WorldWideScience

Sample records for mixed fissile materials

  1. 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

  2. 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

  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. 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

  5. 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.

  6. 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

  7. 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

  8. 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

  9. 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.

  10. 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

  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. 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

  14. 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

  15. 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

  16. 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

  17. 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

  18. 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.

  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. 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.

  3. 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

  4. 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

  5. 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

  6. 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

  7. 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.

  8. 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

  9. 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

  10. 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

  11. 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

  12. 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

  13. 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

  14. 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

  15. 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

  16. 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.

  17. 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

  18. 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

  19. 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

  20. 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)

  1. 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)

  2. 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

  3. 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

  4. 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

  5. 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)

  6. 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

  7. 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.)

  8. 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

  9. 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

  10. 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

  11. 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

  12. 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

  13. 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

  14. 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

  15. 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

  16. 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

  17. 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.)

  18. 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

  19. 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

  20. 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.

  1. 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)

  2. 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

  3. 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

  4. 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

  5. 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

  6. 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)

  7. 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

  8. 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

  9. 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

  10. 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)

  11. 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.

  12. 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)

  13. 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

  14. 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

  15. 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.

  16. 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)

  17. 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)

  18. 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

  19. 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.

  20. 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...

  1. 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

  2. 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

  3. 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

  4. 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

  5. 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

  6. 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

  7. 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.

  8. 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)

  9. 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

  10. 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)

  11. 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)

  12. 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

  13. 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

  14. 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

  15. 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

  16. 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

  17. 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)

  18. 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

  19. 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

  20. 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.

  1. 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

  2. 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

  3. 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

  4. 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

  5. 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)

  6. 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

  7. 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

  8. 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.

  9. 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)

  10. 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)

  11. 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)

  12. 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.

  13. 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

  14. 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.

  15. 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

  16. 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

  17. 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.

  18. 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

  19. 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)

  20. 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)

  1. 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

  2. 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

  3. 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.)

  4. 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

  5. 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

  6. 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

  7. 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

  8. 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)

  9. 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

  10. 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)

  11. 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)

  12. 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

  13. 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.)

  14. 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

  15. 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.

  16. 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.

  17. 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.

  18. 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

  19. 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

  20. 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

  1. 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

  2. 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

  3. 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

  4. 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

  5. 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

  6. 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)

  7. 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

  8. 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

  9. 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

  10. 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)

  11. 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)

  12. 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

  13. 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.

  14. 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

  15. 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)

  16. 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

  17. 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.

  18. 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)

  19. 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

  20. 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

  1. 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)

  2. 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)

  3. 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)

  4. 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)

  5. 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

  6. 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)

  7. 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

  8. 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

  9. 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.)

  10. 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

  11. 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

  12. 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

  13. 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

  14. 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.)

  15. 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)

  16. 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

  17. 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)

  18. 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...

  19. 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.

  20. 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

  1. 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

  2. Material Barriers to Diffusive Mixing

    Science.gov (United States)

    Haller, George; Karrasch, Daniel

    2017-11-01

    Transport barriers, as zero-flux surfaces, are ill-defined in purely advective mixing in which the flux of any passive scalar is zero through all material surfaces. For this reason, Lagrangian Coherent Structures (LCSs) have been argued to play the role of mixing barriers as most repelling, attracting or shearing material lines. These three kinematic concepts, however, can also be defined in different ways, both within rigorous mathematical treatments and within the realm of heuristic diagnostics. This has lead to a an ever-growing number of different LCS methods, each generally identifying different objects as transport barriers. In this talk, we examine which of these methods have actual relevance for diffusive transport barriers. The latter barriers are arguably the practically relevant inhibitors in the mixing of physically relevant tracers, such as temperature, salinity, vorticity or potential vorticity. We demonstrate the role of the most effective diffusion barriers in analytical examples and observational data. Supported in part by the DFG Priority Program on Turbulent Superstructures.

  3. 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

  4. 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

  5. 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

  6. 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

  7. 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.

  8. 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

  9. 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

  10. 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

  11. 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

  12. 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

  13. 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.

  14. 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

  15. 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)

  16. 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

  17. 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.

  18. 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

  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. Dry powder mixes comprising phase change materials

    Science.gov (United States)

    Salyer, Ival O.

    1992-01-01

    Free flowing, conformable powder-like mix of silica particles and a phase change material (p.c.m.) is disclosed. The silica particles have a critical size of about 7.times.10.sup.-3 to about 7.times.10.sup.-2 microns and the pcm must be added to the silica in an amount of 80 wt. % or less pcm per combined weight of silica and pcm. The powder-like mix can be used in tableware items, medical wraps, tree wraps, garments, quilts and blankets, and in cementitious compositions of the type in which it is beneficial to use a pcm material. The silica-pcm mix can also be admixed with soil to provide a soil warming effect and placed about a tree, flower, or shrub.

  3. Dry powder mixes comprising phase change materials

    Science.gov (United States)

    Salyer, I.O.

    1994-02-01

    Free flowing, conformable powder-like mix of silica particles and a phase change material (PCM) is provided. The silica particles have a critical size of about 0.005 to about 0.025 microns and the PCM must be added to the silica in an amount of 75% or less PCM per combined weight of silica and PCM. The powder-like mix can be used in tableware items, medical wraps, tree wraps, garments, quilts and blankets, and in cementitious compositions of the type in which it is beneficial to use a PCM material. The silica-PCM mix can also be admixed with soil to provide a soil warming effect and placed about a tree, flower, or shrub. 2 figures.

  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. Co-disposal of mixed waste materials

    International Nuclear Information System (INIS)

    Phillips, S.J.; Alexander, R.G.; Crane, P.J.; England, J.L.; Kemp, C.J.; Stewart, W.E.

    1993-08-01

    Co-disposal of process waste streams with hazardous and radioactive materials in landfills results in large, use-efficiencies waste minimization and considerable cost savings. Wasterock, produced from nuclear and chemical process waste streams, is segregated, treated, tested to ensure regulatory compliance, and then is placed in mixed waste landfills, burial trenches, or existing environmental restoration sites. Large geotechnical unit operations are used to pretreat, stabilize, transport, and emplace wasterock into landfill or equivalent subsurface structures. Prototype system components currently are being developed for demonstration of co-disposal

  7. 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

  8. 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

  9. 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.

  10. 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

  11. 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

  12. 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

  13. Extrusion-mixing compared with hand-mixing of polyether impression materials?

    Science.gov (United States)

    McMahon, Caroline; Kinsella, Daniel; Fleming, Garry J P

    2010-12-01

    The hypotheses tested were two-fold (a) whether altering the base:catalyst ratio influences working time, elastic recovery and strain in compression properties of a hand-mixed polyether impression material and (b) whether an extrusion-mixed polyether impression material would have a significant advantage over a hand-mixed polyether impression material mixed to the optimum base:catalyst ratio. The polyether was hand-mixed at the optimum (manufacturers recommended) base:catalyst ratios (7:1) and further groups were made by increasing or decreasing the catalyst length by 25%. Additionally specimens were also made from an extrusion-mixed polyether impression material and compared with the optimum hand-mixed base:catalyst ratio. A penetrometer assembly was used to measure the working time (n=5). Five cylindrical specimens for each hand-mixed and extrusion mixed group investigated were employed for elastic recovery and strain in compression testing. Hand-mixing polyether impression materials with 25% more catalyst than that recommended significantly decreased the working time while hand-mixing with 25% less catalyst than that recommended significantly increased the strain in compression. The extrusion-mixed polyether impression material provided similar working time, elastic recovery and strain in compression to the hand-mixed polyether mixed at the optimum base:catalyst ratio.

  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. 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.

  16. 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.

  17. 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

  18. 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.

  19. 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

  20. Mixed plasma-facing materials research at INEEL

    International Nuclear Information System (INIS)

    Anderl, R.A.; Longhurst, G.R.; Pawelko, R.J.

    2001-01-01

    Mixed-materials research at the Idaho National Engineering and Environmental Laboratory (INEEL) has focused on Be-C and W-C systems. The purpose of this work was to investigate hydrogen isotope retention in these systems. Plasma-mixed material layers using carbon coated Be and W specimens that were heat-treated and tungsten carbide specimens prepared by chemical vapor deposition (CVD) were simulated. Hydrogen isotope retention was investigated by means of thermal desorption spectroscopy (TDS) measurements on deuterium implanted samples

  1. A global renewable mix with proven technologies and common materials

    International Nuclear Information System (INIS)

    García-Olivares, Antonio; Ballabrera-Poy, Joaquim; García-Ladona, Emili; Turiel, Antonio

    2012-01-01

    A global alternative mix to fossil fuels is proposed, based on proven renewable energy technologies that do not use scarce materials. The mix consists of a combination of onshore and offshore wind turbines, concentrating solar power stations, hydroelectricity and wave power devices attached to the offshore turbines. Solar photovoltaic power could contribute to the mix if its dependence on scarce materials is solved. The most adequate deployment areas for the power stations are studied, as well as the required space. Material requirements are studied for the generation, power transport and for some future transport systems. The order of magnitude of copper, aluminium, neodymium, lithium, nickel, zinc and platinum that may be required for the proposed solution is obtained and compared with available reserves. Overall, the proposed global alternative to fossil fuels seems technically feasible. However, lithium, nickel and platinum could become limiting materials for future vehicles fleet if no global recycling systems were implemented and rechargeable zinc–air batteries would not be developed; 60% of the current copper reserves would have to be employed in the implementation of the proposed solution. Altogether, they may become a long-term physical constraint, preventing the continuation of the usual exponential growth of energy consumption. - Highlights: ▶ A global renewable mix with proven energy technologies and common materials. ▶ Wind turbines, concentrating solar power, hydroelectricity and wave attenuators. ▶ Mix technically feasible. Lithium, nickel and platinum may limit vehicles fleet. ▶ Sixty per cent of copper reserves used in the mix and in societal electrification. ▶ Power cannot growth exponentially. Future “spaceship economy” scenario expected.

  2. 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

  3. Characterization of Concrete Mixes Containing Phase Change Materials

    Science.gov (United States)

    Paksoy, H.; Kardas, G.; Konuklu, Y.; Cellat, K.; Tezcan, F.

    2017-10-01

    Phase change materials (PCM) can be used in passive building applications to achieve near zero energy building goals. For this purpose PCM can be added in building structures and materials in different forms. Direct incorporation, form stabilization and microencapsulation are different forms used for PCM integration in building materials. In addition to thermal properties of PCM itself, there are several other criteria that need to be fulfilled for the PCM enhanced building materials. Mechanical properties, corrosive effects, morphology and thermal buffering have to be determined for reliable and long-term applications in buildings. This paper aims to give an overview of characterization methods used to determine these properties in PCM added fresh concrete mixes. Thermal, compressive strength, corrosion, and microscopic test results for concrete mixes with PCM are discussed.

  4. 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

  5. MIXING OF INCOMPATIBLE MATERIALS IN WASTE TANKS TECHNICAL BASIS DOCUMENT

    International Nuclear Information System (INIS)

    SANDGREN, K.R.

    2006-01-01

    This document presents onsite radiological, onsite toxicological, and offsite toxicological consequences, risk binning, and control decision results for the mixing of incompatible materials in waste tanks representative accident. Revision 4 updates the analysis to consider bulk chemical additions to single shell tanks (SSTs)

  6. Mixed materials for concrete. Concrete yo konwazai ni tsuite

    Energy Technology Data Exchange (ETDEWEB)

    Kono, K [Tokushima Univ., Tokushima (Japan). Faculty of Engineering

    1994-07-05

    The materials except cement, water and aggregate added into the cement paste, mortar or concrete before the execution of smashing are called mixed materials. The mixed materials are indispensable to the concrete for improving the quality of the fresh concrete as well as the hardened concrete; providing the characteristics suitable for the operation; and increasing the economy. In this paper, the main mixed materials including fly ash, which is the by-product in coal thermoelectric power station; silica fume; micropowder of slag in blast furnace; expansive materials and so on are described summarily. Especially, silica fume is the by-product, which are the super micro-powders with the average size around 0.1 micrometer, collected by the dust-collector from the waste gas generated during the manufacture in the electric furnace of ferrosilicon, which is an alloy iron, or silicon metal used as the deacidificating and desulfurizing agents in the steel production. But the most part thereof is depended on the import since the domestic output is low. 38 refs., 19 figs., 6 tabs.

  7. 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

  8. Mixed arrays - problems with current methods and rules

    International Nuclear Information System (INIS)

    Mennerdahl, D.

    1987-01-01

    Simplified methods are used to control the criticality safety of mixed arrays (non-identical units) in storage or in transport. The basis for these methods is that the analyses of arrays of identical units are sufficient for drawing proper conclusions on mixed arrays. In a recent study of the rules for transport, two general flaws in such methods have been identified. One flaw is caused by increased neutron return rate to the central part of the array. The other flaw is caused by increased neutron coupling between two or more fissile units in an array. In both cases, replacement of fissile units with other units, which appear to be less reactive, can lead to criticality. This paper shows that the two flaws are common in also in current methods used for storage of fissile materials. (author)

  9. 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

  10. Nonlinear Wave Mixing Technique for Nondestructive Assessment of Infrastructure Materials

    Science.gov (United States)

    Ju, Taeho

    To operate safely, structures and components need to be inspected or monitored either periodically or in real time for potential failure. For this purpose, ultrasonic nondestructive evaluation (NDE) techniques have been used extensively. Most of these ultrasonic NDE techniques utilize only the linear behavior of the ultrasound. These linear techniques are effective in detecting discontinuities in materials such as cracks, voids, interfaces, inclusions, etc. However, in many engineering materials, it is the accumulation of microdamage that leads to degradation and eventual failure of a component. Unfortunately, it is difficult for linear ultrasonic NDE techniques to characterize or quantify such damage. On the other hand, the acoustic nonlinearity parameter (ANLP) of a material is often positively correlated with such damage in a material. Thus, nonlinear ultrasonic NDE methods have been used in recently years to characterize cumulative damage such as fatigue in metallic materials, aging in polymeric materials, and degradation of cement-based materials due to chemical reactions. In this thesis, we focus on developing a suit of novel nonlinear ultrasonic NDE techniques based on the interactions of nonlinear ultrasonic waves, namely wave mixing. First, a noncollinear wave mixing technique is developed to detect localized damage in a homogeneous material by using a pair of noncollinear a longitudinal wave (L-wave) and a shear wave (S-wave). This pair of incident waves make it possible to conduct NDE from a single side of the component, a condition that is often encountered in practical applications. The proposed noncollinear wave mixing technique is verified experimentally by carrying out measurements on aluminum alloy (AA 6061) samples. Numerical simulations using the Finite Element Method (FEM) are also conducted to further demonstrate the potential of the proposed technique to detect localized damage in structural components. Second, the aforementioned nonlinear

  11. 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

  12. Use of selected waste materials in concrete mixes.

    Science.gov (United States)

    Batayneh, Malek; Marie, Iqbal; Asi, Ibrahim

    2007-01-01

    A modern lifestyle, alongside the advancement of technology has led to an increase in the amount and type of waste being generated, leading to a waste disposal crisis. This study tackles the problem of the waste that is generated from construction fields, such as demolished concrete, glass, and plastic. In order to dispose of or at least reduce the accumulation of certain kinds of waste, it has been suggested to reuse some of these waste materials to substitute a percentage of the primary materials used in the ordinary portland cement concrete (OPC). The waste materials considered to be recycled in this study consist of glass, plastics, and demolished concrete. Such recycling not only helps conserve natural resources, but also helps solve a growing waste disposal crisis. Ground plastics and glass were used to replace up to 20% of fine aggregates in concrete mixes, while crushed concrete was used to replace up to 20% of coarse aggregates. To evaluate these replacements on the properties of the OPC mixes, a number of laboratory tests were carried out. These tests included workability, unit weight, compressive strength, flexural strength, and indirect tensile strength (splitting). The main findings of this investigation revealed that the three types of waste materials could be reused successfully as partial substitutes for sand or coarse aggregates in concrete mixtures.

  13. Use of selected waste materials in concrete mixes

    International Nuclear Information System (INIS)

    Batayneh, Malek; Marie, Iqbal; Asi, Ibrahim

    2007-01-01

    A modern lifestyle, alongside the advancement of technology has led to an increase in the amount and type of waste being generated, leading to a waste disposal crisis. This study tackles the problem of the waste that is generated from construction fields, such as demolished concrete, glass, and plastic. In order to dispose of or at least reduce the accumulation of certain kinds of waste, it has been suggested to reuse some of these waste materials to substitute a percentage of the primary materials used in the ordinary portland cement concrete (OPC). The waste materials considered to be recycled in this study consist of glass, plastics, and demolished concrete. Such recycling not only helps conserve natural resources, but also helps solve a growing waste disposal crisis. Ground plastics and glass were used to replace up to 20% of fine aggregates in concrete mixes, while crushed concrete was used to replace up to 20% of coarse aggregates. To evaluate these replacements on the properties of the OPC mixes, a number of laboratory tests were carried out. These tests included workability, unit weight, compressive strength, flexural strength, and indirect tensile strength (splitting). The main findings of this investigation revealed that the three types of waste materials could be reused successfully as partial substitutes for sand or coarse aggregates in concrete mixtures

  14. Physical Hypermedia: Organising Collections of Mixed Physical and Digital Material

    DEFF Research Database (Denmark)

    Grønbæk, Kaj; Kristensen, Jannie Friis; Ørbæk, Peter

    2003-01-01

    This paper introduces the notion of physical hypermedia, addressing the problem of organizing material in mixed digital and physical environments. Based on empirical studies, we propose concepts for collectional actions and meta-data actions, and present prototypes combining principles from...... augmented reality and hypermedia to support organization of mixtures of digital and physical materials. Our prototype of a physical hypermedia system is running on an augmented architect's desk and digital walls utilizing Radio Frequency Identifier (RFID) tags as well as visual tags tracked by cameras....... It allows users to tag physical materials, and have these tracked by readers (antennas) that may become pervasive in our work environments. In the physical hypermedia system, we work with three categories of RFID tags: simple object tags, collectional tags, and tooltags invoking operations such as grouping...

  15. The use of shale ash in dry mix construction materials

    Science.gov (United States)

    Gulbe, L.; Setina, J.; Juhnevica, I.

    2017-10-01

    The research was made to determine the use of shale ash usage in dry mix construction materials by replacing part of cement amount. Cement mortar ZM produced by SIA Sakret and two types of shale ashes from Narva Power plant (cyclone ash and electrostatic precipitator ash) were used. Fresh mortar properties, hardened mortar bulk density, thermal conductivity (λ10, dry) (table value) were tested in mortar ZM samples and mortar samples in which 20% of the amount of cement was replaced by ash. Compressive strenght, frost resistance and resistance to sulphate salt solutions were checked. It was stated that the use of electrostatic precipitator ash had a little change of the material properties, but the cyclone ash significantly reduced the mechanical strength of the material.

  16. Assessing of bulk materials mixing and sorting by radiotracer methods

    International Nuclear Information System (INIS)

    Thyn, J.

    1983-01-01

    Various applications are indicated of tracer techniques for the evaluation of mixing and sorting of mixtures of solid particles. The evaluation of the process of mixing, i.e., the determination of the homogenization time is done by labelling of the entire volume of the monitored component of the mixture and continuous detection of radiation through the walls of the mixer using one or several detectors. The evaluation of the character of the flow and the evacuation of solid particles from the bin is done by labelling with a radiotracer the material which is spread out on the top along the whole cross-section of the bin, and the concentration is monitored of the tracer in the material outflow. The evaluation of material sorting in bins which takes place during the filling and emptying is done on the basis of significance tests or using self-correlation functions and frequency characteristics. Also monitored was the dependence of the equalizing ability of the continuous gravity mixer at the vertex angle of the tip. (M.D.)

  17. Mixed conducting materials for partial oxidation of hydrocarbons

    Directory of Open Access Journals (Sweden)

    Frade, J. R.

    2004-06-01

    Full Text Available Thermodynamic calculations with additional conditions for the conservation of carbon and hydrogen were used to predict the gas composition obtained by partial oxidation of methane as a function of oxygen partial pressure and temperature; this was used to assess the stability and oxygen permeability requirements of mixed conducting membrane materials proposed for this purpose. A re-examination of known mixed conductors shows that most materials with highest permeability still fail to fulfil the requirements of stability under reducing conditions. Other materials possess sufficient stability but their oxygen permeability is insufficient. Different approaches were thus used to attempt to overcome those limitations, including changes in composition in the A and B site positions of ABO3 perovskites, and tests of materials with different structure types. Promising results were obtained mainly for some materials with perovskite or related K2NiF4-type structures. Limited stability of the most promising materials shows that one should rely mainly on kinetic limitations in the permeate side to protect the mixed conductor from severe reducing conditions.

    Se han usado cálculos termodinámicos con condiciones adicionales para la conservación del carbono e hidrógeno para predecir la composición del gas obtenido mediante la oxidación parcial del metano en función de la presión parcial de oxígeno y de la temperatura; esto se ha usado para asegurar los requerimientos de estabilidad y permeabilidad al oxígeno de los materiales conductores mixtos empleados como membrana para este propósito. Un nuevo exámen de los conductores mixtos conocidos muestra que la mayoría de los materiales con la mayor permeabilidad todavía fallan en el cumplimiento de los requerimientos de estabilidad bajo condiciones reductoras. Otros materiales poseen suficiente estabilidad, pero su permeabilidad al oxígeno es insuficiente. Por ello se han empleado diferentes

  18. 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)

  19. 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)

  20. 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.

  1. 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

  2. Mixing of incompatible materials in waste tanks technical basis document

    International Nuclear Information System (INIS)

    SANDGREN, K.R.

    2003-01-01

    This technical basis document was developed to support the Tank Farms Documented Safety Analysis (DSA) and describes the risk binning process, the technical basis for assigning risk bins, and the controls selected for the mixing of incompatible materials representative accident and associated represented hazardous conditions. The purpose of the risk binning process is to determine the need for safety-significant structures, systems, and components (SSCs) and/or technical safety requirement (TSR)-level controls for a given representative accident or represented hazardous conditions based on an evaluation of the FR-equency and consequence. Note that the risk binning process is not applied to facility workers, because all facility worker hazardous conditions are considered for safety-significant SSCs and/or TSR level controls. Determination of the need for safety-class SSCs was performed in accordance with DOE-STD-3009-94, ''Preparation Guide for US Department of Energy Nonreactor Nuclear Facility Documented Safety Analyses'', as described in this report

  3. 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

  4. 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

  5. 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...

  6. Integrated Data Collection Analysis (IDCA) Program - Mixing Procedures and Materials Compatibility

    Energy Technology Data Exchange (ETDEWEB)

    Olinger, Becky D. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Sandstrom, Mary M. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Warner, Kirstin F. [Naval Surface Warfare Center (NSWC-IHD), Indian Head, MD (United States); Sorensen, Daniel N. [Naval Surface Warfare Center (NSWC-IHD), Indian Head, MD (United States); Remmers, Daniel L. [Naval Surface Warfare Center (NSWC-IHD), Indian Head, MD (United States); Moran, Jesse S. [Naval Surface Warfare Center (NSWC-IHD), Indian Head, MD (United States); Shelley, Timothy J. [Air Force Research Lab. (AFRL), Tyndall AFB, FL (United States); Whinnery, LeRoy L. [Sandia National Lab. (SNL-CA), Livermore, CA (United States); Hsu, Peter C. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Whipple, Richard E. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Kashgarian, Michaele [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Reynolds, John G. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2011-01-14

    Three mixing procedures have been standardized for the IDCA proficiency test—solid-solid, solid-liquid, and liquid-liquid. Due to the variety of precursors used in formulating the materials for the test, these three mixing methods have been designed to address all combinations of materials. Hand mixing is recommended for quantities less than 10 grams and Jar Mill mixing is recommended for quantities over 10 grams. Consideration must also be given to the type of container used for the mixing due to the wide range of chemical reactivity of the precursors and mixtures. Eight web site sources from container and chemical manufacturers have been consulted. Compatible materials have been compiled as a resource for selecting containers made of materials stable to the mixtures. In addition, container materials used in practice by the participating laboratories are discussed. Consulting chemical compatibility tables is highly recommended for each operation by each individual engaged in testing the materials in this proficiency test.

  7. Fatigue behaviour of bituminous materials : from binders to mixes

    OpenAIRE

    SOENEN, H; DE LA ROCHE, C; REDELIUS, P

    2003-01-01

    Test procedures, aiming at measuring fatigue directly on bituminous binders, are increasingly used. The purpose of this paper is to investigate the relevance of this type of binder fatigue tests and to compare the results with laboratory fatigue properties of the corresponding mixes, using one mix composition for all binders, and similar fatigue tests conditions. Eight binders were selected, derived from two crude sources, including an oxidised and two polymer modified samples. All fatigue te...

  8. 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

  9. 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

  10. Determination of optimum mixing time for raw materials with the tracer method in the glass industry

    Energy Technology Data Exchange (ETDEWEB)

    Gallyas, M; Gemesi, J; Kurinka, J

    1983-02-01

    The authors explain how the optimum mixing time for the raw materials for glass manufacture can be determined with the aid of the radioactive tracer method. Basing themselves on measurements, they indicate the change in the degree of mixing of the individual components (soda (Na-24), sodium sulphate, coke (La-140) and bone meal (P-32) as a function of mixing time. The optimum degree of mixing and mixing time for dry and for wet mixing are determined. Finally, data for determining the permissible storage time of the mixture are given.

  11. Effect of mixing techniques on bacterial attachment and disinfection time of polyether impression material.

    Science.gov (United States)

    Guler, Umut; Budak, Yasemin; Ruh, Emrah; Ocal, Yesim; Canay, Senay; Akyon, Yakut

    2013-09-01

    The aim of this study was 2-fold. The first aim was to evaluate the effects of mixing technique (hand-mixing or auto-mixing) on bacterial attachment to polyether impression materials. The second aim was to determine whether bacterial attachment to these materials was affected by length of exposure to disinfection solutions. Polyether impression material samples (n = 144) were prepared by hand-mixing or auto-mixing. Escherichia coli, Staphylococcus aureus and Pseudomonas aeruginosa were used in testing. After incubation, the bacterial colonies were counted and then disinfectant solution was applied. The effect of disinfection solution was evaluated just after the polymerization of impression material and 30 min after polymerization. Differences in adherence of bacteria to the samples prepared by hand-mixing and to those prepared by auto-mixing were assessed by Kruskal-Wallis and Mann-Whitney U-tests. For evaluating the efficiency of the disinfectant, Kruskal-Wallis multiple comparisons test was used. E. coli counts were higher in hand-mixed materials (P 2.394). The methods used for mixing polyether impression material did not affect bacterial attachment to impression surfaces. In contrast, the disinfection procedure greatly affects decontamination of the impression surface.

