WorldWideScience

Sample records for mixed fissile materials

  1. 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.

  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. 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

  4. 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.

  5. 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

  6. 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

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

    International Nuclear Information System (INIS)

    Forsberg, C.W.

    1997-01-01

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

  8. 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

  9. 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.

  10. 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

  11. 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

  12. 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

  13. 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

  14. 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

  15. 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

  16. 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.

  17. 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

  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. 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

  20. 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

  1. 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)

  2. 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

  3. 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

  4. 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

  5. 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

  6. 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)

  7. 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

  8. 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)

  9. 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)

  10. 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.)

  11. 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

  12. 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)

  13. 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.

  14. 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.

  15. 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

  16. 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

  17. 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

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

    International Nuclear Information System (INIS)

    Von Baeckmann, A.; Neuber, J.

    1975-01-01

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

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  1. 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.

  2. 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.

  3. 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

  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. 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 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)

  7. 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

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

    International Nuclear Information System (INIS)

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

    1996-01-01

    The Department of Energy (DOE) is analyzing long-term storage 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

  9. 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

  10. 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

  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. 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

  13. 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

  14. 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

  15. 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)

  16. 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)

  17. 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)

  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. 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. 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

  1. 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.

  2. 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

  3. 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

  4. 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

  5. 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

  6. 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

  7. 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

  8. 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

  9. 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

  10. 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

  11. 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

  12. 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

  13. 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.

  14. 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

  15. 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

  16. 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...

  17. 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)

  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 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)

  19. 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

  20. 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)

  1. 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)

  2. 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.

  3. 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

  4. 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

  5. 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

  6. 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

  7. 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

  8. 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.)

  9. Methodology for interpretation of fissile mass flow measurements

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  10. 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

  11. 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

  12. 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.)

  13. 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

  14. 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

  15. 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.)

  16. 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

  17. 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

  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. 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.

  20. 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

  1. 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

  2. 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.

  3. 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.

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

    International Nuclear Information System (INIS)

    Thomas, J.T.

    1985-01-01

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

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

    International Nuclear Information System (INIS)

    Meyer, C.A.

    1975-06-01

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

  6. 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. 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)

  8. 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)

  9. 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

  10. 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.)

  11. 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

  12. 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.

  13. 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)

  14. 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

  15. 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)

  16. 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

  17. 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

  18. 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.

  19. 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.

  20. 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)

  1. 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

  2. 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.

  3. 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

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

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  5. 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.)

  6. 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

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

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  8. 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

  9. 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

  10. 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

  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. 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

  13. 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)

  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. 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)

  16. 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] ...

  17. 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

  18. 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.

  19. 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

  20. 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.

  1. 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)

  2. 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.

  3. 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

  4. 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

  5. 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

  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. 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

  9. 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...

  10. 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.)

  11. 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

  12. 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

  13. 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

  14. 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

  15. 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

  16. 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

  17. 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

  18. 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)

  19. 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

  20. 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

  1. 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

  2. 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

  3. 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)

  4. 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)

  5. 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)

  6. 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

  7. 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

  8. 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.

  9. 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

  10. 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.

  11. 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.

  12. 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.

  13. 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

  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. 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

  16. 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)

  17. 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)

  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. 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

  20. 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

  1. 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

  2. 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)

  3. 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

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

    International Nuclear Information System (INIS)

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

    1977-10-01

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

  5. 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

  6. 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

  7. 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

  8. 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

  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. 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

  11. 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)

  12. 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)

  13. Nuclear fuel pin

    International Nuclear Information System (INIS)

    Hartley, Kenneth; Moulding, T.L.J.; Rostron, Norman.

    1979-01-01

    Fuel pin for use in fast breeder nuclear reactors containing fissile and fertile areas of which the fissile and fertile materials do not mix. The fissile material takes the shape of large and small diameter microspheres (the small diameter microspheres can pass through the interstices between the large microspheres). The barrier layers being composed of microspheres with a diameter situated between those of the large and small microspheres ensure that the materials do not mix [fr

  14. 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

  15. 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.

  16. 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.

  17. 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

  18. 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)

  19. 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)

  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. 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)

  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. 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

  4. 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)

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

    International Nuclear Information System (INIS)

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

    2000-01-01

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

  6. 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)

  7. 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

  8. 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

  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. 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. 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

  12. 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

  13. Counterstreaming-ion-tokamak fissile breeder

    International Nuclear Information System (INIS)

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

    1976-08-01

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

  14. 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.

  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. 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

  19. Safeguarding nuclear weapon: Usable materials in Russia

    International Nuclear Information System (INIS)

    Cochran, T.

    1998-01-01

    Both the United States and Russia are retaining as strategic reserves more plutonium and HEU for potential reuse as weapons, than is legitimately needed. Both have engaged in discussions and have programs in various stages of development to dispose of excess plutonium and HEU. These fissile material disposition programs will take decades to complete. In the interim there will be, as there is now, hundreds of tons of separated weapon-usable fissile material stored in tens of thousands of transportable canisters, each containing from a few to several tons of kgs of weapon-usable fissile material. This material must be secured against theft and unauthorized use. To have high confidence that the material is secure, one must establish criteria against which the adequacy of the protective systems can be judged. For example, one finds such criteria in US Nuclear Regulatory Commission (USNRC) regulations for the protection of special nuclear materials

  20. 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

  1. 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)

  2. 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)

  3. 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

  4. 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

  5. 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.

  6. 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)

  7. 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

  8. 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

  9. 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

  10. 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

  11. 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.

  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. 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.

  14. 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.

  15. 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.

  16. 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)

  17. 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)

  18. Global nuclear material flow/control model

    International Nuclear Information System (INIS)

    Dreicer, J.S.; Rutherford, D.S.; Fasel, P.K.; Riese, J.M.

    1997-01-01

    This is the final report of a two-year, Laboratory Directed Research and Development (LDRD) project at the Los Alamos National Laboratory (LANL). The nuclear danger can be reduced by a system for global management, protection, control, and accounting as part of an international regime for nuclear materials. The development of an international fissile material management and control regime requires conceptual research supported by an analytical and modeling tool which treats the nuclear fuel cycle as a complete system. The prototype model developed visually represents the fundamental data, information, and capabilities related to the nuclear fuel cycle in a framework supportive of national or an international perspective. This includes an assessment of the global distribution of military and civilian fissile material inventories, a representation of the proliferation pertinent physical processes, facility specific geographic identification, and the capability to estimate resource requirements for the management and control of nuclear material. The model establishes the foundation for evaluating the global production, disposition, and safeguards and security requirements for fissile nuclear material and supports the development of other pertinent algorithmic capabilities necessary to undertake further global nuclear material related studies

  19. Materials of Criticality Safety Concern in Waste Packages

    International Nuclear Information System (INIS)

    Larson, S.L.; Day, B.A.

