An, Hai Phong (Viet Nam); Dam, Hieu Chi [Japan Advanced Institute of Science and Technology, 1-1, Asahidai, Nomi, Ishikawa, 923-1292 Japan (Japan)

    2015-10-15

    Several forms of carbon-based magnetic materials, i.e. single radicals, radical dimers, and alternating stacks of radicals and diamagnetic molecules, have been investigated using density-functional theory with dispersion correction and full geometry optimization. Our calculated results demonstrate that the C{sub 31}H{sub 15} (R{sub 4}) radical has a spin of ½. However, in its [R{sub 4}]{sub 2} dimer structure, the net spin becomes zero due to antiferromagnetic spin-exchange between radicals. To avoid antiferromagnetic spin-exchange of identical face-to-face radicals, eight alternating stacks, R{sub 4}/D{sub 2m}/R{sub 4} (with m = 3-10), were designed. Our calculated results show that charge transfer (Δn) between R{sub 4} radicals and the diamagnetic molecule D{sub 2m} occurs with a mechanism of spin exchange (J) in stacks. The more electrons that transfer from R{sub 4} to D{sub 2m}, the stronger the ferromagnetic spin-exchange in stacks. In addition, our calculated results show that Δn can be tailored by adjusting the electron affinity (E{sub a}) of D{sub 2m}. The correlation between Δn, E{sub a}, m, and J is discussed. These results give some hints for the design of new ferromagnetic carbon-based materials.

  2. Charge transfer processes in hybrid solar cells composed of amorphous silicon and organic materials

    Energy Technology Data Exchange (ETDEWEB)

    Schaefer, Sebastian; Neher, Dieter [Universitaet Potsdam, Inst. Physik u. Astronomie, Karl-Liebknecht-Strasse 24/25, 14467 Potsdam-Golm (Germany); Schulze, Tim; Korte, Lars [Helmholtz Zentrum Berlin, Inst. fuer Silizium Photovoltaik, Kekulestrasse 5, 12489 Berlin (Germany)

    2011-07-01

    The efficiency of hybrid solar cells composed of organic materials and amorphous hydrogenated silicon (a-Si:H) strongly depends upon the efficiency of charge transfer processes at the inorganic-organic interface. We investigated the performance of devices comprising an ITO/a-Si:H(n-type)/a-Si:H(intrinsic)/organic/metal multilayer structure and using two different organic components: zinc phthalocyanine (ZnPc) and poly(3-hexylthiophene) (P3HT). The results show higher power conversion- and quantum efficiencies for the P3HT based cells, compared to ZnPc. This can be explained by larger energy-level offset at the interface between the organic layer and a-Si:H, which facilitates hole transfer from occupied states in the valence band tail to the HOMO of the organic material and additionally promotes exciton splitting. The performance of the a-Si:H/P3HT cells can be further improved by treatment of the amorphous silicon surface with hydrofluoric acid (HF) and p-type doping of P3HT with F4TCNQ. The improved cells reached maximum power conversion efficiencies of 1%.

  3. Hydration induced material transfer in membranes of osmotic pump tablets measured by synchrotron radiation based FTIR.

    Science.gov (United States)

    Wu, Li; Yin, Xianzhen; Guo, Zhen; Tong, Yajun; Feng, Jing; York, Peter; Xiao, Tiqiao; Chen, Min; Gu, Jingkai; Zhang, Jiwen

    2016-03-10

    Osmotic pump tablets are reliable oral controlled drug delivery systems based on their semipermeable membrane coating. This research used synchrotron radiation-based Fourier transform infrared (SR-FTIR) microspectroscopy and imaging to investigate the hydration induced material transfer in the membranes of osmotic pump tablets. SR-FTIR was applied to record and map the chemical information of a micro-region of the membranes, composed of cellulose acetate (CA, as the water insoluble matrix) and polyethylene glycol (PEG, as the soluble pore forming agent and plasticizing agent). The microstructure and chemical change of membranes hydrated for 0, 5, 10 and 30min were measured using SR-FTIR, combined with scanning electronic microscopy and atom force microscopy. The SR-FTIR microspectroscopy results indicated that there was a major change at the absorption range of 2700-3100cm(-1) in the membranes after different periods of hydration time. The absorption bands at 2870-2880cm(-1) and 2950-2960cm(-1) were assigned to represent CA and PEG, respectively. The chemical group signal distribution illustrated by the ratio of PEG to CA demonstrated that the trigger of drug release in the preliminary stage was due to the rapid transfer of PEG into liquid medium with a sharp decrease of PEG in the membranes. The SR-FTIR mapping results have demonstrated the hydration induced material transfer in the membranes of osmotic pump tablets and enabled reassessment of the drug release mechanism of membrane controlled osmotic pump systems. Copyright © 2016 Elsevier B.V. All rights reserved.

  4. UF6 fissile mass flow simulation at Oak Ridge National Laboratory

    International Nuclear Information System (INIS)

    Mihalczo, J.T.; March-Leuba, J.; Valentine, T.E.; Mattingly, J.K.; Uckan, T.; McEvers, J.A.

    1997-01-01

    Basis for measuring fissile mass flow in slurries, liquid, and gaseous streams is activation of a fissile stream by neutrons and then detection of delayed radiation from resulting fission products. This paper describes recent simulation measurements with the first prototype of the system for fissile mass flow measurements with HEU UF 6 gas for use in blenddown facilities. Theory was only 15% higher than actual measured; thus calibration factor would be 0.85. This simulation of HEU gas flow confirms well the understanding of the physical phenomena associated with this measurement system

  5. Accelerator based production of fissile nuclides, threshold uranium price and perspectives

    International Nuclear Information System (INIS)

    Djordjevic, D.; Knapp, V.

    1988-01-01

    Accelerator breeder system characteristics are considered in this work. One such system which produces fissile nuclides can supply several thermal reactors with fissile fuel, so this system becomes analogous to an uranium enrichment facility with difference that fissile nuclides are produced by conversion of U-238 rather than by separation from natural uranium. This concept, with other long-term perspective for fission technology on the basis of development only one simpler technology. The influence of basic system characteristics on threshold uranium price is examined. Conditions for economically acceptable production are established. (author)

  6. Ceramics for Molten Materials Containment, Transfer and Handling on the Lunar Surface

    Science.gov (United States)

    Standish, Evan; Stefanescu, Doru M.; Curreri, Peter A.

    2009-01-01

    As part of a project on Molten Materials Transfer and Handling on the Lunar Surface, molten materials containment samples of various ceramics were tested to determine their performance in contact with a melt of lunar regolith simulant. The test temperature was 1600 C with contact times ranging from 0 to 12 hours. Regolith simulant was pressed into cylinders with the approximate dimensions of 1.25 dia x 1.25cm height and then melted on ceramic substrates. The regolith-ceramic interface was examined after processing to determine the melt/ceramic interaction. It was found that the molten regolith wetted all oxide ceramics tested extremely well which resulted in chemical reaction between the materials in each case. Alumina substrates were identified which withstood contact at the operating temperature of a molten regolith electrolysis cell (1600 C) for eight hours with little interaction or deformation. This represents an improvement over alumina grades currently in use and will provide a lifetime adequate for electrolysis experiments lasting 24 hours or more. Two types of non-oxide ceramics were also tested. It was found that they interacted to a limited degree with the melt resulting in little corrosion. These ceramics, Sic and BN, were not wetted as well as the oxides by the melt, and so remain possible materials for molten regolith handling. Tests wing longer holding periods and larger volumes of regolith are necessary to determine the ultimate performance of the tested ceramics.

  7. Effects of an applied voltage on direct interspecies electron transfer via conductive materials for methane production.

    Science.gov (United States)

    Lee, Jung-Yeol; Park, Jeong-Hoon; Park, Hee-Deung

    2017-10-01

    Direct interspecies electron transfer (DIET) between exoelectrogenic bacteria and methanogenic archaea via conductive materials is reported as an efficient method to produce methane in anaerobic organic waste digestion. A voltage can be applied to the conductive materials to accelerate the DIET between two groups of microorganisms to produce methane. To evaluate this hypothesis, two sets of anaerobic serum bottles with and without applied voltage were used with a pair of graphite rods as conductive materials to facilitate DIET. Initially, the methane production rate was similar between the two sets of serum bottles, and later the serum bottles with an applied voltage of 0.39V showed a 168% higher methane production rate than serum bottles without an applied voltage. In cyclic voltammograms, the characteristic redox peaks for hydrogen and acetate oxidation were identified in the serum bottles with an applied voltage. In the microbial community analyses, hydrogenotrophic methanogens (e.g. Methanobacterium) were observed to be abundant in serum bottles with an applied voltage, while methanogens utilizing carbon dioxide (e.g., Methanosaeta and Methanosarcina) were dominant in serum bottles without an applied voltage. Taken together, the applied voltage on conductive materials might not be effective to promote DIET in methane production. Instead, it appeared to generate a condition for hydrogenotrophic methanogenesis. Copyright © 2017 Elsevier Ltd. All rights reserved.

  8. Quantification of ion or atom transfer phenomena in materials implanted by nuclear methods

    International Nuclear Information System (INIS)

    Oudadesse, Hassane

    1998-01-01

    Knowledge of transfer of the constituents of a system from regions of higher to lower concentration is of interest for implanted bio-materials. It allows determining the rate at which this material is integrated in a living material. To evaluate the ossification kinetics and to study the bio-functionality in corals of Ca and Sr, irradiations with a 10 13 n.cm -2 .s -1 was performed, followed by the examination of changes in the localization of these elements. By using PIXE analysis method the distribution of Ca, P, Sr, Zn and Fe in the implant, bone and bone-implant interfaces were determined. Thus, it was shown that resorption of coral in sheep is achieved in 5 months after implantation and is identical to the cortical tissues 4 months after implantation in animals as for instance in hares. We have analyzed the tissues from around the prostheses extracted from patients. The samples were calcined and reduced to powder weighting some milligrams. We have adopted for this study the PIXE analysis method. The samples were irradiated by a proton beam of 3 MeV and about 400 μm diameter. The results show the presence of the elements Ti, Fe, Cr, Ni or Zn according to the type of the implanted prosthesis. This dispersal of the metallic ions and atoms contaminate the tissues. The transfer factors translate the exchanges between bone and the implanted material. The solvatation phenomenon and the electric charge equilibrium explain the transfer order of cations Mg 2+ , Ca 2+ and Sr 2+ and of the anion PO 4 3- . We have also determined these factors for the elements Ti, Cr and Ni. An original technique to study the bone bio-functionality was used. Use of phosphate derivatives labelled by 99m Tc allows obtaining information about the fixation of radioactive tracer. It was found that only after the eighth month at the implantation the neo-formed bone fixes the MDP (methyl diphosphate) labelled by 99m Tc in a similar way as in the control sample. Starting from this moment the

  9. X-ray spectroscopy studies of nonradiative energy transfer processes in luminescent lanthanide materials

    Science.gov (United States)

    Pacold, Joseph I.

    Luminescent materials play important roles in energy sciences, through solid state lighting and possible applications in solar energy utilization, and in biomedical research and applications, such as in immunoassays and fluorescence microscopy. The initial excitation of a luminescent material leads to a sequence of transitions between excited states, ideally ending with the emission of one or more optical-wavelength photons. It is essential to understand the microscopic physics of this excited state cascade in order to rationally design materials with high quantum efficiencies or with other fine-tuning of materials response. While optical-wavelength spectroscopies have unraveled many details of the energy transfer pathways in luminescent materials, significant questions remain open for many lanthanide-based luminescent materials. For organometallic dyes in particular, quantum yields remain limited in comparison with inorganic phosphors. This dissertation reports on a research program of synchrotron x-ray studies of the excited state electronic structure and energy-relaxation cascade in trivalent lanthanide phosphors and dyes. To this end, one of the primary results presented here is the first time-resolved x-ray absorption near edge spectroscopy studies of the transient 4f excited states in lanthanide-activated luminescent dyes and phosphors. This is a new application of time-resolved x-ray absorption spectroscopy that makes it possible to directly observe and, to some extent, quantify intramolecular nonradiative energy transfer processes. We find a transient increase in 4f spectral weight associated with an excited state confined to the 4f shell of trivalent Eu. This result implies that it is necessary to revise the current theoretical understanding of 4f excitation in trivalent lanthanide activators: either transient 4f-5d mixing effects are much stronger than previously considered, or else the lanthanide 4f excited state has an unexpectedly large contribution

  10. Donor–acceptor graphene-based hybrid materials facilitating photo-induced electron-transfer reactions

    Directory of Open Access Journals (Sweden)

    Anastasios Stergiou

    2014-09-01

    Full Text Available Graphene research and in particular the topic of chemical functionalization of graphene has exploded in the last decade. The main aim is to increase the solubility and thereby enhance the processability of the material, which is otherwise insoluble and inapplicable for technological applications when stacked in the form of graphite. To this end, initially, graphite was oxidized under harsh conditions to yield exfoliated graphene oxide sheets that are soluble in aqueous media and amenable to chemical modifications due to the presence of carboxylic acid groups at the edges of the lattice. However, it was obvious that the high-defect framework of graphene oxide cannot be readily utilized in applications that are governed by charge-transfer processes, for example, in solar cells. Alternatively, exfoliated graphene has been applied toward the realization of some donor–acceptor hybrid materials with photo- and/or electro-active components. The main body of research regarding obtaining donor–acceptor hybrid materials based on graphene to facilitate charge-transfer phenomena, which is reviewed here, concerns the incorporation of porphyrins and phthalocyanines onto graphene sheets. Through illustrative schemes, the preparation and most importantly the photophysical properties of such graphene-based ensembles will be described. Important parameters, such as the generation of the charge-separated state upon photoexcitation of the organic electron donor, the lifetimes of the charge-separation and charge-recombination as well as the incident-photon-to-current efficiency value for some donor–acceptor graphene-based hybrids, will be discussed.

  11. Impact of cementitious materials decalcification on transfer properties: application to radioactive waste deep repository

    International Nuclear Information System (INIS)

    Perlot, C.

    2005-09-01

    Cementitious materials have been selected to compose the engineering barrier system (EBS) of the French radioactive waste deep repository, because of concrete physico-chemical properties: the hydrates of the cementitious matrix and the pH of the pore solution contribute to radionuclides retention; furthermore the compactness of these materials limits elements transport. The confinement capacity of the system has to be assessed while a period at least equivalent to waste activity (up to 100.000 years). His durability was sustained by the evolution of transfer properties in accordance with cementitious materials decalcification, alteration that expresses structure long-term behavior. Then, two degradation modes were carried out, taking into account the different physical and chemical solicitations imposed by the host formation. The first mode, a static one, was an accelerated decalcification test using nitrate ammonium solution. It replicates the EBS alteration dues to underground water. Degradation kinetic was estimated by the amount of calcium leached and the measurement of the calcium hydroxide dissolution front. To evaluate the decalcification impact, samples were characterized before and after degradation in term of microstructure (porosity, pores size distribution) and of transfer properties (diffusivity, gas and water permeability). The influence of cement nature (ordinary Portland cement, blended cement) and aggregates type (lime or siliceous) was observed: experiments were repeated on different mortars mixes. On this occasion, an essential reflection on this test metrology was led. The second mode, a dynamical degradation, was performed with an environmental permeameter. It recreates the EBS solicitations ensured during the re-saturation period, distinguished by the hydraulic pressure imposed by the geologic layer and the waste exothermicity. This apparatus, based on triaxial cell functioning, allows applying on samples pressure drop between 2 and 10 MPa and

  12. Calculated nuclide production yields in relativistic collisions of fissile nuclei

    Energy Technology Data Exchange (ETDEWEB)

    Benlliure, J.; Schmidt, K.H. [Gesellschaft fuer Schwerionenforschung mbH, Darmstadt (Germany); Grewe, A.; Jong, M. de [Technische Univ. Darmstadt (Germany). Inst. fuer Kernphysik; Zhdanov, S. [AN Kazakhskoj SSR, Alma-Ata (USSR). Inst. Yadernoj Fiziki

    1997-11-01

    A model calculation is presented which predicts the complex nuclide distribution resulting from peripheral relativistic heavy-ion collisions involving fissile nuclei. The model is based on a modern version of the abrasion-ablation model which describes the formation of excited prefragments due to the nuclear collisions and their consecutive decay. The competition between the evaporation of different light particles and fission is computed with an evaporation code which takes dissipative effects and the emission of intermediate-mass fragments into account. The nuclide distribution resulting from fission processes is treated by a semiempirical description which includes the excitation-energy dependent influence of nuclear shell effects and pairing correlations. The calculations of collisions between {sup 238}U and different reaction partners reveal that a huge number of isotopes of all elements up to uranium is produced. The complex nuclide distribution shows the characteristics of fragmentation, mass-asymmetric low-energy fission and mass-symmetric high-energy fission. The yields of the different components for different reaction partners are studied. Consequences for technical applications are discussed. (orig.)