  12. Materials dependence of mixed gas plasticization behavior in asymmetric membranes

    NARCIS (Netherlands)

    Visser, Tymen; Masetto, N.; Wessling, Matthias

    2007-01-01

    The mass transport of asymmetric membranes for the separation of carbon dioxide/methane mixtures is determined by competitive sorption and plasticization. With increasing feed pressure in mixed gas experiments, the selectivity decreases due to both effects. Distinction whether one or the other

  13. Mixing core material into the envelopes of red grants

    International Nuclear Information System (INIS)

    Deupree, R.G.

    1986-01-01

    A discussion is presented of calculations of four core helium flashes in red giant stars. The starting point for these calculations is a point source explosion on the polar axis of a two-dimensional finite difference grid. The amount of residue of the core helium flash mixed into and above the hydrogen shell is calculated at four temperatures for the elements carbon, oxygen, neon, magnesium, silicon, and sulfur. 7 refs., 1 tab

  14. Spatial frequency mixing by nonlinear charge transport in photorefractive materials

    DEFF Research Database (Denmark)

    Limeres, J.; Carrascosa, M.; Arizmendi, L.

    2002-01-01

    in the material. The physical origin of the new gratings is extensively discussed. The formalism is applied to investigate multiple recording in LiNbO3 as a material relevant for applications. The influence of the multiple-recording method (either sequential or simultaneous) on the generation of second...

  15. 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)

  16. 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

  17. 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.

  18. Modern methods of material accounting for mixed oxide fuel fabrication facility

    International Nuclear Information System (INIS)

    Eggers, R.F.; Pindak, J.L.; Brouns, R.J.; Williams, R.C.; Brite, D.W.; Kinnison, R.R.; Fager, J.E.

    1981-01-01

    The generic requirements loss detection, and response to alarms of a contemporary material control and accounting (MCandA) philosophy have been applied to a mixed oxide fuel fabrication plant to produce a detailed preliminary MCandA system design that is generally applicable to facilities of this type. This paper summarizes and discusses detailed results of the mixed oxide fuel fabrication plant study

  19. 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)

  20. Evaluation of properties of irreversible hydrocolloid impression materials mixed with disinfectant liquids

    Directory of Open Access Journals (Sweden)

    Arul Amalan

    2013-01-01

    Conclusion: Chlorhexidine solution can be used to mix irreversible hydrocolloid impression materials in regular dental practice as it did not significantly alter the properties. This may ensure effective disinfection of impressions.

  1. Gas adsorption and gas mixture separations using mixed-ligand MOF material

    Science.gov (United States)

    Hupp, Joseph T [Northfield, IL; Mulfort, Karen L [Chicago, IL; Snurr, Randall Q [Evanston, IL; Bae, Youn-Sang [Evanston, IL

    2011-01-04

    A method of separating a mixture of carbon dioxiode and hydrocarbon gas using a mixed-ligand, metal-organic framework (MOF) material having metal ions coordinated to carboxylate ligands and pyridyl ligands.

  2. Radial mixing of material in the asterodial zone

    International Nuclear Information System (INIS)

    Ruzmaikina, T.V.; Safronov, V.S.; Weidenschilling, S.J.

    1989-01-01

    The asteroid belt shows radial zoning of compositional structure. The most abundant types are successively S, C and P types from the inner to the outer parts of the main belt, and D type in the Trojan clouds. Boundaries between compositional zones are not sharp, but gradual transitions over scales ∼1 AU in semimajor axis. The authors examine processes for producing this structure before, during and after the accretion of asteroids. The initial structure is established by temperature and composition gradients in the turbulent solar nebula during the collapse of the presolar cloud. The radial scale of the zoning, comparable to the disk thickness, favors disk models with relatively low turbulent viscosity. Radial decay of solid bodies due to gas drag during settling to the central plane and planetesimal formation probably causes only a small degree of mixing, due to the systematic nature of drag-induced motions

  3. 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

  4. Plasma Interactions with Mixed Materials and Impurity Transport

    Energy Technology Data Exchange (ETDEWEB)

    Rognlien, T. D. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Beiersdorfer, Peter [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Chernov, A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Frolov, T. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Magee, E. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Rudd, R. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Umansky, M. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2016-10-28

    The project brings together three discipline areas at LLNL to develop advanced capability to predict the impact of plasma/material interactions (PMI) on metallic surfaces in magnetic fusion energy (MFE) devices. These areas are (1) modeling transport of wall impurity ions through the edge plasma to the core plasma, (2) construction of a laser blow-off (LBO) system for injecting precise amounts of metallic atoms into a tokamak plasma, and (3) material science analysis of fundamental processes that modify metallic surfaces during plasma bombardment. The focus is on tungsten (W), which is being used for the ITER divertor and in designs of future MFE devices. In area (1), we have worked with the University of California, San Diego (UCSD) on applications of the UEDGE/DUSTT coupled codes to predict the influx of impurity ions from W dust through the edge plasma, including periodic edge-plasma oscillations, and revived a parallel version of UEDGE to speed up these simulations. In addition, the impurity transport model in the 2D UEDGE code has been implemented into the 3D BOUT++ turbulence/transport code to allow fundamental analysis of the impact of strong plasma turbulence on the impurity transport. In area (2), construction and testing of the LBO injection system has been completed. The original plan to install the LBO on the National Spherical Torus Experiment Upgrade (NSTX-U) at Princeton and its use to validate the impurity transport simulations is delayed owing to NSTX-U being offline for substantial magnetic coil repair period. In area (3), an analytic model has been developed to explain the growth of W tendrils (or fuzz) observed for helium-containing plasmas. Molecular dynamics calculations of W sputtering by W and deuterium (D) ions shows that a spatial blending of interatomic potentials is needed to describe the near-surface and deeper regions of the material.

  5. Plasma Interactions with Mixed Materials and Impurity Transport

    International Nuclear Information System (INIS)

    Rognlien, T. D.; Beiersdorfer, Peter; Chernov, A.; Frolov, T.; Magee, E.; Rudd, R.; Umansky, M.

    2016-01-01

    The project brings together three discipline areas at LLNL to develop advanced capability to predict the impact of plasma/material interactions (PMI) on metallic surfaces in magnetic fusion energy (MFE) devices. These areas are (1) modeling transport of wall impurity ions through the edge plasma to the core plasma, (2) construction of a laser blow-off (LBO) system for injecting precise amounts of metallic atoms into a tokamak plasma, and (3) material science analysis of fundamental processes that modify metallic surfaces during plasma bombardment. The focus is on tungsten (W), which is being used for the ITER divertor and in designs of future MFE devices. In area (1), we have worked with the University of California, San Diego (UCSD) on applications of the UEDGE/DUSTT coupled codes to predict the influx of impurity ions from W dust through the edge plasma, including periodic edge-plasma oscillations, and revived a parallel version of UEDGE to speed up these simulations. In addition, the impurity transport model in the 2D UEDGE code has been implemented into the 3D BOUT++ turbulence/transport code to allow fundamental analysis of the impact of strong plasma turbulence on the impurity transport. In area (2), construction and testing of the LBO injection system has been completed. The original plan to install the LBO on the National Spherical Torus Experiment Upgrade (NSTX-U) at Princeton and its use to validate the impurity transport simulations is delayed owing to NSTX-U being offline for substantial magnetic coil repair period. In area (3), an analytic model has been developed to explain the growth of W tendrils (or fuzz) observed for helium-containing plasmas. Molecular dynamics calculations of W sputtering by W and deuterium (D) ions shows that a spatial blending of interatomic potentials is needed to describe the near-surface and deeper regions of the material.

  6. Theoretical Synthesis of Mixed Materials for CO2 Capture Applications

    Energy Technology Data Exchange (ETDEWEB)

    Duan, Yuhua

    2015-01-01

    These pages provide an example of the layout and style required for the preparation of four-page papers for the TechConnect World 2015 technical proceedings.Documents must be submitted in electronic (Adobe PDFfile) format. Please study the enclosed materials beforebeginning the final preparation of your paper. Proofread your paper carefully before submitting (it will appear in the published volume in exactly the same form). Your PDF manuscript must be uploaded online by April 11th, 2015.You will receive no proofs. Begin your paper with an abstract of no more than 18 lines. Thoroughly summarize your article in this section since this text will be used for on-line listing and classification of the publication.

  7. A Systemic and Systematic Approach to the Development of a Policy Mix for Material Resource Efficiency

    Directory of Open Access Journals (Sweden)

    Tomas Ekvall

    2016-04-01

    Full Text Available Increasing material use efficiency is important to mitigate future supply risks and minimize environmental impacts associated with the production of the materials. The policy mix presented in this paper aims to contribute to reducing the use of virgin metals in the EU by 80% by 2050 without significant shifting of burdens to other material resources, environmental impacts, or parts of the world. We used a heuristic framework and a systems perspective for designing the policy mix that combines primary instruments designed to increase material efficiency, recycling and substitution of materials (a materials tax, the extended producer responsibility, technical regulations, and environmental taxes and supportive instruments aimed to reduce barriers to implementing the primary instruments and to contribute towards the policy objectives (e.g., research and development support, and advanced recycling centers. Furthermore, instruments were designed so as to increase political feasibility: e.g., taxes were gradually increased as part of a green fiscal reform, and border-tax adjustments were introduced to reduce impacts on competitiveness. However, even in such a policy mix design ongoing ex-ante assessments indicate that the policy mix will be politically difficult to implement—and also fall short of achieving the 80% reduction target. Nonetheless, we suggest combining primary and supportive instruments into coherent and dynamic policy mixes as a promising step towards system reconfigurations for sustainability.

  8. 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.

  9. 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

  10. 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.

  11. 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

  12. Plant growth response in experimental soilless mixes prepared from coal combustion products and organic waste materials

    Energy Technology Data Exchange (ETDEWEB)

    Bardhan, S.; Watson, M.; Dick, W.A. [Ohio State University, Wooster, OH (United States)

    2008-07-15

    Large quantities of organic materials such as animal manures, yard trimmings, and biosolids are produced each year. Beneficial use options for them are often limited, and composting has been proposed as a way to better manage these organic materials. Similarly, burning of coal created 125 million tons of coal combustion products (CCP) in the United States in 2006. An estimated 53 million tons of CCP were reused, whereas the remainder was deposited in landfills. By combining CCP and composted organic materials (COM), we were able to create soilless plant growth mixes with physicochemical conditions that can support excellent plant growth. An additional benefit is the conservation of natural raw materials, such as peat, which is generally used for making soilless mixes. Experimental mixes were formulated by combining CCP and COM at ratios ranging from 2:8 to 8:2 (vol/vol), respectively. Water content at saturation for the created mixes was 63% to 72%, whereas for the commercial control, it was 77%. pH values for the best performing mixes ranged between 5.9 and 6.8. Electrical conductivity and concentrations of required plant nutrient were also within plant growth recommendations for container media. Significantly (P < 0.0001) higher plant biomass growth (7%-130%) was observed in the experimental mixes compared with a commercial mix. No additional fertilizers were provided during the experiment, and reduced fertilization costs can thus accrue as an added benefit to the grower. In summary, combining CCP and COM, derived from source materials often viewed as wastes, can create highly productive plant growth mixes.

  13. 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.

  14. 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

  15. 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

  16. 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

  17. Investigations of processes of mixing in fabrication of carbonaceous materials and materials of silicon carbide by means of tracer technique

    International Nuclear Information System (INIS)

    Vagner, K.; Bruchin, F.; Ritter, I.; Grech, T.; Tsimmermann, V.; Ebert, G.

    1979-01-01

    A necessary condition for production of carbonic and SiC materials is providing for good mixing of liquid and solid components. Some experiments have been done by application of the method of radioactive marking of one of the components to be mixed for the purpose of investigation of the mixing process in an installation with capacity up to 2000 kg. Marking by the bromine-82 - naphthalene has been found to be suitable. The state of mixture is characterized by the homogeneity of radioactivity distribution which can be determined by measurement of gamma-radiation of series of samples. The mass of sample is from 1 to 100 g, depending on the volume of mixer. Besides it, fine distribution of components was determined by the contact autoradiography of flat surfaces of samples. The results of investigations give possibility to establish technological time of mixing, to compare different mixers and technologies of mixing as well as to determine behaviour of the mixed components. Recommendations have been developed on standardization of this method. Necessary means of radiation protection have been described [ru

  18. Stochastic Modelling and Self Tuning Control of a Continuous Cement Raw Material Mixing System

    Directory of Open Access Journals (Sweden)

    Hannu T. Toivonen

    1980-01-01

    Full Text Available The control of a continuously operating system for cement raw material mixing is studied. The purpose of the mixing system is to maintain a constant composition of the cement raw meal for the kiln despite variations of the raw material compositions. Experimental knowledge of the process dynamics and the characteristics of the various disturbances is used for deriving a stochastic model of the system. The optimal control strategy is then obtained as a minimum variance strategy. The control problem is finally solved using a self-tuning minimum variance regulator, and results from a successful implementation of the regulator are given.

  19. Light water reactor mixed-oxide fuel irradiation experiment

    International Nuclear Information System (INIS)

    Hodge, S.A.; Cowell, B.S.; Chang, G.S.; Ryskamp, J.M.

    1998-01-01

    The United States Department of Energy Office of Fissile Materials Disposition is sponsoring and Oak Ridge National Laboratory (ORNL) is leading an irradiation experiment to test mixed uranium-plutonium oxide (MOX) fuel made from weapons-grade (WG) plutonium. In this multiyear program, sealed capsules containing MOX fuel pellets fabricated at Los Alamos National Laboratory (LANL) are being irradiated in the Advanced Test Reactor (ATR) at the Idaho National Engineering and Environmental Laboratory (INEEL). The planned experiments will investigate the utilization of dry-processed plutonium, the effects of WG plutonium isotopics on MOX performance, and any material interactions of gallium with Zircaloy cladding

  20. 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.

  1. 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)

  2. 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

  3. 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

  4. 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...

  5. Mixed-layered bismuth--oxygen--iodine materials for capture and waste disposal of radioactive iodine

    Science.gov (United States)

    Krumhansl, James L; Nenoff, Tina M

    2015-01-06

    Materials and methods of synthesizing mixed-layered bismuth oxy-iodine materials, which can be synthesized in the presence of aqueous radioactive iodine species found in caustic solutions (e.g. NaOH or KOH). This technology provides a one-step process for both iodine sequestration and storage from nuclear fuel cycles. It results in materials that will be durable for repository conditions much like those found in Waste Isolation Pilot Plant (WIPP) and estimated for Yucca Mountain (YMP). By controlled reactant concentrations, optimized compositions of these mixed-layered bismuth oxy-iodine inorganic materials are produced that have both a high iodine weight percentage and a low solubility in groundwater environments.

  6. Mixed-layered bismuth-oxygen-iodine materials for capture and waste disposal of radioactive iodine

    Science.gov (United States)

    Krumhansl, James L; Nenoff, Tina M

    2013-02-26

    Materials and methods of synthesizing mixed-layered bismuth oxy-iodine materials, which can be synthesized in the presence of aqueous radioactive iodine species found in caustic solutions (e.g. NaOH or KOH). This technology provides a one-step process for both iodine sequestration and storage from nuclear fuel cycles. It results in materials that will be durable for repository conditions much like those found in Waste Isolation Pilot Plant (WIPP) and estimated for Yucca Mountain (YMP). By controlled reactant concentrations, optimized compositions of these mixed-layered bismuth oxy-iodine inorganic materials are produced that have both a high iodine weight percentage and a low solubility in groundwater environments.

  7. Transportation of radioactive, hazardous, and mixed wastes: Material identification is the key

    International Nuclear Information System (INIS)

    Stancell, D.F.; Willaford, D.M.

    1992-01-01

    This paper will discuss how material identification and classification will result in an accurate determination of regulatory requirements, and will assure safe and compliant shipment of radioactive, hazardous, and mixed wastes. The primary focus of the paper is a discussion of lessons learned by the Department of Energy in making waste shipments, and how this can be applied to future mixed waste shipments. There will be a brief discussion of the Department's regulatory compliance program, including a presentation of compliance audit results, and how regulatory issues are addressed through effective information exchange, technical assistance, and compliance training. A detailed discussion will follow, which describes cases involving material identification and classification problems. Examples will include both RCRA waste and uranium mill tailings shipments. The paper will conclude with a discussion concerning the application of these lessons to future mixed waste shipments proposed by the Department. (author)

  8. The influence of core materials and mix on the performance of a 100 kVA three phase transformer core

    Energy Technology Data Exchange (ETDEWEB)

    Snell, David E-mail: dave.snell@cogent-power.com; Coombs, Alan

    2003-01-01

    Various grades of grain-oriented electrical steel, and the effect of mixing domain refined and non-domain refined materials in the same three phase transformer core have been assessed using a developed computer-based test system. Ball unit domain refined material and non-domain refined material can be successfully mixed in the same core, without degrading performance.

  9. A Study of Moisture Induced Material Loss of Hot Mix Asphalt (HMA)

    Science.gov (United States)

    2017-10-31

    Maine Department of Transportation has noticed the partial or complete loss of material within 2-3 years of construction in the traffic wheel path in the presence of moisture in few of their mixes. Regularly used moisture susceptibility tests are una...

  10. 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

  11. 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

  12. 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

  13. Mixed

    Directory of Open Access Journals (Sweden)

    Pau Baya

    2011-05-01

    Full Text Available Remenat (Catalan (Mixed, "revoltillo" (Scrambled in Spanish, is a dish which, in Catalunya, consists of a beaten egg cooked with vegetables or other ingredients, normally prawns or asparagus. It is delicious. Scrambled refers to the action of mixing the beaten egg with other ingredients in a pan, normally using a wooden spoon Thought is frequently an amalgam of past ideas put through a spinner and rhythmically shaken around like a cocktail until a uniform and dense paste is made. This malleable product, rather like a cake mixture can be deformed pulling it out, rolling it around, adapting its shape to the commands of one’s hands or the tool which is being used on it. In the piece Mixed, the contortion of the wood seeks to reproduce the plasticity of this slow heavy movement. Each piece lays itself on the next piece consecutively like a tongue of incandescent lava slowly advancing but with unstoppable inertia.

  14. Iron zircon pigment synthesis: Proposal of a mixing index for the raw materials mixtures

    International Nuclear Information System (INIS)

    Zumaquero, E.; Ortsb, M.J.; Sanz, V.; Mestre, S.

    2017-01-01

    Iron zircon coral pigments are very interesting from an industrial point of view because of their high colouring power and their stability at high temperatures. However, the pigment's synthesis is particularly troublesome due to its specific reaction mechanism. As an encapsulated pigment it becomes very important how the raw materials are distributed in the reaction mixture. To evaluate the effectiveness of the mixing process, it would be convenient to define a parameter, that is the mixing index, to estimate the degree of homogeneity of the system. In the current investigation, a mixing index is proposed derived from the power spectrum of Fourier transform of scanning electron microscope (SEM) images of the raw material mixture. Concretely, the number of pixels in a certain range of values in the image of the power spectrum, seems to behave relatively well as mixing index. This index allows us to distinguish between samples with different zirconia and iron oxide used as precursors. The proposed mixing index seems to be related to the colouring power of the final pigment when the synthesis generates enough zircon to encapsulate hematite particles. [es

  15. Nanocasting of Periodic Mesoporous Materials as an Effective Strategy to Prepare Mixed Phases of Titania

    Directory of Open Access Journals (Sweden)

    Luther Mahoney

    2015-12-01

    Full Text Available Mesoporous titanium dioxide materials were prepared using a nanocasting technique involving silica SBA-15 as the hard-template. At an optimal loading of titanium precursor, the hexagonal periodic array of pores in SBA-15 was retained. The phases of titanium dioxide could be easily varied by the number of impregnation cycles and the nature of titanium alkoxide employed. Low number of impregnation cycles produced mixed phases of anatase and TiO2(B. The mesoporous TiO2 materials were tested for solar hydrogen production, and the material consisting of 98% anatase and 2% TiO2(B exhibited the highest yield of hydrogen from the photocatalytic splitting of water. The periodicity of the pores was an important factor that influenced the photocatalytic activity. This study indicates that mixed phases of titania containing ordered array of pores can be prepared by using the nanocasting strategy.

  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. Feasibility study on diagnosis of material damage using bulk wave mixing technique

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Jeong Seok; Cho, Youn Ho [Pusan National University, Busan (Korea, Republic of)

    2016-02-15

    Ultrasonic nonlinear evaluation is generally utilized for detection of not only defects but also microdamage such as corrosion and plastic deformation. Nonlinearity is determined by the amplitude ratio of primary wave second harmonic wave, and the results of its comparison are used for evaluation. Owing to the experimental features, the experimental nonlinearity result contains system nonlinearity and material nonlinearity. System nonlinearity is that which is unwanted by the user; hence, it acts as an error and interrupts analysis. In this study, a bulk wave mixing technique is implemented in order to minimize the system nonlinearity and obtain the reliable analysis results. The biggest advantage of this technique is that experimental nonlinearity contains less system nonlinearity than that for the conventional nonlinear ultrasonic technique. Theoretical and experimental verifications are performed in this study. By comparing the results of the bulk wave mixing technique with those of the conventional technique, the strengths, weaknesses, and application validity of the bulk wave mixing technique are determined.

  18. Modern methods of material accounting for mixed-oxide fuel-fabrication facility

    International Nuclear Information System (INIS)

    Eggers, R.F.; Brouns, R.J.; Brite, D.W.; Pindak, J.L.

    1981-07-01

    The generic requirements loss detection, and response to alarms of a contemporary material control and accounting (MC and A) philosophy have been applied to a mixed-oxide fuel-fabrication plant to produce a detailed preliminary MC and A system design that is generally applicable to facilities of this type. This paper summarizes and discusses detailed results of the mixed-oxide fuel-fabrication plant study. Topics covered in this paper include: mixed-oxide fuel-fabrication process description, process disaggregation into MC and A system control units, quantitative results of analysis of control units for abrupt and recurring loss-detection capability, impact of short- and long-term holdup on loss-detection capability, response to alarms for abrupt loss, and response to alarms for recurring loss

  19. Consideration of reinforcement mechanism in the short fiber mixing granular materials by granular element simulations

    Science.gov (United States)

    Mori, Kentaro; Kaneko, Kenji; Hashizume, Yutaka

    2017-06-01

    The short fiber mixing method is well known as one of the method to improve the strength of gran- ular soils in geotechnical engineering. Mechanical properties of the short fiber mixing granular materials are influenced by many factors, such as the mixture ratio of the short fiber, the material of short fiber, the length, and the orientation. In particular, the mixture ratio of the short fibers is very important in mixture design. In the past study, we understood that the strength is reduced by too much short fiber mixing by a series of tri-axial compression experiments. Namely, there is "optimum mixture ratio" in the short fiber mixing granular soils. In this study, to consider the mechanism of occurrence of the optimum mixture ratio, we carried out the numerical experiments by granular element method. As the results, we can understand that the strength decrease when too much grain-fiber contact points exist, because a friction coefficient is smaller than the grain-grain contact points.

  20. Coordinated safeguards for materials management in a mixed-oxide fuel facility

    International Nuclear Information System (INIS)

    Shipley, J.P.; Cobb, D.D.; Dietz, R.J.; Evans, M.L.; Schelonka, E.P.; Smith, D.B.; Walton, R.B.

    1977-02-01

    A coordinated safeguards system is described for safeguarding strategic quantities of special nuclear materials in mixed-oxide recycle fuel fabrication facilities. The safeguards system is compatible with industrial process requirements and combines maximum effectiveness consistent with modest cost and minimal process interference. It is based on unit process accounting using a combination of conventional and state-of-the-art NDA measurement techniques. The effectiveness of the system against single and multiple thefts is evaluated using computer modeling and simulation techniques

  1. OFFSITE RADIOLOGICAL CONSEQUENCE CALCULATION FOR THE BOUNDING MIXING OF INCOMPATIBLE MATERIALS ACCIDENT

    International Nuclear Information System (INIS)

    SANDGREN, K.R.

    2006-01-01

    This document quantifies the offsite radiological consequence of the bounding mixing of incompatible materials accident for comparison with the 25 rem Evaluation Guideline established in Appendix A of DOE-STD-3009. The bounding accident is an inadvertent addition of acid to a waste tank. The calculated offsite dose does not challenge the Evaluation Guideline. Revision 4 updates the analysis to consider bulk chemical additions to single shell tanks (SSTs)

  2. Coordinated safeguards for materials management in a mixed-oxide fuel facility

    Energy Technology Data Exchange (ETDEWEB)

    Shipley, J.P.; Cobb, D.D.; Dietz, R.J.; Evans, M.L.; Schelonka, E.P.; Smith, D.B.; Walton, R.B.

    1977-02-01

    A coordinated safeguards system is described for safeguarding strategic quantities of special nuclear materials in mixed-oxide recycle fuel fabrication facilities. The safeguards system is compatible with industrial process requirements and combines maximum effectiveness consistent with modest cost and minimal process interference. It is based on unit process accounting using a combination of conventional and state-of-the-art NDA measurement techniques. The effectiveness of the system against single and multiple thefts is evaluated using computer modeling and simulation techniques.

  3. 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

  4. A manufacturing process for a mixed-oxide type superconducting material

    International Nuclear Information System (INIS)

    Gendre, P.; Regnier, P.; Schmirgeld-Mignot, L.; Marquet, A.

    1995-01-01

    In order to produce high temperature superconducting materials such as YBaCuO and Bi 2 Sr 2 Ca Cu 2 O 8 , a process is presented which consists in an electrodeposition on a conductive substrate of successive layers made of the metallic elements composing the superconductor, with only one element in each layer; between each layer deposition, an intermediary oxide-reaction thermal treatment is carried out; a global oxidation thermal treatment is then finally conducted to produce the mixed oxide material. Narrow superconducting transitions and high critical current densities are possible. 3 refs., 4 figs

  5. 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

  6. 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.

  7. 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

  8. Soiled-based uranium disequilibrium and mixed uranium-thorium series radionuclide reference materials

    International Nuclear Information System (INIS)

    Donivan, S.; Chessmore, R.

    1988-12-01

    The US Department of Energy (DOE) Office of Remedial Action and Waste Technology has assigned the Technical Measurements Center (TMC), located at the DOE Grand Junction Colorado, Projects Office and operated by UNC Geotech (UNC), the task of supporting ongoing remedial action programs by providing both technical guidance and assistance in making the various measurements required in all phases of remedial action work. Pursuant to this task, the Technical Measurements Center prepared two sets of radionuclide reference materials for use by remedial action contractors and cognizant federal and state agencies. A total of six reference materials, two sets comprising three reference materials each, were prepared with varying concentrations of radionuclides using mill tailings materials, ores, and a river-bottom soil diluent. One set (disequilibrium set) contains varying amounts of uranium with nominal amounts of radium-226. The other set (mixed-nuclide set) contains varying amounts of uranium-238 and thorium-232 decay series nuclides. 14 refs., 10 tabs

  9. 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

  10. 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)

  11. Application of Hard Metal Weld Deposit in the Area of Mixing Organic Materials

    Directory of Open Access Journals (Sweden)

    Jiří Votava

    2014-01-01

    Full Text Available Any machine part is subject to degradation processes. Intensive wear occurs either when two bearing surfaces come into contact or when loose particles rub the function surface of a machine part. Soil processing machines are a good example. A similar process of abrasive wear occurs also in mixing machines or lines for material transport, such as worm-conveyors. The experiment part of this paper analyses hard metal weld deposit dedicated for renovation of abrasive stressed surfaces. In order to prolong the service life of a blade disc in a mixing machine Kreis-Biogas-Dissolver, the technology of hard surfacing by an electric arc was used. Tested hard metal electrodes were applied on a steel tape class 11 373. To eliminate mixing with the base material, weld beads were applied in two layers. Firstly, the weld bead was visually analyzed on a binocular microscope. Further, weld bead as well as the base material was analyzed from the metallographic point of view, whose aim was to identify the structure of weld metal and the origin of microcracks in weld bead. Moreover, there was also measured microhardness of weld metal. Abrasive resistance was tested according to the norm ČSN 01 5084, which is an abrasive cloth test. As in the mixing process also erosion wear occurs, there was also processed a test on a Bond device simulating stress of test samples by loose abrasive particles. The abrading agents were formed by broken stones of 8–16 mm in size. Based on the results of the individual tests, the recommendation of usage hard metal electrodes for prolonging service life of machine parts will be made.