    2006-01-01

    10 CFR 71.55 requires in part that the fissile material package remain subcritical when considering 'the most reactive credible configuration consistent with the chemical and physical form of the material'. As waste drums and packages may contain unlimited types of materials, determination of the appropriately bounding moderator and reflector materials to ensure compliance with 71.55 requires a comprehensive analysis. Such an analysis was performed to determine the materials or elements that produce the most reactive configuration with regards to both moderation and reflection of a Pu-239 system. The study was originally performed for the TRUPACT-II shipping package and thus the historical fissile mass limit for the package, 325 g Pu-239, was used [1]. Reactivity calculations were performed with the SCALE package to numerically assess the moderation or reflection merits of the materials [2]. Additional details and results are given in SAIC-1322-001 [3]. The development of payload controls utilizing process knowledge to determine the classification of special moderator and/or reflector materials and the associated fissile mass limit is also addressed. (authors)

  20. 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)

  1. 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

  2. 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.

  3. A novel method to assay special nuclear materials by measuring prompt neutrons from polarized photofission

    Energy Technology Data Exchange (ETDEWEB)

    Mueller, J.M., E-mail: mueller@tunl.duke.edu [Triangle Universities Nuclear Laboratory, Durham, NC 27710 (United States); Department of Physics, Duke University, Durham, NC 27708 (United States); Ahmed, M.W. [Triangle Universities Nuclear Laboratory, Durham, NC 27710 (United States); Department of Physics, Duke University, Durham, NC 27708 (United States); Department of Mathematics and Physics, North Carolina Central University, Durham, NC 27707 (United States); Weller, H.R. [Triangle Universities Nuclear Laboratory, Durham, NC 27710 (United States); Department of Physics, Duke University, Durham, NC 27708 (United States)

    2014-08-01

    A novel method of measuring the enrichment of special nuclear material is presented. Recent photofission measurements using a linearly polarized γ-ray beam were performed on samples of {sup 232}Th, {sup 233,235,238}U, {sup 237}Np, and {sup 239,240}Pu. Prompt neutron polarization asymmetries, defined to be the difference in the prompt neutron yields parallel and perpendicular to the plane of beam polarization divided by their sum, were measured. It was discovered that the prompt neutron polarization asymmetries differed significantly depending on the sample. Prompt neutrons from photofission of even–even (non-fissile) targets had significant polarization asymmetries (∼0.2 to 0.5), while those from odd-A (generally fissile) targets had polarization asymmetries close to zero. This difference in the polarization asymmetries could be exploited to measure the fissile versus non-fissile content of special nuclear materials, and potentially to detect the presence of fissile material during active interrogation. The proposed technique, its expected performance, and its potential applicability are discussed.

  4. A novel method to assay special nuclear materials by measuring prompt neutrons from polarized photofission

    International Nuclear Information System (INIS)

    Mueller, J.M.; Ahmed, M.W.; Weller, H.R.

    2014-01-01

    A novel method of measuring the enrichment of special nuclear material is presented. Recent photofission measurements using a linearly polarized γ-ray beam were performed on samples of 232 Th, 233,235,238 U, 237 Np, and 239,240 Pu. Prompt neutron polarization asymmetries, defined to be the difference in the prompt neutron yields parallel and perpendicular to the plane of beam polarization divided by their sum, were measured. It was discovered that the prompt neutron polarization asymmetries differed significantly depending on the sample. Prompt neutrons from photofission of even–even (non-fissile) targets had significant polarization asymmetries (∼0.2 to 0.5), while those from odd-A (generally fissile) targets had polarization asymmetries close to zero. This difference in the polarization asymmetries could be exploited to measure the fissile versus non-fissile content of special nuclear materials, and potentially to detect the presence of fissile material during active interrogation. The proposed technique, its expected performance, and its potential applicability are discussed

  5. 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

  6. 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.

  7. 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

  8. 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

  9. 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

  10. 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.

  11. Reducing nuclear danger through intergovernmental technical exchanges on nuclear materials safety management

    International Nuclear Information System (INIS)

    Jardine, L.J.; Peddicord, K.L.; Witmer, F.E.; Krumpe, P.F.; Lazarev, L.; Moshkov, M.

    1997-01-01

    The United States and Russia are dismantling nuclear weapons and generating hundreds of tons of excess plutonium and high enriched uranium fissile nuclear materials that require disposition. The U.S. Department of Energy and Russian Minatom organizations.are planning and implementing safe, secure storage and disposition operations for these materials in numerous facilities. This provides a new opportunity for technical exchanges between Russian and Western scientists that can establish an improved and sustained common safety culture for handling these materials. An initiative that develops and uses personal relationships and joint projects among Russian and Western participants involved in fissile nuclear materials safety management contributes to improving nuclear materials nonproliferation and to making a safer world. Technical exchanges and workshops are being used to systematically identify opportunities in the nuclear fissile materials facilities to improve and ensure the safety of workers, the public, and the environment

  12. 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.

  13. 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

  14. 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.

  15. 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

  16. 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)

  17. 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.

  18. The industrial development of atomic energy; Le developpement industriel de l'energie atomique

    Energy Technology Data Exchange (ETDEWEB)

    Kowarski, L [Commissariat a l' Energie Atomique, Paris (France). Centre d' Etudes Nucleaires

    1955-07-01

    Countries with large stock of fissile material and producing large quantity of nuclear pure {sup 235}U and {sup 239}Pu are able to allocate part of the stock to non military research. For countries with low stock of fissile material, all the stock is allocated to military research. An economical and technical solution has to be find to dedicate a part of fissile material to non military research and develop the atomic energy industry. It stated the industrial and economical problems and in particular the choice between the use of enriched fuel with high refining cost or depleted fuel with low production cost. It discusses of four possible utilizations of the natural resources: reactors functioning with pure fissile material ({sup 235}U or {sup 239}Pu) or concentrated material ({sup 235}U mixed with small quantities of {sup 238}U after an incomplete isotopic separation), breeder reactors functioning with enriched material mixed with {sup 238}U or Thorium placed in an appropriate spatial distribution to allow neutrons beam to activate {sup 238}U or Thorium with the regeneration of fissile material in {sup 239}Pu, reactors using natural uranium or low enriched uranium can also produce Plutonium with less efficiency than breeder reactors and the last solution being the use of natural uranium with the only scope of energy production and no production of secondary fissile material. The first class using pure fissile material has a low energy efficiency and is used only by large fissile material stock countries to accumulate energy in small size fuel for nuclear engines researches for submarines and warships. The advantage of the second class of reactors, breeder reactors, is that they produce energy and plutonium. Two type of breeder reactor are considered: breeder reactor using pure fissile material and {sup 238}U or breeder reactor using the promising mixture of pure fissile material and Thorium. Different projects are in phase of development in United States, England

  19. The industrial development of atomic energy

    International Nuclear Information System (INIS)

    Kowarski, L.