  13. Physics design of fissile mass-flow monitoring system

    International Nuclear Information System (INIS)

    Mattingly, J.K.; March-Leuba, J.; Valentine, T.E.; Mihalczo, J.T.; Uckan, T.

    1997-01-01

    The system measures the flow rate and uranium-235 content in liquid or gas streams; it does not penetrate the process piping. A moderated fission neutron source is used to periodicially introduce a burst of thermal neutrons into the fluid stream to induce fission; delayed gamma emissions from the resulting fission fragments are detected by high-efficiency scintillators downstream of the neutron source. The fluid flow rate is measure from the time between initiation of the thermal neutron burst and detection of the fission product gamma emissions, and the U-235 content is inferred from the intensity of the gamma burst detected. Design of the fissile mass flow monitor requires satisfaction of several competing constraints. Efficient operation of the monitor requires that source-induced fission rate and detection efficiency be maximized while the source-induced background rate is simultaneoulsy minimized. Near optical nuclear design of the system was achieved using numerous Monte Carlo calculations and measurements. This paper addresses calculational aspects of the physics design for the system applied to UF 6 gas

  14. Economic evaluation of fissile fuel production using resistive magnet tokamaks

    International Nuclear Information System (INIS)

    Doyle, J.C. Jr.

    1985-06-01

    The application of resistive magnet tokamaks to fissile fuel production has been studied. Resistive magnets offer potential advantages over superconducting magnets in terms of robustness, less technology development required and possibility of demountable joints. Optimization studies within conservatively specified constraints for a compact machine result in a major radius of 3.81 m and 618 MW fusion power and a blanket space envelope of 0.35 m inboard and 0.75 m outboard. This machine is called the Resistive magnet Tokamak Fusion Breeder (RTFB). A computer code was developed to estimate the cost of the resistive magnet tokamak breeder. This code scales from STARFIRE values where appropriate and calculates costs of other systems directly. The estimated cost of the RTFB is $3.01 B in 1984 dollars. The cost of electricity on the same basis as STARFIRE is 42.4 mills/kWhre vs 44.9 mills/kWhre for STARFIRE (this does not include the fuel value or fuel cycle costs for the RTFB). The breakeven cost of U 3 O 8 is $150/lb when compared to a PWR on the once through uranium fuel cycle with no inflation and escalation. On the same basis, the breakeven cost for superconducting tokamak and tandem mirror fusion breeders is $160/lb and $175/lb. Thus, the RTFB appears to be competitive in breakeven U 3 O 8 cost with superconducting magnet fusion breeders and offers the potential advantages of resistive magnet technology

  15. Simulator for an Accelerator-Driven Subcritical Fissile Solution System

    International Nuclear Information System (INIS)

    Klein, Steven Karl; Day, Christy M.; Determan, John C.

    2015-01-01

    LANL has developed a process to generate a progressive family of system models for a fissile solution system. This family includes a dynamic system simulation comprised of coupled nonlinear differential equations describing the time evolution of the system. Neutron kinetics, radiolytic gas generation and transport, and core thermal hydraulics are included in the DSS. Extensions to explicit operation of cooling loops and radiolytic gas handling are embedded in these systems as is a stability model. The DSS may then be converted to an implementation in Visual Studio to provide a design team the ability to rapidly estimate system performance impacts from a variety of design decisions. This provides a method to assist in optimization of the system design. Once design has been generated in some detail the C++ version of the system model may then be implemented in a LabVIEW user interface to evaluate operator controls and instrumentation and operator recognition and response to off-normal events. Taken as a set of system models the DSS, Visual Studio, and LabVIEW progression provides a comprehensive set of design support tools.

  16. Simulator for an Accelerator-Driven Subcritical Fissile Solution System

    Energy Technology Data Exchange (ETDEWEB)

    Klein, Steven Karl [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Day, Christy M. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Determan, John C. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-09-14

    LANL has developed a process to generate a progressive family of system models for a fissile solution system. This family includes a dynamic system simulation comprised of coupled nonlinear differential equations describing the time evolution of the system. Neutron kinetics, radiolytic gas generation and transport, and core thermal hydraulics are included in the DSS. Extensions to explicit operation of cooling loops and radiolytic gas handling are embedded in these systems as is a stability model. The DSS may then be converted to an implementation in Visual Studio to provide a design team the ability to rapidly estimate system performance impacts from a variety of design decisions. This provides a method to assist in optimization of the system design. Once design has been generated in some detail the C++ version of the system model may then be implemented in a LabVIEW user interface to evaluate operator controls and instrumentation and operator recognition and response to off-normal events. Taken as a set of system models the DSS, Visual Studio, and LabVIEW progression provides a comprehensive set of design support tools.

  17. Fast-neutron capture in fissile and fertile nuclides

    International Nuclear Information System (INIS)

    Peelle, R.W.

    1982-01-01

    Extensive graphical and numerical presentations, available to the working group, assisted us in exploring the rich data base established through the labors of many skilled persons. Consistent with the meeting setting, the working group discussion concentrated on data for fast-breeder reactor (FBR) applications. All but 1 to 3% of the magnitude of cross section sensitivities of FBR parameters come from the energy region below approx. = 1.5 MeV, so the statistical model is the relevant theoretical concept. The Meeting emphasizes energies above approx. = 10 keV where resonance fluctuations are not a dominant factor. However, we should remember that approximately half the FBR sensitivity to 238 U capture data, as relfected in integral parameters, lies below 25 keV where resonance fluctuations are strong and resonance self-protection is a most important consideration in reactor physics. There are similar low-energy aspects to 239 Pu capture in that approx. = 30% of the FBR-parameter data sensitivity lies below approx. = 4 keV. Even with the discussion largely cofined to the approx. = 10 to 1500 keV region, the working group could only scratch the surface of the available body of information. The reader is referred to the papers presented at the Meeting and to the references contained therein in order to obtain a more detailed understanding of current issues related to fissile and fertile fast-neutron capture

  18. Photosynthesis Revisited: Optimization of Charge and Energy Transfer in Quantum Materials

    Science.gov (United States)

    Gabor, Nathaniel

    2014-03-01

    The integration of new nano- and molecular-scale quantum materials into ultra-efficient energy harvesting devices presents significant scientific challenges. Of the many challenges, the most difficult is achieving high photon-to-electron conversion efficiency while maintaining broadband absorption. Due to exciton effects, devices composed of quantum materials may allow near-unity optical absorption efficiency yet require the choice of precisely one fundamental energy (HOMO-LUMO gap). To maximize absorption, the simplest device would absorb at the peak of the solar spectrum, which spans the visible wavelengths. If the peak of the solar spectrum spans the visible wavelengths, then why are terrestrial plants green? Here, I discuss a physical model of photosynthetic absorption and photoprotection in which the cell utilizes active feedback to optimize charge and energy transfer, thus maximizing stored energy rather than absorption. This model, which addresses the question of terrestrial greenness, is supported by several recent results that have begun to unravel the details of photoprotection in higher plants. More importantly, this model indicates a novel route for the design of next-generation energy harvesting systems based on nano- and molecular-scale quantum materials.

  19. The contact heat transfer between the heating plate and granular materials in rotary heat exchanger under overloaded condition

    Directory of Open Access Journals (Sweden)

    Luanfang Duan

    2018-03-01

    Full Text Available In the present work, the contact heat transfer between the granular materials and heating plates inside plate rotary heat exchanger (PRHE was investigated. The heat transfer coefficient is dominated by the contact heat transfer coefficient at hot wall surface of the heating plates and the heat penetration inside the solid bed. A plot scale PRHE with a diameter of Do = 273 mm and a length of L = 1000 mm has been established. Quartz sand with dp = 2 mm was employed as the experimental material. The operational parameters were in the range of ω = 1 – 8 rpm, and F = 15, 20, 25, 30%, and the effect of these parameters on the time-average contact heat transfer coefficient was analyzed. The time-average contact heat transfer coefficient increases with the increase of rotary speed, but decreases with the increase of the filling degree. The measured data of time-average heat transfer coefficients were compared with theoretical calculations from Schlünder’s model, a good agreement between the measurements and the model could be achieved, especially at a lower rotary speed and filling degree level. The maximum deviation between the calculated data and the experimental data is approximate 10%. Keywords: Rotary heat exchanger, Contact heat transfer, Granular material, Heating plate, Overloaded

  20. Thermophysical and heat transfer properties of phase change material candidate for waste heat transportation system

    Science.gov (United States)

    Kaizawa, Akihide; Maruoka, Nobuhiro; Kawai, Atsushi; Kamano, Hiroomi; Jozuka, Tetsuji; Senda, Takeshi; Akiyama, Tomohiro

    2008-05-01

    A waste heat transportation system trans-heat (TH) system is quite attractive that uses the latent heat of a phase change material (PCM). The purpose of this paper is to study the thermophysical properties of various sugars and sodium acetate trihydrate (SAT) as PCMs for a practical TH system and the heat transfer property between PCM selected and heat transfer oil, by using differential scanning calorimetry (DSC), thermogravimetry-differential thermal analysis (TG-DTA) and a heat storage tube. As a result, erythritol, with a large latent heat of 344 kJ/kg at melting point of 117°C, high decomposition point of 160°C and excellent chemical stability under repeated phase change cycles was found to be the best PCM among them for the practical TH system. In the heat release experiments between liquid erythritol and flowing cold oil, we observed foaming phenomena of encapsulated oil, in which oil droplet was coated by solidification of PCM.

  1. 3D Multiphysical Modelling of Fluid Dynamics and Mass Transfer in Laser Welding of Dissimilar Materials

    Directory of Open Access Journals (Sweden)

    Jiazhou Wu

    2018-06-01

    Full Text Available A three-dimensional multiphysical transient model was developed to investigate keyhole formation, weld pool dynamics, and mass transfer in laser welding of dissimilar materials. The coupling of heat transfer, fluid flow, keyhole free surface evolution, and solute diffusion between dissimilar metals was simulated. The adaptive heat source model was used to trace the change of keyhole shape, and the Rayleigh scattering of the laser beam was considered. The keyhole wall was calculated using the fluid volume equation, primarily considering the recoil pressure induced by metal evaporation, surface tension, and hydrostatic pressure. Fluid flow, diffusion, and keyhole formation were considered simultaneously in mass transport processes. Welding experiments of 304L stainless steel and industrial pure titanium TA2 were performed to verify the simulation results. It is shown that spatters are shaped during the welding process. The thickness of the intermetallic reaction layer between the two metals and the diffusion of elements in the weld are calculated, which are important criteria for welding quality. The simulation results correspond well with the experimental results.

  2. Transfer of radioactive materials in the fuel cycle. Transportation systems, transportation volume and radiation protection

    International Nuclear Information System (INIS)

    Schwarz, G.

    1997-01-01

    No other aspect of the carriage of hazardous goods has been provoking such long-lived concern in the general public and in the press during the last few years as the transport of spent nuclear fuels and high-level radioactive wastes to the storage facility at Gorleben. One reason for this controversy, besides clear-cut opposition in principal against such transfer activities, is the fact that there is an information gap, so that large parts of the population are not well informed about the relevant legal safety requirements and obligations governing such transports. The article therefore tries to fill this gap, presenting information on the number and necessity of transports of radioactive materials in the nuclear fuel cycle, the relevant scenarios, the transportation systems and packing and shielding requirements, as well as information on the radiological classification and hazardousness of waste forms. (Orig.) [de

  3. Nanostructured 2D cellular materials in silicon by sidewall transfer lithography NEMS

    Science.gov (United States)

    Syms, Richard R. A.; Liu, Dixi; Ahmad, Munir M.

    2017-07-01

    Sidewall transfer lithography (STL) is demonstrated as a method for parallel fabrication of 2D nanostructured cellular solids in single-crystal silicon. The linear mechanical properties of four lattices (perfect and defected diamond; singly and doubly periodic honeycomb) with low effective Young’s moduli and effective Poisson’s ratio ranging from positive to negative are modelled using analytic theory and the matrix stiffness method with an emphasis on boundary effects. The lattices are fabricated with a minimum feature size of 100 nm and an aspect ratio of 40:1 using single- and double-level STL and deep reactive ion etching of bonded silicon-on-insulator. Nanoelectromechanical systems (NEMS) containing cellular materials are used to demonstrate stretching, bending and brittle fracture. Predicted edge effects are observed, theoretical values of Poisson’s ratio are verified and failure patterns are described.

  4. Modeling of Heat Transfer and Ablation of Refractory Material Due to Rocket Plume Impingement

    Science.gov (United States)

    Harris, Michael F.; Vu, Bruce T.

    2012-01-01

    CR Tech's Thermal Desktop-SINDA/FLUINT software was used in the thermal analysis of a flame deflector design for Launch Complex 39B at Kennedy Space Center, Florida. The analysis of the flame deflector takes into account heat transfer due to plume impingement from expected vehicles to be launched at KSC. The heat flux from the plume was computed using computational fluid dynamics provided by Ames Research Center in Moffet Field, California. The results from the CFD solutions were mapped onto a 3-D Thermal Desktop model of the flame deflector using the boundary condition mapping capabilities in Thermal Desktop. The ablation subroutine in SINDA/FLUINT was then used to model the ablation of the refractory material.

  5. Coupled DQ-FE methods for two dimensional transient heat transfer analysis of functionally graded material

    Energy Technology Data Exchange (ETDEWEB)

    Golbahar Haghighi, M.R.; Eghtesad, M. [Department of Mechanical Engineering, School of Engineering, Shiraz University, Shiraz 71348-51154 (Iran, Islamic Republic of); Malekzadeh, P. [Department of Mechanical Engineering, School of Engineering, Persian Gulf University, Boushehr 75169-13798 (Iran, Islamic Republic of)], E-mail: malekzadeh@pgu.ac.ir

    2008-05-15

    In this paper, a mixed finite element (FE) and differential quadrature (DQ) method as a simple, accurate and computationally efficient numerical tool for two dimensional transient heat transfer analysis of functionally graded materials (FGMs) is developed. The method benefits from the high accuracy, fast convergence behavior and low computational efforts of the DQ in conjunction with the advantages of the FE method in general geometry, loading and systematic boundary treatment. Also, the boundary conditions at the top and bottom surfaces of the domain can be implemented more precisely and in strong form. The temporal derivatives are discretized using an incremental DQ method (IDQM), whose numerical stability is not sensitive to time step size. The effects of non-uniform convective-radiative conditions on the boundaries are investigated. The accuracy of the proposed method is demonstrated by comparing its results with those available in the literature. It is shown that using few grid points, highly accurate results can be obtained.

  6. Investigation of Electron Transfer-Based Photonic and Electro-Optic Materials and Devices

    Energy Technology Data Exchange (ETDEWEB)

    Bromenshenk, Jerry J; Abbott, Edwin H; Dickensheets, David; Donovan, Richard P; Hobbs, J D; Spangler, Lee; McGuirl, Michele A; Spangler, Charles; Rebane, Aleksander; Rosenburg, Edward; Schmidt, V H; Singel, David J

    2008-03-28

    Montana's state program began its sixth year in 2006. The project's research cluster focused on physical, chemical, and biological materials that exhibit unique electron-transfer properties. Our investigators have filed several patents and have also have established five spin-off businesses (3 MSU, 2 UM) and a research center (MT Tech). In addition, this project involved faculty and students at three campuses (MSU, UM, MT Tech) and has a number of under-represented students, including 10 women and 5 Native Americans. In 2006, there was an added emphasis on exporting seminars and speakers via the Internet from UM to Chief Dull Knife Community College, as well as work with the MT Department of Commerce to better educate our faculty regarding establishing small businesses, licensing and patent issues, and SBIR program opportunities.

  7. Cross section measurements of fissile nuclei for slow neutrons; Mesures de sections efficaces de noyaux fissiles pour les neutrons lents

    Energy Technology Data Exchange (ETDEWEB)

    Auclair, J M; Hubert, P; Joly, R; Vendryes, G; Jacrot, B; Netter, F [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Galula, M [Centre National de la Recherche Scientifique (CNRS), 91 - Gif-sur-Yvette (France)

    1955-07-01

    It presents the experimental measurements of cross section of fissile nuclei for slow neutrons to improve the understanding of some heavy nuclei of great importance in the study of nuclear reactors. The different experiments are divided in three categories. In the first part, it studied the variation with energy of the cross sections of natural uranium, {sup 233}U, {sup 235}U and {sup 239}Pu. Two measurement techniques are used: the time-of-flight spectrometer and the crystal spectrometer. In a second part, the fission cross sections of {sup 233}U and {sup 239}Pu for thermal neutrons are compared using a neutron flux from EL-2 going through a double fission chamber. The matter quantity contained in each source is measured by counting the {alpha} activity with a solid angle counter. Finally, the average cross section of {sup 236}U for a spectra of neutrons from the reactor is measured by studying the {beta} activity of {sup 237}U formed by the reaction {sup 236}U (n, {gamma}) {sup 237}U in a sample of {sup 236}U irradiated in the Saclay reactor (EL-2). (M.P.)