  12. 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

  13. Mix design and mechanical performance of geopolymer binder for sustainable construction and building material

    Science.gov (United States)

    Saeli, Manfredi; Novais, Rui M.; Seabra, Maria Paula; Labrincha, João A.

    2017-11-01

    Sustainability in construction is a major concern worldwide, due to the huge volume of materials and energy consumed by this sector. Associated supplementing industries (e.g. Portland cement production) constitute a significant source of CO2 emissions and global warming. Valorisation and reuse of industrial wastes and by-products make geopolymers a solid and sustainable via to be followed as a valid alternative to Portland cement. In this work the mix design of a green fly ash-based geopolymer is evaluated as an environmentally friendly construction material. In the pursuit of sustainability, wastes from a regional kraft pulp industry are exploited for the material processing. Furthermore, a simple, reproducible, and low-cost manufacture is used. The mix design is hence optimised in order to improve the desirable mechanical performance of the material intended for structural applications in construction. Tests indicate that geopolymers may efficiently substitute the ordinary Portland cement as a mortar/concrete binder. Furthermore, valorisation and reuse of wastes in geopolymers is a suboptimal way of gaining financial surplus for the involved industrial players, while contributes for the implementation of a desirable circular economy.

  14. The Amster concept: a configuration generating its own uranium with a mixed thorium and uranium support

    International Nuclear Information System (INIS)

    Vergnes, J.; Garzenne, C.; Lecarpentier, D.; Mouney, H.; Delpech, M.

    2001-01-01

    AMSTER is a continuously reloaded, graphite-moderated molten salt critical reactor, using a 238 U or 232 Th fuel support, slightly enriched with 235 U if necessary. Using this concept, one can define a large number of configurations according to the products loaded and recycled. The choice of thorium fuel support leads to two configurations requiring no additional 235 U as fissile material: a configuration with one moderating zone, incinerating Transuranium elements (TRU); a configuration with 2 moderating zones self-consuming TRU and regenerating the fissile uranium ( 233 U). In this configuration, it is even possible to burn 238 U (from depleted uranium) by adding it to the thorium support. These configurations use a minimum amount of fuel (100 kg of 232 Th or 100 kg of a 232 Th- 238 U mix per TWh) and produce very little TRU (a few tens of grams per TWh). (author)

  15. 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

  16. 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)

  17. Raw material utilization in slaughterhouses – optimizing expected profit using mixed-integer programming

    DEFF Research Database (Denmark)

    Jensen, Toke Koldborg; Kjærsgaard, Niels Christian

    Slaughterhouses are major players in the pork supply chain, and supply and demand must be matched in order to generate the highest profit. In particular, carcasses must be sorted in order to produce the “right” final products from the “right” carcasses. We develop a mixed-integer programming (MIP) ...... at slaughterhouses. Finally, we comment on the expected effect of variations in the raw material supply and the demand as well as future research concerning joint modelling of supply chain aspects.......Slaughterhouses are major players in the pork supply chain, and supply and demand must be matched in order to generate the highest profit. In particular, carcasses must be sorted in order to produce the “right” final products from the “right” carcasses. We develop a mixed-integer programming (MIP...

  18. Amine–mixed oxide hybrid materials for carbon dioxide adsorption from CO2/H2 mixture

    Science.gov (United States)

    Ravi, Navin; Aishah Anuar, Siti; Yusuf, Nur Yusra Mt; Isahak, Wan Nor Roslam Wan; Shahbudin Masdar, Mohd

    2018-05-01

    Bio-hydrogen mainly contains hydrogen and high level of carbon dioxide (CO2). High concentration of CO2 lead to a limitation especially in fuel cell application. In this study, the amine-mixed oxide hybrid materials for CO2 separation from bio-hydrogen model (50% CO2:50% H2) have been studied. Fourier-transform infrared spectroscopy (FTIR) and x-ray diffraction (XRD) characterizations showed that the amine–mixed oxide hybrid materials successfully adsorbed CO2 physically with no chemical adsorption evidence. The dry gas of CO2/H2 mixture adsorbed physically on amine–CuO–MgO hybrid material. No carbonates were detected after several times of adsorption, which indicated the good recyclability of adsorbents. The adsorbent system of diethanolamine (DEA)/15% CuO–75% MgO showed the highest CO2 adsorption capacity of 21.2 wt% due to the presence of polar substance on MgO surface, which can adsorb CO2 at ambient condition. The alcohol group of DEA can enhance the CO2 solubility on the adsorbent surface. In the 20% CuO–50% MgO adsorbent system, DEA as amine type showed a high CO2 adsorption of 19.4 wt%. The 10% amine loading system showed that the DEA adsorption system provided high CO2 adsorption. The BET analysis confirmed that a high amine loading contributed to the decrease in CO2 adsorption due to the low surface area of the adsorbent system.

  19. Solidification of radioactive waste resins using cement mixed with organic material

    International Nuclear Information System (INIS)

    Laili, Zalina; Yasir, Muhamad Samudi; Wahab, Mohd Abdul

    2015-01-01

    Solidification of radioactive waste resins using cement mixed with organic material i.e. biochar is described in this paper. Different percentage of biochar (0%, 5%, 8%, 11%, 14% and 18%) was investigated in this study. The characteristics such as compressive strength and leaching behavior were examined in order to evaluate the performance of solidified radioactive waste resins. The results showed that the amount of biochar affect the compressive strength of the solidified resins. Based on the data obtained for the leaching experiments performed, only one formulation showed the leached of Cs-134 from the solidified radioactive waste resins

  20. Solidification of radioactive waste resins using cement mixed with organic material

    Energy Technology Data Exchange (ETDEWEB)

    Laili, Zalina, E-mail: liena@nm.gov.my [Nuclear Science Programme, School of Applied Physics, Faculty of Science and Technology, Universiti Kebangsaan Malaysia (UKM), Bangi, 43600, Selangor Malaysia (Malaysia); Waste and Environmental Technology Division, Malaysian Nuclear Agency (Nuclear Malaysia), Bangi, 43000 Kajang, Selangor (Malaysia); Yasir, Muhamad Samudi [Nuclear Science Programme, School of Applied Physics, Faculty of Science and Technology, Universiti Kebangsaan Malaysia (UKM), Bangi, 43600, Selangor Malaysia (Malaysia); Wahab, Mohd Abdul [Waste and Environmental Technology Division, Malaysian Nuclear Agency (Nuclear Malaysia), Bangi, 43000 Kajang, Selangor (Malaysia)

    2015-04-29

    Solidification of radioactive waste resins using cement mixed with organic material i.e. biochar is described in this paper. Different percentage of biochar (0%, 5%, 8%, 11%, 14% and 18%) was investigated in this study. The characteristics such as compressive strength and leaching behavior were examined in order to evaluate the performance of solidified radioactive waste resins. The results showed that the amount of biochar affect the compressive strength of the solidified resins. Based on the data obtained for the leaching experiments performed, only one formulation showed the leached of Cs-134 from the solidified radioactive waste resins.

  1. Usability of Clay Mixed Red Mud as Building Material in Transdanubian (Hungary) Region

    International Nuclear Information System (INIS)

    Sas, Z.; Somlai, J.J.; Szeiler, G.; Kovacs, T.

    2014-01-01

    The most commonly used building materials in Hungary and in numerous country of the world are the bricks, which made from clays. Due to the congenial internal structure properties of the clays these raw materials can be mixed with other materials, provides great possibility to reuse industrial by-products as additive material. The production and inbuilt of new types of synthetic building materials based on NORM (naturally occurring radioactive materials) by-products is raising concerns among authorities, public and scientists. Several NORM residues produced in large quantity, such as: phospogypsum (phosphate industry), red mud (aluminium processing industry), fly ash, coal slag (coal burning and steelworks) and so on are presently under investigation. The aluminum manufacturing in Ajka (Hungary) started in 1943. As a result of the bauxite refining activities up to now approximately 30 Mt of red mud has been produced in Hungary, stored in reservoirs. The radionuclide content of the bauxite usually exceeds the world average in soils (WA), which entirely remains in the by-product during Bayer process. The exposure pathways in case of application of NORM residues have to be explored in order to reveal the potential risks of NORMs on residents. The gamma radiation originated from the primordial radionuclides (K-40; U-238; Th-232) and their daughter elements found in nature and in building materials as well increase the external dose of the human body. In the EU the Radiation Protection 112 (RP 112) guideline serves for classification of building material, wherein the gamma exposure is limited by I-index

  2. Materials used for indirect pulp treatment in primary teeth: a mixed treatment comparisons meta-analysis

    Directory of Open Access Journals (Sweden)

    Pablo Silveira dos Santos

    2017-12-01

    Full Text Available Abstract: This study aimed to systematically review the literature to address the question regarding the influence of different materials in the clinical and radiographic success of indirect pulp treatment in primary teeth. A literature search was carried out for articles published prior to January 2017 in PubMed/MEDLINE, CENTRAL, Scopus, TRIP and ClinicalTrials databases; relevant articles included randomized clinical trials that compared materials used for indirect pulp treatment in primary teeth. Two reviewers independently selected the studies and extracted the data. The effects of each material on the outcome (clinical and radiographic failures were analyzed using a mixed treatment comparisons meta-analysis. The ranking of treatments according to their probability of being the best choice was also calculated. From 1,088 potentially eligible studies, 11 were selected for full-text analysis, and 4 were included in the meta-analysis. In all papers, calcium hydroxide liner was used as the control group versus an adhesive system, resin-modified glass ionomer cement or placebo. The follow-up period ranged from 24 to 48 months, with dropout rates of 0-25.7%. The material type did not significantly affect the risk of failure of the indirect pulp treatment. However, calcium hydroxide presented a higher probability of failure. In conclusion, there is no scientific evidence showing the superiority of any material used for indirect pulp treatment in primary teeth.

  3. 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

  4. Analysis of Cement as Raw Material Inventory on Ready Mix Production Unit at PT. Perdana Beton in Samarinda

    OpenAIRE

    Wahyuni, Novita Eka

    2014-01-01

    This study aims to identify and analyze the optimal amount of cement supplies in PT. Perdana Beton In Samarinda. Formulation of the problem in this study is whether the supply of cement as raw material for Ready Mix at PT. Perdana Beton In Samarinda already minimizing the cost of inventory cement. The hypothesis in this study is: '' Allegedly cement as raw material supplies ready mix at PT. Perdana Beton In Samarinda not provide minimum inventory costs'. Basic...

  5. 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)

  6. Study on the light-color mixing of rare earth luminescent materials for anti-counterfeiting application

    Science.gov (United States)

    Zhang, Jishu; Zhang, Yingzi; Tao, Jin; Zhu, Yanan

    2018-04-01

    In order to find out the light color mixing mechanism of rare earth luminescent materials used in anti-counterfeiting fibers, we prepared three kinds of rare earth luminescent materials according to RGB tri-primary color, and mixed it together to form different mixtures in certain proportion. The phase structures of the luminescent material monomers were measured by x-ray diffractometer. The photochromic properties of the luminescent materials were tested and analyzed by fluorescence spectrophotometer. The results show that the light color mixing was consistent with the blending principle of additive color, but not the same because of the photochromic properties of rare earth luminescent materials, and we explored the reasons in the light wavelength and intensity. It was found that the enhancement of the luminescence intensity of the mixture on account of the superimposing of luminescence.

  7. 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.

  8. 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

  9. The apical leakage of mineral trioxide aggregate as the retrograde filling material with various mixing agents

    Directory of Open Access Journals (Sweden)

    Ema Mulyawati

    2010-06-01

    Full Text Available Background: Mineral trioxide aggregate (MTA is relatively considered as a new material in endodontic. It even has been used as retrograde filling material due to its biocompatibility, antibacterial effect, sealing ability and anti-moist effect. Some materials have been used as mixing agent to achieve an appropiate setting of MTA. Purpose: The aim of this study is to investigate the effect of the mixing agents of MTA towards the apical leakage when they are used together as retrograde filling materials. Method: The samples of this research consist of 30 human extracted upper central incisors. First, the crown of each tooth is sectioned. The root canals are prepared by using the conventional technique and then are obturated with gutta percha. After cutting the root apex, 2 mm from apical, class 1 cavities are prepared by using fissure bur with the depth of 3 mm. The samples then are divided into 3 groups with 10 teeth for each. Group I uses aquabidest as mixing agent of MTA (MTA-aquabidest, group II uses saline (MTA-saline, while group III uses 0.12% chlorhexidine (MTA-chlorhexidine. The apex of each group then is filled with the mixing MTA determined already. Afterwards, clearing method is used to evaluate the apical leakage. The apical leakage actually is determined by measuring the depth of methylene blue penetration with stereomicroscope. The statictical analyses of the linear dye penetration then are performed with analysis of varians ANOVA. Result: The dye penetration for both MTA-aquadest and MTA-saline groups indicates the lowest penetration, and there is even a significant difference compared with MTA-0.12% chlorhexidine group (p<0.005. Conclusion: It can be concluded that aquabidest and saline as mixing agents of MTA produce less apical leakage compared with 0.12% chlorhexidine.Latar belakang: Mineral trioxide aggregate (MTA merupakan bahan yang relatif baru dalam bidang endodontik. Bahan tersebut diindikasikan sebagai bahan pengisi

  10. The Influence of Unusual Materials as Prospective Fillers in the Hot Mix Asphalt

    Science.gov (United States)

    Cavalcate Ferrão, Wallace; Moizinho, Joel Carlos

    2017-10-01

    Among the factors that influence directly the durability of the asphaltic layer on pavements, the type and percentage of filler in the hot mix asphalt pavement (HMA) is a great player. The most traditional fillers, the Portland cement and the hydrated lime, are well known for resisting to weather variations and adding extra features to the hot mixtures. The glass powder, the cladding waste (gotten from clay bricks), the ashes of rice husks and laterite powder are proposed as substitutes to the traditional ones. The materials have been sieved and classified by fitting the powder on the filler grain size required by Brazilian Rules, eventually they have been tested with asphalt 50/70. The glass powder performed a Thermic Susceptibility Index (IST) of -0.69 for 5% in weight of filler and -0.75 for 10% in weight of filler, proving that this material satisfies the Brazilian specification DNIT-EM 095/2006; on the other hand, the laterite powder presented an IST of -0.61 for 5% and 0.32 for 10%. After executing the Softening Point, Penetration and Flash Point tests, it has been confirmed that the glass and laterite powder are recommended materials as potential substitutes to the Portland cement, however the first one performs better under balmy temperatures due to its negative IST; the cladding powder and the rice husks turns the mixtures too rigid and breakable on percentages close to 10%.

  11. Progress on Perovskite Materials and Solar Cells with Mixed Cations and Halide Anions.

    Science.gov (United States)

    Ono, Luis K; Juarez-Perez, Emilio J; Qi, Yabing

    2017-09-13

    Organic-inorganic halide perovskite materials (e.g., MAPbI 3 , FAPbI 3 , etc.; where MA = CH 3 NH 3 + , FA = CH(NH 2 ) 2 + ) have been studied intensively for photovoltaic applications. Major concerns for the commercialization of perovskite photovoltaic technology to take off include lead toxicity, long-term stability, hysteresis, and optimal bandgap. Therefore, there is still need for further exploration of alternative candidates. Elemental composition engineering of MAPbI 3 and FAPbI 3 has been proposed to address the above concerns. Among the best six certified power conversion efficiencies reported by National Renewable Energy Laboratory on perovskite-based solar cells, five are based on mixed perovskites (e.g., MAPbI 1-x Br x , FA 0.85 MA 0.15 PbI 2.55 Br 0.45 , Cs 0.1 FA 0.75 MA 0.15 PbI 2.49 Br 0.51 ). In this paper, we review the recent progress on the synthesis and fundamental aspects of mixed cation and halide perovskites correlating with device performance, long-term stability, and hysteresis. In the outlook, we outline the future research directions based on the reported results as well as related topics that warrant further investigation.

  12. Mixing of two co-directional Rayleigh surface waves in a nonlinear elastic material.

    Science.gov (United States)

    Morlock, Merlin B; Kim, Jin-Yeon; Jacobs, Laurence J; Qu, Jianmin

    2015-01-01

    The mixing of two co-directional, initially monochromatic Rayleigh surface waves in an isotropic, homogeneous, and nonlinear elastic solid is investigated using analytical, finite element method, and experimental approaches. The analytical investigations show that while the horizontal velocity component can form a shock wave, the vertical velocity component can form a pulse independent of the specific ratios of the fundamental frequencies and amplitudes that are mixed. This analytical model is then used to simulate the development of the fundamentals, second harmonics, and the sum and difference frequency components over the propagation distance. The analytical model is further extended to include diffraction effects in the parabolic approximation. Finally, the frequency and amplitude ratios of the fundamentals are identified which provide maximum amplitudes of the second harmonics as well as of the sum and difference frequency components, to help guide effective material characterization; this approach should make it possible to measure the acoustic nonlinearity of a solid not only with the second harmonics, but also with the sum and difference frequency components. Results of the analytical investigations are then confirmed using the finite element method and the experimental feasibility of the proposed technique is validated for an aluminum specimen.

  13. Evaluation of the Efficacy of Different Mixing Techniques and Disinfection on Microbial Colonization of Polyether Impression Materials: A Comparative Study.

    Science.gov (United States)

    Singla, Youginder; Pachar, Renu B; Poriya, Sangeeta; Mishra, Aalok; Sharma, Rajni; Garg, Anshu

    2018-03-01

    This study aims to determine the role of mixing techniques of polyether impression materials and efficacy of disinfection on microbial colonization of these impression materials. Polyether impression material was mixed using two methods: First by hand mixing (group I) and second using an automixer (group II) with a total of 100 samples. Four microbial strains were studied, which included Escherichia coli, Staphylococcus aureus, Pseudomonas aeruginosa, and Candida albicans. After incubation, the bacterial colonies were counted, and then, disinfectant solution was applied. The effect of disinfection solution was evaluated for each specimen. The surface of polyether impression materials mixed with an automixer has less number of voids and overall a smoother surface as compared with the hand-mixed ones. On comparing the disinfection procedures, i.e., specimens without any disinfection and specimens after disinfection, statistically highly significant difference was seen between all the groups. We can conclude that impression mixing procedures are important in determining the surface characteristics of the impression and ultimately the colonization of bacteria and also determine the importance of disinfection on microbial colonization. This study emphasises the deleterious role of nosocomial infections and specific measures that should be taken regarding the prevention of such diseases. Dental impressions are proved to be a source of such infections and may lead to transmission of such diseases. Thus, proper measures should be taken right from the first step of impression taking to minimizing and preventing such kind of contaminations in clinical practice.

  14. Comment on the interpretation and application of limiting critical concentrations of fissile nuclides in water

    International Nuclear Information System (INIS)

    Clayton, E.D.; Durst, B.M.

    1977-01-01

    Calculations of the infinite multiplication factor for aqueous homogeneous mixtures of mixed oxides of plutonium and natural uranium at low fissile concentrations (7 g Pu/l) disclose a maximum to occur in the value of k/sub infinity/ at a weight fraction, Pu/(Pu + U), of approximately 0.0035. With mixed oxide solutions containing 7 g Pu/l, the value of k/sub infinity/ is estimated to be nearly 1.04, whereas in the absence of the natural uranium, the maximum value of k/sub infinity/ at 7 g Pu/l in water is approximately 4% less or near unity. The occurrence of this peak in value of k/sub infinity/ is due to the 235 U content in the natural uranium. Thus, in the presence of natural uranium, it should be borne in mind that the limiting subcritical concentration of plutonium (given as 7.0 g Pu/l) in water must be reduced to values <7.0 g Pu/l to ensure subcriticality of the mixture

  15. 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

  16. Numerical simulation of multi-material mixing in an inclined interface Richtmyer-Meshkov instability

    Science.gov (United States)

    Subramaniam, Akshay; Lele, Sanjiva K.

    2017-01-01

    In this work, high fidelity simulations of shock induced multi-material mixing between air and SF6 in a shock tube are performed for a Mach 1.5 shock interacting with a planar material interface that is inclined with respect to the shock propagating direction. In the current configuration, unlike the classical perturbed flat interface case, the evolution of the interface is fully non-linear from early time. The simulations attempt to replicate an experiment conducted at the Georgia Tech STAML. Tight coupling between numerics and flow physics and the large range of spatial scales make this a challenging problem to simulate numerically. Often, two dimensional simulations are performed to reduce the computational cost of these simulations. We show here that the effect of small three dimensional perturbations likely to be present in an experimental setting is not negligible. Full 3D simulations would have to be performed to do a proper comparison with experiments. Effect of grid resolution is also studied in the present work. Simulations shown are conducted with an extended version of the Miranda solver developed by Cook et. al [1] which combines high-order compact finite differences [2] with localized non-linear artificial properties for shock and interface capturing [3].

  17. Lithium Storage in Microstructures of Amorphous Mixed-Valence Vanadium Oxide as Anode Materials.

    Science.gov (United States)

    Zhao, Di; Zheng, Lirong; Xiao, Ying; Wang, Xia; Cao, Minhua

    2015-07-08

    Constructing three-dimensional (3 D) nanostructures with excellent structural stability is an important approach for realizing high-rate capability and a high capacity of the electrode materials in lithium-ion batteries (LIBs). Herein, we report the synthesis of hydrangea-like amorphous mixed-valence VOx microspheres (a-VOx MSs) through a facile solvothermal method followed by controlled calcination. The resultant hydrangea-like a-VOx MSs are composed of intercrossed nanosheets and, thus, construct a 3 D network structure. Upon evaluation as an anode material for LIBs, the a-VOx MSs show excellent lithium-storage performance in terms of high capacity, good rate capability, and long-term stability upon extended cycling. Specifically, they exhibit very stable cycling behavior with a highly reversible capacity of 1050 mA h g(-1) at a rate of 0.1 A g(-1) after 140 cycles. They also show excellent rate capability, with a capacity of 390 mA h g(-1) at a rate as high as 10 A g(-1) . Detailed investigations on the morphological and structural changes of the a-VOx MSs upon cycling demonstrated that the a-VOx MSs went through modification of the local VO coordinations accompanied with the formation of a higher oxidation state of V, but still with an amorphous state throughout the whole discharge/charge process. Moreover, the a-VOx MSs can buffer huge volumetric changes during the insertion/extraction process, and at the same time they remain intact even after 200 cycles of the charge/discharge process. Thus, these microspheres may be a promising anode material for LIBs. © 2015 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  18. When Educational Material Is Delivered: A Mixed Methods Content Validation Study of the Information Assessment Method.

    Science.gov (United States)

    Badran, Hani; Pluye, Pierre; Grad, Roland

    2017-03-14

    The Information Assessment Method (IAM) allows clinicians to report the cognitive impact, clinical relevance, intention to use, and expected patient health benefits associated with clinical information received by email. More than 15,000 Canadian physicians and pharmacists use the IAM in continuing education programs. In addition, information providers can use IAM ratings and feedback comments from clinicians to improve their products. Our general objective was to validate the IAM questionnaire for the delivery of educational material (ecological and logical content validity). Our specific objectives were to measure the relevance and evaluate the representativeness of IAM items for assessing information received by email. A 3-part mixed methods study was conducted (convergent design). In part 1 (quantitative longitudinal study), the relevance of IAM items was measured. Participants were 5596 physician members of the Canadian Medical Association who used the IAM. A total of 234,196 ratings were collected in 2012. The relevance of IAM items with respect to their main construct was calculated using descriptive statistics (relevance ratio R). In part 2 (qualitative descriptive study), the representativeness of IAM items was evaluated. A total of 15 family physicians completed semistructured face-to-face interviews. For each construct, we evaluated the representativeness of IAM items using a deductive-inductive thematic qualitative data analysis. In part 3 (mixing quantitative and qualitative parts), results from quantitative and qualitative analyses were reviewed, juxtaposed in a table, discussed with experts, and integrated. Thus, our final results are derived from the views of users (ecological content validation) and experts (logical content validation). Of the 23 IAM items, 21 were validated for content, while 2 were removed. In part 1 (quantitative results), 21 items were deemed relevant, while 2 items were deemed not relevant (R=4.86% [N=234,196] and R=3.04% [n

  19. Molten salt oxidation of mixed wastes: Separation of radioactive materials and Resource Conservation and Recovery Act (RCRA) materials

    International Nuclear Information System (INIS)

    Bell, J.T.; Haas, P.A.; Rudolph, J.C.

    1993-01-01

    The Oak Ridge National Laboratory (ORNL) is involved in a program to apply a molten salt oxidation (MSO) process to the treatment of mixed wastes at Oak Ridge and other Department of Energy (DOE) sites. Mixed wastes are defined as those wastes that contain both radioactive components, which are regulated by the atomic energy legislation, and hazardous waste components, which are regulated under the Resource Conservation and Recovery Act (RCRA). A major part of our ORNL program involves the development of separation technologies that are necessary for the complete treatment of mixed wastes. The residues from the MSO treatment of the mixed wastes must be processed further to separate the radioactive components, to concentrate and recycle residues, or to convert the residues into forms acceptable for final disposal. This paper is a review of the MSO requirements for separation technologies, the information now available, and the concepts for our development studies

  20. 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

  1. Defect physics vis-à-vis electrochemical performance in layered mixed-metal oxide cathode materials

    Science.gov (United States)

    Hoang, Khang; Johannes, Michelle

    Layered mixed-metal oxides with different compositions of (Ni,Co,Mn) [NCM] or (Ni,Co,Al) [NCA] have been used in commercial lithium-ion batteries. Yet their defect physics and chemistry is still not well understood, despite having important implications for the electrochemical performance. In this presentation, we report a hybrid density functional study of intrinsic point defects in the compositions LiNi1/3Co1/3Mn1/3O2 (NCM1/3) and LiNi1/3Co1/3Al1/3O2 (NCA1/3) which can also be regarded as model compounds for NCM and NCA. We will discuss defect landscapes in NCM1/3 and NCA1/3 under relevant synthesis conditions with a focus on the formation of metal antisite defects and its implications on the electrochemical properties and ultimately the design of NCM and NCA cathode materials.