    1955-01-01

    Countries with large stock of fissile material and producing large quantity of nuclear pure 235 U and 239 Pu are able to allocate part of the stock to non military research. For countries with low stock of fissile material, all the stock is allocated to military research. An economical and technical solution has to be find to dedicate a part of fissile material to non military research and develop the atomic energy industry. It stated the industrial and economical problems and in particular the choice between the use of enriched fuel with high refining cost or depleted fuel with low production cost. It discusses of four possible utilizations of the natural resources: reactors functioning with pure fissile material ( 235 U or 239 Pu) or concentrated material ( 235 U mixed with small quantities of 238 U after an incomplete isotopic separation), breeder reactors functioning with enriched material mixed with 238 U or Thorium placed in an appropriate spatial distribution to allow neutrons beam to activate 238 U or Thorium with the regeneration of fissile material in 239 Pu, reactors using natural uranium or low enriched uranium can also produce Plutonium with less efficiency than breeder reactors and the last solution being the use of natural uranium with the only scope of energy production and no production of secondary fissile material. The first class using pure fissile material has a low energy efficiency and is used only by large fissile material stock countries to accumulate energy in small size fuel for nuclear engines researches for submarines and warships. The advantage of the second class of reactors, breeder reactors, is that they produce energy and plutonium. Two type of breeder reactor are considered: breeder reactor using pure fissile material and 238 U or breeder reactor using the promising mixture of pure fissile material and Thorium. Different projects are in phase of development in United States, England and Scotland. The third class of reactor using

  20. Criticality safety margins for mixtures of fissionable materials

    International Nuclear Information System (INIS)

    Williamson, T.G.; Mincey, J.F.

    1992-01-01

    In the determination of criticality safety margins, approximations for combinations of fissile and fissionable isotopes are sometimes used that go by names such as the rule of fractions or equivalency relations. Use of the rule of fractions to ensure criticality safety margins was discussed in an earlier paper. The purpose of this paper is to correct errors and to clarify some of the implications. Deviations of safety margins from those calculated by the rule of fractions are still noted; however, the deviations are less severe. Caution in applying such rules is still urged. In general, these approximations are based on American National Standard ANSI/ANS-8.15, Sec. 5.2. This section allows that ratios of material masses to their limits may be summed for fissile nuclides in aqueous solutions. It also allows the addition of nonfissile nuclides if an aqueous moderator is present and addresses the effects of infinite water or equivalent reflector. Water-reflected binary combinations of aqueous solutions of fissile materials, as well as binary combinations of fissile and fissionable metals, were considered. Some combinations were shown to significantly decrease the margin of subcriticality compared to the single-unit margins. In this study, it is confirmed that some combinations of metal units in an optimum geometry may significantly decrease the margin of subcriticality. For some combinations of aqueous solutions of fissile materials, the margin of subcriticality may also be reduced by very small amounts. The conclusion of Ref. 1 that analysts should be careful in applying equivalency relations for combining materials remains valid and sound advice. The ANSI/ANS standard, which allows the use of ratios of masses to their limits, applies to aqueous, fully water-reflected, single-unit solutions. Extensions to other situations should be considered with extreme care

  1. 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

  2. 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)

  3. 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)

  4. 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.

  5. 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

  6. 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

  7. Test and evaluation of computerized nuclear material accounting methods. Final report

    International Nuclear Information System (INIS)

    1995-01-01

    In accordance with the definition of a Material Balance Area (MBA) as a well-defined geographical area involving an Integral operation, the building housing the BFS-1 and BFS-1 critical facilities is considered to consist of one MBA. The BFS materials are in the form of small disks clad in stainless steel and each disk with nuclear material has its own serial number. Fissile material disks in the BFS MBA can be located at three key monitoring points: BFS-1 facility, BFS-2 facility and main storage of BFS fissile materials (storage 1). When used in the BFS-1 or BFS-2 critical facilities, the fissile material disks are loaded in tubes (fuel rods) forming critical assembly cores. The following specific features of the BFS MBA should be taken into account for the purpose of computerized accounting of nuclear material: (1) very large number of nuclear material items (about 70,000 fissile material items); and (2) periodically very intensive shuffling of nuclear material items. Requirements for the computerized system are determined by basic objectives of nuclear material accounting: (1) providing accurate information on the identity and location of all items in the BFS material balance area; (2) providing accurate information on location and identity of tamper-indicating devices; (3) tracking nuclear material inventories; (4) issuing periodic reports; (5) assisting with the detection of material gains or losses; (6) providing a history of nuclear material transactions; (7) preventing unauthorized access to the system and data falsification. In August 1995, the prototype computerized accounting system was installed on the BFS facility for trial operation. Information on two nuclear material types was entered into the data base: weapon-grade plutonium metal and 36% enriched uranium dioxide. The total number of the weapon-grade plutonium disks is 12,690 and the total number of the uranium dioxide disks is 1,700

  8. 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.

  9. 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

  10. Computerized real-time materials accountability system for safeguards material control

    International Nuclear Information System (INIS)

    Spencer, W.F.; Affel, R.G.; Austin, H.C.; Nichols, J.P.; Stoutt, B.H.; Wachter, J.W.

    1975-01-01

    A real-time, computer-based system is described which provides safeguards material control at the Oak Ridge National Laboratory. Originally installed in 1972 to provide computerized real-time fissile materials accountability for criticality control purposes, the system has been expanded to provide accountability of all source and nuclear materials (SNM) and to utilize the on-line inventory files in support of the Laboratory physical protection and surveillance procedures. (auth)

  11. 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)

  12. 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

  13. 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

  14. 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)

  15. The environmental assessment of nuclear materials disposition options: A transportation perspective

    International Nuclear Information System (INIS)

    Wilson, R.K.; Clauss, D.B.; Moyer, J.W.

    1994-01-01

    The US Department of Energy has undertaken a program to evaluate and select options for the long-term storage and disposition of fissile materials declared surplus to defense needs as a result of the end of the Cold War. The transport of surplus fissile material will be an important and highly visible aspect of the environmental impact studies and other planning documents required for implementation of the disposition options. This report defines the roles and requirements for transportation of fissile materials in the program, and discusses an existing methodology for determining the environmental impact in terms of risk. While it will be some time before specific alternatives are chosen that will permit the completion of detailed risk calculations, the analytical models for performing the probabilistic risk assessments already exist with much of the supporting data related to the transportation system. This report summarizes the various types of data required and identifies sources for that data

  16. LIFE Materials: Overview of Fuels and Structural Materials Issues Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    Farmer, J

    2008-09-08

    blanket in a fusion-fission hybrid system is subcritical, a LIFE engine can burn any fertile or fissile nuclear material, including un-enriched natural or depleted U and SNF, and can extract a very high percentage of the energy content of its fuel resulting in greatly enhanced energy generation per metric ton of nuclear fuel, as well as nuclear waste forms with vastly reduced concentrations of long-lived actinides. LIFE engines could thus provide the ability to generate vast amounts of electricity while greatly reducing the actinide content of any existing or future nuclear waste and extending the availability of low cost nuclear fuels for several thousand years. LIFE also provides an attractive pathway for burning excess weapons Pu to over 99% FIMA (fission of initial metal atoms) without the need for fabricating or reprocessing mixed oxide fuels (MOX). Because of all of these advantages, LIFE engines offer a pathway toward sustainable and safe nuclear power that significantly mitigates nuclear proliferation concerns and minimizes nuclear waste. An important aspect of a LIFE engine is the fact that there is no need to extract the fission fuel from the fission blanket before it is burned to the desired final level. Except for fuel inspection and maintenance process times, the nuclear fuel is always within the core of the reactor and no weapons-attractive materials are available outside at any point in time. However, an important consideration when discussing proliferation concerns associated with any nuclear fuel cycle is the ease with which reactor fuel can be converted to weapons usable materials, not just when it is extracted as waste, but at any point in the fuel cycle. Although the nuclear fuel remains in the core of the engine until ultra deep actinide burn up is achieved, soon after start up of the engine, once the system breeds up to full power, several tons of fissile material is present in the fission blanket. However, this fissile material is widely dispersed

  17. 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

  18. 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)

  19. 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.