  8. Max Phase Materials And Coatings For High Temperature Heat Transfer Applications

    Energy Technology Data Exchange (ETDEWEB)

    Martinez-Rodriguez, M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Garcia-Diaz, B. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Olson, L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Fuentes, R. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Sindelar, R. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-10-19

    Molten salts have been used as heat transfer fluids in a variety of applications within proposed Gen IV nuclear designs and in advanced power system such as Concentrating Solar Power (CSP). However, operating at elevated temperatures can cause corrosion in many materials. This work developed coating technologies for MAX phase materials on Haynes-230 and characterized the corrosion of the coatings in the presence of commercial MgCl2-KCl molten salt. Cold spraying of Ti2AlC and physical vapor deposition (PVD) of Ti2AlC or Zr2AlC were tested to determine the most effective form of coating MAX phases on structural substrates. Corrosion testing at 850°C for 100 hrs showed that 3.9 μm Ti2AlC by PVD was slightly protective while 117 μm Ti2AlC by cold spray and 3.6 μm Zr2AlC by PVD were completely protective. None of the tests showed decomposition of the coating (Ti or Zr) into the salt

  9. Control of radioactive wastes and coupling of neutron/gamma measurements: use of radiative capture for the correction of matrix effects that penalize the fissile mass measurement by active neutron interrogation; Controle des dechets radioactifs et couplage de mesures neutron/gamma: exploitation de la capture radiative pour corriger les effets de matrice penalisant la mesure de la masse fissile par interrogation neutronique active

    Energy Technology Data Exchange (ETDEWEB)

    Loche, F

    2006-10-15

    In the framework of radioactive waste drums control, difficulties arise in the nondestructive measurement of fissile mass ({sup 235}U, {sup 239}Pu..) by Active Neutron Interrogation (ANI), when dealing with matrices containing materials (Cl, H...) influencing the neutron flux. The idea is to use the neutron capture reaction (n,{gamma}) to determine the matrix composition to adjust the ANI calibration coefficient value. This study, dealing with 118 litres, homogeneous drums of density less than 0,4 and composed of chlorinated and/or hydrogenated materials, leads to build abacus linking the {gamma} ray peak areas to the ANI calibration coefficient. Validation assays of these abacus show a very good agreement between the corrected and true fissile masses for hydrogenated matrices (max. relative standard deviation: 23 %) and quite good for chlorinated and hydrogenated matrices (58 %). The developed correction method improves the measured values. It may be extended to 0,45 density, heterogeneous drums. (author)

  10. Theoretical, physical and experimental study of fissile aqueous media; Etudes theorique, physique et experimentale des milieux fissiles aqueux

    Energy Technology Data Exchange (ETDEWEB)

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

    1969-04-01

    This thesis consists of a set of theoretical and experimental studies. a) Theoretical calculation methods used for cross-sections and the critical parameters; b) Comparison of the theoretical and experimental results: it is shown that the agreement between these results cannot be improved above a certain limit because of the accuracy with which are known the composition and the dimensions of the media and the microscopic cross-sections; c) Determination of the ratios {eta}{sup 9}-bar / {eta}{sup 5}-bar, {eta}{sup 3}-bar / {eta}{sup 5}-bar for fissile aqueous media ({eta}-bar: number of neutrons emitted per neutron absorbed, averaged over the reactor neutron spectrum). Evaluation of the accuracy to which these ratios are known; d) Effect of {sup 240}Pu: the measurements are carried out on Pu with a {sup 240}Pu content of 1.5 per cent, 3.11 per cent and 9.95 per cent; Calculation of the resonance integral I240 using the experimental results gives values in reasonable agreement with the results obtained by other more conventional methods. e) Measurement of the spectrum indices for aqueous media containing Pu, U5 and U3. With these latter it is possible to obtain mean fission cross-section ratios {sigma}f239-bar / {sigma}f235-bar for these different spectra. A calculation-experiment comparison is carried out using various theoretical methods. (author) [French] Cette these groupe un ensemble d'etudes theoriques et experimentales. a) Methodes theoriques de calcul utilisees pour les sections efficaces et les parametres critiques; b) Comparaisons des resultats theoriques et experimentaux: on montre que l'accord entre ces resultats ne peut etre ameliore au-dela de certaines limites vu la precision avec laquelle sont connues la composition et les dimensions des milieux et les sections efficaces macroscopiques; c) Determination des rapports {eta}{sup 9}-bar / {eta}{sup 5}-bar, {eta}{sup 3}-bar / {eta}{sup 5}-bar pour les milieux fissiles aqueux ({eta}: nombre de

  11. Theoretical, physical and experimental study of fissile aqueous media; Etudes theorique, physique et experimentale des milieux fissiles aqueux

    Energy Technology Data Exchange (ETDEWEB)

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

    1969-04-01

    This thesis consists of a set of theoretical and experimental studies. a) Theoretical calculation methods used for cross-sections and the critical parameters; b) Comparison of the theoretical and experimental results: it is shown that the agreement between these results cannot be improved above a certain limit because of the accuracy with which are known the composition and the dimensions of the media and the microscopic cross-sections; c) Determination of the ratios {eta}{sup 9}-bar / {eta}{sup 5}-bar, {eta}{sup 3}-bar / {eta}{sup 5}-bar for fissile aqueous media ({eta}-bar: number of neutrons emitted per neutron absorbed, averaged over the reactor neutron spectrum). Evaluation of the accuracy to which these ratios are known; d) Effect of {sup 240}Pu: the measurements are carried out on Pu with a {sup 240}Pu content of 1.5 per cent, 3.11 per cent and 9.95 per cent; Calculation of the resonance integral I240 using the experimental results gives values in reasonable agreement with the results obtained by other more conventional methods. e) Measurement of the spectrum indices for aqueous media containing Pu, U5 and U3. With these latter it is possible to obtain mean fission cross-section ratios {sigma}f239-bar / {sigma}f235-bar for these different spectra. A calculation-experiment comparison is carried out using various theoretical methods. (author) [French] Cette these groupe un ensemble d'etudes theoriques et experimentales. a) Methodes theoriques de calcul utilisees pour les sections efficaces et les parametres critiques; b) Comparaisons des resultats theoriques et experimentaux: on montre que l'accord entre ces resultats ne peut etre ameliore au-dela de certaines limites vu la precision avec laquelle sont connues la composition et les dimensions des milieux et les sections efficaces macroscopiques; c) Determination des rapports {eta}{sup 9}-bar / {eta}{sup 5}-bar, {eta}{sup 3}-bar / {eta}{sup 5}-bar pour les milieux fissiles aqueux ({eta}: nombre de neutrons emis

  12. Heat transfer enhancement of phase change materials by fins under simultaneous charging and discharging

    International Nuclear Information System (INIS)

    Joybari, Mahmood Mastani; Haghighat, Fariborz; Seddegh, Saeid; Al-Abidi, Abduljalil A.

    2017-01-01

    Highlights: • CFD simulation of a finned triplex tube heat exchanger with PCM under simultaneous charging and discharging. • Developed fin configurations for SCD, compatible with natural convection. • More fins enhanced the heat transfer as long as natural convection was not suppressed. • Longer fins enhanced the heat transfer as long as natural convection was not suppressed. • The effect of fin thickness was negligible, similar to non-SCD conditions. - Abstract: Due to the inherent intermittency of renewable energy sources such as solar, latent heat thermal energy storage in phase change materials (PCMs) has received considerable attention. Among several techniques to enhance PCMs’ thermal conductivity, the majority of studies have focused on fin integration due to its simplicity, ease of manufacturing, and low cost. In this study, utilization of extended surfaces (by longitudinal fins) was investigated by development of a numerical model to study the performance of a triplex tube heat exchanger (TTHX) equipped with a PCM under simultaneous charging and discharging (SCD). Governing equations were developed and numerically solved using ANSYS Fluent v16.2. Three conventional fin geometries and six developed fin configurations were compared based on the temperature, liquid fraction, and natural convection behavior under both SCD and non-SCD conditions. The intensity of natural convection was investigated for different fins for the inside heating/outside cooling scenario based on the solid–liquid interface evolution over time. The results indicated that since the buoyancy forces induce upward melted PCM motion, the inner hot tube requires fins on its lower half, while the outer cold one should be extended from its upper half. It was concluded that the case with 3 hot tube fins and 1 cold tube fin is most compatible with natural convection and provides the best performance under SCD conditions.

  13. Heat transfer and thermal management of electric vehicle batteries with phase change materials

    Energy Technology Data Exchange (ETDEWEB)

    Ramandi, M.Y.; Dincer, I.; Naterer, G.F. [University of Ontario Institute of Technology, Faculty of Engineering and Applied Science, Oshawa, ON (Canada)

    2011-07-15

    This paper examines a passive thermal management system for electric vehicle batteries, consisting of encapsulated phase change material (PCM) which melts during a process to absorb the heat generated by a battery. A new configuration for the thermal management system, using double series PCM shells, is analyzed with finite volume simulations. A combination of computational fluid dynamics (CFD) and second law analysis is used to evaluate and compare the new system against the single PCM shells. Using a finite volume method, heat transfer in the battery pack is examined and the results are used to analyse the exergy losses. The simulations provide design guidelines for the thermal management system to minimize the size and cost of the system. The thermal conductivity and melting temperature are studied as two important parameters in the configuration of the shells. Heat transfer from the surroundings to the PCM shell in a non-insulated case is found to be infeasible. For a single PCM system, the exergy efficiency is below 50%. For the second case for other combinations, the exergy efficiencies ranged from 30-40%. The second shell content did not have significant influence on the exergy efficiencies. The double PCM shell system showed higher exergy efficiencies than the single PCM shell system (except a case for type PCM-1). With respect to the reference environment, it is found that in all cases the exergy efficiencies decreased, when the dead-state temperatures rises, and the destroyed exergy content increases gradually. For the double shell systems for all dead-state temperatures, the efficiencies were very similar. Except for a dead-state temperature of 302 K, with the other temperatures, the exergy efficiencies for different combinations are well over 50%. The range of exergy efficiencies vary widely between 15 and 85% for a single shell system, and between 30-80% for double shell systems. (orig.)

  14. Heat transfer and thermal management of electric vehicle batteries with phase change materials

    Science.gov (United States)

    Ramandi, M. Y.; Dincer, I.; Naterer, G. F.

    2011-07-01

    This paper examines a passive thermal management system for electric vehicle batteries, consisting of encapsulated phase change material (PCM) which melts during a process to absorb the heat generated by a battery. A new configuration for the thermal management system, using double series PCM shells, is analyzed with finite volume simulations. A combination of computational fluid dynamics (CFD) and second law analysis is used to evaluate and compare the new system against the single PCM shells. Using a finite volume method, heat transfer in the battery pack is examined and the results are used to analyse the exergy losses. The simulations provide design guidelines for the thermal management system to minimize the size and cost of the system. The thermal conductivity and melting temperature are studied as two important parameters in the configuration of the shells. Heat transfer from the surroundings to the PCM shell in a non-insulated case is found to be infeasible. For a single PCM system, the exergy efficiency is below 50%. For the second case for other combinations, the exergy efficiencies ranged from 30-40%. The second shell content did not have significant influence on the exergy efficiencies. The double PCM shell system showed higher exergy efficiencies than the single PCM shell system (except a case for type PCM-1). With respect to the reference environment, it is found that in all cases the exergy efficiencies decreased, when the dead-state temperatures rises, and the destroyed exergy content increases gradually. For the double shell systems for all dead-state temperatures, the efficiencies were very similar. Except for a dead-state temperature of 302 K, with the other temperatures, the exergy efficiencies for different combinations are well over 50%. The range of exergy efficiencies vary widely between 15 and 85% for a single shell system, and between 30-80% for double shell systems.

  15. A model for radiative heat transfer in mixtures of a hot solid or molten material with water and steam

    International Nuclear Information System (INIS)

    Vaeth, L.

    1997-05-01

    A model has been devised for describing the radiative heat transfer in mixtures of a hot radiant material with water and steam, to be used, e.g., in the framework of a multiphase, multicomponent flow simulation. The main features of the model are: 1. The radiative heat transfer is modelled for a homogeneous mixture of one continuous material with droplets/bubbles of the other two, of the kind normally assumed for the material distribution in one cell of a bigger calculational problem. Neither the heat transfer over the cell boundaries nor the finite dimensions of the cell are taken into account. 2. The geometry of the mixture (radiant material continuous or discontinuous, droplet/bubble diameters and number densities) is taken into account. 3. The optical properties of water and water vapour are modelled as functions of the temperature of the radiant and, in the case of water vapour, also of the absorbing material. 4. The model distinguishes between heat transfer to the surface of the water (leading to evaporation) and into the bulk of the water (pure heating). (orig./DG) [de

  16. Transfer, attachment, and formation of biofilms by Escherichia coli O157:H7 on meat-contact surface materials.

    Science.gov (United States)

    Simpson Beauchamp, Catherine; Dourou, Dimitra; Geornaras, Ifigenia; Yoon, Yohan; Scanga, John A; Belk, Keith E; Smith, Gary C; Nychas, George-John E; Sofos, John N

    2012-06-01

    Studies examined the effects of meat-contact material types, inoculation substrate, presence of air at the liquid-solid surface interface during incubation, and incubation substrate on the attachment/transfer and subsequent biofilm formation by Escherichia coli O157:H7 on beef carcass fabrication surface materials. Materials studied as 2 × 5 cm coupons included stainless steel, acetal, polypropylene, and high-density polyethylene. A 6-strain rifampicin-resistant E. coli O157:H7 composite was used to inoculate (6 log CFU/mL, g, or cm²) tryptic soy broth (TSB), beef fat/lean tissue homogenate (FLH), conveyor belt-runoff fluids, ground beef, or beef fat. Coupons of each material were submerged (4 °C, 30 min) in the inoculated fluids or ground beef, or placed between 2 pieces of inoculated beef fat with pressure (20 kg) applied. Attachment/transfer of the pathogen was surface material and substrate dependent, although beef fat appeared to negate differences among surface materials. Beef fat was the most effective (P transfer and subsequent biofilm formation by E. coli O157:H7. The results highlight the importance of thoroughly cleaning soiled surfaces to remove all remnants of beef fat or other organic material that may harbor or protect microbial contaminants during otherwise lethal antimicrobial interventions. © 2012 Institute of Food Technologists®

  17. Experimental determination of the heat transfer and cold storage characteristics of a microencapsulated phase change material in a horizontal tank

    International Nuclear Information System (INIS)

    Allouche, Yosr; Varga, Szabolcs; Bouden, Chiheb; Oliveira, Armando C.

    2015-01-01

    Highlights: • Cold storage characteristics in latent and sensible heat storage mediums were studied. • Thermo-physical characterization of the phase change material was carried out. • A non-Newtonian shear thickening behavior of the phase change material was observed. • An energy storage enhancement (53%) was observed in the latent heat storage medium. - Abstract: In the present paper, the performance of a microencapsulated phase change material (in 45% w/w concentration) for low temperature thermal energy storage, suitable for air conditioning applications is studied. The results are compared to a sensible heat storage unit using water. Thermo-physical properties such as the specific heat, enthalpy variation, thermal conductivity and density are also experimentally determined. The non-Newtonian shear-thickening behavior of the phase change material slurry is quantified. Thermal energy performance is experimentally determined for a 100 l horizontal tank. The heat transfer between the heat transfer fluid and the phase change material was provided by a tube-bundle heat exchanger inside the tank. The results show that the amount of energy stored using the phase change material is 53% higher than for water after 10 h of charging, for the same storage tank volume. It was found that the heat transfer coefficient between the phase change material and the tube wall increases during the phase change temperature range, however it remains smaller than the values obtained for water

  18. Heat transfer and thermal storage performance of an open thermosyphon type thermal storage unit with tubular phase change material canisters

    International Nuclear Information System (INIS)

    Wang, Ping-Yang; Hu, Bo-Wen; Liu, Zhen-Hua

    2015-01-01

    Highlights: • A novel open heat pipe thermal storage unit is design to improve its performance. • Mechanism of its operation is phase-change heat transfer. • Tubular canisters with phase change material were placed in thermal storage unit. • Experiment and analysis are carried out to investigate its operation properties. - Abstract: A novel open thermosyphon-type thermal storage unit is presented to improve design and performance of heat pipe type thermal storage unit. In the present study, tubular canisters filled with a solid–liquid phase change material are vertically placed in the middle of the thermal storage unit. The phase change material melts at 100 °C. Water is presented as the phase-change heat transfer medium of the thermal storage unit. The tubular canister is wrapped tightly with a layer of stainless steel mesh to increase the surface wettability. The heat transfer mechanism of charging/discharging is similar to that of the thermosyphon. Heat transfer between the heat resource or cold resource and the phase change material in this device occurs in the form of a cyclic phase change of the heat-transfer medium, which occurs on the surface of the copper tubes and has an extremely high heat-transfer coefficient. A series of experiments and theoretical analyses are carried out to investigate the properties of the thermal storage unit, including power distribution, start-up performance, and temperature difference between the phase change material and the surrounding vapor. The results show that the whole system has excellent heat-storage/heat-release performance

  19. Improved resonance formulas for cross sections of fissile elements

    International Nuclear Information System (INIS)

    Segev, M.