  2. 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

  3. 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)

  4. A decision analysis method for selection of waste minimization process options for TRU mixed material at Rocky Flats

    International Nuclear Information System (INIS)

    Williams, R.E.; Dustin, D.F.

    1994-01-01

    When plutonium production operations were halted at the Rocky Flats Plant, there remained a volume of material that was retained in order that its plutonium content could be reclaimed. This material, known as residue, is transuranic and mixed transuranic material with a plutonium content above what was called the ''economic discard limit,'' or EDL. The EDL was defined in terms of each type of residue material, and each type of material is given an Item Description Code, or IDC. Residue IDCs have been grouped into general category descriptions which include plutonium (Pu) nitrate solutions, Pu chloride solutions, salts, ash, metal, filters, combustibles, graphite, crucibles, glass, resins, gloves, firebrick, and sludges. Similar material exists both below and above the EDL, with material with the (previous) economic potential for reclamation of plutonium classified as residue

  5. Exposure to selected fragrance materials. A case study of fragrance-mix-positive eczema patients

    DEFF Research Database (Denmark)

    Johansen, J D; Rastogi, Suresh Chandra; Menné, T

    1996-01-01

    . In all cases, the use of these cosmetics completely or partly explained present or past episodes of eczema. Between 1 to 6 constituents of the fragrance mix were found in 22 out of 23 products. The cosmetics of all the patients sensitive to hydroxycitronellal, eugenol, cinnamic alcohol and alpha......The aim of the present study was to assess exposure to constituents of the fragrance mix from cosmetic products used by fragrance-mix-positive eczema patients. 23 products, which had either given a positive patch and/or use test in a total of 11 fragrance-mix-positive patients, were analyzed....... It is concluded that exposure to constituents of the fragrance mix is common in fragrance-allergic patients with cosmetic eczema, and that the fragrance mix is a good reflection of actual exposure....

  6. Investigation of plasma interaction with carbon based and mixed materials related to next-generation fusion devices

    International Nuclear Information System (INIS)

    Guseva, M.I.; Martynenko, Yu.V.; Korshunov, S.N.

    2003-01-01

    Carbon-carbon composites, tungsten and beryllium are considered at present as candidate-materials for International Thermonuclear Experimental Reactor (ITER). The presence of various materials, as the divertor and the first wall components, will unavoidably result in the formation of mixed layers on the surfaces of plasma facing components. In this review, processes of plasma interaction with these materials and layers formed by mixing of the materials are considered. Mixed W-Be and W-C layers were prepared by deposition of two species atoms upon a substrate under simultaneous sputtering of two targets by 20 keV Ar + -ions. The thickness of the deposited mixed layers was 100-500 nm. The most important processes investigated here are: a) erosion at threshold energies and at various temperatures, b) erosion at plasma disruption, c) surface modification at normal operation regime and disruption, d) the influence of the surface modification on material erosion, e) erosion product formation at plasma disruption (dust creation), f) hydrogen isotopes retention in materials. An experimental method of determination of sputtering yield under ion bombardment in the near-threshold energy range has been developed. The method is based on the use of special regimes of field ion microscopic analysis. The method has been used for measurement of the sputtering yield of C-C composite, technically pure tungsten, tungsten oxide and mixed W-C layer on the tungsten by deuterium ions. The energy dependences of the sputtering yield of those materials by deuterium ions at energies ranging from 10 to 500 eV was investigated. Temperature dependences of pure and B-doped C-C composites erosion by deuterium ions were investigated. Material erosion was studied in a steady state plasma at the LENTA facility with parameters close to those expected at normal operation of ITER, and in the MKT plasma accelerator simulating plasma disruption. Surface modifications of graphite materials and tungsten

  7. 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.)

  8. 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

  9. 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

  10. 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.

  11. 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.

  12. 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

  13. Removal of Hazardous Pollutants from Wastewaters: Applications of TiO2-SiO2 Mixed Oxide Materials

    Directory of Open Access Journals (Sweden)

    Shivatharsiny Rasalingam

    2014-01-01

    Full Text Available The direct release of untreated wastewaters from various industries and households results in the release of toxic pollutants to the aquatic environment. Advanced oxidation processes (AOP have gained wide attention owing to the prospect of complete mineralization of nonbiodegradable organic substances to environmentally innocuous products by chemical oxidation. In particular, heterogeneous photocatalysis has been demonstrated to have tremendous promise in water purification and treatment of several pollutant materials that include naturally occurring toxins, pesticides, and other deleterious contaminants. In this work, we have reviewed the different removal techniques that have been employed for water purification. In particular, the application of TiO2-SiO2 binary mixed oxide materials for wastewater treatment is explained herein, and it is evident from the literature survey that these mixed oxide materials have enhanced abilities to remove a wide variety of pollutants.

  14. The primary results for the mixed carbon material used for high flux steady-state tokamak operation in China

    International Nuclear Information System (INIS)

    Guo, Q.G.; Li, J.G.; Zhai, G.T.; Liu, L.; Song, J.R.; Zhang, L.F.; He, Y.X.; Chen, J.L.

    2001-01-01

    Several types of carbon mixed materials have been developed in China to be used for high flux steady-state tokamak operation. Performance evaluation of these materials is necessary to determine their applicability as PFCs for high flux steady state. This paper describes the primary results of carbon mixed materials and the effects of dopants on properties are primarily discussed. Test results reveal that bulk boronized graphite has excellent physical and mechanical properties while their thermal conductivity is no more than 73 W/m K due to the formation of a uniform boron-carbon solid solution. In case of multi-element doped graphite, titanium dopant or a decreased boron content is favorable to enhance thermal conductivity. A kind of doped graphite has been developed with thermal conductivity as high as 278 W/m K by optimizing the compositions. Correlations among compositions, microstructure and properties of such doped graphite are discussed

  15. 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.)

  16. Neutronics Benchmarks for the Utilization of Mixed-Oxide Fuel: Joint U.S./Russian Progress Report for Fiscal Year 1997 Volume 2-Calculations Performed in the United States

    Energy Technology Data Exchange (ETDEWEB)

    Primm III, RT

    2002-05-29

    This volume of the progress report provides documentation of reactor physics and criticality safety studies conducted in the US during fiscal year 1997 and sponsored by the Fissile Materials Disposition Program of the US Department of Energy. Descriptions of computational and experimental benchmarks for the verification and validation of computer programs for neutron physics analyses are included. All benchmarks include either plutonium, uranium, or mixed uranium and plutonium fuels. Calculated physics parameters are reported for all of the computational benchmarks and for those experimental benchmarks that the US and Russia mutually agreed in November 1996 were applicable to mixed-oxide fuel cycles for light-water reactors.

  17. Neutronics Benchmarks for the Utilization of Mixed-Oxide Fuel: Joint U.S./Russian Progress Report for Fiscal Year 1997 Volume 2-Calculations Performed in the United States

    International Nuclear Information System (INIS)

    Primm III, RT

    2002-01-01

    This volume of the progress report provides documentation of reactor physics and criticality safety studies conducted in the US during fiscal year 1997 and sponsored by the Fissile Materials Disposition Program of the US Department of Energy. Descriptions of computational and experimental benchmarks for the verification and validation of computer programs for neutron physics analyses are included. All benchmarks include either plutonium, uranium, or mixed uranium and plutonium fuels. Calculated physics parameters are reported for all of the computational benchmarks and for those experimental benchmarks that the US and Russia mutually agreed in November 1996 were applicable to mixed-oxide fuel cycles for light-water reactors

  18. BIOREMEDIATION OF BTEX, NAPTHALENE, AND PHENANTHRENE IN AQUIFER MATERIAL USING MIXED OXYGEN/NITRATE ELECTRON

    Science.gov (United States)

    The goal of the research described herein was to examine the feasibility of biodegradation of mono and polycyclic aromatic hydrocarbons typically present in a manufactured gas processing (MGP) site groundwater and subsurface sediments under mixed oxygen/denitrifying conditions. ...

  19. A Review of the Application of Zeolite Materials in Warm Mix Asphalt Technologies

    Directory of Open Access Journals (Sweden)

    Agnieszka Woszuk

    2017-03-01

    Full Text Available Among warm mix asphalt (WMA technologies, asphalt foaming techniques offer high potential in terms of decreasing production temperature. Reluctance of manufacturers to introduce this technology is connected with the concerns of a large investment costs. However, there are known additives which, through asphalt foaming, allow a decrease in temperatures by approximately 30 °C; the use of these additives do not involve expensive investment in order to change the asphalt mix production method. These additives are zeolites, that is, minerals of the aluminosilicate group, the crystalline structure of which contains water bound in a specific way. Its release, at mix asphalt production temperatures, causes asphalt foaming. It is currently known that zeolites can be used in WMA, including natural and synthetic zeolites obtained using chemical reagents and waste. This review presents the results of studies of WMA technology, including the effects of zeolite addition on asphalt properties and mix asphalt, as well as related environmental, economic, and technological benefits.

  20. Studying the Behavior of Asphalt Mix and Their Properties in the Presence of Nano Materials

    OpenAIRE

    Aman Patidar; Dipankar Sarkar; Manish Pal

    2017-01-01

    Due to rapid development, increase in the traffic load, higher traffic volume and seasonal variation in temperature, asphalt pavement shows distresses like rutting, fatigue and thermal cracking etc. because of this pavement fails during service life so that bitumen needs to be modified with some additive. In this study VG30 grade bitumen modify with addition of nanosilica with 1% to 5% (increment of 1%) by weight of bitumen. Hot mix asphalt (HMA) have higher mixing, laying and rolling tempera...

  1. A comparative study of recycled aggregates from concrete and mixed debris as material for unbound road sub-base

    International Nuclear Information System (INIS)

    Jimenez, J. R.; Agrela, F.; Ayuso, J.; Lopez, M.

    2011-01-01

    Seven different types of recycled aggregates from construction and demolition waste (CDW) have been evaluated as granular materials for unbound road sub bases construction. The results showed that recycled concrete aggregates complied with all specifications for using in the construction of unbound structural layers (sub-base) for T3 and T4 traffic categories according to the Spanish General Technical Specification for Road Construction (PG-3). Some mixed recycled aggregates fell short of some specifications due to a high content of sulphur compounds and poor fragmentation resistance. Sieving off the fine fraction prior to crushing the mixed CDW reduce the total sulphur content and improve the quality of the mixed recycled aggregates, by contrast, pre-sieving concrete CDW had no effect on the quality of the resulting aggregates. The results were compared with a crushed limestone as natural aggregate. (Author) 23 refs.

  2. 1st IAEA research co-ordination meeting on 'plasma-material interaction data for mixed plasma facing materials in fusion reactors'. Summary report

    International Nuclear Information System (INIS)

    Janev, R.K.; Longhurst, G.

    1998-12-01

    The proceedings and conclusions of the 1st IAEA Research Co-ordination Meeting on 'Plasma-Material Interaction Data for Mixed Plasma Facing Materials in Fusion Reactors', held on December 19 and 20, 1998 at the IAEA Headquarters in Vienna, are briefly described. This report includes a summary of the presentations made by meeting participants, a review of the data availability and data needs in the areas from the scope of the Co-ordinated Research Project (CRP) on the subject of the meeting, and recommendations regarding the future work within this CRP. (author)

  3. 2nd (final) IAEA research co-ordination meeting on 'plasma-material interaction data for mixed plasma facing materials in fusion reactors'. Summary report

    International Nuclear Information System (INIS)

    Clark, R.E.H.

    2001-11-01

    The proceedings and conclusions of the 2nd Research Co-ordination Meeting on 'Plasma-Material Interaction Data for Mixed Plasma Facing Materials in Fusion Reactors', held on October 16 and 17, 2000 at the IAEA Headquarters in Vienna, are briefly described. This report includes a summary of the presentations made by the meeting participants and a review of the accomplishments of the Co-ordinated Research Project (CRP). In addition, short summaries from the participants are included indicating the specific research completed in support of this CRP. (author)

  4. 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)

  5. Advanced research workshop: nuclear materials safety

    International Nuclear Information System (INIS)

    Jardine, L J; Moshkov, M M.

    1999-01-01

    The Advanced Research Workshop (ARW) on Nuclear Materials Safety held June 8-10, 1998, in St. Petersburg, Russia, was attended by 27 Russian experts from 14 different Russian organizations, seven European experts from six different organizations, and 14 U.S. experts from seven different organizations. The ARW was conducted at the State Education Center (SEC), a former Minatom nuclear training center in St. Petersburg. Thirty-three technical presentations were made using simultaneous translations. These presentations are reprinted in this volume as a formal ARW Proceedings in the NATO Science Series. The representative technical papers contained here cover nuclear material safety topics on the storage and disposition of excess plutonium and high enriched uranium (HEU) fissile materials, including vitrification, mixed oxide (MOX) fuel fabrication, plutonium ceramics, reprocessing, geologic disposal, transportation, and Russian regulatory processes. This ARW completed discussions by experts of the nuclear materials safety topics that were not covered in the previous, companion ARW on Nuclear Materials Safety held in Amarillo, Texas, in March 1997. These two workshops, when viewed together as a set, have addressed most nuclear material aspects of the storage and disposition operations required for excess HEU and plutonium. As a result, specific experts in nuclear materials safety have been identified, know each other from their participation in t he two ARW interactions, and have developed a partial consensus and dialogue on the most urgent nuclear materials safety topics to be addressed in a formal bilateral program on t he subject. A strong basis now exists for maintaining and developing a continuing dialogue between Russian, European, and U.S. experts in nuclear materials safety that will improve the safety of future nuclear materials operations in all the countries involved because of t he positive synergistic effects of focusing these diverse backgrounds of

  6. Treatment of mixed F006 contaminated material to meet the new EPA debris rule at the Savannah River Site

    International Nuclear Information System (INIS)

    Pickett, J.B.; Diener, G.A.; Carroll, S.J.; Steingard, J.M.

    1993-01-01

    The Westinghouse Savannah River Company (WSRC), as the operating contractor for the Department of Energy (DOE) at the Savannah River Site (SRS) has demonstrated a procedure to clean mixed (radioactive/hazardous) materials to meet the criteria in the recently promulgated Land Disposal Restrictions ''debris'' rule. The material was equipment (steel piping, transfer pumps valves) which had been used in industrial wastewater treatment facility to transfer listed F006 wastewater treatment plating line sludges to a RCRA storage tank complex. When the equipment needed to be replaced/repaired, it was concluded that the resulting debris would have to be managed as a mixed waste, due to the fact that the solid waste ''contained'' the listed hazardous waste

  7. Molten salt oxidation of mixed wastes: Separation of radioactive materials and Resource Conservation and Recovery Act (RCRA) materials

    International Nuclear Information System (INIS)

    Bell, J.T.; Haas, P.A.; Rudolph, J.C.

    1995-01-01

    The Oak Ridge National Laboratory (ORNL) is participating in a program to apply a molten salt oxidation (MSO) process to treatment of mixed (radioactive and RCRA) wastes. The salt residues from the MSO treatment will require further separations or other processing to prepare them for final disposal. A bench-scale MSO apparatus is being installed at ORNL and will be operated on real Oak Ridge wastes. The treatment concepts to be tested and demonstrated on the salt residues from real wastes are described

  8. The accretion of solar material onto white dwarfs: No mixing with core material implies that the mass of the white dwarf is increasing

    Directory of Open Access Journals (Sweden)

    Sumner Starrfield

    2014-02-01

    Full Text Available Cataclysmic Variables (CVs are close binary star systems with one component a white dwarf (WD and the other a larger cooler star that fills its Roche Lobe. The cooler star is losing mass through the inner Lagrangian point of the binary and some unknown fraction of this material is accreted by the WD. One consequence of the WDs accreting material, is the possibility that they are growing in mass and will eventually reach the Chandrasekhar Limit. This evolution could result in a Supernova Ia (SN Ia explosion and is designated the Single Degenerate Progenitor (SD scenario. This paper is concerned with the SD scenario for SN Ia progenitors. One problem with the single degenerate scenario is that it is generally assumed that the accreting material mixes with WD core material at some time during the accretion phase of evolution and, since the typical WD has a carbon-oxygen CO core, the mixing results in large amounts of carbon and oxygen being brought up into the accreted layers. The presence of enriched carbon causes enhanced nuclear fusion and a Classical Nova explosion. Both observations and theoretical studies of these explosions imply that more mass is ejected than is accreted. Thus, the WD in a Classical Nova system is losing mass and cannot be a SN Ia progenitor. However, the composition in the nuclear burning region is important and, in new calculations reported here, the consequences to the WD of no mixing of accreted material with core material have been investigated so that the material involved in the explosion has only a Solar composition. WDs with a large range in initial masses and mass accretion rates have been evolved. I find that once sufficient material has been accreted, nuclear burning occurs in all evolutionary sequences and continues until a thermonuclear runaway (TNR occurs and the WD either ejects a small amount of material or its radius grows to about 1012 cm and the evolution is ended. In all cases where mass ejection occurs

  9. 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

  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 neutrons emis

  11. Certified reference material for radionuclides in fish flesh sample IAEA-414 (mixed fish from the Irish Sea and North Sea)

    DEFF Research Database (Denmark)

    Pham, M.K.; Sanchez-Cabeza, J.A.; Povinec, P.P.

    2006-01-01

    A certified reference material (CRM) for radionuclides in fish sample IAEA-414 (mixed fish from the Irish Sea and North Seas) is described and the results of the certification process are presented. Nine radionuclides (K-40, Cs-137, Th-232, U-234, U-235, U-238, Pu-238, Pu239+240 and Am-241) were...... ratios are also included. The CRM can be used for quality assurance/quality control of the analysis of radionuclides in fish sample, for the development and validation of analytical methods and for training purposes. The material is available from IAEA, Vienna, in 100 g units. (c) 2006 Elsevier Ltd. All...

  12. Mixed phase-amplitude holographic gratings recorded in bleached silver halide materials

    International Nuclear Information System (INIS)

    Neipp, C.; Pascual, C.; Belendez, A.

    2002-01-01

    The coupled wave theory of Kogelnik has given a well-established basis for the comprehension of how light propagates inside a volume hologram. This theory gives a good approximation for the diffraction efficiency of both volume phase holograms and volume absorption holograms. Mixed holograms (phase and absorption) have also been dealt with from the point of view of the coupled wave theory. In this paper we use Kogelnik's coupled wave theory to give quantitative information about the mechanisms which produce mixed gratings in photographic emulsions. In particular, we demonstrate that mixed amplitude-phase gratings are recorded on photographic emulsions when fixation-free bleaching techniques are used to obtain volume holograms. We will prove that the oxidation products of the bleach can give rise to an absorption modulation at high values of exposure and high concentrations of potassium bromide in the bleach bath. We will also give quantitative data regarding the absorption created by these oxidation products. (author)

  13. On a separating method for mixed-modes crack growth in wood material using image analysis

    Science.gov (United States)

    Moutou Pitti, R.; Dubois, F.; Pop, O.

    2010-06-01

    Due to the complex wood anatomy and the loading orientation, the timber elements are subjected to a mixed-mode fracture. In these conditions, the crack tip advance is characterized by mixed-mode kinematics. In order to characterize the fracture process function versus the loading orientation, a new mixed-mode crack growth timber specimen is proposed. In the present paper, the design process and the experimental validation of this specimen are proposed. Using experimental results, the energy release rate is calculated for several modes. The calculi consist on the separation of each fracture mode. The design of the specimen is based on the analytical approach and numerical simulation by finite element method. The specimen particularity is the stability of the crack propagation under a force control.

  14. Hydration rate and strength development of low-heat type portland cement mortar mixed with pozzolanic materials

    International Nuclear Information System (INIS)

    Matsui, Jun

    1998-01-01

    Recently, low-heat type Portland cement was specified in Japan Industrial Standards (JIS). Its hydration proceeds slowly. The results of the research so far obtained indicate that slow hydration of cement and mixing of pozzolanic materials with cement make micro-structure of harded cement paste dense and durable. In this study, a blended cement using low-heat type Portland cement and some of pozzolanic materials has been newly developed and its strength property and hydration ratio were checked. The followings are conclusion. (1) Hydration rate of cement paste varies with the replacement ratio of pozzolanic materials. (2) A good liner relationship between strength and total hydration rate of cement paste was observed. (3) A proper replacement ratio of both base-cement and pozzolanic material for manufacturing a blended cement is 50%. (author)

  15. Reduction of minimum required weight of cementitious materials in WisDOT concrete mixes.

    Science.gov (United States)

    2011-12-01

    This project was designed to explore the feasibility of lowering the cementitious materials content : (CMC) used in Wisconsin concrete pavement construction. The cementitious materials studied included : portland cement, fly ash, and ground granulate...

  16. Study on HDPE Mixed with Sand as Backfilled Material on Retaining Structure

    Science.gov (United States)

    Talib, Z. A.

    2018-04-01

    The failure of the retaining wall is closely related to backfill material. Granular soils such as sand and gravel are most suitable backfill material because of its drainage properties. However two basic materials are quite heavy and contribute high amount of lateral loads. This study was to determine the effectiveness High Density Polyethylene (HDPE) as a backfill material. HDPE has a lighter weight compare to the sand. It makes HDPE has potential to be used as backfill material. The objective of this study is to identify the most effective percentage of HDPE to replace sand as a backfill material. The percentage of HDPE used in this study was 20%, 30%, 50%, 75% and also 100%. Testing involved in this study were sieve analysis test, constant head permeability test, direct shear test and relative density test. The result shows that the HDPE can be used as backfilled material and save the cost of backfill material

  17. Reduction of minimum required weight of cementitious materials in WisDOT concrete mixes.

    Science.gov (United States)

    2011-12-01

    "This project was designed to explore the feasibility of lowering the cementitious materials content : (CMC) used in Wisconsin concrete pavement construction. The cementitious materials studied included : portland cement, fly ash, and ground granulat...

  18. Preparation and characterisation of mixed silicon oxycarbide materials; Preparacion y caracterizacion de materiales de oxicarburo de silicio mixtos

    Energy Technology Data Exchange (ETDEWEB)

    Tellez, L.; Tamayo, A.; Mazo, M. A.; Rubio, F.; Rubio, J.

    2010-07-01

    In this work different mixed Silicon oxicarbide materials have been prepared. Si, Si-Ti, Si-Zr and Si-Al oxicarbide materials have been obtained from pyrolisis at 1000 degree centigrade and 1300 degree centigrade of the respective preceramic materials. After pyrolisis X, D and T units of the oxycarbide structure have been observed in such materials. They show the presence of Si-C and Si-O bonds in a given material. The characterization has been carried out by means of FT-IR, Raman NMR {sup 2}9Si, NMR {sup 1}3C and XRD. The formation of Si-Ti, Si-Zr and Si-Al bonds has been estimated in accordance with the decrease of the Si-O-Si wave number observed in the FT-IR spectra. Si and Si-Ti oxycarbide materials do not lead to crystallisation after pyrolisis at highest temperatures, however for Si-Zr and Si-Al oxycarbide materials different crystalline phases have been observed. All pyrolised materials present free and carbidic carbon. After pyrolisis at 1300 degree centigrade the free carbon reacts with Si-O bonds to form SiC{sub 4} groups which must be assigned to nucleus of the {beta}-SiC crystals. (Author)

  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. Lifetime improvement mechanism in organic light-emitting diodes with mixed materials at a heterojunction interface

    Science.gov (United States)

    Minagawa, Masahiro; Takahashi, Noriko

    2016-02-01

    To investigate the lifetime improvement mechanism caused by mixing at the heterojunction interface, organic light-emitting diodes (OLEDs) with stacked and mixed 4,4‧-bis[N-(1-naphthyl)-N-phenyl-amino]-biphenyl (α-NPD)/tris(8-hydroxyquinoline)aluminum (Alq3) interfaces were fabricated, and changes in their displacement current due to continuous operation were measured. A decrease in accumulated holes at the α-NPD/Alq3 interface was observed in the stacked configuration devices over longer operations. These results indicate that the injected hole density was reduced during continuous operation, implying that the carrier balance became uneven in the emission region. However, few accumulated holes and changes in the displacement current due to continuous operation were observed in the devices having the mixed layer. Therefore, it was deduced that the number of holes concentrated between the α-NPD and Alq3 layers was decreased by mixing at the heterojunction interface, and that the change in the number of holes was smaller during continuous operation, resulting in less degradation.

  1. The direct incorporation of micro-encapsulated phase change materials in the concrete mixing process

    NARCIS (Netherlands)

    Hunger, M.; Entrop, A.G.; Mandilaras, I.; Brouwers, H.J.H.; Founti, M.; Durmisevic, E.

    2009-01-01

    The present study refers to a set of tests using different amounts of microencapsulated PCM directly mixed into self-compacting concrete. This SCC is investigated regarding its fresh and hardened properties. It will be shown that increasing PCM amounts lead to lower thermal conductivity and

  2. 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

  3. Determination of renewable energy yield from mixed waste material from the use of novel image analysis methods.

    Science.gov (United States)

    Wagland, S T; Dudley, R; Naftaly, M; Longhurst, P J

    2013-11-01

    Two novel techniques are presented in this study which together aim to provide a system able to determine the renewable energy potential of mixed waste materials. An image analysis tool was applied to two waste samples prepared using known quantities of source-segregated recyclable materials. The technique was used to determine the composition of the wastes, where through the use of waste component properties the biogenic content of the samples was calculated. The percentage renewable energy determined by image analysis for each sample was accurate to within 5% of the actual values calculated. Microwave-based multiple-point imaging (AutoHarvest) was used to demonstrate the ability of such a technique to determine the moisture content of mixed samples. This proof-of-concept experiment was shown to produce moisture measurement accurate to within 10%. Overall, the image analysis tool was able to determine the renewable energy potential of the mixed samples, and the AutoHarvest should enable the net calorific value calculations through the provision of moisture content measurements. The proposed system is suitable for combustion facilities, and enables the operator to understand the renewable energy potential of the waste prior to combustion. Copyright © 2013 Elsevier Ltd. All rights reserved.

  4. 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)

  5. Certified reference material for radionuclides in fish flesh sample IAEA-414 (mixed fish from the Irish Sea and North Sea)

    International Nuclear Information System (INIS)

    Pham, M.K.; Sanchez-Cabeza, J.A.; Povinec, P.P.; Arnold, D.; Benmansour, M.; Bojanowski, R.; Carvalho, F.P.; Kim, C.K.; Esposito, M.; Gastaud, J.; Gasco, C.L.; Ham, G.J.; Hegde, A.G.; Holm, E.; Jaskierowicz, D.; Kanisch, G.; Llaurado, M.; La Rosa, J.; Lee, S.-H.; Liong Wee Kwong, L.; Le Petit, G.; Maruo, Y.; Nielsen, S.P.; Oh, J.-S.; Oregioni, B.; Palomares, J.; Pettersson, H.B.L.; Rulik, P.; Ryan, T.P.; Sato, K.; Schikowski, J.; Skwarzec, B.; Smedley, P.A.; Tarjan, S.; Vajda, N.; Wyse, E.