  20. 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

  1. 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

  2. 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

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

    International Nuclear Information System (INIS)

    Nakajima, Ken; Ohnishi, Nobuaki

    1986-01-01

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

  4. Identification of High-Z Materials With Photoneutrons Driven by a Low-Energy Electron Linear Accelerator

    Science.gov (United States)

    Yang, Yigang; Zhang, Zhi; Chen, Huaibi; Li, Yulan; Li, Yuanjing

    2017-07-01

    Contraband-detection systems can use X-rays and photoneutrons delivered from the same 7-MeV electron linear accelerator (e-LINAC) to stimulate and extract information from inspected materials. The X-ray attenuation information is used to measure the mass thickness, which is combined with the photoneutron attenuation information to categorize inspected materials as common organic materials, metals, and heavy metals. Once a heavy metal is found, the beta-delayed neutrons stimulated by the (γ,fission) reaction are measured by a polyethylene-moderated 3He counter to clarify if the material is fissile. The presence of neutron events 2000 μs after the X-ray pulse confirms the existence of the fissile material. The isotopes in the material are then identified using the time-of-flight method to analyze the resonant attenuation of the fissile material to the 10-1-102 eV photoneutrons emitted from and thermalized by the D2O photonto-neutron convertor, which converts X-rays to photoneutrons. Eight high-Z simulants are tested to confirm the feasibility of identifying the isotopes from the photoneutron resonance. The underlying principles and experimental results are discussed.

  5. 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

  6. 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

  7. 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.

  8. Chemical, mass spectrometric, and spectrochemical analysis of nuclear-grade mixed oxides [(U,Pu)O2

    International Nuclear Information System (INIS)

    Anon.

    1981-01-01

    Mixed oxide, a mixture of uranium and plutonium oxides, is used as a nuclear-reactor fuel in the form of pellets. The plutonium content may be up to 10 wt %, and the diluent uranium may be of any U-235 enrichment. In order to be suitable for use as a nuclear fuel, the material must meet certain criteria for combined uranium and plutonium content, effective fissile content, and impurity content. Analytical procedures used to determine if mixed oxides comply with specifications are: uranium by controlled-potential coulometry; plutonium by controlled-potential coulometry; plutonium by amperometric titration with iron (II); nitrogen by distillation spectrophotometry using Nessler reagent; carbon (total) by direct combustion-thermal-conductivity; total chlorine and fluorine by pyrohydrolysis; sulfur by distillation-spectrophotometry; moisture by the coulometric, electrolytic moisture analyzer; isotopic composition by mass spectrometry; rare earths by copper spark spectroscopy; trace impurities by carrier distillation spectroscopy; impurities by spark-source mass spectrography; total gas in reactor-grade mixed dioxide pellets; tungsten by dithiol-spectrophotometry; rare earth elements by spectroscopy; plutonium-238 isotopic abundance by alpha spectrometry; uranium and plutonium isotopic analysis by mass spectrometry; oxygen-to-metal atom ratio by gravimetry

  9. 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

  10. The regulation concerning transportation of radioactive materials by vehicles

    International Nuclear Information System (INIS)

    1978-01-01

    The Regulation is established on the basis of The law for the regulations of nuclear source materials, nuclear fuel materials and reactors'' and the ''Law for the prevention of radiation injuries due to radioisotopes.'' The prescriptions cover the transport of radioactive materials by railway, street rail way, ropeway, trolley buses, motorcars and light vehicles. Terms are explained, such as nuclear fuel materials, radioisotopes, radioactive substances, transported radioactive things, transported fissile things, vehicles, containers, exclusive loading, surrounding inspection area. Four types of transported radioactive things are specified, L and A types being less dangerous and BM and BU being more dangerous. Transported fissile things are classified to three kinds according to the safety to criticality of such things. Transported radioactive things except those of L type and containers with transported fissile things shall not be loaded or unloaded at the places where persons other than those concerned come in usually. Loading and unloading of such things shall be carried out so that the safety of such things is not injured. The maximum dose rate of radiation of the containers with transported radioactive things shall not be more than 200 millirem per hour on the surface and 10 millirem per hour at the distance of 1 meter. Specified transported radioactive things shall be particularly marked by the letter of ''radioactive'' or other signs indicating as such. (Okada, K.)

  11. 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

  12. 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)

  13. 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.

  14. 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

  15. Design of active-neutron fuel rod scanner

    International Nuclear Information System (INIS)

    Griffith, G.W.; Menlove, H.O.

    1996-01-01

    An active-neutron fuel rod scanner has been designed for the assay of fissile materials in mixed oxide fuel rods. A 252 Cf source is located at the center of the scanner very near the through hole for the fuel rods. Spontaneous fission neutrons from the californium are moderated and induce fissions within the passing fuel rod. The rod continues past a combined gamma-ray and neutron shield where delayed gamma rays above 1 MeV are detected. We used the Monte Carlo code MCNP to design the scanner and review optimum materials and geometries. An inhomogeneous beryllium, graphite, and polyethylene moderator has been designed that uses source neutrons much more efficiently than assay systems using polyethylene moderators. Layers of borated polyethylene and tungsten are used to shield the detectors. Large NaI(Tl) detectors were selected to measure the delayed gamma rays. The enrichment zones of a thermal reactor fuel pin could be measured to within 1% counting statistics for practical rod speeds. Applications of the rod scanner include accountability of fissile material for safeguards applications, quality control of the fissile content in a fuel rod, and the verification of reactivity potential for mixed oxide fuels. (orig.)

  16. 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)

  17. 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

  18. Testing in support of transportation of residues in the pipe overpack container

    International Nuclear Information System (INIS)

    Ammerman, D.J.; Bobbe, J.G.; Arviso, M.; Bronowski, D.R.

    1997-04-01

    The disposition of the large back-log of plutonium residues at the Rocky Flats Environmental Technology Site (Rocky Flats) will require interim storage and subsequent shipment to a waste repository. Current plants call for disposal at the Waste Isolation Pilot Plant (WIPP) and the transportation to WIPP in the TRUPACT-II. The transportation phase will require the residues to be packaged in a container that is more robust than a standard 55-gallon waste drum. Rocky Flats has designed the Pipe Overpack Container to meet this need. The tests described here were performed to qualify the Pipe Overpack Container as a waste container for shipment in the TRUPACT-II. Using a more robust container will assure the fissile materials in each container can not be mixed with the fissile material from the other containers and will provide criticality control. This will allow an increase in the payload of the TRUPACT-II from 325 fissile gram equivalents to 2,800 fissile gram equivalents

  19. 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

  20. 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

  1. Preliminary concepts for materials measurement and accounting in critical facilities

    International Nuclear Information System (INIS)

    Cobb, D.D.; Sapir, J.L.