    1978-01-01

    The Adler--Adler cross-section formalism with energy-dependent parameters is a practical approximation to the R-matrix formalism, on the basis of the smallness of the s-wave neutron width in fissile elements. Attempts were made to represent experimental cross sections by the Adler--Adler formulas through an initial representation by the Reich--Moore approximation of R-matrix and a subsequent conversion of the Reich--Moore formulas to the Adler--Adler formulas. Adler and Adler foresaw difficulties in associating their formulas with approximate R-matrix theories such as those of Reich and Moore. Indeed, it is shown that, due to the nonunitarity of the Adler--Adler formalism on the one hand and the unitarity, by definition, of the Reich--Moore formalism on the other hand, the conversion from the latter to the former is ambiguous. Examples are shown to demonstrate that this ambiguity results in numerical inaccuracies, sometimes very large ones, for neutron widths that are not extremely small. Improved Adler--Adler-type formulas have been derived from the R-matrix formalism. In these formulas, the multipliers of the Breit--Wigner resonance lines exhibit more explicit energy dependence than their original counterparts, mainly in the form of additional terms in the formula for the total cross section. The conversion from Reich--Moore cross sections to the improved resonance formulas is shown to be much less ambiguous and to produce very accurate cross sections. In particular, the inaccuracies encountered with the Reich--Moore to Adler--Adler conversion are eliminated. A computer code, PEDRA, was written to perform the conversion from a given set of Reich--Moore parameters to the parameters required in the improved formulas. The numerical algorithm of this code is based on an adaptation with modifications of the numerical approach of de Saussure--Perez in the POLLA code, which converts Reich--Moore parameters to Adler--Adler parameters. 7 figures, 1 table

  20. Development of a poly(dimethylacrylamide) based matrix material for solid phase high density peptide array synthesis employing a laser based material transfer

    International Nuclear Information System (INIS)

    Ridder, Barbara; Foertsch, Tobias C.; Welle, Alexander; Mattes, Daniela S.; Bojnicic-Kninski, Clemens M. von; Loeffler, Felix F.; Nesterov-Mueller, Alexander; Meier, Michael A.R.; Breitling, Frank

    2016-01-01

    Highlights: • New matrix material for peptide array synthesis from a ‘solid solvent’. • Resolution was increased with possible spot densities of up to 20.000 spots per cm"2. • The coupling depth and the effectiveness of washing steps analyzed by ToF-SIMS. • Adaptations and custom changes of the matrix material are possible. - Abstract: Poly(dimethylacrylamide) (PDMA) based matrix materials were developed for laser-based in situ solid phase peptide synthesis to produce high density arrays. In this specific array synthesis approach, amino acid derivatives are embedded into a matrix material, serving as a “solid” solvent material at room temperature. Then, a laser pulse transfers this mixture to the target position on a synthesis slide, where the peptide array is synthesized. Upon heating above the glass transition temperature of the matrix material, it softens, allowing diffusion of the amino acid derivatives to the synthesis surface and serving as a solvent for peptide bond formation. Here, we synthesized PDMA six-arm star polymers, offering the desired matrix material properties, using atom transfer radical polymerization. With the synthesized polymers as matrix material, we structured and synthesized arrays with combinatorial laser transfer. With densities of up to 20,000 peptide spots per cm"2, the resolution could be increased compared to the commercially available standard matrix material. Time-of-Flight Secondary Ion Mass Spectrometry experiments revealed the penetration behavior of an amino acid derivative into the prepared acceptor synthesis surface and the effectiveness of the washing protocols.

  1. Development of a poly(dimethylacrylamide) based matrix material for solid phase high density peptide array synthesis employing a laser based material transfer

    Energy Technology Data Exchange (ETDEWEB)

    Ridder, Barbara [Institute of Microstructure Technology (IMT), Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Institute of Organic Chemistry (IOC), Karlsruhe Institute of Technology (KIT), Fritz-Haber-Weg 6, 76131 Karlsruhe (Germany); Foertsch, Tobias C. [Institute of Microstructure Technology (IMT), Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Welle, Alexander [Karlsruhe Nano Micro Facility (KNMF), Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Mattes, Daniela S. [Institute of Microstructure Technology (IMT), Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Institute of Organic Chemistry (IOC), Karlsruhe Institute of Technology (KIT), Fritz-Haber-Weg 6, 76131 Karlsruhe (Germany); Bojnicic-Kninski, Clemens M. von; Loeffler, Felix F.; Nesterov-Mueller, Alexander [Institute of Microstructure Technology (IMT), Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Meier, Michael A.R., E-mail: m.a.r.meier@kit.edu [Institute of Organic Chemistry (IOC), Karlsruhe Institute of Technology (KIT), Fritz-Haber-Weg 6, 76131 Karlsruhe (Germany); Breitling, Frank, E-mail: frank.breitling@kit.edu [Institute of Microstructure Technology (IMT), Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany)

    2016-12-15

    Highlights: • New matrix material for peptide array synthesis from a ‘solid solvent’. • Resolution was increased with possible spot densities of up to 20.000 spots per cm{sup 2}. • The coupling depth and the effectiveness of washing steps analyzed by ToF-SIMS. • Adaptations and custom changes of the matrix material are possible. - Abstract: Poly(dimethylacrylamide) (PDMA) based matrix materials were developed for laser-based in situ solid phase peptide synthesis to produce high density arrays. In this specific array synthesis approach, amino acid derivatives are embedded into a matrix material, serving as a “solid” solvent material at room temperature. Then, a laser pulse transfers this mixture to the target position on a synthesis slide, where the peptide array is synthesized. Upon heating above the glass transition temperature of the matrix material, it softens, allowing diffusion of the amino acid derivatives to the synthesis surface and serving as a solvent for peptide bond formation. Here, we synthesized PDMA six-arm star polymers, offering the desired matrix material properties, using atom transfer radical polymerization. With the synthesized polymers as matrix material, we structured and synthesized arrays with combinatorial laser transfer. With densities of up to 20,000 peptide spots per cm{sup 2}, the resolution could be increased compared to the commercially available standard matrix material. Time-of-Flight Secondary Ion Mass Spectrometry experiments revealed the penetration behavior of an amino acid derivative into the prepared acceptor synthesis surface and the effectiveness of the washing protocols.

  2. Numerical studies on heat transfer and pressure drop characteristics of flat finned tube bundles with various fin materials

    Science.gov (United States)

    Peng, Y.; Zhang, S. J.; Shen, F.; Wang, X. B.; Yang, X. R.; Yang, L. J.

    2017-11-01

    The air-cooled heat exchanger plays an important role in the field of industry like for example in thermal power plants. On the other hand, it can be used to remove core decay heat out of containment passively in case of a severe accident circumstance. Thus, research on the performance of fins in air-cooled heat exchangers can benefit the optimal design and operation of cooling systems in nuclear power plants. In this study, a CFD (Computational Fluid Dynamic) method is implemented to investigate the effects of inlet velocity, fin spacing and tube pitch on the flow and the heat transfer characteristics of flat fins constructed of various materials (316L stainless steel, copper-nickel alloy and aluminium). A three dimensional geometric model of flat finned tube bundles with fixed longitudinal tube pitch and transverse tube pitch is established. Results for the variation of the average convective heat transfer coefficient with respect to cooling air inlet velocity, fin spacing, tube pitch and fin material are obtained, as well as for the pressure drop of the cooling air passing through finned tube. It is shown that the increase of cooling air inlet velocity results in enhanced average convective heat transfer coefficient and decreasing pressure drop. Both fin spacing and tube pitch engender positive effects on pressure drop and have negative effects on heat transfer characteristics. Concerning the fin material, the heat transfer performance of copper-nickel alloy is superior to 316L stainless steel and inferior to aluminium.

  3. High order statistical signatures from source-driven measurements of subcritical fissile systems

    International Nuclear Information System (INIS)

    Mattingly, J.K.

    1998-01-01

    This research focuses on the development and application of high order statistical analyses applied to measurements performed with subcritical fissile systems driven by an introduced neutron source. The signatures presented are derived from counting statistics of the introduced source and radiation detectors that observe the response of the fissile system. It is demonstrated that successively higher order counting statistics possess progressively higher sensitivity to reactivity. Consequently, these signatures are more sensitive to changes in the composition, fissile mass, and configuration of the fissile assembly. Furthermore, it is shown that these techniques are capable of distinguishing the response of the fissile system to the introduced source from its response to any internal or inherent sources. This ability combined with the enhanced sensitivity of higher order signatures indicates that these techniques will be of significant utility in a variety of applications. Potential applications include enhanced radiation signature identification of weapons components for nuclear disarmament and safeguards applications and augmented nondestructive analysis of spent nuclear fuel. In general, these techniques expand present capabilities in the analysis of subcritical measurements

  4. Numerical Heat Transfer Studies of a Latent Heat Storage System Containing Nano-Enhanced Phase Change Material

    Directory of Open Access Journals (Sweden)

    S F Hosseinizadeh

    2011-01-01

    Full Text Available The heat transfer enhancement in the latent heat thermal energy storage system through dispersion of nanoparticle is reported. The resulting nanoparticle-enhanced phase change materials (NEPCM exhibit enhanced thermal conductivity in comparison to the base material. The effects of nanoparticle volume fraction and some other parameters such as natural convection are studied in terms of solid fraction and the shape of the solid-liquid phase front. It has been found that higher nanoparticle volume fraction result in a larger solid fraction. The present results illustrate that the suspended nanoparticles substantially increase the heat transfer rate and also the nanofluid heat transfer rate increases with an increase in the nanoparticles volume fraction. The increase of the heat release rate of the NEPCM shows its great potential for diverse thermal energy storage application.

  5. Experimental evaluation on natural convection heat transfer of microencapsulated phase change materials slurry in a rectangular heat storage tank

    International Nuclear Information System (INIS)

    Zhang Yanlai; Rao Zhonghao; Wang Shuangfeng; Zhang Zhao; Li Xiuping

    2012-01-01

    Highlights: ► It gives heat transfer characteristics in a rectangular heat storage tank as the basic unit for reservoir of thermal storage. ► Onset of natural convection gets easier for the MPCMS with a higher mass concentration. ► It enhances the heat transfer ability of natural convection for the MPCMS. ► Obtained the relationship between Ra and Nu of the MPCMS. - Abstract: The main purpose of this experiment is to evaluate natural convection heat transfer characteristics of microencapsulated PCM (phase change material) slurry (MPCMS) during phase change process in a rectangular heat storage tank heated from the bottom and cooled at the top. The microencapsulated PCM is several material compositions of n-paraffin waxes (mainly nonadecane) as the core materials, outside a layer of a melamine resin wrapped. In the present study, its slurry is used mixing with water. And the specific heat capacity with latent heat shows a peak value at the temperature of about T = 31 °C. We investigate the influences of the phase change process of the MPCMS on natural convection heat transfer. The experimental results indicate that phase change process of the MPCMS promote natural convection heat transfer. The local maximum heat transfer enhancement occurs at approximately T H = 34 °C corresponding to the heated plate temperature. With high mass concentration C m , the onset of natural convection gets easier for the MPCMS. The temperature gradient is larger near top plate and bottom plate of a rectangular heat storage tank. Heat transfer coefficient increases with the phase change of the PCM. And it summarizes that the phase change process of the PCM promote the occurrence of natural convection.

  6. Determination of enthalpy, temperature, surface tension and geometry of the material transfer in PGMAW for the system argon–iron

    International Nuclear Information System (INIS)

    Siewert, E; Schein, J; Forster, G

    2013-01-01

    The metal transfer is a fundamental process in gas metal arc welding, which substantially determines the shape of the weld seam and strongly influences arc formation and stability. In this investigation the material transfer from the wire electrode (anode) to the workpiece (cathode) is analysed experimentally with high accuracy using various innovative diagnostic techniques for a pulsed gas metal arc welding (PGMAW) process. A high-speed two-colour pyrometer, a calorimeter, thermocouples, a stereo optical setup and a droplet oscillation technique are used to analyse a precisely defined PGMAW process. Thus, results obtained are verified by different measurement techniques and enable a comprehensive description of the material transfer procedure. The surface temperature of both electrodes as well as the droplet temperature, enthalpy and surface tension were determined. Furthermore, the geometry of the arc, wire, droplets and weld pool were extracted in three dimensions in order to describe the interaction between the material transfer and the formation of the weld seam. The experiments are performed using argon as shielding gas and pure iron as filler and base material to reduce complex chemical processes. It turned out that the wire feed rate has the biggest influence on droplet temperature and detachment. A correlation between weld pool formation and weld pool surface temperature gradient was observed, which is mainly a function of welding speed and wire feed rate. The experimental results obtained provide a detailed data pool for use in modelling. (paper)

  7. Royal order relating to the transfer of nuclear materials and technology to non-nuclear weapon states

    International Nuclear Information System (INIS)

    1989-05-01

    In implementation of the Act of 1981 on conditions for the export of nuclear materials, equipment and technological data, this Order sets down the detailed mechanisms for such transfers. Its object is to ensure that they will be carried out exclusively for peaceful purposes and in conformity with the NPT [fr

  8. Experience of work with radioactive materials and nuclear fuel at the reactor WWR-K

    International Nuclear Information System (INIS)

    Maltseva, R.M.; Petukhov, V.K.

    1998-01-01

    In the report there are considered questions concerning the handling with fresh and spent fuel, experimental devices, containing high enriched uranium, being fissile materials of the bulk form, radioisotopes, obtained in the reactor, and radioactive waste, formed during the operation of the reactor, and organization of storage, account and control of radioactive and fissile materials is described. (author)

  9. Global nuclear material control model

    International Nuclear Information System (INIS)

    Dreicer, J.S.; Rutherford, D.A.

    1996-01-01

    The nuclear danger can be reduced by a system for global management, protection, control, and accounting as part of a disposition program for special nuclear materials. The development of an international fissile material management and control regime requires conceptual research supported by an analytical and modeling tool that treats the nuclear fuel cycle as a complete system. Such a tool must represent the fundamental data, information, and capabilities of the fuel cycle including an assessment of the global distribution of military and civilian fissile material inventories, a representation of the proliferation pertinent physical processes, and a framework supportive of national or international perspective. They have developed a prototype global nuclear material management and control systems analysis capability, the Global Nuclear Material Control (GNMC) model. The GNMC model establishes the framework for evaluating the global production, disposition, and safeguards and security requirements for fissile nuclear material

  10. Effect of fissile isotope burnup on criticality safety for stored disintegrated fuel rods

    International Nuclear Information System (INIS)

    Heaberlin, S.W.; Selby, G.P.

    1978-09-01

    If the fuel rods were to disintegrate and water added, a criticality could occur in a 13-in. PWR canister with fresh fuel enriched to 3.5 wt % 235 U. The question is, ''If credit could be taken for burnup, could this indicate a subcritical condition.'' In attempting to answer this question, a series of calculations were performed. A set of isotopic concentrations were generated for 5,000, 10,000, 15,000, and 20,000 MWD/MTU burnup levels. Four reflector materials, water, concrete and two types of soil, were considered. Results indicate that allowing credit for fissile isotope burnup does not completely remove the concern for criticality safety in the event of rod disintegration. Reactivities which are ''subcritical'' (k/sub eff/ = 0.95) would not occur for three of the four reflector materials at even the 20,000 MWD/MTU burnup level in the 13-in. canister. The water reflected canister would achieve the k/sub eff/ = 0.95 level near 18,000 MWD/MTU. A smaller canister could be postulated. If a quarter inch gap is allowed, a Westinghouse 17 x 17 PWR assembly requires a 12 1 / 4 inch diameter canister. For such a canister with water reflection the ''subcritical'' (k/sub eff/ = 0.95) level would be reached near 15,000 MWD/MTU. The soil reflected canisters would reach this level between 18,000 and 19,000 MWD/MTU. Considering the difficulties in taking credit for burnup, such modest gains in apparent safety are not encouraging. This situation might be improved, however, if credit were also taken for neutron absorption by fission product poisons produced during burnup. It is strongly recommended that other approaches to a solution of the criticality safety problem be considered

  11. Temperature Profile of the Solution Vessel of an Accelerator-Driven Subcritical Fissile Solution System

    Energy Technology Data Exchange (ETDEWEB)

    Klein, Steven Karl [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Determan, John C. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-09-14

    Dynamic System Simulation (DSS) models of fissile solution systems have been developed and verified against a variety of historical configurations. DSS techniques have been applied specifically to subcritical accelerator-driven systems using fissile solution fuels of uranium. Initial DSS models were developed in DESIRE, a specialized simulation scripting language. In order to tailor the DSS models to specifically meet needs of system designers they were converted to a Visual Studio implementation, and one of these subsequently to National Instrument’s LabVIEW for human factors engineering and operator training. Specific operational characteristics of subcritical accelerator-driven systems have been examined using a DSS model tailored to this particular class using fissile fuel.