    2006-01-01

    A certified reference material (CRM) for radionuclides in fish sample IAEA-414 (mixed fish from the Irish Sea and North Seas) is described and the results of the certification process are presented. Nine radionuclides ( 4 K, 137 Cs, 232 Th, 234 U, 235 U, 238 U, 238 Pu, 239+24 Pu and 241 Am) were certified for this material. Information on massic activities with 95% confidence intervals is given for six other radionuclides ( 9 Sr, 21 Pb( 21 Po), 226 Ra, 239 Pu, 24 Pu 241 Pu). Less frequently reported radionuclides ( 99 Tc, 129 I, 228 Th, 23 Th and 237 Np) and information on some activity and mass ratios are also included. The CRM can be used for quality assurance/quality control of the analysis of radionuclides in fish sample, for the development and validation of analytical methods and for training purposes. The material is available from IAEA, Vienna, in 100 g units

  6. 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

  7. 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

  8. Potential benefits of the recovery of strategic/critical materials from mixed wastes

    International Nuclear Information System (INIS)

    Acton, C.F.; Hofman, L.A.

    1986-01-01

    The United States is currently dependent on foreign suppliers for many of its strategic and critical materials. Chrome, platinum, and manganese are primarily supplied by South Africa and the Soviet Union. Zaire and Zambia are the major sources of cobalt. Budgetary cutbacks over the last 20 yr have reduced the stockpiles maintained for defense use. Many are currently in deficit. Considering the tenuous relationships the United States maintains with many of the countries relied on for these strategic materials, an alternative source of materials would be welcomed. One potential source of some of these strategic and critical materials is uranium mill tailings. Substantial quantities of uranium, vanadium, molybdenum, and cobalt have been successfully recovered from uranium mill tailings at Durango, Old Rifle, Grand Junction, Colorado, and Mexican Hat, Utah. The national defense stockpile currently has a deficit of cobalt and vanadium. Other strategic materials currently in deficit (nickel, platinum, palladium, etc.) also may be recoverable from mill tailings

  9. 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.

  10. 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.

  11. 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

  12. Vestas Pinaria Region: Original Basaltic Achondrite Material Derived from Mixing Upper and Lower Crust

    Science.gov (United States)

    Mcfadden, L. A.; Combe, Jean-Philippe; Ammannito, Eleonora; Frigeri, Alessandro; Stephan, Katrin; Longobardo, Andrea; Palomba, Ernesto; Tosi, Federico; Zambon, Francesca; Krohn, Katrin; hide

    2015-01-01

    Analysis of data from the Dawn mission shows that the Pinaria region of Vesta spanning a portion of the rim of the Rheasilvia basin is bright and anhydrous. Reflectance spectra, absorption band centers, and their variations, cover the range of pyroxenes from diogenite-rich to howardite and eucrite compositions, with no evidence of olivine in this region. By examining band centers and depths of the floor, walls and rims of six major craters in the region, we find a lane of diogenite-rich material next to howardite-eucrite material that does not follow the local topography. The source of this material is not clear and is probably ejecta from post-Rheasilvia impacts. Material of a howardite-eucrite composition originating from beyond the Rheasilvia basin is evident on the western edge of the region. Overall, the Pinaria region exposes the complete range of basaltic achondrite parent body material, with little evidence of contamination of non-basaltic achondrite material. With both high reflectance and low abundance of hydrated material, this region of Vesta may be considered the "Pinaria desert".

  13. Sol-gel preparation of cobalt manganese mixed oxides for their use as electrode materials in lithium cells

    International Nuclear Information System (INIS)

    Lavela, P.; Tirado, J.L.; Vidal-Abarca, C.

    2007-01-01

    An ethanol dehydration procedure has been used to precipitate gel-like citrate precursors containing cobalt and manganese transition metal ions. Further annealing led to the Mn x Co 3-x O 4 spinel oxide series (x: 1, 1.5, 2, 3). Annealing temperature and treatment time were also evaluated to optimize the performance of the oxides as active electrode materials in lithium cells. The manganese-cobalt mixed oxides obtained by this procedure were cubic or tetragonal phases depending on the cobalt content. SEM images showed spherical macroporous aggregates for MnCo 2 O 4 and hollow spheres for manganese oxides. The galvanostatic cycling of lithium cells assembled with these materials demonstrated a simultaneous reduction of cobalt and manganese during the first discharge and separation of cobalt- and manganese-based products on further cycling. As compared with binary manganese oxides, a notorious electrochemical improvement was observed in the mixed oxides. This behavior is a consequence of the synergistic effect of both transition metal elements, associated with the in-situ formation of a nanocomposite electrode material when cobalt is introduced in the manganese oxide composition. Values higher than 400 mAh/g were sustained after 50 cycles for MnCo 2 O 4

  14. Materials testing for in situ stabilization treatability study of INEEL mixed wastes soils

    International Nuclear Information System (INIS)

    Heiser, J.; Fuhrmann, M.

    1997-09-01

    This report describes the contaminant-specific materials testing phase of the In Situ Stabilization Comprehensive Environment Response, Compensation, and Liability Act (CERCLA) Treatability Study (TS). The purpose of materials testing is to measure the effectiveness of grouting agents to stabilize Idaho National Engineering and Environmental Laboratory (INEEL) Acid Pit soils and select a grout material for use in the Cold Test Demonstration and Acid Pit Stabilization Treatability Study within the Subsurface Disposal Area (SDA) at the Radioactive Waste Management Complex (RWMC). Test results will assist the selecting a grout material for the follow-on demonstrations described in Test Plan for the Cold Test Demonstration and Acid Pit Stabilization Phases of the In Situ Stabilization Treatability Study at the Radioactive Waste Management Complex

  15. Rate sensitivity of mixed mode interface toughness of dissimilar metallic materials: Studied at steady state

    DEFF Research Database (Denmark)

    Nielsen, Kim Lau; Niordson, Christian Frithiof

    2012-01-01

    the SSV model [Suo, Z., Shih, C., Varias, A., 1993. A theory for cleavage cracking in the presence of plastic flow. Acta Metall. Mater. 41, 1551–1557] embedded in a steady state finite element formulation, here assuming plane strain conditions and small-scale yielding. Results are presented for a wide......Crack propagation in metallic materials produces plastic dissipation when material in front for the crack tip enters the active plastic zone traveling with the tip, and later ends up being part of the residual plastic strain wake. Thus, the macroscopic work required to advance the crack...... is typically much larger than the work needed in the near tip fracture process. For rate sensitive materials, the amount of plastic dissipation typically depends on the rate at which the material is deformed. A dependency on the crack velocity should therefore be expected. The objective of this paper...

  16. 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)

  17. 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

  18. New Mixed Conductivity Mechanisms in the Cold Plasma Device Based on Silver-Modified Zeolite Microporous Electronic Materials

    Science.gov (United States)

    Koç, Sevgul Ozturk; Galioglu, Sezin; Ozturk, Seckin; Kurç, Burcu Akata; Koç, Emrah; Salamov, Bahtiyar G.

    2018-02-01

    We have analyzed the interaction between microdischarge and microporous zeolite electronic materials modified by silver (Ag0) nanoparticles (resistivity 1011 to 106 Ω cm) on the atmospheric pressure cold plasma generation in air. The generation and maintenance of stable cold plasma is studied according to the effect of the Ag0 nanoparticles. The role of charge carriers in mixed conductivity processes and electrical features of zeolite from low pressure to atmospheric pressure is analyzed in air microplasmas for both before and after breakdown regimes. The results obtained from the experiments indicate that Ag0 nanoparticles play a significant role in considerably reducing the breakdown voltage in plasma electronic devices with microporous zeolite electronic materials.

  19. High temperature vitrification of surrogate Savannah River Site (SRS) mixed waste materials

    International Nuclear Information System (INIS)

    Applewhite-Ramsey, A.; Schumacher, R.F.; Spatz, T.L.; Newsom, R.A.; Circeo, L.J.; Danjaji, M.B.

    1995-01-01

    The Savannah River Technology Center (SRTC) has been funded through the DOE Office of Technology Development (DOE-OTD) to investigate high-temperature vitrification technologies for the treatment of diverse low-level and mixed wastes. High temperature vitrification is a likely candidate for processing heterogeneous solid wastes containing low levels of activity. Many SRS wastes fit into this category. Plasma torch technology is one high temperature vitrification method. A trial demonstration of plasma torch processing is being performed at the Georgia Institute of Technology on surrogate SRS wastes. This effort is in cooperation with the Engineering Research and Development Association of Georgia Universities (ERDA) program. The results of phase 1 of these plasma torch trials will be presented

  20. Dark fermentative hydrogen production by defined mixed microbial cultures immobilized on ligno-cellulosic waste materials

    Energy Technology Data Exchange (ETDEWEB)

    Patel, Sanjay K.S. [Microbial Biotechnology and Genomics, Institute of Genomics and Integrative Biology (IGIB), CSIR, Delhi University Campus, Mall Road, Delhi 110007 (India); Department of Biotechnology, University of Pune, Pune 411007 (India); Purohit, Hemant J. [Environmental Genomics Unit, National Environmental Engineering Research Institute (NEERI), CSIR, Nehru Marg, Nagpur 440020 (India); Kalia, Vipin C. [Microbial Biotechnology and Genomics, Institute of Genomics and Integrative Biology (IGIB), CSIR, Delhi University Campus, Mall Road, Delhi 110007 (India)

    2010-10-15

    Mixed microbial cultures (MMCs) based on 11 isolates belonging to Bacillus spp. (Firmicutes), Bordetella avium, Enterobacter aerogenes and Proteus mirabilis (Proteobacteria) were employed to produce hydrogen (H{sub 2}) under dark fermentative conditions. Under daily fed culture conditions (hydraulic retention time of 2 days), MMC6 and MMC4, immobilized on ligno-cellulosic wastes - banana leaves and coconut coir evolved 300-330 mL H{sub 2}/day. Here, H{sub 2} constituted 58-62% of the total biogas evolved. It amounted to a H{sub 2} yield of 1.54-1.65 mol/mol glucose utilized over a period of 60 days of fermentation. The involvement of various Bacillus spp. -Bacillus sp., Bacillus cereus, Bacillus megaterium, Bacillus pumilus and Bacillus thuringiensis as components of the defined MMCs for H{sub 2} production has been reported here for the first time. (author)

  1. Outstanding visible photocatalytic activity of a new mixed bismuth titanatate material

    Energy Technology Data Exchange (ETDEWEB)

    Zambrano, P. [Instituto de Ciencia de Materiales de Sevilla, CSIC-Universidad de Sevilla, Americo Vespucio 49, 410092, Sevilla (Spain); Departamento Cristalografía, Mineralogía y Química Agrícola, Universidad de Sevilla, C/Profesor García González s/n, 41012 Sevilla (Spain); Sayagués, M.J.; Navío, J.A. [Instituto de Ciencia de Materiales de Sevilla, CSIC-Universidad de Sevilla, Americo Vespucio 49, 410092, Sevilla (Spain); Hidalgo, M.C., E-mail: carmen.hidalgo@csic.es [Instituto de Ciencia de Materiales de Sevilla, CSIC-Universidad de Sevilla, Americo Vespucio 49, 410092, Sevilla (Spain)

    2017-02-01

    Highlights: • Photocatalyst based on bismuth titanates with high visible activity. • Its visible activity as high as UV activity of TiO{sub 2} P25 for phenol degradation. • Photocatalyst is majority of phase Bi{sub 20}TiO{sub 32} with Bi{sub 4}Ti{sub 3}O{sub 12} and amorphous TiO{sub 2}. • High visible activity due to low BG, interconnected phases and high surface area. - Abstract: In this work, a new photocatalyst based on bismuth titanates with outstanding visible photocatalytic activity was prepared by a facile hydrothermal method. The synthesised material showed visible activity as high as UV activity of commercial TiO{sub 2} P25 under the same experimental conditions for phenol degradation. A wide characterisation of the photocatalyst was performed. The material was composed of three phases; majority of Bi{sub 20}TiO{sub 32} closely interconnected to Bi{sub 4}Ti{sub 3}O{sub 12} and amorphous TiO{sub 2}. The high visible activity showed by this material could be ascribed to a combination of several features; i.e. low band gap energy value (2.1 eV), a structure allowing a good separation path for visible photogenerated electron-holes pairs and a relatively high surface area. This photocatalyst appeared as a promising material for solar and visible applications of photocatalysis.

  2. The influence of fusion-relevant D2-0.1He plasma on Be-W mixed-materials

    International Nuclear Information System (INIS)

    Jepu, I.; Baldwin, M.J.; Nishijima, D.; Doerner, R.P.; Porosnicu, C.; Lungu, C.P.; Dinca, P.; Marin, A.

    2017-01-01

    Compositionally homogeneous mixed-material layers of Be and W, ∼2 μm thick, and of various W contents (spanning 0–100 at.% W), are produced by thermionic vacuum arc deposition. The layers are exposed to low energy (∼50 eV), high ion flux (6–8 × 10 22 m −2 s −1 ) D 2 -0.1He (10% He) mixed species plasma at 473 K and 1073 K to simulate ITER-like first wall and divertor plasma-material interaction. Surface morphological and compositional analysis is conducted prior and subsequent to plasma exposure. The action of the plasma produces almost complete depletion of Be from the near surface, and this is found regardless of the exposure temperature or initial composition. In the exposure case of 473 K, thermal desorption spectrometry reveals decreased D retention in W rich compositions compared to similar exposure in D 2 plasma without He, but no discernable difference is noted in Be rich compositions. At 1073 K, surface enrichment in W, by depletion of the Be, results in He induced W “fuzz” structures for all Be-W compositions explored.

  3. Synthesis and characterization of mesoporous ceria/alumina nanocomposite materials via mixing of the corresponding ceria and alumina gel precursors.

    Science.gov (United States)

    Khalil, Kamal M S

    2007-03-01

    Mesoporous ceria/alumina, CeO(2)/Al(2)O(3), composites containing 10, 20 and 30% (w/w) ceria were prepared by a novel gel mixing method. In the method, ceria gel (formed via hydrolysis of ammonium cerium(IV) nitrate by aqueous ammonium carbonate solution) and alumina gel (formed via controlled hydrolysis of aluminum tri-isopropoxide) were mixed together. The mixed gel was subjected to subsequent drying and calcination for 3 h at 400, 600, 800 and 1000 degrees C. The uncalcined (dried at 110 degrees C) and the calcined composites were investigated by different techniques including TGA, DSC, FTIR, XRD, SEM and nitrogen adsorption/desorption isotherms. Results indicated that composites calcined for 3 h at 800 degrees C mainly kept amorphous alumina structure and gamma-alumina formed only upon calcinations at 1000 degrees C. On the other hand, CeO(2) was found to crystallize in the common ceria, cerinite, phase and it kept this structure over the entire calcination range (400-1000 degrees C). Therefore, high surface areas, stable surface textures, and non-aggregated nano-sized ceria dispersions were obtained. A systematic texture change based on ceria ratio was observed, however in all cases mesoporous composite materials exposing thermally stable texture and structure were obtained. The presented method produces composite ceria/alumina materials that suit different applications in the field of catalysis and membranes technology, and throw some light on physicochemical factors that determine textural morphology and thermal stability of such important composite.

  4. Radiological emergencies due to postulated events of melted radioactive material mixed in steel reaching public domain

    International Nuclear Information System (INIS)

    Meena, T.R.; Anoj Kumar; Patra, R.P.; Vikas; Patil, S.S.; Chatterjee, M.K.; Sharma, Ranjit; Murali, S.

    2014-01-01

    National level response mechanism is developed at emergency response centres of DAE (DAE-ERCs) at 22 different locations spread all over the country and National Disaster Response Forces with National Disaster Management Authority (NDMA). ERCs are equipped with radiation monitors, radionuclide identifinders, Personnel Radiation Dosimeters (PRD) with monitoring capabilities of the order of tens of nGy/h (μR/hr) above the radiation background at any suspected locations. Even if small amounts of radioactive material is smuggled and brought in some other form into public domain, ERCs are capable to detect, identify and segregate the radioactive material from any inactive scrap. DAE-ERCs have demonstrated their capability in source search, detection, identification and recovery during the radiological emergency at Mayapuri, New Delhi

  5. Radiological emergencies due to postulated events of melted radioactive material mixed in steel reaching public domain

    Energy Technology Data Exchange (ETDEWEB)

    Meena, T. R.; Kumar, Anoj; Patra, R. P.; Vikas,; Patil, S. S.; Chatterjee, M. K.; Sharma, Ranjit; Murali, S., E-mail: tejram@barc.gov.in [Radiation Safety Systems Division, Bhabha Atomic Research Centre, Mumbai (India)

    2014-07-01

    National level response mechanism is developed at emergency response centres of DAE (DAE-ERCs) at 22 different locations spread all over the country and National Disaster Response Forces with National Disaster Management Authority (NDMA). ERCs are equipped with radiation monitors, radionuclide identifinders, Personnel Radiation Dosimeters (PRD) with monitoring capabilities of the order of tens of nGy/h (μR/hr) above the radiation background at any suspected locations. Even if small amounts of radioactive material is smuggled and brought in some other form into public domain, ERCs are capable to detect, identify and segregate the radioactive material from any inactive scrap. DAE-ERCs have demonstrated their capability in source search, detection, identification and recovery during the radiological emergency at Mayapuri, New Delhi.

  6. New Portland cement-based materials for endodontics mixed with articaine solution: a study of cellular response.

    Science.gov (United States)

    Gandolfi, Maria Giovanna; Perut, Francesca; Ciapetti, Gabriela; Mongiorgi, Romano; Prati, Carlo

    2008-01-01

    The biocompatibility of innovative tetrasilicate cements proposed for root-end filling restorations was tested. White ProRoot-MTA and AH Plus were used as control. The new cements were mixed with a local anesthetic solution (4% articaine) to form a paste. Human osteoblast-like cells Saos-2 were challenged in short-term cultures (72 hours) with solid materials and with material extracts prepared in culture medium. Cell growth and viability, cellular attachment, and morphologic features were assessed to verify cell/material interactions. No acute toxicity was exerted by the experimental cements in the assay systems. On solid samples Saos-2 adhered and proliferated on all the experimental cements and on MTA. The ultrastructural findings revealed that Saos-2 were able to adhere and to spread. The maintenance of the osteoblastic phenotype on the innovative cements was confirmed by the alkaline phosphatase assay. All experimental cements prepared with articaine supported the growth of bone-like cells, showing suitable properties to be used as canal sealers and root-end filling materials.

  7. Evaluation of performance indicators applied to a material recovery facility fed by mixed packaging waste.

    Science.gov (United States)

    Mastellone, Maria Laura; Cremiato, Raffaele; Zaccariello, Lucio; Lotito, Roberta

    2017-06-01

    Most of the integrated systems for municipal solid waste management aim to increase the recycling of secondary materials by means of physical processes including sorting, shredding and reprocessing. Several restrictions prevent from reaching a very high material recycling efficiency: the variability of the composition of new-marketed materials used for packaging production and its shape and complexity are critical issues. The packaging goods are in fact made of different materials (aluminium, polymers, paper, etc.), possibly assembled, having different shape (flat, cylindrical, one-dimensional, etc.), density, colours, optical properties and so on. These aspects limit the effectiveness and efficiency of the sorting and reprocessing plants. The scope of this study was to evaluate the performance of a large scale Material Recovery Facility (MRF) by utilizing data collected during a long period of monitoring. The database resulted from the measured data has been organized in four sections: (1) data related to the amount and type of inlet waste; (2) amount and composition of output products and waste; (3) operating data (such as worked hours for shift, planned and unscheduled maintenance time, setting parameters of the equipment, and energy consumption for shift); (4) economic data (value of each product, disposal price for the produced waste, penalty for non-compliance of products and waste, etc.). A part of this database has been utilized to build an executive dashboard composed by a set of performance indicators suitable to measure the effectiveness and the efficiency of the MRF operations. The dashboard revealed itself as a powerful tool to support managers and engineers in their decisions in respect to the market demand or compliance regulation variation as well as in the designing of the lay-out improvements. The results indicated that the 40% of the input waste was recovered as valuable products and that a large part of these (88%) complied with the standards of

  8. Fabrication of uranium-americium mixed oxide fuels: thermodynamical modeling and materials properties

    International Nuclear Information System (INIS)

    Prieur, D.

    2011-01-01

    Fuel irradiation in pressurized water reactors lead to the formation of fission products and minor actinides (Np, Am, Cm) which can be transmuted in fast neutrons reactors. In this context, the aim of this work was to study the fabrication conditions of the U 1-y Am y O 2+x fuels which exhibit particular thermodynamical properties requiring an accurate monitoring of the oxygen potential during the sintering step. For this reason, a thermodynamical model was developed to assess the optimum sintering conditions for these materials. From these calculations, U 1-y Am y O 2+x (y=0.10; 0.15; 0.20; 0.30) were sintered in two range of atmosphere. In hyper-stoichiometric conditions at low temperature, porous and multiphasic compounds are obtained whereas in reducing conditions at high temperature materials are dense and monophasic. XAFS analyses were performed in order to obtain additional experimental data for the thermodynamical modeling refinement. These characterizations also showed the reduction of Am(+IV) to Am(+III) and the partial oxidation of U(+IV) to U(+V) due to a charge compensation mechanism occurring during the sintering. Finally, taking into account the high - activity of Am, self-irradiation effects were studied for two types of microstructures and two Am contents (10 and 15%). For each composition, a lattice parameter increase was observed without structural change coupled with a macroscopic swelling of the pellet diameter up to 1.2% for the dense compounds and 0.6% for the tailored porosity materials. (author) [fr

  9. 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

  10. Mixed Reload Design Using MOX and UOX Fuel Assemblies

    International Nuclear Information System (INIS)

    Ramon, Ramirez Sanchez J.; Perry, R.T.

    2002-01-01

    As part of the studies involved in plutonium utilization assessment for a Boiling Water Reactor, a conceptual design of MOX fuel was developed, this design is mechanically the same design of 10 X 10 BWR fuel assemblies but different fissile material. Several plutonium and gadolinium concentrations were tested to match the 18 months cycle length which is the current cycle length of LVNPP, a reference UO 2 assembly was modeled to have a full cycle length to compare results, an effective value of 0.97 for the multiplication factor was set as target for 470 Effective Full Power days for both cycles, here the gadolinium concentration was a key to find an average fissile plutonium content of 6.55% in the assembly. A reload of 124 fuel assemblies was assumed to simulate the complete core, several load fractions of MOX fuel mixed with UO 2 fresh fuel were tested to verify the shutdown margin, the UO 2 fuel meets the shutdown margin when 124 fuel assemblies are loaded into the core, but it does not happen when those 124 assemblies are replaced with MOX fuel assemblies, so the fraction of MOX was reduced step by step up to find a mixed load that meets both length cycle and shutdown margin. Finally the conclusion is that control rods losses some of their worth in presence of plutonium due to a more hardened neutron spectrum in MOX fuel and this fact limits the load of MOX fuel assemblies in the core, this results are shown in this paper. (authors)

  11. BRICKS WITH TOTAL REPLACEMENT OF CLAY BY FLY ASH MIXED WITH DIFFERENT MATERIALS

    OpenAIRE

    J.N Akhtar; J.Alam; M.N Akhtar

    2011-01-01

    Fly ash is a powdery substance obtained from the dust collectors in the Thermal power plants that use coal as fuel. From the cement point of view the mineralogy of Fly ash is important as it contains 80% - 90% of glass. The impurities in coal-mostly clays, shale’s, limestone & dolomite; they cannot be burned so they turn up as ash. The Fly ash of class C category was used as a raw material to total replacement of clay for making Fly ash bricks. In present study the effect of Fly ash with high...

  12. 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

  13. Polyimide-polyether mixed conductors as switchable materials for electrochromic devices

    Energy Technology Data Exchange (ETDEWEB)

    Michot, C; Baril, D; Armand, M [Laboratoire d` Ionique et d` Electrochimie du Solide, ENSEEG, Institut National Polytechnique de Grenoble, Saint-Martin d` Heres (France)

    1995-12-01

    A new family of redox-active polymers have been obtained by polycondensation of {alpha},{omega}-diamino oligopolyethers with various aromatic tetracarboxylic acid anhydrides. The polyether blocks retain the usual cation co-ordination ability leading to solid-state ionic conduction while the relatively large electron affinity of the bis-imide moieties formed in the block polymers allows the reversible formation of stable radical anions ({sup -} and {sup 2-}) in the range 2-2.6 V vs. Li. Either slow-scan or microelectrode voltammetry indicate that the totality of the redox centres in such materials are readily accessible for all polyether spacer lengths tested (6-22 ether fragments), suggesting both a fast electron-exchange mechanism between anion radicals and a tendency for self assembly (stacking) of the planar aromatic groups. When increasing the number of atoms in the aromatic ring, the peak separation became smaller (benzene>naphthalene>perylene), resulting ultimately in simultaneous 2-electrons injection. These materials, due to the high coloration efficiency of the reduced state and their ready processability in thin films are excellent candidate for the realisation of laminated solid-state electrochromic devices

  14. Thermo-mechanical properties of mixed ion-electron conducting membrane materials

    Energy Technology Data Exchange (ETDEWEB)

    Huang, Bingxin

    2011-07-01

    The thesis presents thermo-mechanical properties of La{sub 0.58}Sr{sub 0.4}Co{sub 0.2}Fe{sub 0.8}O{sub 3-{delta}} (LSCF) and Ba{sub 0.5}Sr{sub 0.5}Co{sub 0.8}Fe{sub 0.2}O{sub 3-{delta}} (BSCF) perovskite materials, which are considered as oxygen transport membranes (OTM) for gas separation units. Ring-on-ring bending test with disk-shaped samples and depth-sensitive micro-indentation have been used as macroscopic and microscopic tests, respectively. In addition, the thermo-mechanical properties of a third OTM candidate material La{sub 2}NiO{sub 4+{delta}} (LNO) were investigated. The results of the thermo-mechanical measurements with the BSCF revealed an anomaly between 200 C and 400 C. In particular, the temperature dependence of Young's modulus shows a minimum at {proportional_to} 200 C. Fracture stress and toughness exhibit a qualitatively similar behavior with a minimum between 200 C and 400 C, before recovering between 500 C and 800 C. X-ray diffraction analyses verified that BSCF remains cubic in the relevant temperature range. Hence the anomalies were assumed to be related to the transition of Co{sup 3+} spin states reported for other Co-containing perovskites. This assumption could be experimentally confirmed by magnetic susceptibility measurements. The fracture surfaces of the specimens are not affected by the mechanical anomalies at intermediate temperatures, since only a transgranular fracture mode has been observed. Complementary to the mechanical characterization of BSCF, also the temperature dependency of fracture stress and elastic behavior of LSCF have been determined. Phase compositions of LSCF have been studied by in-situ high temperature XRD. Changes in phase composition with temperature are observed. At ambient temperature the LSCF perovskite material comprises two phases: rhombohedral and cubic symmetry. The ratio of the two phases depends on both cooling rate and atmosphere. The transition of rhombohedral to cubic occurs between 700 C and

  15. 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)

  16. 256-pixel microcalorimeter array for high-resolution γ-ray spectroscopy of mixed-actinide materials

    Energy Technology Data Exchange (ETDEWEB)

    Winkler, R., E-mail: rwinkler@lanl.gov [Los Alamos National Laboratory, Los Alamos, NM (United States); Hoover, A.S.; Rabin, M.W. [Los Alamos National Laboratory, Los Alamos, NM (United States); Bennett, D.A.; Doriese, W.B.; Fowler, J.W.; Hays-Wehle, J.; Horansky, R.D.; Reintsema, C.D.; Schmidt, D.R.; Vale, L.R.; Ullom, J.N. [National Institute of Standards and Technology, Boulder, CO (United States)

    2015-01-11

    The application of cryogenic microcalorimeter detectors to γ-ray spectroscopy allows for measurements with unprecedented energy resolution. These detectors are ideally suited for γ-ray spectroscopy applications for which the measurement quality is limited by the spectral overlap of many closely spaced transitions using conventional detector technologies. The non-destructive analysis of mixed-isotope Pu materials is one such application where the precision can be potentially improved utilizing microcalorimeter detectors compared to current state-of-the-art high-purity Ge detectors (HPGe). The LANL-NIST γ-ray spectrometer, a 256-pixel microcalorimeter array based on transition-edge sensors (TESs), was recently commissioned and used to collect data on a variety of Pu isotopic standards to characterize the instrument performance. These measurements represent the first time the simultaneous readout of all 256 pixels for measurements of mixed-isotope Pu materials has been achieved. The LANL-NIST γ-ray spectrometer has demonstrated an average pixel resolution of 55 eV full-width-at-half-maximum at 100 keV, nearly an order of magnitude better than HPGe detectors. Some challenges of the analysis of many-channel ultra-high resolution data and the techniques used to produce quality spectra for isotopic analysis will be presented. The LANL-NIST γ-ray spectrometer has also demonstrated stable operation and obtained high resolution measurements at total array event rates beyond 1 kHz. For a total event rate of 1.25 kHz, approximately 5.6 cps/pixel, a 72.2 eV average FWHM for the 103 keV photopeak of {sup 153}Gd was achieved.