    1978-01-01

    Preliminary concepts are presented for improved materials measurement and accounting in large critical facilities. These concepts will be developed as part of a study that will emphasize international safeguarding of critical facilities. The major safeguards problem is the timely verification of in-reactor inventory during periods of reactor operation. This will require a combination of measurement, statistical sampling, and data analysis techniques. Promising techniques include integral measurements of reactivity and other reactor parameters that are sensitive to the total fissile inventory, and nondestructive assay measurements of the fissile material in reactor fuel drawers and vault storage canisters coupled with statistical sampling plans tailored for the specific application. The effectiveness of proposed measurement and accounting strategies will be evaluated during the study

  2. Mobility and Criticality of Plutonium in a Repository

    International Nuclear Information System (INIS)

    Kienzler, Bernhard; Loida, Andreas; Maschek, Werner; Rineiski, Andrei

    2003-01-01

    In an underground repository for spent fuel, criticality is excluded initially by compliance with the disposal conditions. In the long term, critical accumulations of fissile material can be formed only by mobilization of uranium and plutonium from the waste forms and subsequent precipitation or sorption of these elements. This paper presents an overview of mechanisms relevant for mobilization and possible accumulation of U and Pu from disposed mixed-oxide fuel elements. Concentrations of fissile materials observed in laboratory corrosion experiments together with model approaches are applied to determine the degree of fissile material accumulation and the risk of a sustained nuclear chain reaction. A prerequisite of criticality in a repository is an accumulation of fissile materials. Since geometry, moderation, and neutron absorption properties cannot be forecast, the neutron multiplication factor k inf is used (instead of k eff ) as a measure of the incidence of criticality. The factor k inf is derived for several scenarios. Required critical masses and critical volumes are evaluated.The accumulation of Pu onto solids is considered, and it is shown how selective enrichment of Pu and U may affect the risk of criticality. It is also shown that the criterion for criticality would be met only in the unrealistic case of selective sorption of 239 Pu. Realistic sorption densities are too low to provide sufficient accumulation of fissile materials for criticality. This is particularly true if high Cl concentrations are present

  3. 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.)

  4. 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.

  5. 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.

  6. Nuclear materials management for safety and efficiency

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1965-12-15

    The use of nuclear materials in industrial processes presents management with some special problems which are peculiar to the atomic energy industry. If reactor fuel costs are to be kept low, too, each fuel element must yield the maximum economic 'bum-up' before it is withdrawn from service, and this calls for reliable non-destructive methods of measurement of 'burn-up' and appropriate records and fuel-changing schedules. The special hazards of radioactive materials call for special precautions and appropriate systems of handling and storage. A further danger unique to atomic energy is that of criticality - the possibility that an excessive concentration of fissile material may result in a chain reaction. Every part of the processing plant must be surveyed and checked to ensure that there is no build-up of fissile residues; in storage or transit there must be no aggregation of small lots. In the nuclear energy industry, too, the standards of purity required are much higher than in most other large-scale operation, so that stringent quality checks are needed

  7. 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

  8. 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.

  9. 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.

  10. 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)

  11. Manufacturing method for fuel assembly

    International Nuclear Information System (INIS)

    Yamaguchi, Takashi.

    1997-01-01

    In an FBR type reactor, uranium/plutonium mixed oxide fuels (MOX fuels) are used. Nuclear fuel materials containing uranium and plutonium are filled to a portion or all of a plurality of fuel rods. In this case, an equivalent fissile coefficient (B) based on a combustion guarantee method defined by the formula: (B) = (M) · (F) is determined. (M) is a combustion matrix constituted based on the solution of equation of combustion which is a differential equation representing change with time of each of nuclear fuel materials during combustion. (F) is an equivalent fissile coefficient based on a reactivity keeping method which is a coefficient representing a reactivity worth equivalent with plutonium-239. The content of each of the nuclear fuel materials is determined so that the effective multiplication factor at the final stage of the operation cycle is substantially constant by using the equivalent fissile coefficient (B) based on the combustion guarantee method. (I.N.)

  12. 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.

  13. Safety evaluation for packaging (onsite) concrete-lined waste packaging

    Energy Technology Data Exchange (ETDEWEB)

    Romano, T.

    1997-09-25

    The Pacific Northwest National Laboratory developed a package to ship Type A, non-transuranic, fissile excepted quantities of liquid or solid radioactive material and radioactive mixed waste to the Central Waste Complex for storage on the Hanford Site.

  14. Safety evaluation report for packaging (onsite) concrete-lined waste packaging

    International Nuclear Information System (INIS)

    Romano, T.

    1997-01-01

    The Pacific Northwest National Laboratory developed a package to ship Type A, non-transuranic, fissile excepted quantities of liquid or solid radioactive material and radioactive mixed waste to the Central Waste Complex for storage on the Hanford Site

  15. 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)

  16. 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.

  17. Law of substitution for mixed arrays

    International Nuclear Information System (INIS)

    Koudelka, A.J.

    1987-01-01

    The nuclear safety justification of a mixed array of dissimilar fissile units of metal units and dilute solution units, according to Clayton, has been a persistent and nagging problem. Dissimilar uranium metal or dissimilar uranium solution units in a mixed array can also create a modeling nightmare for the nuclear criticality safety engineer. Now, a calculational method known as the Law of Substitution has been developed to ensure that the k/sub eff/ of an array of uranium metal and uranium solution units will satisfy any k/sub eff/ limit set by the nuclear safety engineer. The nuclear criticality safety engineer can utilize the Law of Substitution to safely mix or substitute different uranium metal units, different uranium solution units, and more importantly, uranium metal and dilute UO 2 solution units in an array. The Law of Substitution is as follows: (1) calculate the k/sub eff/ of each unit type in its own infinite planar array. (2) Determine the edge-to-edge spacing of the infinite planar array of each type of unit to satisfy a desired k/sub eff/. (3) Select the largest edge-to-edge spacing from among the similar units in their infinite planar arrays and use that spacing for the finite or infinite planar array of mixed units

  18. SOR/89-426, Transport Packaging of Radioactive Materials Regulations, amendment

    International Nuclear Information System (INIS)

    1989-01-01

    These Regulations of 24 August 1989 amend the Transport Packaging of Radioactive Materials Regulations by clarifying the text and specifying certain requirements. In particular certain definitions have been replaced, namely those of ''Fissile Class III package'' and ''Special form radioactive material''. Also, this latter material may not be carried without a certificate attesting that it meets the requirements of the Regulations. (NEA)

  19. 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

  20. 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

  1. Kalman filter analysis of delayed neutron nondestructive assay measurements

    International Nuclear Information System (INIS)

    Aumeier, S. E.

    1998-01-01

    The ability to nondestructively determine the presence and quantity of fissile and fertile nuclei in various matrices is important in several nuclear applications including international and domestics safeguards, radioactive waste characterization and nuclear facility operations. Material irradiation followed by delayed neutron counting is a well known and useful nondestructive assay technique used to determine the fissile-effective content of assay samples. Previous studies have demonstrated the feasibility of using Kalman filters to unfold individual isotopic contributions to delayed neutron measurements resulting from the assay of mixes of uranium and plutonium isotopes. However, the studies in question used simulated measurement data and idealized parameters. We present the results of the Kalman filter analysis of several measurements of U/Pu mixes taken using Argonne National Laboratory's delayed neutron nondestructive assay device. The results demonstrate the use of Kalman filters as a signal processing tool to determine the fissile and fertile isotopic content of an assay sample from the aggregate delayed neutron response following neutron irradiation

  2. 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)

  3. 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.