  12. Temperature Profile of the Solution Vessel of an Accelerator-Driven Subcritical Fissile Solution System

    International Nuclear Information System (INIS)

    Klein, Steven Karl; Determan, John C.

    2015-01-01

    Dynamic System Simulation (DSS) models of fissile solution systems have been developed and verified against a variety of historical configurations. DSS techniques have been applied specifically to subcritical accelerator-driven systems using fissile solution fuels of uranium. Initial DSS models were developed in DESIRE, a specialized simulation scripting language. In order to tailor the DSS models to specifically meet needs of system designers they were converted to a Visual Studio implementation, and one of these subsequently to National Instrument's LabVIEW for human factors engineering and operator training. Specific operational characteristics of subcritical accelerator-driven systems have been examined using a DSS model tailored to this particular class using fissile fuel.

  13. An approximate method to estimate the minimum critical mass of fissile nuclides

    International Nuclear Information System (INIS)

    Wright, R.Q.; Jordan, W.C.

    1999-01-01

    When evaluating systems in criticality safety, it is important to approximate the answer before any analysis is performed. There is currently interest in establishing the minimum critical parameters for fissile actinides. The purpose is to describe the OB-1 method for estimating the minimum critical mass for thermal systems based on one-group calculations and 235 U spheres fully reflected by water. The observation is made that for water-moderated, well-thermalized systems, the transport and leakage from the system are dominated by water. Under these conditions two fissile mixtures will have nearly the same critical volume provided the infinite media multiplication factor (k ∞ ) for the two systems is the same. This observation allows for very simple estimates of critical concentration and mass as a function of the hydrogen-to-fissile (H/X) moderation ratio by comparison to the known 235 U system

  14. Laser-induced forward transfer of intact, solid-phase inorganic materials

    NARCIS (Netherlands)

    Feinäugle, Matthias

    2014-01-01

    Laser-induced forward transfer (LIFT) is a technique for the micro- and nanofabrication of photonic, electronic and biomedical devices. Compared to conventional methods of device microfabrication, LIFT offers the unique features of transfer of functional and sensitive thin films with a minimum of

  15. Fissility of actinide nuclei induced by 60-130 MeV photons

    International Nuclear Information System (INIS)

    Morcelle, Viviane; Tavares, Odilon A.P.

    2004-06-01

    Nuclear fissilities obtained from recent photofission reaction cross section measurements carried out at Saskatchewan Accelerator Laboratory (Saskatoon, Canada) in the energy range 60-130 MeV for 232 Th, 233 U, 235 U, 238 U, and 237 Np nuclei have been analysed in a systematic way. To this aim, a semiempirical approach has been developed based on the quasi-deuteron nuclear photoabsorption model followed by the process of competition between neutron evaporation and fission for the excited nucleus. The study reproduces satisfactorily well the increasing trend of nuclear fissility with parameter Z 2 =A. (author)

  16. Humidity adsorption and transfer in hygroscopic materials. Percolation-type approach and experimentation

    International Nuclear Information System (INIS)

    Quenard, Daniel

    1989-01-01

    Water vapor adsorption and transfer in microporous media are studied by using a 3 level hierarchical approach. At the microscopic level (pore size), we describe the basic phenomena (adsorption/desorption, capillary condensation, molecular and Knudsen diffusion, Hagen-Poiseuille flow) that occur during the isotherm water vapor transport in a single cylindrical pore, at the steady state. The transport through a condensed pore is taken into account by its 'vapor equivalent flow' and we underline that capillary condensation may cause vapor flow amplification of several orders of magnitude. We suggest to use an electrical analogy between a cylindrical pore and a Zener diode. Then at the mesoscopic level (material size), we introduce pore networks to provide use with a simplified description of the microstructure. Three types of networks are studied: square, triangular and honeycomb. By using a random distribution of the single cylindrical pores on the 2D networks, we are able to estimate the sorption isotherms and the water vapor permeability which are the two essential characteristics to understand the behaviour of materials towards humidity. To develop this approach we refer to the percolation concept and we use most of its principal results. To estimate the adsorption isotherms we introduce a surface adsorption model and we use the KELVIN-LAPLACE equation. Hysteresis appears naturally thanks to the 'ink-bottle' phenomenon and it is all the more important since the network is ill-connected. The water vapor permeability is calculated thanks to the electrical analogy (cylindrical pore-Zener diode). We emphasize an important amplification of the equivalent permeability when the relative humidity reaches a threshold value. This phenomenon provides use with a possible explanation of numerous experimental results. The respective effects of pore size distribution and temperature, on sorption isotherms and permeability, are presented. We present several

  17. Solution-processed, molecular photovoltaics that exploit hole transfer from non-fullerene, n-type materials

    KAUST Repository

    Douglas, Jessica D.

    2014-05-12

    Solution-processed organic photovoltaic devices containing p-type and non-fullerene n-type small molecules obtain power conversion efficiencies as high as 2.4%. The optoelectronic properties of the n-type material BT(TTI-n12)2 allow these devices to display high open-circuit voltages (>0.85 V) and generate significant charge carriers through hole transfer in addition to the electron-transfer pathway, which is common in fullerene-based devices. © 2014 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  18. Using a quasi-heat-pulse method to determine heat and moisture transfer properties for porous orthotropic wood products or cellular solid materials

    Science.gov (United States)

    M. A. Dietenberger

    2006-01-01

    Understanding heat and moisture transfer in a wood specimen as used in the K-tester has led to an unconventional numerical solution arid intriguing protocol to deriving the transfer properties. Laplace transform solutions of Luikov’s differential equations are derived for one-dimensional heat and moisture transfer in porous hygroscopic orthotropic materials and for a...

  19. Tank 40 final sludge batch 9 chemical and fissile radionuclide characterization results

    Energy Technology Data Exchange (ETDEWEB)

    Bannochie, C. J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Kubilius, W. P. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Pareizs, J. M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-06-26

    A sample of Sludge Batch (SB) 9 was pulled from Tank 40 in order to obtain radionuclide inventory analyses necessary for compliance with the Waste Acceptance Product Specifications (WAPS)i. The SB9 WAPS sample was also analyzed for chemical composition, including noble metals, and fissile constituents, and these results are reported here. These analyses along with the WAPS radionuclide analyses will help define the composition of the sludge in Tank 40 that is fed to the Defense Waste Processing Facility (DWPF) as SB9. At the Savannah River National Laboratory (SRNL), the 3-L Tank 40 SB9 sample was transferred from the shipping container into a 4-L high density polyethylene bottle and solids were allowed to settle. Supernate was then siphoned off and circulated through the shipping container to complete the transfer of the sample. Following thorough mixing of the 3-L sample, a 547 g sub-sample was removed. This sub-sample was then utilized for all subsequent slurry sample preparations. Eight separate aliquots of the slurry were digested, four with HNO3/HCl (aqua regiaii) in sealed Teflon® vessels and four with NaOH/Na2O2 (alkali or peroxide fusioniii) using Zr crucibles. Three Analytical Reference Glass – 1iv (ARG-1) standards were digested along with a blank for each preparation. Each aqua regia digestion and blank was diluted to 1:100 with deionized water and submitted to Analytical Development (AD) for inductively coupled plasma – atomic emission spectroscopy (ICP-AES) analysis, inductively coupled plasma – mass spectrometry (ICP-MS) analysis, atomic absorption spectroscopy (AA) for As and Se, and cold vapor atomic absorption spectroscopy (CV-AA) for Hg. Equivalent dilutions of the alkali fusion digestions and blank were submitted to AD for ICP-AES analysis. Tank 40 SB9 supernate was collected from a mixed slurry sample in the SRNL Shielded Cells and submitted to AD for ICP-AES, ion chromatography (IC), total base/free OH-/other base, total inorganic

  20. A Polarizable and Transferable PHAST CO 2 Potential for Materials Simulation

    KAUST Repository

    Mullen, Ashley L.; Pham, Tony; Forrest, Katherine A.; Cioce, Christian R.; McLaughlin, Keith; Space, Brian

    2013-01-01

    Reliable PHAST (Potentials with High Accuracy Speed and Transferability) intermolecular potential energy functions for CO2 have been developed from first principles for use in heterogeneous systems, including one with explicit polarization

  1. Separation of silicon carbide-coated fertile and fissile particles by gas classification

    International Nuclear Information System (INIS)

    Vaughen, V.C.A.

    1976-07-01

    The separation of 235 U and 233 U in the reprocessing of HTGR fuels is a key feature of the feed-breed fuel cycle concept. This is attained in the Fort St. Vrain (FSV) reactor by coating the fissile (Th- 235 U) particles and the fertile (Th- 233 U) particles separately with silicon carbide (SiC) layers to contain the fission products and to protect the kernels from burning in the head-end reprocessing steps. Pneumatic (gas) classification based on size and density differences is the reference process for separating the SiC-coated particles into fissile and fertile streams for subsequent handling. Terminal velocities have been calculated for the +- 2 sigma ranges of particle sizes and densities for ''Fissile B''--''Fertile A'' particles used in the FSV reactor. Because of overlapping particle fractions, a continuous pneumatic separator appears infeasible; however, a batch separation process can be envisioned. Changing the gas from air to CO 2 and/or the temperature to 300 0 C results in less than 10 percent change in calculated terminal velocities. Recently reported work in gas classification is discussed in light of the theoretical calculations. The pneumatic separation of fissile and fertile particles needs more study, specifically with regard to (1) measuring the recoveries and separation efficiencies of actual fissile and fertile fractions in the tests of the pneumatic classifiers; and (2) improving the contactor design or flowsheet to avoid apparent flow separation or flooding problems at the feed point when using the feed rates required for the pilot plant

  2. Study of relationships between microstructures and service properties, of U(Mo) fissile alloys particles

    International Nuclear Information System (INIS)

    Champion, G.

    2013-01-01

    This thesis enters in the Material and Testing Reactors (MTRs) framework where the necessity to use a Low- Enriched Uranium (LEU) fuel has led to the development of a dense fissile material based on U(Mo) alloys. The designed fuel is a composite material, made of dispersed U(Mo) particles embedded in an Al based matrix. Post- Irradiation Examinations of these LEU fuel plates showed that the irradiation behaviour of the fuel is not fit for purpose yet. This is mainly due to the growth of an interaction layer between the fuel and the matrix and to the bad gas retention efficiency of the fuel particles. This thesis had for purpose the development of several solutions in order to modify and/or decrease or even inhibit the fuel/matrix interaction and to increase the gas retention capacities of the fuel. In order to achieve so, two solutions have been tested during this thesis, (i) optimization of the U(Mo) alloy intrinsic microstructural properties and (ii) modification of the fuel meat/matrix interface, through the deposition of a layer acting as a 'diffusion barrier'. Concerning the first axis of study, a characterization campaign of the reference powders has been performed, as a first step, in order to identify the key parameters for the development of products showing an 'optimized' microstructure. Two novel products have then been developed: one based on a combined process associating 'atomization + grinding' and another, which consists in a magnesiothermy process. These products were subjected to characterization: X-Ray and neutron diffraction, electron backscattered diffraction and transmission electron microscopy have been performed in particular. We managed to show that these powders can be an advantage concerning the issue with the gas retention capacities of the fuel. Concerning the growth of the interaction layer, a third product has been developed: an U(Mo) atomized powder, coated with an alumina layer. We managed to show that a thickness between 100 and

  3. 2D materials in electro-optic modulation: energy efficiency, electrostatics, mode overlap, material transfer and integration

    Science.gov (United States)

    Ma, Zhizhen; Hemnani, Rohit; Bartels, Ludwig; Agarwal, Ritesh; Sorger, Volker J.

    2018-02-01

    Here we discuss the physics of electro-optic modulators deploying 2D materials. We include a scaling laws analysis and show how energy-efficiency and speed change for three underlying cavity systems as a function of critical device length scaling. A key result is that the energy-per-bit of the modulator is proportional to the volume of the device, thus making the case for submicron-scale modulators possible deploying a plasmonic optical mode. We then show how Graphene's Pauli-blocking modulation mechanism is sensitive to the device operation temperature, whereby a reduction of the temperature enables a 10× reduction in modulator energy efficiency. Furthermore, we show how the high-index tunability of graphene is able to compensate for the small optical overlap factor of 2D-based material modulators, which is unlike classical silicon-based dispersion devices. Lastly, we demonstrate a novel method towards a 2D material printer suitable for cross-contamination free and on-demand printing. The latter paves the way to integrate 2D materials seamlessly into taped-out photonic chips.

  4. Academic Entrepreneurship and Exchange of Scientific Resources: Material Transfer in Life and Materials Sciences in Japanese Universities

    Science.gov (United States)

    Shibayama, Sotaro; Walsh, John P.; Baba, Yasunori

    2012-01-01

    This study uses a sample of Japanese university scientists in life and materials sciences to examine how academic entrepreneurship has affected the norms and behaviors of academic scientists regarding sharing scientific resources. Results indicate that high levels of academic entrepreneurship in a scientific field are associated with less reliance…

  5. The transfer of diatoms from freshwater to footwear materials: An experimental study assessing transfer, persistence, and extraction methods for forensic reconstruction.

    Science.gov (United States)

    Levin, E A; Morgan, R M; Scott, K R; Jones, V J

    2017-09-01

    In recent years there has been growing interest in environmental forms of trace evidence, and ecological trace evidence collected from footwear has proved valuable within casework. Simultaneously, there has been growing awareness of the need for empirical experimentation to underpin forensic inferences. Diatoms are unicellular algae, and each cell (or 'frustule') consists of two valves which are made of silica, a robust material that favours their preservation both in sediments and within forensic scenarios. A series of experiments were carried out to investigate the transfer and persistence of diatoms upon common footwear materials, a recipient surface that has historically been overlooked by studies of persistence. The effectiveness of two novel extraction techniques (jet rinsing, and heating and agitation with distilled water) was compared to the established extraction technique of hydrogen peroxide digestion, for a suite of five common footwear materials: canvas, leather, and 'suede' (representing upper materials), and rubber and polyurethane (representing sole materials). It was observed that the novel extraction technique of heating and agitation with distilled water did not extract fewer diatom valves, or cause increased fragmentation of valves, when compared to peroxide digestion, suggesting that the method may be viable where potentially hazardous chemical reactions may be encountered with the peroxide digestion method. Valves could be extracted from all five footwear materials after 3min of immersion, and more valves were extracted from the rougher, woven upper materials than the smoother sole materials. Canvas yielded the most valves (a mean of 2511/cm 2 ) and polyurethane the fewest (a mean of 15/cm 2 ). The persistence of diatoms on the three upper materials was addressed with a preliminary pilot investigation, with ten intervals sampled between 0 and 168h. Valves were seen to persist in detectable quantities after 168h on all three upper materials

  6. Graphene-assisted near-field radiative heat transfer between corrugated polar materials

    International Nuclear Information System (INIS)

    Liu, X. L.; Zhang, Z. M.

    2014-01-01

    Graphene has attracted great attention in nanoelectronics, optics, and energy harvesting. Here, the near-field radiative heat transfer between graphene-covered corrugated silica is investigated based on the exact scattering theory. It is found that graphene can improve the radiative heat flux between silica gratings by more than one order of magnitude and alleviate the performance sensitivity to lateral shift. The underlying mechanism is mainly attributed to the improved photon tunneling of modes away from phonon resonances. Besides, coating with graphene leads to nonlocal radiative transfer that breaks Derjaguin's proximity approximation and enables corrugated silica to outperform bulk silica in near-field radiation.

  7. Wheeled Vehicle Clutches, Transmissions, and Transfers. Military Curriculum Materials for Vocational and Technical Education.

    Science.gov (United States)

    Army Ordnance Center and School, Aberdeen Proving Ground, MD.

    This course is one of several subcourses that make up the entire Army correspondence course on wheeled vehicle maintenance. The subcourse is designed to provide the student with information about the operation, malfunction diagnosis, maintenance, and repair of wheeled vehicle clutches, transmissions, and transfer cases. It provides the basic…

  8. Control of radioactive wastes and coupling of neutron/gamma measurements: use of radiative capture for the correction of matrix effects that penalize the fissile mass measurement by active neutron interrogation

    International Nuclear Information System (INIS)

    Loche, F.