  17. 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)

  18. Process for recycling mixed-cathode materials from spent lithium-ion batteries and kinetics of leaching.

    Science.gov (United States)

    Li, Li; Bian, Yifan; Zhang, Xiaoxiao; Guan, Yibiao; Fan, Ersha; Wu, Feng; Chen, Renjie

    2018-01-01

    A "grave-to-cradle" process for the recycling of spent mixed-cathode materials (LiCoO 2 , LiCo 1/3 Ni 1/3 Mn 1/3 O 2 , and LiMn 2 O 4 ) has been proposed. The process comprises an acid leaching followed by the resynthesis of a cathode material from the resulting leachate. Spent cathode materials were leached in citric acid (C 6 H 8 O 7 ) and hydrogen peroxide (H 2 O 2 ). Optimal leaching conditions were obtained at a leaching temperature of 90 °C, a H 2 O 2 concentration of 1.5 vol%, a leaching time of 60 min, a pulp density of 20 g L -1 , and a citric acid concentration of 0.5 M. The leaching efficiencies of Li, Co, Ni, and Mn exceeded 95%. The leachate was used to resynthesize new LiCo 1/3 Ni 1/3 Mn 1/3 O 2 material by using a sol-gel method. A comparison of the electrochemical properties of the resynthesized material (NCM-spent) with that synthesized directly from original chemicals (NCM-syn) indicated that the initial discharge capacity of NCM-spent at 0.2 C was 152.8 mA h g -1 , which was higher than the 149.8 mA h g -1 of NCM-syn. After 160 cycles, the discharge capacities of the NCM-spent and NCM-syn were 140.7 mA h g -1 and 121.2 mA h g -1 , respectively. After discharge at 1 C for 300 cycles, the NCM-spent material remained a higher capacity of 113.2 mA h g -1 than the NCM-syn (78.4 mA h g -1 ). The better performance of the NCM-spent resulted from trace Al doping. A new formulation based on the shrinking-core model was proposed to explain the kinetics of the leaching process. The activation energies of the Li, Co, Ni, and Mn leaching were calculated to be 66.86, 86.57, 49.46, and 45.23 kJ mol -1 , respectively, which indicates that the leaching was a chemical reaction-controlled process. Copyright © 2017 Elsevier Ltd. All rights reserved.

  19. The macro- and micro properties of cement pastes with silica-rich materials cured by wet-mixed steaming injection

    International Nuclear Information System (INIS)

    Wu, D.S.; Peng, Y.N.

    2003-01-01

    This research used cement pastes with a low water/blaine ratio (W/b=0.27). Rice husk ashes (RHA) burned at 700 and 850 deg. C, silica fume, silica sand (Ottawa standard sand), etc., were the added ingredients. Wet-mixed steam injection (WMSI) was at five different temperatures: 65, 80, 120, 150 and 180 deg. C. We investigated cement pastes with added silica-rich materials. For different WMSI temperatures and times, we explored the relations between compressive strength, hydration products, and pozzolanic reaction mechanism. From scanning electron microscopy (SEM) and EDS, we know that hydration products become very complicated, depending on the WMSI temperatures and times. It is difficult to determine the direct effects on the strength based on changes in the products. Experimental results, however, clearly showed that the compressive strength was worst for 80 deg. C and best for 180 deg. C. High-temperature WMSI is best with 4-h presteaming period and 8-h retention time. Curing in saturated limewater for 28 days did not increase the strength. The three types of silica-rich materials used in this research all participated in the reaction during high-temperature WMSI; they helped to increase the strength. Addition of Ottawa standard sand resulted in the best strength, followed by addition of RHA, while addition of silica fume was worse than the others. Specimens treated with high-temperature WMSI would swell slightly if they were placed in air. This was different from normal-temperature curing

  20. 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

  1. 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

  2. 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

  3. Excess Foundry Sand Characterization and Experimental Investigation in Controlled Low-Strength Material and Hot-Mixing Asphalt

    Energy Technology Data Exchange (ETDEWEB)

    Pauul J. Tikalsky

    2004-10-31

    This report provides technical data regarding the reuse of excess foundry sand. The report addresses three topics: (1) a statistically sound evaluation of the characterization of foundry sand, (2) a laboratory investigation to qualify excess foundry sand as a major component in controlled low-strength material (CLSM), and (3) the identification of the best methods for using foundry sand as a replacement for natural aggregates for construction purposes, specifically in asphalt paving materials. The survival analysis statistical technique was used to characterize foundry sand over a full spectrum of general chemical parameters, metallic elements, and organic compounds regarding bulk analysis and leachate characterization. Not limited to characterization and environmental impact, foundry sand was evaluated by factor analyses, which contributes to proper selection of factor and maximization of the reuse marketplace for foundry sand. Regarding the integration of foundry sand into CLSM, excavatable CLSM and structural CLSM containing different types of excess foundry sands were investigated through laboratory experiments. Foundry sand was approved to constitute a major component in CLSM. Regarding the integration of foundry sand into asphalt paving materials, the optimum asphalt content was determined for each mixture, as well as the bulk density, maximum density, asphalt absorption, and air voids at N{sub ini}, N{sub des}, and N{sub max}. It was found that foundry sands can be used as an aggregate in hot-mix asphalt production, but each sand should be evaluated individually. Foundry sands tend to lower the strength of mixtures and also may make them more susceptible to moisture damage. Finally, traditional anti-stripping additives may decrease the moisture sensitivity of a mixture containing foundry sand, but not to the level allowed by most highway agencies.

  4. Excess Foundry Sand Characterization and Experimental Investigation in Controlled Low-Strength Material and Hot-Mixing Asphalt

    Energy Technology Data Exchange (ETDEWEB)

    Tikalsky, Paul J. [Pennsylvania State Univ., University Park, PA (United States); Bahia, Hussain U. [Univ. of Wisconsin, Madison, WI (United States); Deng, An [Pennsylvania State Univ., University Park, PA (United States); Snyder, Thomas [Univ. of Wisconsin, Madison, WI (United States)

    2004-10-15

    This report provides technical data regarding the reuse of excess foundry sand. The report addresses three topics: a statistically sound evaluation of the characterization of foundry sand, a laboratory investigation to qualify excess foundry sand as a major component in controlled low-strength material (CLSM), and the identification of the best methods for using foundry sand as a replacement for natural aggregates for construction purposes, specifically in asphalt paving materials. The survival analysis statistical technique was used to characterize foundry sand over a full spectrum of general chemical parameters, metallic elements, and organic compounds regarding bulk analysis and leachate characterization. Not limited to characterization and environmental impact, foundry sand was evaluated by factor analyses, which contributes to proper selection of factor and maximization of the reuse marketplace for foundry sand. Regarding the integration of foundry sand into CLSM, excavatable CLSM and structural CLSM containing different types of excess foundry sands were investigated through laboratory experiments. Foundry sand was approved to constitute a major component in CLSM. Regarding the integration of foundry sand into asphalt paving materials, the optimum asphalt content was determined for each mixture, as well as the bulk density, maximum density, asphalt absorption, and air voids at Nini, Ndes, and Nmax. It was found that foundry sands can be used as an aggregate in hot-mix asphalt production, but each sand should be evaluated individually. Foundry sands tend to lower the strength of mixtures and also may make them more susceptible to moisture damage. Finally, traditional anti-stripping additives may decrease the moisture sensitivity of a mixture containing foundry sand, but not to the level allowed by most highway agencies.

  5. Self-immobilization of acidogenic mixed consortia on mesoporous material (SBA-15) and activated carbon to enhance fermentative hydrogen production

    Energy Technology Data Exchange (ETDEWEB)

    Mohan, S. Venkata; Mohanakrishna, G.; Sarma, P.N. [Bioengineering and Environmental Centre, Indian Institute of Chemical Technology, Hyderabad 500607 (India); Reddy, S. Sreevardhan; Raju, B. David; Rao, K.S. Rama [Catalysis and Physical Chemistry Division, Indian Institute of Chemical Technology, Hyderabad 500607 (India)

    2008-11-15

    The influence of self-immobilization of enriched acidogenic mixed consortia on fermentative hydrogen (H{sub 2}) production was studied on different supporting materials [SBA-15 (mesoporous) and activated carbon (granular; GAC and powder; PAC)] using chemical wastewater as substrate. Batch fermentation experiments were performed with same substrate at different organic loading rates (OLRs) under acidophilic microenvironment (pH 5.5) and room temperature (28 {+-} 2 C). Experimental data evidenced the effectiveness of attached growth on both the H{sub 2} yields and substrate degradation efficiency, particularly at higher loading rates. Among the three materials evaluated, immobilization on SBA-15 material showed comparatively effective performance in enhancing both H{sub 2} yield and substrate degradation. Suspended growth (SG-control) culture showed inhibition in terms of both H{sub 2} production and substrate degradation especially at applied higher loading rates. Immobilization on SBA-15 resulted in nine times higher H{sub 2} production (7.29 mol/kg COD{sub R}-day at OLR of 0.83 kg COD/m{sup 3}-day) than the lowest yield observed (suspended growth at OLR of 2.55 kg COD/m{sup 3}-day). Maximum substrate degradation rate (SDR) of 0.96 kg COD/m{sup 3}-day (OLR 2.55 kg COD/m{sup 3}-day) was also observed with SBA-15 immobilization, which is 1.62 times higher than the lowest substrate degradation observed with SG-control experiments with the same OLR. Attached growth on GAC and PAC also showed remarkable improvement in the process performance at higher OLRs compared to SG-control. (author)

  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. A comparative study of recycled aggregates from concrete and mixed debris as material for unbound road sub-base

    Directory of Open Access Journals (Sweden)

    Jiménez, J. R.

    2011-06-01

    Full Text Available Seven different types of recycled aggregates from construction and demolition waste (CDW have been evaluated as granular materials for unbound road sub-bases construction. The results showed that recycled concrete aggregates complied with all specifications for using in the construction of unbound structural layers (sub-base for T3 and T4 traffic categories according to the Spanish General Technical Specification for Road Construction (PG-3. Some mixed recycled aggregates fell short of some specifications due to a high content of sulphur compounds and poor fragmentation resistance. Sieving off the fine fraction prior to crushing the mixed CDW reduce the total sulphur content and improve the quality of the mixed recycled aggregates, by contrast, pre-sieving concrete CDW had no effect on the quality of the resulting aggregates. The results were compared with a crushed limestone as natural aggregate.

    Siete áridos reciclados de residuos de construcción y demolición (RCD se han evaluado como zahorras para la construcción de sub-bases de carreteras. Los resultados muestran que los áridos reciclados de hormigón cumplen todas las especificaciones del Pliego de Prescripciones Técnicas Generales para Obras de Carreteras de España (PG-3 para su uso en capas estructurales (sub-base de las categorías de tráfico T3 y T4. Algunos áridos reciclados mixtos no cumplen por escaso margen algunas de las especificaciones, debido a un alto contenido de compuestos de azufre y a una menor resistencia a la fragmentación. El precribado de la fracción fina antes de la trituración de los RCD mixtos reduce el contenido de azufre total y mejora la calidad, por el contrario, el precribado de los RCD de hormigón no tiene ningún efecto sobre la calidad de los áridos reciclados. Los resultados se compararon con una zahorra artificial caliza como árido natural.

  8. Materials for nuclear reactors

    International Nuclear Information System (INIS)

    Banerjee, S.; Kamath, H.S.

    2005-01-01

    The improved performance of present generation nuclear reactors and the realization of advanced reactor concepts, both, require development of better materials. Physical metallurgy/materials science principles which have been exploited in meeting the exacting requirements of nuclear reactor materials (fuels and structural materials), are outlined citing a few specific examples. While the incentive for improvement of traditional fuels (e.g., UO 2 fuel) is primarily for increasing the average core burn up, the development of advanced fuels (e.g., MOX, mixed carbide, nitride, silicide and dispersion fuels) are directed towards better utilization of fissile and fertile inventories through adaptation of innovative fuel cycles. As the burn up of UO 2 fuel reaches higher levels, a more detailed and quantitative understanding of the phenomena such as fission gas release, fuel restructuring induced by radiation and thermal gradients and pellet-clad interaction is being achieved. Development of zirconium based alloys for both cladding and pressure tube applications is discussed with reference to their physical metallurgy, fabrication techniques and in-reactor degradation mechanisms. The issue of radiation embrittlement of reactor pressure vessels (RPVs) is covered drawing a comparison between the western and eastern specifications of RPV steels. The search for new materials which can stand higher rates of atomic displacement due to radiation has led to the development of swelling resistant austenitic and ferritic stainless steels for fast reactor applications as exemplified by the development of the D-9 steel for Indian fast breeder reactor. The presentation will conclude by listing various materials related phenomena, which have a strong bearing on the successful development of future nuclear energy systems. (author)

  9. Multi and mixed 3D-printing of graphene-hydroxyapatite hybrid materials for complex tissue engineering.

    Science.gov (United States)

    Jakus, Adam E; Shah, Ramille N

    2017-01-01

    With the emergence of three-dimensional (3D)-printing (3DP) as a vital tool in tissue engineering and medicine, there is an ever growing need to develop new biomaterials that can be 3D-printed and also emulate the compositional, structural, and functional complexities of human tissues and organs. In this work, we probe the 3D-printable biomaterials spectrum by combining two recently established functional 3D-printable particle-laden biomaterial inks: one that contains hydroxyapatite microspheres (hyperelastic bone, HB) and another that contains graphene nanoflakes (3D-graphene, 3DG). We demonstrate that not only can these distinct, osteogenic, and neurogenic inks be co-3D-printed to create complex, multimaterial constructs, but that composite inks of HB and 3DG can also be synthesized. Specifically, the printability, microstructural, mechanical, electrical, and biological properties of a hybrid material comprised of 1:1 HA:graphene by volume is investigated. The resulting HB-3DG hybrid exhibits mixed characteristics of the two distinct systems, while maintaining 3D-printability, electrical conductivity, and flexibility. In vitro assessment of HB-3DG using mesenchymal stem cells demonstrates the hybrid material supports cell viability and proliferation, as well as significantly upregulates both osteogenic and neurogenic gene expression over 14 days. This work ultimately demonstrates a significant step forward towards being able to 3D-print graded, multicompositional, and multifunctional constructs from hybrid inks for complex composite tissue engineering. © 2016 Wiley Periodicals, Inc. J Biomed Mater Res Part A: 105A: 274-283, 2017. © 2016 Wiley Periodicals, Inc.

  10. 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

  11. 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

  12. 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.

  13. 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

  14. 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

  15. 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

  16. 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

  17. 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)

  18. 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

  19. 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.

  20. 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

  1. 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

  2. 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

  3. 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

  4. Overview of material migration and mixing, fuel retention and cleaning of ITER-like castellated structures in TEXTOR

    International Nuclear Information System (INIS)

    Litnovsky, A.; Philipps, V.; Wienhold, P.; Kreter, A.; Kirschner, A.; Matveev, D.; Brezinsek, S.; Sergienko, G.; Pospieszczyk, A.; Schweer, B.; Schulz, C.; Schmitz, O.; Coenen, J.W.; Samm, U.; Krieger, K.; Hirai, T.; Emmoth, B.; Rubel, M.; Bazylev, B.; Breuer, U.

    2011-01-01

    Plasma-facing components (PFCs) in ITER will be castellated by splitting them into small-size blocks to maintain the thermo-mechanical stability. However, there are concerns in particular on retention of codeposited radioactive fuel in the gaps. An R and D program is underway in TEXTOR addressing this acute issue of castellation. Material migration and fuel inventory are investigated using long- and short-term discharge-resolved experiments with castellated structures in TEXTOR. Significant impurity transport to the gaps was detected and results were in part quantitatively reproduced with 3D-GAPS code. Deposits containing up to 70 at.% of tungsten on the gap areas closest to the plasma were detected in recent experiments. Deposition in the gaps accompanied by metal mixing demand for development of effective cleaning techniques. In experiments with ITER-like castellation, the gaps were cleaned from carbonaceous deposits using oxygen plasmas at 350 deg. C. This contribution contains an overview of experimental and modeling results along with recommendations for PFCs in ITER.

  5. Technical specification: Mixed-oxide pellets for the light-water reactor irradiation demonstration test

    International Nuclear Information System (INIS)

    Cowell, B.S.

    1997-06-01

    This technical specification is a Level 2 Document as defined in the Fissile Materials Disposition Program Light-Water Reactor Mixed-oxide Fuel Irradiation Test Project Plan. It is patterned after the pellet specification that was prepared by Atomic Energy of Canada, Limited, for use by Los Alamos National Laboratory in fabrication of the test fuel for the Parallex Project, adjusted as necessary to reflect the differences between the Canadian uranium-deuterium reactor and light-water reactor fuels. This specification and the associated engineering drawing are to be utilized only for preparation of test fuel as outlined in the accompanying Request for Quotation and for additional testing as directed by Oak Ridge National Laboratory or the Department of Energy

  6. Mixed Uranium/Refractory Metal Carbide Fuels for High Performance Nuclear Reactors

    International Nuclear Information System (INIS)

    Knight, Travis; Anghaie, Samim

    2002-01-01

    Single phase, solid-solution mixed uranium/refractory metal carbides have been proposed as an advanced nuclear fuel for advanced, high-performance reactors. Earlier studies of mixed carbides focused on uranium and either thorium or plutonium as a fuel for fast breeder reactors enabling shorter doubling owing to the greater fissile atom density. However, the mixed uranium/refractory carbides such as (U, Zr, Nb)C have a lower uranium densities but hold significant promise because of their ultra-high melting points (typically greater than 3700 K), improved material compatibility, and high thermal conductivity approaching that of the metal. Various compositions of (U, Zr, Nb)C were processed with 5% and 10% metal mole fraction of uranium. Stoichiometric samples were processed from the constituent carbide powders, while hypo-stoichiometric samples with carbon-to-metal (C/M) ratios of 0.92 were processed from uranium hydride, graphite, and constituent refractory carbide powders. Processing techniques of cold uniaxial pressing, dynamic magnetic compaction, sintering, and hot pressing were investigated to optimize the processing parameters necessary to produce high density (low porosity), single phase, solid-solution mixed carbide nuclear fuels for testing. This investigation was undertaken to evaluate and characterize the performance of these mixed uranium/refractory metal carbides for high performance, ultra-safe nuclear reactor applications. (authors)

  7. 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

  8. Angular-resolution and material-characterization measurements for a dual-particle imaging system with mixed-oxide fuel

    Energy Technology Data Exchange (ETDEWEB)

    Poitrasson-Rivière, Alexis, E-mail: alexispr@umich.edu [Department of Nuclear Engineering & Radiological Sciences, University of Michigan, Ann Arbor, MI 48109 (United States); Polack, J. Kyle; Hamel, Michael C.; Klemm, Dietrich D.; Ito, Kai; McSpaden, Alexander T.; Flaska, Marek; Clarke, Shaun D.; Pozzi, Sara A. [Department of Nuclear Engineering & Radiological Sciences, University of Michigan, Ann Arbor, MI 48109 (United States); Tomanin, Alice [Lainsa-Italia S.R.L., Via E. Fermi 2749, 21027 Ispra, VA (Italy); Peerani, Paolo [European Commission, Joint Research Centre, Institute for Transuranium Elements, 21027 Ispra, VA (Italy)

    2015-10-11

    A dual-particle imaging (DPI) system, capable of simultaneously imaging fast neutrons and gamma rays, has been operated in the presence of mixed-oxide (MOX) fuel to assess the system's angular resolution and material-characterization capabilities. The detection principle is based on the scattering physics of neutrons (elastic scattering) and gamma rays (Compton scattering) in organic and inorganic scintillators. The detection system is designed as a combination of a two-plane Compton camera and a neutron-scatter camera. The front plane consists of EJ-309 liquid scintillators and the back plane consists of interleaved EJ-309 and NaI(Tl) scintillators. MCNPX-PoliMi was used to optimize the geometry of the system and the resulting prototype was built and tested using a Cf-252 source as an SNM surrogate. A software package was developed to acquire and process data in real time. The software was used for a measurement campaign to assess the angular resolution of the imaging system with MOX samples. Measurements of two MOX canisters of similar isotopics and intensity were performed for 6 different canister separations (from 5° to 30°, corresponding to distances of 21 cm and 131 cm, respectively). The measurements yielded a minimum separation of 20° at 2.5 m (86-cm separation) required to see 2 separate hot spots. Additionally, the results displayed good agreement with MCNPX-PoliMi simulations. These results indicate an angular resolution between 15° and 20°, given the 5° step. Coupled with its large field of view, and its capability to differentiate between spontaneous fission and (α,n) sources, the DPI system shows its potential for nuclear-nonproliferation applications.

  9. Effects of strain rate, mixing ratio, and stress-strain definition on the mechanical behavior of the polydimethylsiloxane (PDMS) material as related to its biological applications.

    Science.gov (United States)

    Khanafer, Khalil; Duprey, Ambroise; Schlicht, Marty; Berguer, Ramon

    2009-04-01

    Tensile tests on Polydimethylsiloxane (PDMS) materials were conducted to illustrate the effects of mixing ratio, definition of the stress-strain curve, and the strain rate on the elastic modulus and stress-strain curve. PDMS specimens were prepared according to the ASTM standards for elastic materials. Our results indicate that the physiological elastic modulus depends strongly on the definition of the stress-strain curve, mixing ratio, and the strain rate. For various mixing ratios and strain rates, true stress-strain definition results in higher stress and elastic modulus compared with engineering stress-strain and true stress-engineering strain definitions. The elastic modulus increases as the mixing ratio increases up-to 9:1 ratio after which the elastic modulus begins to decrease even as the mixing ratio continues to increase. The results presented in this study will be helpful to assist the design of in vitro experiments to mimic blood flow in arteries and to understand the complex interaction between blood flow and the walls of arteries using PDMS elastomer.

  10. Prompt neutron fission spectrum mean energies for the fissile nuclides and 252Cf

    International Nuclear Information System (INIS)

    Holden, N.E.

    1985-01-01

    The international standard for a neutron spectrum is that produced from the spontaneous fission of 252 Cf, while the thermal neutron induced fission neutron spectra for the four fissile nuclides, 233 U, 235 U, 239 Pu, and 241 Pu are of interest from the standpoint of nuclear reactors. The average neutron energies of these spectra are tabulated. The individual measurements are recorded with the neutron energy range measured, the method of detection as well as the average neutron energy for each author. Also tabulated are the measurements of the ratio of mean energies for pairs of fission neutron spectra. 75 refs., 9 tabs

  11. Development of a new simulation code for evaluation of criticality transients involving fissile solution boiling

    International Nuclear Information System (INIS)

    Basoglu, Benan; Yamamoto, Toshihiro; Okuno, Hiroshi; Nomura, Yasushi

    1998-03-01

    In this work, we report on the development of a new computer code named TRACE for predicting the excursion characteristics of criticality excursions involving fissile solutions. TRACE employs point neutronics coupled with simple thermal-hydraulics. The temperature, the radiolytic gas effects, and the boiling phenomena are estimated using the transient heat conduction equation, a lumped-parameter energy model, and a simple boiling model, respectively. To evaluate the model, we compared our results with the results of CRAC experiments. The agreement in these comparisons is quite satisfactory. (author)

  12. FUP1--an unified program for calculating all fast neutron data of fissile nucleus

    International Nuclear Information System (INIS)

    Cai Chonghai; Zuo Yixin

    1990-01-01

    FUP1 is the first edition of an unified program for calculating all the fast neutron data in ENDF/B-4 format for fissile nucleus. Following data are calculated with FUP1 code: the total cross section, elastic scattering cross section, nonelastic cross section, total including up to 40 isolated levels and continuum state inelastic cross sections. In FUP1 the energy region of incident neutron is restricted to 10 Kev to 20 Mev. The advantages of this program are its perfect function, convenient to users and running very fast

  13. Nuclear dissipation effects on fission and evaporation in systems of intermediate fissility

    Directory of Open Access Journals (Sweden)

    Gelli N.

    2010-03-01

    Full Text Available The systems of intermediate fissility 132Ce and 158Er have been studied experimentally and theoretically in order to investigate the dissipation properties of nuclear matter. Cross sections of fusion-fission and evaporation residues channels together with charged particles multiplicities in both channels, their spectra, angular correlations and mass-energy distribution of fission fragments have been measured. Theoretical analysis has been performed using multi-dimensional stochastic approach with realistic treatment of particle evaporation. The results of analysis show that full one-body or unusually strong two-body dissipation allows to reproduce experimental data. No temperature dependent dissipation was needed.

  14. International safeguards of fissionable material

    International Nuclear Information System (INIS)

    Tempus, P.