  4. Nuclear data of the major actinide fuel materials

    Energy Technology Data Exchange (ETDEWEB)

    Poenitz, W.P.; Saussure, G. De

    1984-01-01

    The effect of nuclear data of the major actinide fuel materials on the design accuracy, economics and safety of nuclear power systems is discussed. Since most of the data are measured relative to measurement standards, in particular the fission cross-section of /sup 235/U, data must be examined to ensure that absolute measurements and relative measurements are correctly handled. Nuclear data of fissile materials, fertile materials and minor plutonium isotopes are discussed.

  5. Method and device for fabricating dispersion fuel comprising fission product collection spaces

    Science.gov (United States)

    Shaber, Eric L; Fielding, Randall S

    2015-05-05

    A method of fabricating a nuclear fuel comprising a fissile material, one or more hollow microballoons, a phenolic resin, and metal matrix. The fissile material, phenolic resin and the one or more hollow microballoons are combined. The combined fissile material, phenolic resin and the hollow microballoons are heated sufficiently to form at least some fissile material carbides creating a nuclear fuel particle. The resulting nuclear fuel particle comprises one or more fission product collection spaces. In a preferred embodiment, the fissile material, phenolic resin and the one or more hollow microballoons are combined by forming the fissile material into microspheres. The fissile material microspheres are then overcoated with the phenolic resin and microballoon. In another preferred embodiment, the fissile material, phenolic resin and the one or more hollow microballoons are combined by overcoating the microballoon with the fissile material, and phenolic resin.

  6. The prospect of uranium nitride (UN) and mixed nitride fuel (UN-PuN) for pressurized water reactor

    International Nuclear Information System (INIS)

    Syarifah, Ratna Dewi; Suud, Zaki

    2015-01-01

    Design study of small Pressurized Water Reactors (PWRs) core loaded with uranium nitride fuel (UN) and mixed nitride fuel (UN-PuN), Pa-231 as burnable poison, and Americium has been performed. Pa-231 known as actinide material, have large capture cross section and can be converted into fissile material that can be utilized to reduce excess reactivity. Americium is one of minor actinides with long half life. The objective of adding americium is to decrease nuclear spent fuel in the world. The neutronic analysis results show that mixed nitride fuel have k-inf greater than uranium nitride fuel. It is caused by the addition of Pu-239 in mixed nitride fuel. In fuel fraction analysis, for uranium nitride fuel, the optimum volume fractions are 45% fuel fraction, 10% cladding and 45% moderator. In case of UN-PuN fuel, the optimum volume fractions are 30% fuel fraction, 10% cladding and 60% coolant/ moderator. The addition of Pa-231 as burnable poison for UN fuel, enrichment U-235 5%, with Pa-231 1.6% has k-inf more than one and excess reactivity of 14.45%. And for mixed nitride fuel, the lowest value of reactivity swing is when enrichment (U-235+Pu) 8% with Pa-231 0.4%, the excess reactivity value 13,76%. The fuel pin analyze for the addition of Americium, the excess reactivity value is lower than before, because Americium absorb the neutron. For UN fuel, enrichment U-235 8%, Pa-231 1.6% and Am 0.5%, the excess reactivity is 4.86%. And for mixed nitride fuel, when enrichment (U-235+Pu) 13%, Pa-231 0.4% and Am 0.1%, the excess reactivity is 11.94%. For core configuration, it is better to use heterogeneous than homogeneous core configuration, because the radial power distribution is better

  7. The prospect of uranium nitride (UN) and mixed nitride fuel (UN-PuN) for pressurized water reactor

    Science.gov (United States)

    Syarifah, Ratna Dewi; Suud, Zaki

    2015-09-01

    Design study of small Pressurized Water Reactors (PWRs) core loaded with uranium nitride fuel (UN) and mixed nitride fuel (UN-PuN), Pa-231 as burnable poison, and Americium has been performed. Pa-231 known as actinide material, have large capture cross section and can be converted into fissile material that can be utilized to reduce excess reactivity. Americium is one of minor actinides with long half life. The objective of adding americium is to decrease nuclear spent fuel in the world. The neutronic analysis results show that mixed nitride fuel have k-inf greater than uranium nitride fuel. It is caused by the addition of Pu-239 in mixed nitride fuel. In fuel fraction analysis, for uranium nitride fuel, the optimum volume fractions are 45% fuel fraction, 10% cladding and 45% moderator. In case of UN-PuN fuel, the optimum volume fractions are 30% fuel fraction, 10% cladding and 60% coolant/ moderator. The addition of Pa-231 as burnable poison for UN fuel, enrichment U-235 5%, with Pa-231 1.6% has k-inf more than one and excess reactivity of 14.45%. And for mixed nitride fuel, the lowest value of reactivity swing is when enrichment (U-235+Pu) 8% with Pa-231 0.4%, the excess reactivity value 13,76%. The fuel pin analyze for the addition of Americium, the excess reactivity value is lower than before, because Americium absorb the neutron. For UN fuel, enrichment U-235 8%, Pa-231 1.6% and Am 0.5%, the excess reactivity is 4.86%. And for mixed nitride fuel, when enrichment (U-235+Pu) 13%, Pa-231 0.4% and Am 0.1%, the excess reactivity is 11.94%. For core configuration, it is better to use heterogeneous than homogeneous core configuration, because the radial power distribution is better.

  8. 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

  9. Material control and accountability aspects of safeguards for the USA 233U/Th fuel recycle plant

    International Nuclear Information System (INIS)

    Carpenter, J.A. Jr.; McNeany, S.R.; Angelini, P.; Holder, N.D.; Abraham, L.

    1978-01-01

    The materials control and accountability aspects of the reprocessing and refabrication of a conceptual large-scale HTGR fuel recycle plant have been discussed. Two fuel cycles were considered. The traditional highly enriched uranium cycle uses an initial or makeup fuel element with a fissile enrichment of 93% 235 U. The more recent medium enriched uranium cycle uses initial or makeup fuel elements with a fissile enrichment less than 20% 235 U. In both cases, 233 U bred from the fertile thorium is recycled. Materials control and accountability in the plant will be by means of a real-time accountability method. Accountability data will be derived from monitoring of total material mass through the processes and a system of numerous assays, both destructive and nondestructive

  10. Dual-energy X-ray radiography for automatic high-Z material detection

    International Nuclear Information System (INIS)

    Chen Gongyin; Bennett, Gordon; Perticone, David

    2007-01-01

    There is an urgent need for high-Z material detection in cargo. Materials with Z > 74 can indicate the presence of fissile materials or radiation shielding. Dual (high) energy X-ray material discrimination is based on the fact that different materials have different energy dependence in X-ray attenuation coefficients. This paper introduces the basic physics and analyzes the factors that affect dual-energy material discrimination performance. A detection algorithm is also discussed

  11. Aims and methods of nuclear materials management

    International Nuclear Information System (INIS)

    Leven, D.; Schier, H.