    2006-10-01

    In the framework of radioactive waste drums control, difficulties arise in the nondestructive measurement of fissile mass ( 235 U, 239 Pu..) by Active Neutron Interrogation (ANI), when dealing with matrices containing materials (Cl, H...) influencing the neutron flux. The idea is to use the neutron capture reaction (n,γ) to determine the matrix composition to adjust the ANI calibration coefficient value. This study, dealing with 118 litres, homogeneous drums of density less than 0,4 and composed of chlorinated and/or hydrogenated materials, leads to build abacus linking the γ ray peak areas to the ANI calibration coefficient. Validation assays of these abacus show a very good agreement between the corrected and true fissile masses for hydrogenated matrices (max. relative standard deviation: 23 %) and quite good for chlorinated and hydrogenated matrices (58 %). The developed correction method improves the measured values. It may be extended to 0,45 density, heterogeneous drums. (author)

  9. Accelerator based production of fissile nuclides, threshold uranium price and perspectives; Akceleratorska proizvodnja fisibilnih nuklida, granicna cijena urana i perspektive

    Energy Technology Data Exchange (ETDEWEB)

    Djordjevic, D [INIS-Inzenjering, Sarajevo (Yugoslavia); Knapp, V [Elektrotehnicki fakultet, zagreb (Yugoslavia)

    1988-07-01

    Accelerator breeder system characteristics are considered in this work. One such system which produces fissile nuclides can supply several thermal reactors with fissile fuel, so this system becomes analogous to an uranium enrichment facility with difference that fissile nuclides are produced by conversion of U-238 rather than by separation from natural uranium. This concept, with other long-term perspective for fission technology on the basis of development only one simpler technology. The influence of basic system characteristics on threshold uranium price is examined. Conditions for economically acceptable production are established. (author)

  10. Exchange of notes constituting an implementing arrangement, concerning international obligation exchanges, to the agreement between the Government of Australia and the European Atomic Energy Community (EURATOM) concerning transfers of nuclear material of 21 September 1981

    International Nuclear Information System (INIS)

    1993-01-01

    The implementing arrangement which entered into force on 8 September 1993, concerns the safeguard obligations attaching to nuclear material transferred or re transferred pursuant to the Agreement on Nuclear Transfers between Australia and the European Atomic Energy Community

  11. Heat transfer and material flow during laser assisted multi-layer additive manufacturing

    International Nuclear Information System (INIS)

    Manvatkar, V.; De, A.; DebRoy, T.

    2014-01-01

    A three-dimensional, transient, heat transfer, and fluid flow model is developed for the laser assisted multilayer additive manufacturing process with coaxially fed austenitic stainless steel powder. Heat transfer between the laser beam and the powder particles is considered both during their flight between the nozzle and the growth surface and after they deposit on the surface. The geometry of the build layer obtained from independent experiments is compared with that obtained from the model. The spatial variation of melt geometry, cooling rate, and peak temperatures is examined in various layers. The computed cooling rates and solidification parameters are used to estimate the cell spacings and hardness in various layers of the structure. Good agreement is achieved between the computed geometry, cell spacings, and hardness with the corresponding independent experimental results.

  12. A one-dimensional material transfer model for HECTR version 1.5

    International Nuclear Information System (INIS)

    Geller, A.S.; Wong, C.C.

    1991-08-01

    HECTR (Hydrogen Event Containment Transient Response) is a lumped-parameter computer code developed for calculating the pressure-temperature response to combustion in a nuclear power plant containment building. The code uses a control-volume approach and subscale models to simulate the mass, momentum, and energy transfer occurring in the containment during a loss-of-collant-accident (LOCA). This document describes one-dimensional subscale models for mass and momentum transfer, and the modifications to the code required to implement them. Two problems were analyzed: the first corresponding to a standard problem studied with previous HECTR versions, the second to experiments. The performance of the revised code relative to previous HECTR version is discussed as is the ability of the code to model the experiments. 8 refs., 5 figs., 3 tabs

  13. Industry to Education Technology Transfer Program. Composite Materials--Personnel Development. Final Report.

    Science.gov (United States)

    Tomezsko, Edward S. J.

    A composite materials education program was established to train Boeing Helicopter Company employees in the special processing of new filament-reinforced polymer composite materials. During the personnel development phase of the joint Boeing-Penn State University project, an engineering instructor from Penn State completed a 5-month, full-time…

  14. Low cost transportable device for transference of atmosphere sensitive materials from glove box to SEM

    DEFF Research Database (Denmark)

    Bentzen, Janet Jonna; Saxild, Finn B.

    Moisture or air sensitive materials are often encountered within several highly important fields such as catalyst R&D, pharmaceutical R&D, and battery R&D. Essential to all materials research and development is microstructure characterization, which often implies electron microscopy. Entering the...

  15. Merton and Ziman's modes of science: the case of biological and similar material transfer agreements

    NARCIS (Netherlands)

    Rodriguez, V.F.

    2007-01-01

    This paper makes a connection between recent studies on research materials exchange and its effect on the progress of science. Academia fears that scientific development could be hampered by the privatised practices of research material exchange. Since post-academic science represents a sufficient

  16. The influence of physical properties of materials used for slide rings on the process of heat transfer in the non-contacting face seals

    Directory of Open Access Journals (Sweden)

    Blasiak Slawomir

    2017-01-01

    Full Text Available The paper presents the results of analytical solution of the model of heat transfer for non-contacting face seals. Comparative analyses were performed for various physical properties of materials used for slide rings. A mathematical model includes a series of differential equations of partial derivatives with generally used boundary conditions, i.e. the Reynold’s equation, energy equation and heat transfer equations, which describe the heat transfer in sealing rings with surrounding medium. Heat transfer equation is written in the Cartesian coordinate system and solved using the Green’s functions method. Theoretical studies made it possible to draw a number of practical conclusions on the phenomena of heat transfer in the node seal. The presented model will allow more accurate identification of the heat transfer mechanism in the node seal. The results will help to select appropriate materials for sealing rings, depending on operating conditions of non-contacting face seals.

  17. Empirical Validation of Heat Transfer Performance Simulation of Graphite/PCM Concrete Materials for Thermally Activated Building System

    Directory of Open Access Journals (Sweden)

    Jin-Hee Song

    2017-01-01

    Full Text Available To increase the heat capacity in lightweight construction materials, a phase change material (PCM can be introduced to building elements. A thermally activated building system (TABS with graphite/PCM concrete hollow core slab is suggested as an energy-efficient technology to shift and reduce the peak thermal load in buildings. An evaluation of heat storage and dissipation characteristics of TABS in graphite/PCM concrete has been conducted using dynamic simulations, but empirical validation is necessary to acceptably predict the thermal behavior of graphite/PCM concrete. This study aimed to validate the thermal behavior of graphite/PCM concrete through a three-dimensional transient heat transfer simulation. The simulation results were compared to experimental results from previous studies of concrete and graphite/PCM concrete. The overall thermal behavior for both materials was found to be similar to experiment results. Limitations in the simulation modeling, which included determination of the indoor heat transfer coefficient, assumption of constant thermal conductivity with temperature, and assumption of specimen homogeneity, led to slight differences between the measured and simulated results.

  18. Transfer of impact ejecta material from the surface of Mars to Phobos and Deimos.

    Science.gov (United States)

    Chappaz, Loïc; Melosh, Henry J; Vaquero, Mar; Howell, Kathleen C

    2013-10-01

    The Russian Phobos-Grunt spacecraft originally planned to return a 200 g sample of surface material from Phobos to Earth. Although it was anticipated that this material would mainly be from the body of Phobos, there is a possibility that such a sample may also contain material ejected from the surface of Mars by large impacts. An analysis of this possibility is completed by using current knowledge of aspects of impact cratering on the surface of Mars and the production of high-speed ejecta that might reach Phobos or Deimos.

  19. Safeguarding nuclear weapon: Usable materials in Russia

    International Nuclear Information System (INIS)

    Cochran, T.

    1998-01-01

    Both the United States and Russia are retaining as strategic reserves more plutonium and HEU for potential reuse as weapons, than is legitimately needed. Both have engaged in discussions and have programs in various stages of development to dispose of excess plutonium and HEU. These fissile material disposition programs will take decades to complete. In the interim there will be, as there is now, hundreds of tons of separated weapon-usable fissile material stored in tens of thousands of transportable canisters, each containing from a few to several tons of kgs of weapon-usable fissile material. This material must be secured against theft and unauthorized use. To have high confidence that the material is secure, one must establish criteria against which the adequacy of the protective systems can be judged. For example, one finds such criteria in US Nuclear Regulatory Commission (USNRC) regulations for the protection of special nuclear materials

  20. Industry to Education Technical Transfer Program & Composite Materials. Composite Materials Course. Fabrication I Course. Fabrication II Course. Composite Materials Testing Course. Final Report.

    Science.gov (United States)

    Massuda, Rachel

    These four reports provide details of projects to design and implement courses to be offered as requirements for the associate degree program in composites and reinforced plastics technology. The reports describe project activities that led to development of curricula for four courses: composite materials, composite materials fabrication I,…

  1. Optimization and management of materials in earthwork construction : tech transfer summary.

    Science.gov (United States)

    2010-05-01

    This research provides solutions to identified problems through better : management and optimization of the available pavement geotechnical : materials and through ground improvement, soil reinforcement, : and other soil treatment techniques. : Objec...

  2. Representation of the neutron cross sections of several fertile and fissile nuclei in the resonance regions

    International Nuclear Information System (INIS)

    de Saussure, G.; Perez, R.B.

    1981-01-01

    Several aspects of the measurement, analysis and evaluation of the cross sections of the fertile and fissile nuclides in the resonance regions are discussed. In the resolved range, for the fertile nuclides it is thought that the principal requirement for improved evaluations is for a practical methodology to deal with systematic errors and their correlations. For the fissile nuclides 235 U and 239 Pu, the ENDF/B-V evaluations are not consistent with ENDF/B procedures recommendations and fall short of the goals of resonance analysis. New evaluations of these two isotopes should be performed. In the unresolved resonance region it is shown that the ENDF/B representation is ambiguous and is not theoretically justified. A better representation may be desirable, and a validation of the representation with experimental self-shielding and transmission measurements is certainly required. 105 references

  3. Disposal criticality analysis methodology for fissile waste forms

    International Nuclear Information System (INIS)

    Davis, J.W.; Gottlieb, P.

    1998-03-01

    A general methodology has been developed to evaluate the criticality potential of the wide range of waste forms planned for geologic disposal. The range of waste forms include commercial spent fuel, high level waste, DOE spent fuel (including highly enriched), MOX using weapons grade plutonium, and immobilized plutonium. The disposal of these waste forms will be in a container with sufficiently thick corrosion resistant barriers to prevent water penetration for up to 10,000 years. The criticality control for DOE spent fuel is primarily provided by neutron absorber material incorporated into the basket holding the individual assemblies. For the immobilized plutonium, the neutron absorber material is incorporated into the waste form itself. The disposal criticality analysis methodology includes the analysis of geochemical and physical processes that can breach the waste package and affect the waste forms within. The basic purpose of the methodology is to guide the criticality control features of the waste package design, and to demonstrate that the final design meets the criticality control licensing requirements. The methodology can also be extended to the analysis of criticality consequences (primarily increased radionuclide inventory), which will support the total performance assessment for the respository

  4. Hydrogen transfer reduction of polyketones catalyzed by iridium complexes: a novel route towards more biocompatible materials.

    Science.gov (United States)

    Milani, Barbara; Crottib, Corrado; Farnetti, Erica

    2008-09-14

    Transfer hydrogenation from 2-propanol to CO/4-methylstyrene and CO/styrene polyketones was catalyzed by [Ir(diene)(N-N)X] (N-N = nitrogen chelating ligand; X = halogen) in the presence of a basic cocatalyst. The reactions were performed using dioxane as cosolvent, in order to overcome problems due to low polyketone solubility. The polyalcohols were obtained in yields up to 95%, the conversions being markedly dependent on the nature of the ligands coordinated to iridium as well as on the experimental conditions.

  5. Agreement between the government of Australia and the government of the Republic of Korea concerning cooperation in peaceful uses of nuclear energy and the transfer of nuclear material

    International Nuclear Information System (INIS)

    1979-01-01

    The agreement contains fourteen articles under which the parties will cooperate in the peaceful uses of nuclear energy, including transfer of nuclear materials, research and development, exchange of unclassified information, technical training, visits by scientists and projects of mutual interest

  6. Addendum 2 to CSER 79-002: Extension of the 150 gram fissile limit used in room 187 of PFP

    International Nuclear Information System (INIS)

    Friar, D.E.

    1994-01-01

    The PFP operating organization requests that the limit set permitting 150 grams fissile be extended to the Hoods 4 and 5 of Room 187. The request for the limit change is explained in the attached request for analysis

  7. Four-electron transfer tandem tetracyanoquinodimethane for cathode-active material in lithium secondary battery

    Science.gov (United States)

    Kurimoto, Naoya; Omoda, Ryo; Mizumo, Tomonobu; Ito, Seitaro; Aihara, Yuichi; Itoh, Takahito

    2018-02-01

    Quinoid compounds are important candidates of organic active materials for lithium-ion batteries. However, its high solubility to organic electrolyte solutions and low redox potential are known as their major drawbacks. To circumvent these issues, we have designed and synthesized a tandem-tetracyanoquinonedimethane type cathode-active material, 11,11,12,12,13,13,14,14-octacyano-1,4,5,8-anthradiquinotetramethane (OCNAQ), that has four redox sites per molecule, high redox potential and suppressed solubility to electrolyte solution. Synthesized OCNAQ has been found to have two-step redox reactions by cyclic voltammetry, and each step consists of two-electron reactions. During charge-discharge tests using selected organic cathode-active materials with a lithium metal anode, the cell voltages obtained from OCNAQ are higher than those for 11,11-dicyanoanthraquinone methide (AQM) as expected, due to the strong electron-withdrawing effect of the cyano groups. Unfortunately, even with the use of the organic active material, the issue of dissolution to the electrolyte solution cannot be suppressed completely; however, appropriate choice of the electrolyte solutions, glyme-based electrolyte solutions in this study, give considerable improvement of the cycle retention (98% and 56% at 10 and 100 cycles at 0.5C, respectively). The specific capacity and energy density obtained in this study are 206 mAh g-1 and 554 mWh g-1 with respect to the cathode active material.

  8. Airborne radionuclides in the proglacial environment as indicators of sources and transfers of soil material.

    Science.gov (United States)

    Łokas, Edyta; Wachniew, Przemysław; Jodłowski, Paweł; Gąsiorek, Michał

    2017-11-01

    A survey of artificial ( 137 Cs, 238 Pu, 239+240 Pu, 241 Am) and natural ( 226 Ra, 232 Th, 40 K, 210 Pb) radioactive isotopes in proglacial soils of an Arctic glacier have revealed high spatial variability of activity concentrations and inventories of the airborne radionuclides. Soil column 137 Cs inventories range from below the detection limit to nearly 120 kBq m -2 , this value significantly exceeding direct atmospheric deposition. This variability may result from the mixing of materials characterised by different contents of airborne radionuclides. The highest activity concentrations observed in the proglacial soils may result from the deposition of cryoconites, which have been shown to accumulate airborne radionuclides on the surface of glaciers. The role of cryoconites in radionuclide accumulation is supported by the concordant enrichment of the naturally occurring airborne 210 Pb in proglacial soil cores showing elevated levels of artificial radionuclides. The lithogenic radionuclides show less variability than the airborne radionuclides because their activity concentrations are controlled only by the mixing of material derived from the weathering of different parent rocks. Soil properties vary little within and between the profiles and there is no unequivocal relationship between them and the radionuclide contents. The inventories reflect the pathways and time variable inputs of soil material to particular sites of the proglacial zone. Lack of the airborne radionuclides reflects no deposition of material exposed to the atmosphere after the 1950s or its removal by erosion. Inventories above the direct atmospheric deposition indicate secondary deposition of radionuclide-bearing material. Very high inventories indicate sites where transport pathways of cryoconite material terminated. Copyright © 2017 Elsevier Ltd. All rights reserved.