    1991-01-01

    From the very beginning nuclear fissile materials have been subject to state and - outside nuclear weapon states - also to international monitoring. The latter was a principal task of the International Atomic Energy Agency, a UN affiliated organisation formed in 1957 based in Vienna. The legal, technical and political aspects of its monitoring activity are explained

  15. 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

  16. Mixed bi-material electrodes based on LiMn2O4 and activated carbon for hybrid electrochemical energy storage devices

    International Nuclear Information System (INIS)

    Cericola, Dario; Novak, Petr; Wokaun, Alexander; Koetz, Ruediger

    2011-01-01

    Highlights: → Bi-material electrodes for electrochemical hybrid devices were characterized. → Bi-material electrodes have higher specific charge than capacitor electrodes. → Bi-material electrodes have better rate capability than battery electrodes. → Bi-material systems outperform batteries and capacitors in pulsed applications. - Abstract: The performance of mixed bi-material electrodes composed of the battery material, LiMn 2 O 4 , and the electrochemical capacitor material, activated carbon, for hybrid electrochemical energy storage devices is investigated by galvanostatic charge/discharge and pulsed discharge experiments. Both, a high and a low conductivity lithium-containing electrolyte are used. The specific charge of the bi-material electrode is the linear combination of the specific charges of LiMn 2 O 4 and activated carbon according to the electrode composition at low discharge rates. Thus, the specific charge of the bi-material electrode falls between the specific charge of the activated carbon electrode and the LiMn 2 O 4 battery electrode. The bi-material electrodes have better rate capability than the LiMn 2 O 4 battery electrode. For high current pulsed applications the bi-material electrodes typically outperform both the battery and the capacitor electrode.

  17. Tank 241-AZ-101 criticality assessment resulting from pump jet mixing: Sludge mixing simulation

    Energy Technology Data Exchange (ETDEWEB)

    Onishi, Y.; Recknagle, K.

    1997-04-01

    Tank 241-AZ-101 (AZ-101) is one of 28 double-shell tanks located in the AZ farm in the Hanford Site`s 200 East Area. The tank contains a significant quantity of fissile materials, including an estimated 9.782 kg of plutonium. Before beginning jet pump mixing for mitigative purposes, the operations must be evaluated to demonstrate that they will be subcritical under both normal and credible abnormal conditions. The main objective of this study was to address a concern about whether two 300-hp pumps with four rotating 18.3-m/s (60-ft/s) jets can concentrate plutonium in their pump housings during mixer pump operation and cause a criticality. The three-dimensional simulation was performed with the time-varying TEMPEST code to determine how much the pump jet mixing of Tank AZ-101 will concentrate plutonium in the pump housing. The AZ-101 model predicted that the total amount of plutonium within the pump housing peaks at 75 g at 10 simulation seconds and decreases to less than 10 g at four minutes. The plutonium concentration in the entire pump housing peaks at 0.60 g/L at 10 simulation seconds and is reduced to below 0.1 g/L after four minutes. Since the minimum critical concentration of plutonium is 2.6 g/L, and the minimum critical plutonium mass under idealized plutonium-water conditions is 520 g, these predicted maximums in the pump housing are much lower than the minimum plutonium conditions needed to reach a criticality level. The initial plutonium maximum of 1.88 g/L still results in safety factor of 4.3 in the pump housing during the pump jet mixing operation.

  18. Tank 241-AZ-101 criticality assessment resulting from pump jet mixing: Sludge mixing simulation

    International Nuclear Information System (INIS)

    Onishi, Y.; Recknagle, K.

    1997-04-01

    Tank 241-AZ-101 (AZ-101) is one of 28 double-shell tanks located in the AZ farm in the Hanford Site's 200 East Area. The tank contains a significant quantity of fissile materials, including an estimated 9.782 kg of plutonium. Before beginning jet pump mixing for mitigative purposes, the operations must be evaluated to demonstrate that they will be subcritical under both normal and credible abnormal conditions. The main objective of this study was to address a concern about whether two 300-hp pumps with four rotating 18.3-m/s (60-ft/s) jets can concentrate plutonium in their pump housings during mixer pump operation and cause a criticality. The three-dimensional simulation was performed with the time-varying TEMPEST code to determine how much the pump jet mixing of Tank AZ-101 will concentrate plutonium in the pump housing. The AZ-101 model predicted that the total amount of plutonium within the pump housing peaks at 75 g at 10 simulation seconds and decreases to less than 10 g at four minutes. The plutonium concentration in the entire pump housing peaks at 0.60 g/L at 10 simulation seconds and is reduced to below 0.1 g/L after four minutes. Since the minimum critical concentration of plutonium is 2.6 g/L, and the minimum critical plutonium mass under idealized plutonium-water conditions is 520 g, these predicted maximums in the pump housing are much lower than the minimum plutonium conditions needed to reach a criticality level. The initial plutonium maximum of 1.88 g/L still results in safety factor of 4.3 in the pump housing during the pump jet mixing operation

  19. First Industrial Tests of a Drum Monitor Matrix Correction for the Fissile Mass Measurement in Large Volume Historic Metallic Residues with the Differential Die-away Technique

    Energy Technology Data Exchange (ETDEWEB)

    Antoni, R.; Passard, C.; Perot, B.; Batifol, M.; Vandamme, J.C. [CEA, DEN, Cadarache, Nuclear Measurement Laboratory, F-13108 St Paul-lez-Durance, (France); Grassi, G. [AREVA NC, 1 place Jean-Millier, 92084 Paris-La-Defense cedex (France)

    2015-07-01

    The fissile mass in radioactive waste drums filled with compacted metallic residues (spent fuel hulls and nozzles) produced at AREVA La Hague reprocessing plant is measured by neutron interrogation with the Differential Die-away measurement Technique (DDT. In the next years, old hulls and nozzles mixed with Ion-Exchange Resins will be measured. The ion-exchange resins increase neutron moderation in the matrix, compared to the waste measured in the current process. In this context, the Nuclear Measurement Laboratory (NML) of CEA Cadarache has studied a matrix effect correction method, based on a drum monitor ({sup 3}He proportional counter inside the measurement cavity). A previous study performed with the NML R and D measurement cell PROMETHEE 6 has shown the feasibility of method, and the capability of MCNP simulations to correctly reproduce experimental data and to assess the performances of the proposed correction. A next step of the study has focused on the performance assessment of the method on the industrial station using numerical simulation. A correlation between the prompt calibration coefficient of the {sup 239}Pu signal and the drum monitor signal was established using the MCNPX computer code and a fractional factorial experimental design composed of matrix parameters representative of the variation range of historical waste. Calculations have showed that the method allows the assay of the fissile mass with an uncertainty within a factor of 2, while the matrix effect without correction ranges on 2 decades. In this paper, we present and discuss the first experimental tests on the industrial ACC measurement system. A calculation vs. experiment benchmark has been achieved by performing dedicated calibration measurement with a representative drum and {sup 235}U samples. The preliminary comparison between calculation and experiment shows a satisfactory agreement for the drum monitor. The final objective of this work is to confirm the reliability of the

  20. First Industrial Tests of a Drum Monitor Matrix Correction for the Fissile Mass Measurement in Large Volume Historic Metallic Residues with the Differential Die-away Technique

    International Nuclear Information System (INIS)

    Antoni, R.; Passard, C.; Perot, B.; Batifol, M.; Vandamme, J.C.; Grassi, G.

    2015-01-01

    The fissile mass in radioactive waste drums filled with compacted metallic residues (spent fuel hulls and nozzles) produced at AREVA La Hague reprocessing plant is measured by neutron interrogation with the Differential Die-away measurement Technique (DDT. In the next years, old hulls and nozzles mixed with Ion-Exchange Resins will be measured. The ion-exchange resins increase neutron moderation in the matrix, compared to the waste measured in the current process. In this context, the Nuclear Measurement Laboratory (NML) of CEA Cadarache has studied a matrix effect correction method, based on a drum monitor ( 3 He proportional counter inside the measurement cavity). A previous study performed with the NML R and D measurement cell PROMETHEE 6 has shown the feasibility of method, and the capability of MCNP simulations to correctly reproduce experimental data and to assess the performances of the proposed correction. A next step of the study has focused on the performance assessment of the method on the industrial station using numerical simulation. A correlation between the prompt calibration coefficient of the 239 Pu signal and the drum monitor signal was established using the MCNPX computer code and a fractional factorial experimental design composed of matrix parameters representative of the variation range of historical waste. Calculations have showed that the method allows the assay of the fissile mass with an uncertainty within a factor of 2, while the matrix effect without correction ranges on 2 decades. In this paper, we present and discuss the first experimental tests on the industrial ACC measurement system. A calculation vs. experiment benchmark has been achieved by performing dedicated calibration measurement with a representative drum and 235 U samples. The preliminary comparison between calculation and experiment shows a satisfactory agreement for the drum monitor. The final objective of this work is to confirm the reliability of the modeling approach

  1. 76 FR 50331 - Hazardous Materials Regulations; Compatibility With the Regulations of the International Atomic...

    Science.gov (United States)

    2011-08-12

    ... geometry requirements applicable to tested fissile material packages. This TS-R-1 change is applicable to... percussion test.) The TS-R-1 revisions pertaining to the solar insolation conditions to be assumed in...

  2. Long-Term Effects on Graphene Supercapacitors of Using a Zirconia Bowl and Zirconia Balls for Ball-Mill mixing of Active Materials

    Science.gov (United States)

    Song, Dae-Hoon; Kim, Jin-Young; Kahng, Yung Ho; Cho, Hoonsung; Kim, Eung-Sam

    2018-04-01

    Improving the energy storage performance of supercapacitor electrodes based on reduced graphene oxide (RGO) is one of the main subjects in this research field. However, when a zirconia bowl and zirconia balls were used for ball-mill mixing of the active materials for RGO supercapacitors, the energy storage performance deteriorated over time. Our study revealed that the source of the problem was the inclusion of zirconia bits from abrasion of the bowl and the balls during the ballmill mixing, which increased during a period of 1 year. We probed two solutions to this problem: 1) hydrofluoric (HF) acid treatment of the RGO supercapacitors and 2) use of a tempered steel bowl and tempered steel balls for the mixing. For both cases, the energy storage performance was restored to near the initial level, showing a specific capacitance ( C sp ) of 200 F/g. Our results should lead to progress in research on RGO supercapacitors.

  3. Carbon nanotube-like materials in the exhaust from a diesel engine using gas oil/ethanol mixing fuel with catalysts and sulfur.

    Science.gov (United States)

    Suzuki, Shunsuke; Mori, Shinsuke

    2017-08-01

    Particulate matter from a diesel engine, including soot and carbon nanomaterials, was collected on a sampling holder and the structure of the materials was studied by transmission electron microscopy (TEM) and scanning electron microscopy (SEM). As a result of employing gas oil/ethanol mixing fuel with sulfur and ferrocene/molybdenum as catalyst sources, formation of carbon nanotubes (CNT)-like materials in addition to soot was observed in the exhaust gas from a diesel engine. It was revealed that CNT-like materials were included among soot in our system only when the following three conditions were satisfied simultaneously: high ethanol fraction in fuel, high sulfur loading, and presence of catalyst sources in fuel. This study confirmed that if at least one of these three conditions was not satisfied, CNT-like materials were not observed in the exhaust from a diesel engine. These experimental results shown in this work provide insights into understanding CNT-like material formation mechanism in a diesel engine. Recent papers reported that carbon nanotube-like materials were included in the exhaust gas from engines, but conditions for carbon nanotube-like material formation have not been well studied. This work provides the required conditions for carbon nanotube-like material growth in a diesel engine, and this will be helpful for understanding the carbon nanotube-like material formation mechanism and taking countermeasures to preventing carbon nanotube-like material formation in a diesel engine.

  4. Electronuclear fissile fuel production. Linear accelerator fuel regenerator and producer LAFR and LAFP

    International Nuclear Information System (INIS)

    Steinberg, M.; Powell, J.R.; Takahashi, H.; Grand, P.; Kouts, H.J.C.

    1978-04-01

    A linear accelerator fuel generator is proposed to enrich naturally occurring fertile U-238 or thorium 232 with fissile Pu-239 or U-233 for use in LWR power reactors. High energy proton beams in the range of 1 to 3 GeV energy are made to impinge on a centrally located dispersed liquid lead target producing spallation neutrons which are then absorbed by a surrounding assembly of fabricated LWR fuel elements. The accelerator-target design is reviewed and a typical fuel cycle system and economic analysis is presented. One 300 MW beam (300 ma-1 GeV) linear accelerator fuel regenerator can provide fuel for 3 to 1000 MW(e) LWR power reactors over its 30-year lifetime. There is a significant saving in natural uranium requirement which is a factor of 4.5 over the present LWR fuel requirement assuming the restraint of no fissile fuel recovery by reprocessing. A modest increase (approximately 10%) in fuel cycle and power production cost is incurred over the present LWR fuel cycle cost. The linear accelerator fuel regenerator and producer assures a long-term supply of fuel for the LWR power economy even with the restraint of the non-proliferation policy of no reprocessing. It can also supply hot-denatured thorium U-233 fuel operating in a secured reprocessing fuel center

  5. Influence of the Density Law on Various Fissile Single Unit and Array Storage Methods

    International Nuclear Information System (INIS)

    Huang, S.T.

    2011-01-01

    The advancement of computational technology has resulted in the wide-spread availability of powerful radiation transport Monte Carlo codes. Prevailing practices today rely heavily on Monte Carlo codes to provide the basis for assessing the reactivity of various fissile systems for nuclear criticality safety (NCS). In 1958, Weinberg and Wigner expressed their concerns on a 'deplorable trend in reactor design - the tendency to substitute a code for a theory'. Unfortunately, their concerns have largely become a reality in many modern NCS practices. lacking the time or information to understand the underlying neutron physics of the fissile system under consideration is indeed a deplorable trend. The purpose of this paper is to demonstrate that many features of criticality hand calculation methods are indeed based upon the fundamentals of the density law and that many correlations of important physics parameters can be more easily understood from such a perspective. Historically, the density law was recognized by many pioneers in the field, including during the Manhattan Project. However, it was by and large an 'oral tradition' in that bits and pieces of great physical insights of the pioneers were scattered in many earlier publications. This paper attempts to bring together some of the 'jewels' of the pioneers which might have been lost or forgotten.

  6. Comparison of thorium-based fuels with different fissile components in existing BWRs

    International Nuclear Information System (INIS)

    Bjoerk, Klara Insulander; Fhager, Valentin; Demaziere, Christophe

    2009-01-01

    Three different types of thorium based BWR fuel have been developed, in each of which thorium was combined with a different fissile component, the three components being reactor grade plutonium, uranium enriched to 20% in uranium 235 and pure uranium 233. A BWR nuclear bundle design, based on the geometrical fuel assembly design GE14, was developed for each of these fissile components. The properties and performance of the corresponding fuel assemblies were investigated via full core calculations carried out for an existing BWR and compared with the ones of an ordinary Low Enriched Uranium (LEU) fuel, which was developed for reference. The fuel assemblies and cores were designed to meet existing fuel design criteria, and were then analyzed with regards to reactivity coefficients, delayed neutron fractions, control rod worths and shutdown margins. The results show that all three alternatives seem to be feasible, although some difficulties remain with complying with the thermal limits, and with the moderator temperature and coolant void coefficients of the U-233 containing fuel being positive under some circumstances. (author)

  7. Some methods for the detection of fissionable matter; Quelques methodes de detection des corps fissiles

    Energy Technology Data Exchange (ETDEWEB)

    Guery, M [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1967-03-01

    A number of equipments or processes allowing to detect uranium or plutonium in industrial plants, and in particular to measure solution concentrations, are studied here. Each method has its own field of applications and has its own performances, which we have tried to define by calculations and by experiments. The following topics have been treated: {gamma} absorptiometer with an Am source, detection test by neutron multiplication, apparatus for the measurement of the {alpha} activity of a solution, fissionable matter detection by {gamma} emission, fissionable matter detection by neutron emission. (author) [French] On examine ici plusieurs appareils ou procedes qui permettent de detecter l'uranium ou le plutonium dans les installations industrielles, et en particulier de mesurer les concentrations de solutions. Chacune des methodes a son domaine d'application et ses performances, qu'on a tente de definir par le calcul et par des experiences. Les sujets traites sont les suivants: absorptiometre {gamma} a source d'americium, essais de detection par multiplication neutronique, appareil de mesure de l'activite {alpha} d'une solution, detection des matieres fissiles par leur emission {gamma}, detection des matieres fissiles par leur emission neutronique. (auteur)

  8. Strategy to Boost the Efficiency of Mixed-Ion Perovskite Solar Cells: Changing Geometry of the Hole Transporting Material.

    Science.gov (United States)

    Zhang, Jinbao; Xu, Bo; Johansson, Malin B; Vlachopoulos, Nick; Boschloo, Gerrit; Sun, Licheng; Johansson, Erik M J; Hagfeldt, Anders

    2016-07-26

    The hole transporting material (HTM) is an essential component in perovskite solar cells (PSCs) for efficient extraction and collection of the photoinduced charges. Triphenylamine- and carbazole-based derivatives have extensively been explored as alternative and economical HTMs for PSCs. However, the improvement of their power conversion efficiency (PCE), as well as further investigation of the relationship between the chemical structure of the HTMs and the photovoltaic performance, is imperatively needed. In this respect, a simple carbazole-based HTM X25 was designed on the basis of a reference HTM, triphenylamine-based X2, by simply linking two neighboring phenyl groups in a triphenylamine unit through a carbon-carbon single bond. It was found that a lowered highest occupied molecular orbital (HOMO) energy level was obtained for X25 compared to that of X2. Besides, the carbazole moiety in X25 improved the molecular planarity as well as conductivity property in comparison with the triphenylamine unit in X2. Utilizing the HTM X25 in a solar cell with mixed-ion perovskite [HC(NH2)2]0.85(CH3NH3)0.15Pb(I0.85Br0.15)3, a highest reported PCE of 17.4% at 1 sun (18.9% under 0.46 sun) for carbazole-based HTM in PSCs was achieved, in comparison of a PCE of 14.7% for triphenylamine-based HTM X2. From the steady-state photoluminescence and transient photocurrent/photovoltage measurements, we conclude that (1) the lowered HOMO level for X25 compared to X2 favored a higher open-circuit voltage (Voc) in PSCs; (2) a more uniform formation of X25 capping layer than X2 on the surface of perovskite resulted in more efficient hole transport and charge extraction in the devices. In addition, the long-term stability of PSCs with X25 is significantly enhanced compared to X2 due to its good uniformity of HTM layer and thus complete coverage on the perovskite. The results provide important information to further develop simple and efficient small molecular HTMs applied in solar cells.

  9. Assessment of radioactive material released from a fuel fabrication plant under accidental conditions

    International Nuclear Information System (INIS)

    1981-01-01

    This report evaluates the amounts of fissile material released both inside and outside a mixed oxide fuel fabrication plant (MOFFP) for light water reactors. The first section begins with a descriptive study of fissile material containment systems, and the methods available for quantifying accident occurrence probabilities. In addition to accidents common to all industrial facilities, other much rarer accidents were considered, such as aircraft crashes. The minimum occurrence probability limit for consideration in this study was set at 10 -6 per annum. The second part of this report attempts to assess the consequences of the accidents considered (i.e. with occurrence probabilities exceeding 10 -6 per annum) by determining maximum values for such accidents. Acts of sabotage and other accidents of this type are beyond the scope of this study and were not taken into consideration. The most serious potential accident would be a fire involving all of the glove boxes in the PuO 2 powder calcination and preparation cell, which could release 76.5 mg of PuO 2 powder into the atmosphere; the occurrence probability of such an accident, however, is slight (less than 10 -5 per annum). The second possibility, is a specially nuclear hazard that would release fission products into the atmosphere. The occurrence probability of such an accident is currently evaluated at 10 -3 per annum

  10. Physics concept on the constellation type fissile fuels and its application to the prospective Th-232U Reactor

    International Nuclear Information System (INIS)

    Zhang, Jiahua

    1994-01-01

    In contrast with the conventional nuclear reactor which usually fuelled with on single fissile nuclide, a constellation type fissile fuels reactor consists of a parent nuclide such as 232 Th or 238 U and its whole family of neutron generated daughter nuclides. All of them are regarded as fissile fuels but of quite different fission ability. The concentration of each daughter nuclide is determined by its saturate concentration ratio with the parent nuclide. In such fuel system, the whole fuel consumed by neutron reaction almost completely results in fission products. In this article, some properties of such fuel system, determination of the saturate concentration of each daughter nuclide and applicability to Th- 233 U fueled reactor will be discussed. 3 refs., 1 tab., 2 figs

  11. Continuous mixing of solids

    NARCIS (Netherlands)

    Raouf, M.S.

    1963-01-01

    The most important literature on theoretical aspects of mixing solids was reviewed.

    Only when the mixed materials showed no segregation it was possible to analyse the mixing process quantitatively. In this case the mixture could be described by the 'χ' Square test. Longitudinal mixing could be

  12. Perspectives on nuclear material safety management methods at DOE sites

    International Nuclear Information System (INIS)

    Hyder, M.L.

    1997-01-01

    The management of nuclear materials, and fissile materials in particular, at the USDOE facilities is undergoing significant changes. These result in large part from decreasing requirements for these materials in the US weapons program. Not only is new production no longer required, but returns must be handled and safely stored. Eventually surplus fissile material will be used for power production, or else put into a form suitable for long term disposition. In the meanwhile concentrates must be stored with protection against releases of radioactive material to the environment, and also against theft or deliberate dispersion. In addition, cleaning up large volumes of materials contaminated with fissile isotopes will be a major activity, and there will also be some quantity of spent fuel containing enriched uranium that cannot readily be processed. All these activities pose safety problems, some of which are addressed here

  13. Mixed Movements

    DEFF Research Database (Denmark)

    Brabrand, Helle

    2010-01-01

    levels than those related to building, and this exploration is a special challenge and competence implicit artistic development work. The project Mixed Movements generates drawing-material, not primary as representation, but as a performance-based media, making the body being-in-the-media felt and appear...... as possible operational moves....

  14. Polyfunctional inorganic-organic hybrid materials: an unusual kind of NLO active layered mixed metal oxalates with tunable magnetic properties and very large second harmonic generation.

    Science.gov (United States)

    Cariati, Elena; Macchi, Roberto; Roberto, Dominique; Ugo, Renato; Galli, Simona; Casati, Nicola; Macchi, Piero; Sironi, Angelo; Bogani, Lapo; Caneschi, Andrea; Gatteschi, Dante

    2007-08-01

    Mixed M(II)/M(III) metal oxalates, as "stripes" connected through strong hydrogen bonding by para-dimethylaminobenzaldeide (DAMBA) and water, form an organic-inorganic 2D network that enables segregation in layers of the cationic organic NLO-phore trans-4-(4-dimethylaminostyryl)-1-methylpyridinium, [DAMS+]. The crystalline hybrid materials obtained have the general formula [DAMS]4[M2M'(C2O4)6].2DAMBA.2H2O (M = Rh, Fe, Cr; M' = Mn, Zn), and their overall three-dimensional packing is non-centrosymmetric and polar, therefore suitable for second harmonic generation (SHG). All the compounds investigated are characterized by an exceptional SHG activity, due both to the large molecular quadratic hyperpolarizability of [DAMS+] and to the efficiency of the crystalline network which organizes [DAMS+] into head-to-tail arranged J-type aggregates. The tunability of the pairs of metal ions allows exploiting also the magnetic functionality of the materials. Examples containing antiferro-, ferro-, and ferri-magnetic interactions (mediated by oxalato bridges) are obtained by coupling proper M(III) ions (Fe, Cr, Rh) with M(II) (Mn, Zn). This shed light on the role of weak next-nearest-neighbor interactions and main nearest-neighbor couplings along "stripes" of mixed M(II)/M(III) metal oxalates of the organic-inorganic 2D network, thus suggesting that these hybrid materials may display isotropic 1D magnetic properties along the mixed M(II)/M(III) metal oxalates "stripes".

  15. A comparative study of recycled aggregates from concrete and mixed debris as material for unbound road sub-base; Estudio comparativo de los aridos reciclados de hormigon y mixtos como material para sub-bases de carreteras

    Energy Technology Data Exchange (ETDEWEB)

    Jimenez, J. R.; Agrela, F.; Ayuso, J.; Lopez, M.

    2011-07-01

    Seven different types of recycled aggregates from construction and demolition waste (CDW) have been evaluated as granular materials for unbound road sub bases construction. The results showed that recycled concrete aggregates complied with all specifications for using in the construction of unbound structural layers (sub-base) for T3 and T4 traffic categories according to the Spanish General Technical Specification for Road Construction (PG-3). Some mixed recycled aggregates fell short of some specifications due to a high content of sulphur compounds and poor fragmentation resistance. Sieving off the fine fraction prior to crushing the mixed CDW reduce the total sulphur content and improve the quality of the mixed recycled aggregates, by contrast, pre-sieving concrete CDW had no effect on the quality of the resulting aggregates. The results were compared with a crushed limestone as natural aggregate. (Author) 23 refs.

  16. Ministerial Decree of 27 July 1966; procedure for the notification of holding and accounting radioactive material within the meaning and in implementation of Section 30 of the Decree of the President of the Republic No. 185 of 13 February 1964, and determination of the value of the total quantity of radioactivity in radioactive material within the meaning and in implementation of Sections 3 and 13 of Act No. 1860 of 31 December 1962 respectively amended by Sections 1 and 3 of the Decree of the President of the Republic No. 1704 of 30 December 1965

    International Nuclear Information System (INIS)

    1966-01-01

    This Decree establishes the levels of radioactivity and the procedure for applying the system of notification and accounting of radioactive materials with the exception of special fissile materials, source materials and ores. (NEA) [fr

  17. Smuggling special nuclear materials

    International Nuclear Information System (INIS)

    Lazaroiu, Gheorghe

    1999-01-01

    Ever since the collapse of the former Soviet Union reports have circulated with increasing frequency concerning attempts to smuggle materials from that country's civil and military nuclear programs. Such an increase obviously raises a number of concerns (outlined in the author's introduction), chief among which is the possibility that these materials might eventually fall into the hands of proliferant states or terrorist groups. The following issues are presented: significance of materials being smuggled; sources and smuggling routes; potential customers; international efforts to reduce nuclear smuggling; long-term disposition of fissile materials. (author)

  18. Representation of the neutron cross sections of several fertile and fissile nuclei in the resonance regions

    Energy Technology Data Exchange (ETDEWEB)

    de Saussure, G.; Perez, R.B. (Oak Ridge National Lab., TN (USA))

    1982-01-01

    Several problems related to the measurement, analysis and evaluation of the neutron cross sections of the main fertile and fissile nuclides in the resonance region are reviewed. In particular the ENDF/B-V representation of these cross sections is discussed. In recent years little progress has been made in improving our knowledge of the resolved resonance parameters of the fertile nuclei. It is suggested that this absence of progress is due to a lack of adequate methodologies to deal with the systematic errors arising from uncertainties in the analysis of the measurements. The ENDF/B treatment of the unresolved resonance region is commented on and the authors recommend the validation of the unresolved resonance range evaluations with appropriate transmission and self-indication measurements.

  19. A rational approximation to Reich-Moore collision matrix of non-fissile nuclides

    International Nuclear Information System (INIS)

    Devan, K.; Keshavamurthy, R.S.