    1979-05-01

    Whilst international safeguarding of fissile materials against abuse has been the subject of extensive debate, little public attention has so far been devoted to the internal security of these materials. All countries using nuclear energy for peaceful purposes have laid down appropriate regulations. In the Federal Republic of Germany safeguards are required, for instance, by the Atomic Energy Act, and are therefore a prerequisite for licensing. The aims and methods of national nuclear materials management are contrasted with viewpoints on international safeguards

  12. 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

  13. Russian-U.S. joint program on the safe management of nuclear materials

    International Nuclear Information System (INIS)

    Witmer, F.E.; Krumpe, P.F.; Carlson, D.D.

    1997-12-01

    The Russian-US joint program on the safety of nuclear materials was initiated in response to the 1993 Tomsk-7 accident. The bases for this program are the common technical issues confronting the US and Russia in the safe management of excess weapons grade nuclear materials. The US and Russian weapons dismantlement process is producing hundreds of tons of excess Pu and HEU fissile materials. The US is on a two path approach for disposition of excess Pu: (1) use Pu in existing reactors and/or (2) immobilize Pu in glass or ceramics followed by geologic disposal. Russian plans are to fuel reactors with excess Pu. US and Russia are both converting and blending HEU into LEU for use in existing reactors. Fissile nuclear materials storage, handling, processing, and transportation will be occurring in both countries for tens of years. A table provides a history of the major events comprising the Russian-US joint program on the safety of nuclear materials. A paper delineating program efforts was delivered at the SPECTRUM '96 conference. This paper provides an update on program activities since then

  14. 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.

  15. Characterization of the solid low level mixed waste inventory for the solid waste thermal treatment activity - III

    Energy Technology Data Exchange (ETDEWEB)

    Place, B.G., Westinghouse Hanford

    1996-09-24

    The existing thermally treatable, radioactive mixed waste inventory is characterized to support implementation of the commercial, 1214 thermal treatment contract. The existing thermally treatable waste inventory has been identified using a decision matrix developed by Josephson et al. (1996). Similar to earlier waste characterization reports (Place 1993 and 1994), hazardous materials, radionuclides, physical properties, and waste container data are statistically analyzed. In addition, the waste inventory data is analyzed to correlate waste constituent data that are important to the implementation of the commercial thermal treatment contract for obtaining permits and for process design. The specific waste parameters, which were analyzed, include the following: ``dose equivalent`` curie content, polychlorinated biphenyl (PCB) content, identification of containers with PA-related mobile radionuclides (14C, 12 79Se, 99Tc, and U isotopes), tritium content, debris and non-debris content, container free liquid content, fissile isotope content, identification of dangerous waste codes, asbestos containers, high mercury containers, beryllium dust containers, lead containers, overall waste quantities, analysis of container types, and an estimate of the waste compositional split based on the thermal treatment contractor`s proposed process. A qualitative description of the thermally treatable mixed waste inventory is also provided.

  16. Nuclear materials identification by photon interrogation

    International Nuclear Information System (INIS)

    Pozzi, S.A.; Monville, M.; Padovani, E.

    2005-01-01

    We describe a preliminary modification to the Monte Carlo codes MCNP-X and MCNP-PoliMi that is aimed at simulating the neutron and photon field generated by interrogating fissile (and non-fissile) material with a high energy photon source. Photo-atomic and photo-nuclear collisions are modeled, with particular emphasis on the generation of secondary particles that are emitted as a result of these interactions. The simulations can be used to design and analyze measurements that are performed in a wide variety of scenarios. An application of the methodology to the interrogation of packages on a luggage belt conveyor is presented. Preliminary results show that it is possible to detect 5 Kg of highly enriched uranium in a package by measuring the correlation function between 2 detectors. This correlation function is based on the detection of prompt radiation from photonuclear events

  17. 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

  18. 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)

  19. An evaluation of the deployment of AIROX-recycled fuel in pressurized water reactors

    International Nuclear Information System (INIS)

    Jahshan, S.N.; McGeehan, T.J.

    1994-01-01

    An analytical evaluation is made of the pressurized water reactor (PWR) in-core performance of recycled light water reactor fuel that has been Atomics International reduction oxidation (AIROX) reprocessed and reenriched with fissile materials. The neutronics performance is shown to lie within the neutronics performance of existing high-performance and high-burnup fuels. Three AIROX-recycled fuels are compared with a high-burnup virgin fuel and an equivalent mixed-oxide (MOX) fuel. The AIROX-recycled fuel neutronics performance lies consistently between the virgin and the MOX fuel for both the pin power peaking and the reactivity response characteristics in PWRs. Among the attractive features of AIROX-recycled fuel is that it can optimize fissile and fertile fuel use, minimize final fuel disposal impact on the environment, and provide energy in the process of denaturing weapons-grade fissile materials. The fuel material performance may be anticipated from high-burnup virgin fuel and from MOX fuel performance. Recommendations for lead rod testing and for optimization of the AIROX-processing and resintering techniques are made

  20. 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.

  1. 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

  2. Accelerator-based approach experiments for remote identification of fissionable and other materials

    International Nuclear Information System (INIS)

    Chuvilo, I.V.; Danilov, M.M.; Katarzhnov, Yu.D.; Kushin, V.V.; Nedopekin, V.G.; Plotnikov, S.V.; Rogov, V.I.

    1998-01-01

    Recently there has been a great deal of interest in studying possible methods for remote non-destructive material composition testing, for example, for cargo identification at transportation, neutron logging etc., by means of nuclear detection (D.R. Brown, T. Gozani (1995)). Of current concern are the applications of pulsed fast neutron analysis in determining the composition of fissile objects (I.I. Zaliubovskiy et al. (1993)). In this paper the observed experimental results are discussed indicating the possibility of practical realization of the method for remote material identification. The approach is based on measuring gamma ray spectra from an object to be examined after its irradiation with short neutron pulses produced by an accelerator. The obtained time and energy gamma spectra are used for material inspection. The information is obtained by using time-of-flight (TOF) analysis between the accelerator pulse and the arrival of gamma rays in NaI detectors located far enough from an object to be examined. The method seems to be the most effective for fissile materials identification. (orig.)

  3. 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

  4. Measures for prevention illicit trafficking of nuclear and radioactive materials

    International Nuclear Information System (INIS)

    Strezov, A.

    2002-01-01

    Full text: In the early 1990ies the number of illicit trafficking cases with nuclear material and radioactive sources began to appear in the press more often than before. This fact became of great concern among international organizations and different states that the nuclear material subjected to trafficking might become in possession of rogue states and be implicated in weapons production or that stolen radioactive sources may cause health and safety effects to the population or to the environment. The creation and proposition of a model scheme procedure for the developing countries is important for starting the initial process of preventing and combating the illicit traffic of nuclear materials. Particular efforts have been directed for the protection of fissile materials. The reported incidents for diversion of nuclear materials have raised the problem of potential nuclear terrorism and also for countries of proliferation to take a short cut to the bomb. There is a need of rapid implementation of comprehensive, mutually reinforcing strategy to control the existing stockpiles of fissile material and to lower the future production and use of such materials. The illicit traffic of nuclear materials is a new threat, which requires new efforts, new approaches and coordination of services and institutions and even new legislation. The propositions of a model-procedure will allow better and quicker upgrade of developing countries capabilities for combating illicit nuclear trafficking. (author)

  5. 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.

  6. 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.