  9. Study of charge transfer processes in porphyrins- and phthalocyanins-based materials: from the liquid phase to the solid state

    International Nuclear Information System (INIS)

    Fournier, Thierry

    1994-01-01

    In order to efficiently conceive and build supramolecular materials for molecular electronics and optoelectronics, one need to have access to a large data base on the interactions between the elementary pieces of the material. Such a data base can be established only through the study of model Systems and model media. Oligomers of porphyrins and phthalocyanines constitute models of choice: due to the chemical versatility of the compounds, their physical and photophysical properties can be adjusted to produce a targeted function. The first part of this thesis is concerned with double- and triple-Decker mixed porphyrin and Phthalocyanines sandwich compounds of cerium. Then we study the photophysical properties of complexes formed by pairing in solution porphyrins and phthalocyanines bearing oppositely charged substituents. The charge transfer reactions and geminated recombinations are investigated by time-resolved absorption spectroscopy (from the femto- to millisecond time scales) for excited complexes either in solution, or confined in sol-gel matrices or in Langmuir-Blodgett films. The results obtained in the various media are compared and analysed by the Marcus theory. They allow to show that, for strongly coupled complexes, the solvent does not play any key role in the forward and backward electron transfer. We conclude this work by introducing a few targeted projects based on of the photophysical properties of these complexes, namely photodynamic therapy of cancers, nonlinear optics and the generation of photovoltage. (author) [fr

  10. Experimental study of plasma energy transfer and material erosion under ELM-like heat loads

    Energy Technology Data Exchange (ETDEWEB)

    Garkusha, I.E., E-mail: garkusha@ipp.kharkov.u [Institute of Plasma Physics of the NSC KIPT, Akademicheskaya 1, 61108 Kharkov (Ukraine); Makhlaj, V.A.; Chebotarev, V.V. [Institute of Plasma Physics of the NSC KIPT, Akademicheskaya 1, 61108 Kharkov (Ukraine); Landman, I. [Forschungszentrum Karlsruhe, IHM, 76021 Karlsruhe (Germany); Tereshin, V.I.; Aksenov, N.N.; Bandura, A.N. [Institute of Plasma Physics of the NSC KIPT, Akademicheskaya 1, 61108 Kharkov (Ukraine)

    2009-06-15

    Main features of plasma-surface interaction and energy transfer to tokamak plasma facing components are studied at different heat loads in ELM simulation experiments with the plasma gun QSPA Kh-50. Repetitive plasma exposures of tungsten, graphite and different combined W-C targets were performed at the pulse duration of 0.25 ms and the heat loads varied in the range 0.2-2.5 MJ/m{sup 2}. The onset of vapor shield in front of the surface was investigated. The evaporation is immediately followed by a saturation of surface heat load if further increasing the impact energy. The presence of graphite essentially decreases the heat flux to the nearby tungsten surface, which is due to the carbon vapor shield. Droplet splashing at the tungsten surface and formation of hot spots on the graphite surface are discussed.

  11. Experimental study of plasma energy transfer and material erosion under ELM-like heat loads

    International Nuclear Information System (INIS)

    Garkusha, I.E.; Makhlaj, V.A.; Chebotarev, V.V.; Landman, I.; Tereshin, V.I.; Aksenov, N.N.; Bandura, A.N.

    2009-01-01

    Main features of plasma-surface interaction and energy transfer to tokamak plasma facing components are studied at different heat loads in ELM simulation experiments with the plasma gun QSPA Kh-50. Repetitive plasma exposures of tungsten, graphite and different combined W-C targets were performed at the pulse duration of 0.25 ms and the heat loads varied in the range 0.2-2.5 MJ/m 2 . The onset of vapor shield in front of the surface was investigated. The evaporation is immediately followed by a saturation of surface heat load if further increasing the impact energy. The presence of graphite essentially decreases the heat flux to the nearby tungsten surface, which is due to the carbon vapor shield. Droplet splashing at the tungsten surface and formation of hot spots on the graphite surface are discussed.

  12. Near field heat transfer between random composite materials. Applications and limitations

    Energy Technology Data Exchange (ETDEWEB)

    Santiago, Eva Yazmin; Esquivel-Sirvent, Raul [Univ. Nacional Autonoma de Mexico (Mexico). Inst. de Fisica

    2017-05-01

    We present a theoretical study of the limits and bounds of using effective medium approximations in the calculation of the near field radiative heat transfer between a composite system made of Au nanoparticles in a SiC host and an homogeneous SiC slab. The effective dielectric function of the composite slab is calculated using three different approximations: Maxwell-Garnett, Bruggeman, and Looyenga's. In addition, we considered an empirical fit to the effective dielectric function by Grundquist and Hunderi. We show that the calculated value of the heat flux in the near field is dependent on the model, and the difference in the effective dielectric function is larger around the plasmonic response of the Au nanoparticles. This, in turn, accounts for the difference in the near field radiative heat flux. For all values of filling fractions, the Looyenga approximation gives a lower bound for the heat flux.

  13. Charge transfer from and to manganese phthalocyanine: bulk materials and interfaces

    Directory of Open Access Journals (Sweden)

    Florian Rückerl

    2017-08-01

    Full Text Available Manganese phthalocyanine (MnPc is a member of the family of transition-metal phthalocyanines, which combines interesting electronic behavior in the fields of organic and molecular electronics with local magnetic moments. MnPc is characterized by hybrid states between the Mn 3d orbitals and the π orbitals of the ligand very close to the Fermi level. This causes particular physical properties, different from those of the other phthalocyanines, such as a rather small ionization potential, a small band gap and a large electron affinity. These can be exploited to prepare particular compounds and interfaces with appropriate partners, which are characterized by a charge transfer from or to MnPc. We summarize recent spectroscopic and theoretical results that have been achieved in this regard.

  14. Surface PEGylation of mesoporous silica materials via surface-initiated chain transfer free radical polymerization: Characterization and controlled drug release.

    Science.gov (United States)

    Huang, Long; Liu, Meiying; Mao, Liucheng; Huang, Qiang; Huang, Hongye; Wan, Qing; Tian, Jianwen; Wen, Yuanqing; Zhang, Xiaoyong; Wei, Yen

    2017-12-01

    As a new type of mesoporous silica materials with large pore diameter (pore size between 2 and 50nm) and high specific surface areas, SBA-15 has been widely explored for different applications especially in the biomedical fields. The surface modification of SBA-15 with functional polymers has demonstrated to be an effective way for improving its properties and performance. In this work, we reported the preparation of PEGylated SBA-15 polymer composites through surface-initiated chain transfer free radical polymerization for the first time. The thiol group was first introduced on SBA-15 via co-condensation with γ-mercaptopropyltrimethoxysilane (MPTS), that were utilized to initiate the chain transfer free radical polymerization using poly(ethylene glycol) methyl ether methacrylate (PEGMA) and itaconic acid (IA) as the monomers. The successful modification of SBA-15 with poly(PEGMA-co-IA) copolymers was evidenced by a series of characterization techniques, including 1 H NMR, FT-IR, TGA and XPS. The final SBA-15-SH- poly(PEGMA-co-IA) composites display well water dispersity and high loading capability towards cisplatin (CDDP) owing to the introduction of hydrophilic PEGMA and carboxyl groups. Furthermore, the CDDP could be released from SBA-15-SH-poly(PEGMA-co-IA)-CDDP complexes in a pH dependent behavior, suggesting the potential controlled drug delivery of SBA-15-SH-poly(PEGMA-co-IA). More importantly, the strategy should be also useful for fabrication of many other functional materials for biomedical applications owing to the advantages of SBA-15 and well monomer adoptability of chain transfer free radical polymerization. Copyright © 2017 Elsevier B.V. All rights reserved.

  15. Probing photoinduced electron-transfer in graphene-dye hybrid materials for DSSC

    NARCIS (Netherlands)

    Guarracino, Paola; Gatti, Teresa; Canever, Nicolo; Abdu-Aguye, Mustapha; Loi, Maria Antonietta; Menna, Enzo; Franco, Lorenzo

    2017-01-01

    We investigated the photophysical properties of a newly synthesized hybrid material composed of a triphenylamine dye covalently bound to reduced graphene oxide, potentially relevant as a stable photosensitizer in dye-sensitized solar cells. The photophysical characterization has been carried out by

  16. Commissioning Measurements and Experience Obtained from the Installation of a Fissile Mass Flow monitor in the URAL Electrochemical Integrated Plant (UEIP) in Novouralsk

    International Nuclear Information System (INIS)

    March-Leuba, J.; Mastal, E.; Powell, D.; Sumner, J.; Uckan, T.; Vines, V.

    1999-01-01

    The Blend Down Monitoring System (BDMS) equipment sent earlier to the Ural Electrochemical Integrated Plant (UEIP) at Novouralsk, Russia, was installed and implemented successfully on February 2, 1999. The BDMS installation supports the highly enriched uranium (HEU) Transparency Implementation Program for material subject to monitoring under the HEU purchase agreement between the United States of America (USA) and the Russian Federation (RF). The BDMS consists of the Oak Ridge National Laboratory (ORNL) Fissile (uranium-235) Mass Flow Monitor (FMFM) and the Los Alamos National Laboratory (LANL) Enrichment Monitor (EM). Two BDMSs for monitoring the Main and Reserve HEU blending process lines were installed at UEIP. Independent operation of the FMFM Main and FMFM Reserve was successfully demonstrated for monitoring the fissile mass flow as well as the traceability of HEU to the product low enriched uranium. The FMFM systems failed when both systems were activated during the calibration phase due to a synchronization problem between the systems. This operational failure was caused by the presence of strong electromagnetic interference (EMI) in the blend point. The source-modulator shutter motion of the two FMFM systems was not being properly synchronized because of EMI producing a spurious signal on the synchronization cable connecting the two FMFM cabinets. The signature of this failure was successfully reproduced at ORNL after the visit. This unexpected problem was eliminated by a hardware modification and software improvements during a recent visit (June 9-11, 1999) to UEIP, and both systems are now operating as expected

  17. Fissile material holdup monitoring in the PREPP [Process Experimental Pilot Plant] process

    International Nuclear Information System (INIS)

    Becker, G.K.; Pawelko, R.J.

    1989-01-01

    The Process Experimental Pilot Plant (PREPP) is an incineration system designed to thermally process mixed transuranic (TRU) waste and TRU contaminated low-level waste. The TRU isotopic composition is that of weapons grade plutonium (Pu) which necessitates that criticality prevention measures by incorporated into the plant design and operation. Criticality safety in the PREPP process is assured through the utilization of mass and moderation control in conjunction with favorable vessel geometries. The subject of this paper concerns the Pu mass holdup instrumentation system which is an integral part of the inprocess mass control strategy. Plant vessels and components requiring real-time mass holdup measurements were selected based on their evaluated potential for achieving physically credible Pu mass loadings and associated parameters which could lead to a criticality event. If the parameters requisite to a criticality occurrence could not physically be achieved under credible plant conditions, the particular location only required periodic portable holdup monitoring. Based on these analyses five real-time holdup monitoring locations were identified for criticality assurance purposes. An additional real-time instrument is part of the system but serves primarily in the capacity of providing operational support data. 1 fig

  18. Measurements on an inventory of mixed fissile materials in shipping containers

    International Nuclear Information System (INIS)

    Rinard, P.M.; Krick, M.S.; Kelley, T.A.

    1997-09-01

    An inventory contained a large number of previously unmeasured items, many with both uranium and plutonium. We have assembled a suite of instruments and measured the items in a variety of ways. This report first considers the measurements and deduced results in detail before summarizing the important differences with the declarations of the inventory's database. The appendices referred to in this report are part of a classified version only and are not attached to this unclassified version. The classified report is by the same authors as this report, has the same title (which is unclassified), and is classified as open-quotes SRD.close quotes

  19. Interaction of Radiation with Graphene Based Nanomaterials for Sensing Fissile Materials

    Science.gov (United States)

    2016-03-01

    kelvin ( K ) Radiation curie (Ci) [activity of radionuclides] 3.7 × 10 10 per second (s –1 ) [becquerel (Bq)] roentgen (R) [air exposure] 2.579...quantum dots/ nanoparticles . Photosensitive hybrid devices made of CVD graphene decorated with cadmium selenide quantum dots (CdSe QDs) have been...valuable for understanding Raman spectra and electron-phonon physics in doped and disordered graphene. This study provides us valuable information about

  20. High-pressure {sup 4}He drift tubes for fissile material detection

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Zhehui, E-mail: zwang@lanl.gov [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Morris, Christopher L. [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Gray, F.E. [Regis University, Denver, CO 80221 (United States); Bacon, J.D.; Brockwell, M.I.; Chang, D.Y.; Chung, K.; Dai, W.G.; Greene, S.J.; Hogan, G.E.; Lisowski, P.W.; Makela, M.F.; Mariam, F.G.; McGaughey, P.L. [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Mendenhall, M. [California Institute of Technology, Pasadena, CA 91125 (United States); Milner, E.C.; Miyadera, H.; Murray, M.M.; Perry, J.O.; Roybal, J.D. [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); and others

    2013-03-01

    A detector efficiency model based on energy extraction from neutrons is described and used to compare {sup 4}He detectors with liquid scintillators (EJ301/NE-213). Detector efficiency can be divided into three regimes: single neutron scattering, multiple neutron scattering, and a transition regime in-between. For an average fission neutron of 2 MeV, the amount of {sup 4}He needed would be about 1/4 of the amount of the mass of EJ301/NE-213 in the single-scattering regime. For about 50% neutron energy extraction (1 MeV out of 2 MeV), the two types of detectors ({sup 4}He in the transition regime, EJ301 still in the single-scattering regime) have comparable mass, but {sup 4}He detectors can be much larger depending on the number density. A six-tube 11-bar-pressure {sup 4}He detector prototype is built and tested. Individual electrical pulses from the detector are recorded using a 12-bit digitizer. Differences in pulse rise time and amplitudes, due to different energy loss of neutrons and gamma rays, are used for neutron/gamma separation. Several energy spectra are also obtained and analyzed.

  1. Criticality safety of pipe systems which contain solutions of fissile materials

    International Nuclear Information System (INIS)

    Santos, R. dos.

    1982-03-01

    Criticality calculations for geometric configurations here studied make use of the neutron transport equation in its multigroup formulation, which is solved by the Monte Carlos statistical-probabilistic method. The computational code KENO IV, which use the Monte Carlo method, was utilized in all criticality calculations. All calculations were restricted to plutonium nitrate solutions, 100w% concentration of Pu-239, in water. Calculations were performed to obtain critical dimensions (radius) of a bare infinite cylinder and the effect produced by the addition of a 0.32 cm stainless steel cladding analyzed. Then, The most simple pipe intersection system is examined: the addition, of another cylinder to the one studied in the preceding case, constituting the type 'T' intersection. Further addition of a second cylinder, to the T-Type system is investigated; this is the cross-intersection type. Next, the effect produced by the introduction of a second central column to type 'T' system is analyzed. The effect of the introduction of several arms in the same quadrant is also studied. Infinite cylinders and cross-intersection type systems are analyzed in their nominal and maximum reflection conditions. (E.G.) [pt

  2. Immobilization as a route to surplus fissile materials disposition. Revision 1

    International Nuclear Information System (INIS)

    Gray, L.W.; Kan, T.; McKibben, J.M.

    1996-01-01

    The safe management of surplus weapons plutonium is a very important and urgent task with profound environmental, national and international security implications. In the aftermath of the Cold War, Presidential Police Directive 13 and various analysis by renown scientific, technical and international policy organizations have brought about a focused effort within the Department of Energy to identify and implement paths forward for the long term disposition of surplus weapons usable plutonium. The central, overarching goal is to render surplus weapons plutonium as inaccessible and unattractive for reuse in nuclear weapons, as the much larger and growing stock of plutonium contained in civilian spent reactor fuel. One disposition alternative considered for surplus Pu is immobilization, in which plutonium would be emplaced in glass, ceramic or glass-bonded zeolite. This option, along with some of the progress over the last year is discussed

  3. Addendum 3 to CSAR 80-027, Use of calorimeter 109B for fissile material measurement

    International Nuclear Information System (INIS)

    Chiao, T.

    1994-01-01

    This modification to the Plutonium Finishing Plant (PFP) calorimeter system involves removing current calorimeter No. 3 from the water bath and replacing it with a calorimeter that can accommodate larger diameter items (an oversize can). The inside diameters of both the sample and the reference cells will be increased to 5.835 inches at the top opening and to 5.22 inches at the bottom, the 8 inch high measurement zone. This Addendum 3 to Criticality Safety Analysis Report 80-027 examines criticality safety during the use of the modified calorimeter (Calorimeter 109B) with enlarged cell tube diameters to assure that an adequate margin of subcriticality is maintained for all normal and contingency conditions

  4. Deep borehole disposition of surplus fissile materials-The site selection process

    International Nuclear Information System (INIS)

    Heiken, G.; WoldeGabriel, G.; Morley, R.; Plannerer, H.

    1996-01-01

    One option for disposing of excess weapons plutonium is to place it near the base of deep boreholes in stable crystalline rocks. The technology exists to immediately begin the design of this means of disposition and there are many attractive sites available within the conterminous US. The borehole system utilizes mainly natural barriers to preven migration of Pu and U to the Earth's surface. Careful site selection ensures favorable geologic conditions that provide natural long-lived migration barriers; they include deep, extremely stable rock formations, strongly reducing brines that exhibit increasing salinity with depth, and most importantly, demonstrated isolation or non-communication of deep fluids with the biosphere for millions of years. This isolation is the most important characteristic, with the other conditions mainly being those that will enhance the potential of locating and maintaining the isolated zones. Candidate sites will probably be located on the craton in very old Precambrian crystalline rocks, most likely the center of a granitic pluton. The sites will be located in tectonically stable areas with no recent volcanic or seismic activity, and situated away from tectonic features that might become active in the near geologic future

  5. Fissile material management, an international approach of the future of plutonium

    International Nuclear Information System (INIS)

    Michel, A.; Schryvers, V.; Vanderborck, Y.