    1999-01-01

    The cross sections of many important nuclides are represented in Reich-Moore (RM) formalism in the recent American Evaluated Nuclear Data file, ENDF/B-VI. Processing of cross sections with RM resonance parameters is much more difficult than the other multilevel formalisms such as MLBW and Adler-Adler. In this paper, we derive a rational approximation to the RM collision matrix in the vicinity of a resonance. This simplifies the cross section processing. The energy range of the validity of this approximation in the vicinity of a resonance is also derived. Choosing Ni 58 as an example, results of our approximation for a non-fissile nuclide are given for two typical s-wave resonances. Our rational approximation method is found to work with good accuracies in the vicinity of resonances

  20. Experimental spectrum of reactor antineutrinos and spectra of main fissile isotopes

    Energy Technology Data Exchange (ETDEWEB)

    Sinev, V. V., E-mail: vsinev@pcbai10.inr.ruhep.ru [Russian Academy of Sciences, Institute for Nuclear Research (Russian Federation)

    2013-05-15

    Within the period between the years 1988 and 1990, the spectrum of positrons from the inverse-beta-decay reaction on a proton was measured at the Rovno atomic power plant in the course of experiments conducted there. The measured spectrum has the vastest statistics in relation to other neutrino experiments at nuclear reactors and the lowest threshold for positron detection. An experimental reactor-antineutrino spectrum was obtained on the basis of this positron spectrum and was recommended as a reference spectrum. The spectra of individual fissile isotopes were singled out from the measured antineutrino spectrum. These spectra can be used to analyze neutrino experiments performed at nuclear reactors for various compositions of the fuel in the reactor core.

  1. Criticality safety calculations of 'poison tube tank' compared with annular tanks for storing fissile solutions

    International Nuclear Information System (INIS)

    Gopalakrishnan, C.R.; Joseph, G.

    1995-01-01

    A comparative study of the shielded area space required for storing fissile solution by the conventional annular tank and by poison tube tank is made. Poison tube tank is similar to commercial heat exchanger. The neutron poisons studied are gadolinium oxide and borax. Variation of multiplication factor for an array of annular tanks containing uranium nitrate or plutonium nitrate solutions are presented for annular widths of 10, 7.5 and 5 cm. It is concluded that for the given concentration, 5 cm annular width tanks are safe at a pitch distance of 120 and 90 cm for uranium and plutonium solutions respectively. Using these, as reference values, it is found that the shielded area saving for the poison tube tank is a factor of 12 and 8 for the given concentration of uranium and plutonium solutions respectively. (author)

  2. 15 year's summary report on blanket technology and materials of mixed fuel reactor research sponsored by national '863' projects

    International Nuclear Information System (INIS)

    Xu Zengyu; Chen Jiming; Liu Xiang

    2001-01-01

    15 year's achievements of Southwestern Institute of Physics, China, in fusion technology and materials research sponsored by National '863' Engineering Projects are summarized. Many scientific and technical achievements are obtained in the researches on tritium production and recovery, doped carbon basic materials, V-alloys, 316L SS irradiation performance, B 4 C and TiC coatings, etc. Some facilities were built and some were improved for materials research. 108 references are annexed

  3. Analysis of the differences in breeding ratio and fissile inventory between heterogeneous and homogeneous liquid-metal fast breeder reactors

    International Nuclear Information System (INIS)

    Tzanos, C.P.

    1980-01-01

    The differences in fissile inventory and breeding ratio, with respect to the differences in fertile inventory and neutron spectrum, between equivalent heterogeneous and homogeneous configurations were analyzed. To quantify the effect of spectral changes on reaction rate ratios, a calculational scheme based on properly prepared one-group cross-section sets was used

  4. Atomic-Resolution Visualization of Distinctive Chemical Mixing Behavior of Ni, Co and Mn with Li in Layered Lithium Transition-Metal Oxide Cathode Materials

    Energy Technology Data Exchange (ETDEWEB)

    Yan, Pengfei; Zheng, Jianming; Lv, Dongping; Wei, Yi; Zheng, Jiaxin; Wang, Zhiguo; Kuppan, Saravanan; Yu, Jianguo; Luo, Langli; Edwards, Danny J.; Olszta, Matthew J.; Amine, Khalil; Liu, Jun; Xiao, Jie; Pan, Feng; Chen, Guoying; Zhang, Jiguang; Wang, Chong M.

    2015-07-06

    Capacity and voltage fading of layer structured cathode based on lithium transition metal oxide is closely related to the lattice position and migration behavior of the transition metal ions. However, it is scarcely clear about the behavior of each of these transition metal ions. We report direct atomic resolution visualization of interatomic layer mixing of transition metal (Ni, Co, Mn) and lithium ions in layer structured oxide cathodes for lithium ion batteries. Using chemical imaging with aberration corrected scanning transmission electron microscope (STEM) and DFT calculations, we discovered that in the layered cathodes, Mn and Co tend to reside almost exclusively at the lattice site of transition metal (TM) layer in the structure or little interlayer mixing with Li. In contrast, Ni shows high degree of interlayer mixing with Li. The fraction of Ni ions reside in the Li layer followed a near linear dependence on total Ni concentration before reaching saturation. The observed distinctively different behavior of Ni with respect to Co and Mn provides new insights on both capacity and voltage fade in this class of cathode materials based on lithium and TM oxides, therefore providing scientific basis for selective tailoring of oxide cathode materials for enhanced performance.

  5. The use of amorphous silica-alumina-based additive in the adhesive dry mixes of building materials

    Directory of Open Access Journals (Sweden)

    Loganina VI

    2018-01-01

    Full Text Available Proved the possibility of using amorphous aluminosilicate as a modifying agent for the adhesive dry mixes. Are given the data on the microstructure and chemical composition of the amorphous aluminosilicates. Installed , that the microstructure of the synthetic additives is characterized by particles of round shape, dimensions 5,208-5,704 μm, Also there are particles of elongated shape in size 7.13-8.56 μm. Predominate chemical elements O, Si, Na, S, and Al in quantity 60.69%, 31.26%, 24.23%, 18.69% and 8.29% respectively. Described the character changes in the rheological properties of cement-sand mortar, depending on the percentage of additives. Determined, that the introduction in the cement-sand mortar the additive based on amorphous aluminosilicate leads to higher values of plastic strength. Are given the model of cement stone strength using synthetic additives in the formulation. The results of the evaluation of the frost resistance of cement-based tile adhesives with the use of amorphous aluminosilicates as a modifying additive are presented. In the article is determined the mark on frost resistance of tile glue and frost resistance of the contact zone of tile glue. The evaluation of the performance properties of the layer of tile adhesive on the basis of cement, dry mixes. The calculation of the value of displacement of the adhesive layer made on the basis of the developed recipes cement dry mixes applied to a vertical surface. Experimental data obtained values of displacement tiles relative to the substrate. Described the results of physical and mechanical properties of tile adhesive made on the basis of the developed adhesive dry mix formulations.

  6. Stb 342 - Decree of 4 June 1987 amending the Decree on the transport of fissionable materials, ores and radioactive substances

    International Nuclear Information System (INIS)

    1987-01-01

    The 1969 transport Decree governs all modes of transport of fissile and radioactive materials as well as ores in and to and from the Netherlands. The 1987 Decree amends it, in particular, for modernization purposes. (NEA) [fr

  7. Studies relating to construction materials to be used in different options for head end treatment in reprocessing of mixed carbide fuel of plutonium and uranium

    International Nuclear Information System (INIS)

    Rajan, S.K.; Palamalai, A.; Ravi, T.N.; Sampath, M.; Raman, V.R.; Balasubramanian, G.R.

    1993-01-01

    Mixed carbide of uranium and plutonium has been chosen as the fuel for the first core of Fast Breeder Test Reactor, installed in the Indira Gandhi Centre for Atomic Research. Reprocessing of this fuel is one of the vital steps to prove the viability of the fuel cycle. The head end treatment process introduces constraints in the reprocessing of carbide fuel when compared to the commonly used mixed oxide fuel. Three head end processes, namely direct oxidation, pyrohydrolysis and direct dissolution in nitric acid with oxidation of organic acids were considered for study for exercising the choice. The paper briefly describes the three processes. In each process probable material of construction and related problems are discussed. (author). 3 refs, 5 figs, 7 tabs

  8. Effect of the raw materials and mixing ratio of composted wastes on the dynamic of organic matter stabilization and nitrogen availability in composts of Sub-Saharan Africa.

    Science.gov (United States)

    Kaboré, Théodore Wind-Tinbnoma; Houot, Sabine; Hien, Edmond; Zombré, Prosper; Hien, Victor; Masse, Dominique

    2010-02-01

    The effect of raw materials and their proportions in initial mixtures on organic matter (OM) stabilization and nitrogen (N) availability during pit composting in Sub-Saharan Africa was assessed using biochemical fractionation and laboratory incubations to characterize composts sampled throughout the composting process. Stabilization of OM occurred more rapidly in mixtures with slaughter-house wastes, it was progressive in mixture with household refuses while tree leaves compost remained unstable. Carbon mineralization from compost samples was positively correlated to water soluble and hemicellulose-like organic fractions. Mixtures containing large proportions of household refuses reached the highest stability and total N but available N remained weak. Slaughter-house wastes in the initial mixtures made possible to reach good OM stabilization and the largest N availability. The nature of initial mixing influenced composting parameters, OM stabilization and N availability. It is suggested mixing household refuses and slaughter-house wastes with tree leaves to reach better amending and fertilizer qualities of composts.

  9. Protective potential of non-dialyzable material fraction of cranberry juice on the virulence of P. gingivalis and F. nucleatum mixed infection.

    Science.gov (United States)

    Polak, David; Naddaf, Raja; Shapira, Lior; Weiss, Ervin I; Houri-Haddad, Yael

    2013-07-01

    Periodontitis is a polymicrobial infectious disease. A novel potential chemical treatment modality may lie in bacterial anti-adhesive materials, such as cranberry juice fractions. The aim of this study is to explore the effect of high molecular weight cranberry constituent (non-dialyzable material [NDM]) on the virulence of a mixed infection with Porphyromonas gingivalis and Fusobacterium nucleatum in mice. In vitro, the anti-adhesive property of NDM was validated on epithelial cell culture, and inhibition of coaggregation was tested using a coaggregation assay. The in vivo effect was tested on the outcome of experimental periodontitis induced by a P. gingivalis and F. nucleatum mixed infection, and also on the local host response using the subcutaneous chamber model of infection. Phagocytosis was also tested on RAW macrophages by the use of fluorescent-labeled bacteria. NDM was found to inhibit the adhesion of both species of bacteria onto epithelial cells and to inhibit coaggregation in a dose-dependent manner. NDM consumption by mice attenuated the severity of experimental periodontitis compared with a mixed infection without NDM treatment. In infected subcutaneous chambers, NDM alone reduced tumor necrosis factor-α (TNF-α) levels induced by the mixed infection. In vitro, NDM eliminated TNF-α expression by macrophages that were exposed to P. gingivalis and F. nucleatum, without impairing their viability. Furthermore, NDM increased the phagocytosis of P. gingivalis. The results indicate that the use of NDM may hold potential protective and/or preventive modalities in periodontal disease. Underlying mechanisms for this trait may perhaps be the anti-adhesive properties of NDM or its potential effect on inflammation.

  10. Nutritional Evaluation of Raw Materials Entering the Structure to Mixed Fodder for the Specie Poecilia reticulata (Guppy

    Directory of Open Access Journals (Sweden)

    Adrian Gruber

    2011-05-01

    Full Text Available In the ornamental pisciculture is a especial emphasis on the exterior shape and color to the fishes, issues that are dependent directly to the structure of compound feeds in relation to the nutritional characteristics of the raw materials.Own research or focused on analyzing the crude chemical composition with Weende scheme (water content and dry matter, crude ash, crude protein, crude fat, crude fiber, SEN of raw materials can be used in the structure of a compound feeds for the Poecilia reticulata (guppy species, for most of these materials there are no current data in the literature.These materials were analyzed: gelatin, wheat flour, sunflower meal, soybean meal, meal Spirulina platensis, carrot (Daucus carota, Pangasius fillet, Daphnia pulex, grount dandelion (Taraxacum officinale, grount nettle (Urtica dioica and yeast.

  11. M-edge x-ray absorption spectroscopy: A new tool for dilute mixed-valent materials

    International Nuclear Information System (INIS)

    Kaindl, G.; Brewer, W.D.; Kalkowski, G.; Holtzberg, F.

    1983-01-01

    The valence of Tm compounds is derived from M/sub V/ x-ray absorption spectra recorded by total electron yield under ultra-high-vacuum conditions. For mixed-valent systems the spectra are superpositions of Tm 3+ (three lines) and Tm 2+ (one line) components, providing accurate mean valence values even in highly dilute systems, such as Tm/sub x/Y/sub 1-x/Se, which agree well with those from lattice constant systematics. A surface valence change on TmS(100) is identified as an initial-state effect

  12. Analysis of Environmental Impact for Concrete Using LCA by Varying the Recycling Components, the Compressive Strength and the Admixture Material Mixing

    Directory of Open Access Journals (Sweden)

    Taehyoung Kim

    2016-04-01

    Full Text Available Concrete is a type of construction material in which cement, aggregate, and admixture materials are mixed. When cement is produced, large amounts of substances that impact the environment are emitted during limestone extraction and clinker manufacturing. Additionally, the extraction of natural aggregate causes soil erosion and ecosystem destruction. Furthermore, in the process of transporting raw materials such as cement and aggregate to a concrete production company, and producing concrete in a batch plant, substances with an environmental impact are emitted into the air and water system due to energy use. Considering the fact that the process of producing concrete causes various environmental impacts, an assessment of various environmental impact categories is needed. This study used a life cycle assessment (LCA to evaluate the environmental impacts of concrete in terms of its global warming potential, acidification potential, eutrophication potential, ozone depletion potential, photochemical ozone creation potential, and abiotic depletion potential (GWP, AP, EP, ODP, POCP, ADP. The tendency was that the higher the strength of concrete, the higher the GWP, POCP, and ADP indices became, whereas the AP and EP indices became slightly lower. As the admixture mixing ratio of concrete increased, the GWP, AP, ODP, ADP, and POCP decreased, but EP index showed a tendency to increase slightly. Moreover, as the recycled aggregate mixing ratio of concrete increased, the AP, EP, ODP, and ADP decreased, while GWP and POCP increased. The GWP and POCP per unit compressed strength (1 MPa of high strength concrete were found to be about 13% lower than that for its normal strength concrete counterpart. Furthermore, in the case of AP, EP, ODP, and ADP per unit compressed strength (1 MPa, high-strength concrete was found to be about 10%~25% lower than its normal strength counterpart. Among all the environmental impact categories, ordinary cement was found to have

  13. European mixed forests

    DEFF Research Database (Denmark)

    Bravo-Oviedo, Andres; Pretzsch, Hans; Ammer, Christian

    2014-01-01

    Aim of study: We aim at (i) developing a reference definition of mixed forests in order to harmonize comparative research in mixed forests and (ii) review the research perspectives in mixed forests. Area of study: The definition is developed in Europe but can be tested worldwide. Material...... and Methods: Review of existent definitions of mixed forests based and literature review encompassing dynamics, management and economic valuation of mixed forests. Main results: A mixed forest is defined as a forest unit, excluding linear formations, where at least two tree species coexist at any...... density in mixed forests, (iii) conversion of monocultures to mixed-species forest and (iv) economic valuation of ecosystem services provided by mixed forests. Research highlights: The definition is considered a high-level one which encompasses previous attempts to define mixed forests. Current fields...

  14. Comparative measurements on different thermoluminescence materials for the dosimetry in mixed radiation fields; Vergleichende Messungen an unterschiedlichen Thermolumineszenzmaterialien fuer die Dosimetrie in gemischten Strahlenfeldern

    Energy Technology Data Exchange (ETDEWEB)

    Pillath, J.; Uray, I.

    2006-05-15

    LiF-based thermoluminescence materials (TL materials) have proved efficient and become established for dosimetry for many years now. In combination with suitable filtrations or moderators, they permit high-precision measurements of the radiation doses of all types of radiation relevant in radiation protection. The development of highly sensitive LiF materials with a doping of Mg, P and Cu has made it possible to measure doses down to one {mu}Sv. The measurements presented here specifically serve to phenomenologically analyse the glow curve structure of normally and highly sensitive LiF-TL materials in mixed gamma-neutron radiation fields. The sensitivity of the different TL materials to different types of radiation and the influence of the radiation types and of the evaluation parameters on the structure of the glow curves is examined. It is made apparent how information about the composition and doses of the individual components of a radiation field can be obtained by decomposing the glow curve into its individual peaks. (orig.)

  15. FORMATION CONDITIONS OF ICY MATERIALS IN COMET C/2004 Q2 (MACHHOLZ). I. MIXING RATIOS OF ORGANIC VOLATILES

    International Nuclear Information System (INIS)

    Kobayashi, Hitomi; Kawakita, Hideyo

    2009-01-01

    We observed comet C/2004 Q2 (Machholz) with the Keck II telescope in late 2005 January and we obtained the spectra of C/2004 Q2 including many emission lines of volatile species such as H 2 O, HCN, C 2 H 2 , NH 3 , CH 4 , C 2 H 6 , CH 3 OH, and H 2 CO with high-signal-to-noise ratios. Based on our observations, we determined the mixing ratios of the molecules relative to H 2 O in C/2004 Q2. Since C/2004 Q2 is one of Oort Cloud comets, it is interesting to compare our results with other Oort Cloud comets. The mixing ratios of C 2 H 2 /H 2 O and C 2 H 6 /H 2 O in C/2004 Q2 are lower than typical Oort Cloud comets. Especially, C 2 H 2 /H 2 O ratio in C/2004 Q2 is as lower as Jupiter Family comets. However, mixing ratios of other molecules in C/2004 Q2 are similar to typical Oort Cloud comets. C/2004 Q2 might be the intermediate type between Oort Cloud and Jupiter Family comets. To investigate the formation conditions of such intermediate type comet, we focused on the (C 2 H 2 +C 2 H 6 )/H 2 O ratios and C 2 H 6 /(C 2 H 6 +C 2 H 2 ) ratios in comets from the viewpoint of conversion from C 2 H 2 to C 2 H 6 in the precometary ices. We found that (C 2 H 2 +C 2 H 6 )/H 2 O ratio in C/2004 Q2 is lower than the ratio in typical Oort Cloud comets while C 2 H 6 /(C 2 H 6 +C 2 H 2 ) ratio in C/2004 Q2 is consistent with the ratio of the typical Oort Cloud comets and Jupiter family comets. If we assume that the cometary volatiles such as H 2 O, CH 4 , and C 2 H 2 formed similar environment, the C 2 H 6 /(C 2 H 6 +C 2 H 2 ) ratio might not be sensitive in the temperature range where hydrogen-addition reactions occurred and cometesimals formed (∼30 K). We employed the dynamical-evolutional model and the chemical-evolutional model to determine the formation region of C/2004 Q2 more precisely. We found that comet C/2004 Q2 might have formed in relatively inner region of the solar nebula than the typical Oort Cloud comet (but slightly further than 5 AU from the proto-Sun).

  16. Few layer graphene wrapped mixed phase TiO2 nanofiber as a potential electrode material for high performance supercapacitor applications

    Science.gov (United States)

    Thirugnanam, Lavanya; Sundara, Ramaprabhu

    2018-06-01

    A combination of favorable composition and optimized anatase/rutile mixed-phase TiO2 (MPTNF)/Hydrogen exfoliated graphene (HEG) composite nanofibers (MPTNF/HEG) and anatase/rutile mixed-phase TiO2/reduced graphene oxide (rGO) composite nanofibers (MPTNF/rGO) have been reported to enhance the electrochemical properties for supercapacitor applications. These composite nanofibers have been synthesized by an efficient route of electrospinning together with the help of easy chemical methods. Both the composites exhibit good charge storage capability with enhanced pseudocapacitance and electric double-layer capacitance (EDLC) as confirmed by cyclic voltammetry studies. MPTNF/HEG composite showed maximum specific capacitance of 210.5 F/g at the current density of 1 A/g, which was mainly due to its availability of the more active sites for ions adsorption on a few layers of graphene wrapped TiO2 nanofiber surface. The synergistic effect of anatase/rutile mixed phase with one dimensional nanostructure and the electronic interaction between TiO2 and few layer graphene provided the subsequent improvement of ion adsorption capacity. Also exhibit excellent electrochemical performance to improve the capacitive properties of TiO2 electrode materials which is required for the development of flexible electrodes in energy storage devices and open up new opportunities for high performance supercapacitors.

  17. Fissile fuel breeding and minor actinide transmutation in the life engine

    International Nuclear Information System (INIS)

    Sahin, Suemer; Khan, Mohammad Javed; Ahmed, Rizwan

    2011-01-01

    zone (50 cm), containing MA as fissionable fuel. A 3rd ODS layer (2 cm) separates the molten salt zone on the right side from the graphite reflector (30 cm). Calculations have been conducted for a fusion driver power of 500 MW th in S 8 -P 3 approximation using 238-neutron groups. Minor actinides (MA) out of the nuclear waste of LWRs are used as fissile carbide fuel in TRISO particles with volume fractions of 0, 2, 3, 4 and 5% have been dispersed homogenously in the Flibe coolant. For these cases, tritium breeding at startup is calculated as TBR = 1.134, 1.286, 1.387, 1.52 and 1.67, respectively. In the course of plant operation, TBR and fissile neutron multiplication factor decrease gradually. For a self-sustained reactor, TBR > 1.05 can be kept for all cases over 8 years. Higher fissionable fuel content in the molten salt leads also to higher blanket energy multiplication, namely M = 3.3, 4.6, 6.15 and 8.1 with 2, 3, 4 and 5% TRISO volume fraction at start up, respectively. For all investigated cases, fissile burn up exceeds 400 000 MW D/MT. Major damage mechanisms have been calculated as DPA = 50 and He = 176 appm per year. This implies a replacement of the first wall every 3 years.

  18. The development of fuel pins and material specimens mixed loading irradiation test rig in the experimental fast reactor Joyo. The development of the fuel-material hybrid rig

    International Nuclear Information System (INIS)

    Oyamatsu, Yasuko; Someya, Hiroyuki

    2013-02-01

    In the experimental fast reactor Joyo, there were many tests using the irradiation rigs that it was possible to be set irradiation conditions for each compartment independently. In case of no alternative fuel element to irradiate after unloading the irradiated compartments, the irradiation test was restarted with the dummy compartment which the fuel elements was not mounted. If the material specimens are mounted in this space, it is possible to use the irradiation space effectively. For these reasons, the irradiation rig (hybrid rig) is developed that is consolidated with material specimens compartment and fuel elements compartment. Fuel elements and material specimens differ greatly with heat generation, so that the most important issue in developing of hybrid rig is being able to distribute appropriately the coolant flow which satisfies irradiation conditions. The following is described by this report. (1) It was confirmed that the flow distribution of loading the same irradiation rig with the compartment from which a flow demand differs could be satisfied. (2) It was confirmed that temperature setting range of hybrid rig could be equivalent to that of irradiation condition. (3) By standardizing the coolant entrance structure of the compartment lower part, the prospect which can perform easily recombination of the compartment from which a type differs between irradiation rigs was acquired. (author)

  19. The influence of fusion-relevant D{sub 2}-0.1He plasma on Be-W mixed-materials

    Energy Technology Data Exchange (ETDEWEB)

    Jepu, I., E-mail: ionut.jepu@inflpr.ro [National Institute for Lasers, Plasma and Radiation Physics, NILPRP, Magurele (Romania); Baldwin, M.J.; Nishijima, D.; Doerner, R.P. [Center for Energy Research, University of California at San Diego, La Jolla, CA (United States); Porosnicu, C.; Lungu, C.P. [National Institute for Lasers, Plasma and Radiation Physics, NILPRP, Magurele (Romania); Dinca, P. [National Institute for Lasers, Plasma and Radiation Physics, NILPRP, Magurele (Romania); University of Bucharest, Faculty of Physics, Magurele, Bucharest (Romania); Marin, A. [Institute of Physical Chemistry, “Ilie Murgulescu”, Bucharest (Romania); Institute for Nuclear Research, Mioveni, Arges (Romania)

    2017-02-15

    Compositionally homogeneous mixed-material layers of Be and W, ∼2 μm thick, and of various W contents (spanning 0–100 at.% W), are produced by thermionic vacuum arc deposition. The layers are exposed to low energy (∼50 eV), high ion flux (6–8 × 10{sup 22} m{sup −2} s{sup −1}) D{sub 2}-0.1He (10% He) mixed species plasma at 473 K and 1073 K to simulate ITER-like first wall and divertor plasma-material interaction. Surface morphological and compositional analysis is conducted prior and subsequent to plasma exposure. The action of the plasma produces almost complete depletion of Be from the near surface, and this is found regardless of the exposure temperature or initial composition. In the exposure case of 473 K, thermal desorption spectrometry reveals decreased D retention in W rich compositions compared to similar exposure in D{sub 2} plasma without He, but no discernable difference is noted in Be rich compositions. At 1073 K, surface enrichment in W, by depletion of the Be, results in He induced W “fuzz” structures for all Be-W compositions explored.

  20. Elaboration of hybrid materials by templating with mineral liquid crystals stabilization of a mixed sol of YSZ nanoparticles and V2O5 ribbon-like colloids

    International Nuclear Information System (INIS)

    Guiot, C.

    2009-01-01

    The purpose of this PhD was to investigate innovative soft chemistry ways to prepare hybrid materials with ordered nano-structures. Concretely, research were conducted on the development of a hybrid material made of an yttria-stabilized zirconia (YSZ) matrix templated by a mineral liquid crystal, namely V 2 O 5 . In aqueous solutions, vanadium oxide exhibits ribbon-like colloids of typical dimensions 1 nm x 25 nm x 500 nm, stabilized by a strong negative surface charge. Above a critical concentration, the anisotropic colloids assemble into a nematic liquid crystal, whose domains can be oriented within the same direction over a macroscopic range under a weak magnetic field. The idea is to use V 2 O 5 anisotropic colloids as a template for a hybrid material, taking advantage of their ordering behavior. Preliminary experiments revealed a strong reactivity between molecular compounds of zirconium and vanadium oxide. Therefore, the studies were directed toward the preparation of a mixed colloidal sol containing YSZ nanoparticles and vanadium oxide ribbon-like colloids, as a precursor sol for the intended hybrid material. The YSZ nanoparticles are obtained through an outstanding hydrothermal synthesis leading to a stable suspension of nanocrystalline particles of ca. 5 nm, in pure water. Providing a mixed sol of YSZ and V 2 O 5 is a key challenge for it implies the co-stabilization of two types of colloids having different shape, size and surface properties. Besides, the existence of V 2 O 5 in its ribbon-like form requires acidic conditions and very low ionic strength. The first part of this work was then dedicated to the study of electro-steric stabilization of zirconia suspension by addition of acidic poly-electrolytes. Different polymers with carboxylic and/or sulfonic acidic functions were investigated. Based on zeta potential measurements and adsorption isotherms, the influence of molecular weight and polymer charge were discussed. Among the studied polymers, poly