    2000-01-01

    Plutonium management is a crucial issue in any discussion on the future of nuclear energy: plutonium is indeed a normal by-product of nuclear electricity generation. As a result of long-term reprocessing strategies and recent decisions on the dismantling of nuclear weapons, separated plutonium stockpiles are increasing. Observing this situation, the Belgian Nuclear Society decided that the turn of the century was the right time to invite all the parties involved in decision making on this question to confront their decisions or the absence of it. As an international program committee was created, interested companies and institutions delegated high level experts to it and a comprehensive program was put together. This program covers: - Prospects for nuclear energy; - Public perception of plutonium; - The civil plutonium cycle; - The management of surplus military plutonium; - Non-proliferation and safeguards; - The reasons to improve the plutonium fuels performance. The conference is not scientific but strategic. It does not cover too many technical aspects but looks at the managerial questions. It is devoted to the reasons why things are done much more than how things are done. It allows to confront opinions with a mind open to all and a desire to make strategies transparent, even to the least informed public. The present paper has been written before the conference takes place in early October 2000 and describes the orientations prepared by the Programme committee. The oral presentation to Atalante 2000 will report in full over the Pu 2000 conference. (authors)

  6. Basic Research on Remote Sensing of Fissile Materials utilizing Gamma-rays and Neutrons

    Science.gov (United States)

    2017-02-01

    years, and an atomic mass number less than 60 was initially created for the proposal. From that list the student eliminated all reactions with an...decided that a lithium film grown on a nickel substrate and covered with a nickel film would be a good approach. They have learnt how to prepare thin...Nuclear Instruments and Methods in Physics Research A. 505 (2003) 1-4. 0 1 2 3 4 Shown in Figure 3 are key data taken with the lithium glass detectors

  7. Development of detection techniques for a single-particle of fissile material(II)

    Energy Technology Data Exchange (ETDEWEB)

    Sohn, S. C.; Kim, W. H.; Park, Y. J.; Song, B. C.; Jeon, Y. S.; Jee, K. Y.; Pyo, H. Y.; Kwack, E. H

    2001-06-01

    The Analytical methods and detection limit of signatures, and the particle discrimination techniques of unknown particles by microscope were investigated in this technical report. In connection with pre-treatment of swipe samples, sampling and treatment of particles, etching method, fission track observation and the preparation of sample for the neutron activation analysis were also described in this thchnical report.

  8. Proposed measurements of production of fissile materials from protons and deuterons

    International Nuclear Information System (INIS)

    Cork, B.

    1977-01-01

    There is a real need to improve the measurements of the multiplicities of neutrons produced in reactions suggested for electronuclear breeding, in order to choose the best target and projectile. Experiments to make these measurements are proposed. 2 figures

  9. Open Skies and monitoring a fissile materials cut-off treaty

    International Nuclear Information System (INIS)

    Allentuck, J.; Lemley, J.R.

    1995-01-01

    The Treaty on Open Skies (Open Skies) is intended among other things to provide, in the words of its preamble, means ''to facilitate the monitoring of compliance with existing or future arms control agreements.'' Open Skies permits overflights of the territory of member states by aircraft equipped with an array of sensors of various types. Their types and capabilities are treaty-limited. To find useful application in monitoring a cut-off treaty Open Skies would need to be amended. The number of signatories would need to be expanded so as to provide greater geographical coverage, and restrictions on sensor-array capabilities would need to be relaxed. To facilitate the detection of impending violations of a cut-off convention by Open Skies overflights, the data base provided by parties to the former should include among other things an enumeration of existing and former fuel cycle and research facilities including those converted to other uses, their precise geographic location, and a site plan

  10. TID Environmental Performance Testing In Support of the Mayak Fissile Material Storage Facility

    International Nuclear Information System (INIS)

    Tanner, Jennifer E.; Undem, Halvor A.; Roberts, Bruce A.; Griggs, James R.; Pratt, Sharon L.; Smith, Matthew H.

    2001-01-01

    The purpose of the test and evaluation of tamper indicating devices (TIDs) described in this report is to assure that the recommended TID technologies are acceptable for use at the Mayak FMSF. TID acceptance is based on TID performance with respect to mutually agreed functional, operational, and security requirements for the FMSF, taking into account both the United States and the Russian Federation views. Although some Russian views have been documented, very little information at the level required for formal test planning had been received prior to the start of the testing campaign. Consequently, this report currently represents US recommendations for an arms control and/or safeguards and security application. Acceptance of these test results and recommendations by those Russian Federation entities responsible for the Mayak FMSF will be required before implementing any TID regime at Mayak FMSF

  11. Development of detection techniques for a single-particle of fissile material(II)

    International Nuclear Information System (INIS)

    Sohn, S. C.; Kim, W. H.; Park, Y. J.; Song, B. C.; Jeon, Y. S.; Jee, K. Y.; Pyo, H. Y.; Kwack, E. H.

    2001-06-01

    The Analytical methods and detection limit of signatures, and the particle discrimination techniques of unknown particles by microscope were investigated in this technical report. In connection with pre-treatment of swipe samples, sampling and treatment of particles, etching method, fission track observation and the preparation of sample for the neutron activation analysis were also described in this thchnical report

  12. Forced convection heat transfer with slurry of phase change material in circular ducts: A phenomenological approach

    International Nuclear Information System (INIS)

    Royon, Laurent; Guiffant, Gerard

    2008-01-01

    A model describing the thermal behaviour of a slurry of phase change material flow in a circular duct is presented. Reactors connected in series are considered for the representation of the circular duct with constant wall temperature. A phenomenological equation is formulated to take account of the heat generation due to phase change in the particles. Results of the simulation present a plateau of temperature along the longitudinal direction, characteristic of the phase change. The effect of different parameters such as the Reynolds number, the weight fraction and the temperature of the cold spring on the length of the plateau is analysed. A correlation resulting from numerical results is proposed for use in the determination of the characteristics of the exchanger for a phase change material slurry

  13. Advances in Spiropyrans/Spirooxazines and Applications Based on Fluorescence Resonance Energy Transfer (FRET with Fluorescent Materials

    Directory of Open Access Journals (Sweden)

    Hongyan Xia

    2017-12-01

    Full Text Available Studies on the following were reviewed: (1 the structure of spiropyrans and spirooxazines (two kinds of spiro compounds under external stimuli and (2 the construction and applications of composite systems based on fluorescence resonance energy transfer (FRET with fluorescent materials. When treated with different stimuli (light, acids and bases, solvents, metal ions, temperature, redox potential, and so on, spiropyrans/spirooxazines undergo transformations between the ring-closed form (SP, the ring-opened merocyanine (MC form, and the protonated ring-opened form (MCH. This is due to the breakage of the spiro C–O bond and the protonation of MC, along with a color change. Various novel, multifunctional materials based on photochromic spiropyrans and spirooxazines have been successfully developed because of the vastly differently physiochemical properties posssed by the SP, MC and MCH forms. Among the three different structural forms, the MC form has been studied most extensively. The MC form not only gives complexes with various inorganic particles, biological molecules, and organic chemicals but also acts as the energy acceptor (of energy from fluorescent molecules during energy transfer processes that take place under proper conditions. Furthermore, spiropyran and spirooxazine compounds exhibit reversible physicochemical property changes under proper stimuli; this provides more advantages compared with other photochromic compounds. Additionally, the molecular structures of spiropyrans and spirooxazines can be easily modified and extended, so better compounds can be obtained to expand the scope of already known applications. Described in detail are: (1 the structural properties of spiropyrans and spirooxazines and related photochromic mechanisms; (2 composite systems based on spiropyrans and spirooxazines, and (3 fluorescent materials which have potential applications in sensing, probing, and a variety of optical elements.

  14. High pressure study of high temperatures superconductors: Material base, universal Tc-behavior, and charge transfer

    International Nuclear Information System (INIS)

    Chu, C.W.; Hor, P.H.; Lin, J.G.; Xiong, Q.; Huang, Z.J.; Meng, R.L.; Xue, Y.Y.; Jean, Y.C.

    1991-01-01

    The superconducting transition temperature (T c ) has been measured in YBa 2 Cu 3 O 6.7 , YBa 2 Cu 3 O 7 , Y 2 Ba 4 Cu 7 O 15 , YBa 2 Cu 4 O 8 , Tl 2 Ba 2 Ca n-1 Cu n O n+4-δ , La 2-x Sr x CuO 4 , and La 2-x Ba x CuO 4 under high pressures. The pressure effect on the positron lifetime (τ) has also been determined in the first four compounds. Based on these and other high pressure data, the authors suggest that (1) all known cuprate high temperature superconductors (HTS's) may be no more than mere modifications of either 214-T, 214-T', 123, or a combination of 214-T' and 123, (2) a nonmonotonic T c -behavior may govern the T c -variation of all hole cuprate HTS's and (3) pressure can induce charge transfer leading to a T c -change. The implications of these suggestions will also be discussed

  15. Obtaining the Bidirectional Transfer Distribution Function ofIsotropically Scattering Materials Using an Integrating Sphere

    Energy Technology Data Exchange (ETDEWEB)

    Jonsson, Jacob C.; Branden, Henrik

    2006-10-19

    This paper demonstrates a method to determine thebidirectional transfer distribution function (BTDF) using an integratingsphere. Information about the sample's angle dependent scattering isobtained by making transmittance measurements with the sample atdifferent distances from the integrating sphere. Knowledge about theilluminated area of the sample and the geometry of the sphere port incombination with the measured data combines to an system of equationsthat includes the angle dependent transmittance. The resulting system ofequations is an ill-posed problem which rarely gives a physical solution.A solvable system is obtained by using Tikhonov regularization on theill-posed problem. The solution to this system can then be used to obtainthe BTDF. Four bulk-scattering samples were characterised using both twogoniophotometers and the described method to verify the validity of thenew method. The agreement shown is great for the more diffuse samples.The solution to the low-scattering samples contains unphysicaloscillations, butstill gives the correct shape of the solution. Theorigin of the oscillations and why they are more prominent inlow-scattering samples are discussed.

  16. Evaluation of Glass Density to Support the Estimation of Fissile Mass Loadings from Iron Concentrations in SB6 Glasses

    Energy Technology Data Exchange (ETDEWEB)

    Edwards, T.; Peeler, D.

    2010-12-15

    The Department of Energy - Savannah River (DOE-SR) previously provided direction to Savannah River Remediation (SRR) to maintain fissile concentration in glass below 897 g/m{sup 3}. In support of the guidance, the Savannah River National Laboratory (SRNL) provided a technical basis and a supporting Microsoft{reg_sign} Excel{reg_sign} spreadsheet for the evaluation of fissile loading in Sludge Batch 5 glass based on the Fe concentration in glass as determined by the measurements from the Slurry Mix Evaporator (SME) acceptability analysis. SRR has since requested that SRNL provide the necessary information to allow SRR to update the Excel spreadsheet so that it may be used to maintain fissile concentration in glass below 897 g/m{sup 3} during the processing of Sludge Batch 6 (SB6). One of the primary inputs into the fissile loading spreadsheet includes a bounding density for SB6-based glasses. Based on the measured density data of select SB6 variability study glasses, SRNL recommends that SRR utilize the 99/99 Upper Tolerance Limit (UTL) density value at 38% WL (2.823 g/cm{sup 3}) as a bounding density for SB6 glasses to assess the fissile concentration in this glass system. That is, the 2.823 g/cm{sup 3} is recommended as a key (and fixed) input into the fissile concentration spreadsheet for SB6 processing. It should be noted that no changes are needed to the underlying structure of the Excel based spreadsheet to support fissile assessments for SB6. However, SRR should update the other key inputs to the spreadsheet that are based on fissile and Fe concentrations reported from the SB6 Waste Acceptance Product Specification (WAPS) sample. The purpose of this technical report is to present the density measurements that were determined for the SB6 variability study glasses and to conduct a statistical evaluation of these measurements to provide a bounding density value that may be used as input to the Excel{reg_sign} spreadsheet to be employed by SRR to maintain the

  17. Physical properties and heat transfer characteristics of materials for krypton-85 storage

    International Nuclear Information System (INIS)

    Christensen, A.B.

    1977-09-01

    Krypton-85 decay results in heat generation, and the subsequent temperature increase in the krypton-85 storage media must be evaluated. This report compiles the physical properties of krypton and of potential krypton-85 storage materials which are required to calculate the maximum temperature developed during storage. Temperature calculations were made for krypton-85 stored as a gas or immobilized solid in steel storage cylinders. The effects of krypton-85 loading, cylinder radius, storage media properties, and exterior cooling on storage temperature were shown

  18. High-flexibility combinatorial peptide synthesis with laser-based transfer of monomers in solid matrix material.

    Science.gov (United States)

    Loeffler, Felix F; Foertsch, Tobias C; Popov, Roman; Mattes, Daniela S; Schlageter, Martin; Sedlmayr, Martyna; Ridder, Barbara; Dang, Florian-Xuan; von Bojničić-Kninski, Clemens; Weber, Laura K; Fischer, Andrea; Greifenstein, Juliane; Bykovskaya, Valentina; Buliev, Ivan; Bischoff, F Ralf; Hahn, Lothar; Meier, Michael A R; Bräse, Stefan; Powell, Annie K; Balaban, Teodor Silviu; Breitling, Frank; Nesterov-Mueller, Alexander

    2016-06-14

    Laser writing is used to structure surfaces in many different ways in materials and life sciences. However, combinatorial patterning applications are still limited. Here we present a method for cost-efficient combinatorial synthesis of very-high-density peptide arrays with natural and synthetic monomers. A laser automatically transfers nanometre-thin solid material spots from different donor slides to an acceptor. Each donor bears a thin polymer film, embedding one type of monomer. Coupling occurs in a separate heating step, where the matrix becomes viscous and building blocks diffuse and couple to the acceptor surface. Furthermore, we can consecutively deposit two material layers of activation reagents and amino acids. Subsequent heat-induced mixing facilitates an in situ activation and coupling of the monomers. This allows us to incorporate building blocks with click chemistry compatibility or a large variety of commercially available non-activated, for example, posttranslationally modified building blocks into the array's peptides with >17,000 spots per cm(2).

  19. Status of radioactive material transport

    International Nuclear Information System (INIS)

    Kueny, Laurent

    2012-01-01

    As about 900.000 parcels containing radioactive materials are transported every year in France, the author recalls the main risks and safety principles associated with such transport. He indicates the different types of parcels defined by the regulation: excepted parcels, industrial non fissile parcels (type A), type B and fissile parcels, and highly radioactive type C parcels. He briefly presents the Q-system which is used to classify the parcels. He describes the role of the ASN in the control of transport safety, and indicates the different contracts existing between France or Areva and different countries (Germany, Japan, Netherlands, etc.) for the processing of used fuels in La Hague

  20. Mechanics of mass, energy and momentum transfer in complex textured materials at micro/nanoscales

    Science.gov (United States)

    Raman, Srikar

    The aim of this work is the investigation of the physical properties associated with nanostructured materials for various advanced applications which include controlled drug release, pressure driven nanofluidics, spray cooling etc. Polymer nanofibers (monolithic or core-shell) and turbostatic carbon nanotube bundles fabricated through electrospinning and co-electrospinning respectively were used as the key materials in this work. For controlled release applications, a model fluorescent dye Rhodamine 610 chloride, proteins, drugs or antigens encapsulated inside electrospun polymer nanofibers and its release to a buffer medium was analyzed. As a result of these experiments, it was discovered that the release process is limited by desorption process from nanopore surfaces. The experimental results were used as foundation as novel theory of release process and also allowed characterization of the relevant physical parameters of different compounds involved. In addition, thermal characterization of these electrospun polymer nanofibers was carried out to investigate their creep properties. The aim of this part was in the establishment of a detailed mechanism responsible for shrinkage of nanofiber mats at elevated temperatures and elucidation of its relation to the microscopic thermally-induced changes occurring in the polymer structure. In particular, thermal behavior of Poly(epsilon-caprolactone) (PCL), Poly(methylmethacrylate) (PMMA), Polyacrylonitrile (PAN) and Polyurethane (PU) in electrospun nanofibers and original pellets were studied using Differential Scanning Calorimetry (DSC) and linked to the onset of thermally-induced shrinkage of nanofiber mats. The elctrospinning setup was then extended to Co-electrospinning process for fabricating Turbostratic Carbon Nanotube Bundles, for pressure driven flow of suspensions. Using a model water soluble compound, fluorescent dye Rhodamine 610 chloride, it was shown that deposit buildup on the inner walls of the delivery