  7. Nuclear materials accountancy in an industrial MOX fuel fabrication plant safeguards versus commercial aspects

    International Nuclear Information System (INIS)

    Canck, H. de; Ingels, R.; Lefevre, R.

    1991-01-01

    In a modern MOX Fuel Fabrication Plant, with a large throughput of nuclear materials, computerized real-time accountancy systems are applied. Following regulations and prescriptions imposed by the Inspectorates EURATOM-IAEA, the State and also by internal plant safety rules, the accountancy is kept in plutonium element, uranium element and 235 U for enriched uranium. In practice, Safeguards Authorities are concerned with quantities of the element (U tot , Pu tot ) and to some extent with its fissile content. Custom Authorities are for historical reasons, interested in fissile quantities (U fiss , Pu fiss ) whereas owners wish to recover the energetic value of their material (Pu equivalent). Balancing the accountancy simultaneously in all these related but not proportional units is a new problem in a MOX-plant where pool accountancy is applied. This paper indicates possible ways to solve the balancing problem created by these different units used for expressing nuclear material quantities

  8. 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...

  9. 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

  10. 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...

  11. 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.

  12. 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...

  13. 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.

  14. Routine inspection effort required for verification of a nuclear material production cutoff convention

    International Nuclear Information System (INIS)

    Dougherty, D.; Fainberg, A.; Sanborn, J.; Allentuck, J.; Sun, C.

    1996-11-01

    On 27 September 1993, President Clinton proposed open-quotes... a multilateral convention prohibiting the production of highly enriched uranium or plutonium for nuclear explosives purposes or outside of international safeguards.close quotes The UN General Assembly subsequently adopted a resolution recommending negotiation of a non-discriminatory, multilateral, and internationally and effectively verifiable treaty (hereinafter referred to as open-quotes the Cutoff Conventionclose quotes) banning the production of fissile material for nuclear weapons. The matter is now on the agenda of the Conference on Disarmament, although not yet under negotiation. This accord would, in effect, place all fissile material (defined as highly enriched uranium and plutonium) produced after entry into force (EIF) of the accord under international safeguards. open-quotes Productionclose quotes would mean separation of the material in question from radioactive fission products, as in spent fuel reprocessing, or enrichment of uranium above the 20% level, which defines highly enriched uranium (HEU). Facilities where such production could occur would be safeguarded to verify that either such production is not occurring or that all material produced at these facilities is maintained under safeguards

  15. Validation of the 3D transport monte carlo code TRIPOLI-4.3 for moderated and unmoderated metallic fissile media configurations with JEF2.2 and ENDF/B-VI.4 cross section evaluations

    International Nuclear Information System (INIS)

    Gagnier, E.; Lee, Y.K.; Aguiar, L.; Vedrenne, N.

    2003-01-01

    This paper presents an extended validation of TRIPOLI-4.3 covering all the metallic fissile media configurations present in the CEA facilities. More than 300 ICSBEP benchmarks have been calculated with TRIPOLI-4.3 and compared to the experimental results. These benchmarks include high-enriched uranium fissile media and plutonium fissile media with a low content of plutonium 240. The configurations are calculated with continuous-energy cross-section libraries JEF2.2 and ENDF/B-VI.4 and compared to MCNP or SCALE results presented in the ICSBEP reports. (author)

  16. 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

  17. 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)

  18. 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.

  19. 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.

  20. 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

  1. 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

  2. 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

  3. 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

  4. Fission of 209 Bi by 60-270 MeV tagged photons: cross section measurement and analysis of photo fissility

    International Nuclear Information System (INIS)

    Terranova, M.L.; Tavares, O.A.P.

    1996-07-01

    Tagged photons produced by the ROKK-2 facility have been used to measure the photofission cross section of 209 Bi in the energy range 60-270 MeV. Photofission events were detected by using a nuclear fragment detector designed for fission experiments, based on multiwire spark counters. Fissility values have been deduced and compared with available data obtained in other laboratories by using monochromatic photons. These data, together with early measurements obtained near photofission threshold, have been analysed in the framework of a two-step model which considers the primary photo interaction occurring via the quasi-deuteron and/or photo mesonic processes, followed by a mechanism of evaporation-fission competition for the excited residual nucleus. The model was found to reproduce the main experimental features of 209 Bi photo fissility up to 300 MeV. (author). 52 refs., 7 figs., 2 tabs

  5. 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.

  6. 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

  7. Better materials for nuclear energy

    International Nuclear Information System (INIS)

    Banerjee, 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 systems comprising fuels, structural materials, moderators and coolants 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, in-reactor degradation mechanisms, and in-service inspection. 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. New challenges are thrown to material scientists for the development of materials suitable for high temperature reactors, which have a potential for providing primary heat for thermo chemical dissociation of water. Development of several ceramic materials, carbon based materials, dissimilar

  8. 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.

  9. Multi-Detector Analysis System for Spent Nuclear Fuel Characterization

    Energy Technology Data Exchange (ETDEWEB)

    Reber, Edward Lawrence; Aryaeinejad, Rahmat; Cole, Jerald Donald; Drigert, Mark William; Jewell, James Keith; Egger, Ann Elizabeth; Cordes, Gail Adele

    1999-09-01

    The Spent Nuclear Fuel (SNF) Non-Destructive Analysis (NDA) program at INEEL is developing a system to characterize SNF for fissile mass, radiation source term, and fissile isotopic content. The system is based on the integration of the Fission Assay Tomography System (FATS) and the Gamma-Neutron Analysis Technique (GNAT) developed under programs supported by the DOE Office of Non-proliferation and National Security. Both FATS and GNAT were developed as separate systems to provide information on the location of special nuclear material in weapons configuration (FATS role), and to measure isotopic ratios of fissile material to determine if the material was from a weapon (GNAT role). FATS is capable of not only determining the presence and location of fissile material but also the quantity of fissile material present to within 50%. GNAT determines the ratios of the fissile and fissionable material by coincidence methods that allow the two prompt (immediately) produced fission fragments to be identified. Therefore, from the combination of FATS and GNAT, MDAS is able to measure the fissile material, radiation source term, and fissile isotopics content.

  10. Proposal for Analysis of the Safeguarded Nuclear Materials 235U and 239Pu by Delayed Neutrons Technique

    International Nuclear Information System (INIS)

    El-Mongy, S.A.

    2000-01-01

    This paper introduces, describes and initiates a very sensitive and rapid non-destructive technique to be used for analysis of the safeguarded nuclear materials 235 U and 239 Pu. The technique is based on fission of the nuclear material by neutrons and then measuring the delayed neutrons produced from the neutron rich fission products. By this technique, fissile isotope content ( 235 U) can be determined in the presence of the other fissile (e.g. 239 Pu) or fertile isotopes (e.g. 238 U) in fresh and spent fuel. The time consumed for analysis of bulk materials by this technique is only 4 minutes. The method is also used for analysis of uranium in rock, sediment, soil, meteorites, lunar, biological, urine, archaeological, zircon sand and seawater samples. The method enables uranium in a sample to be measured without respect to its oxidation state, organic and inorganic elements

  11. 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

  12. Calculus of a reactor VVER-1000 benchmark; Calcul d'un benchmark de reacteur VVER-1000

    Energy Technology Data Exchange (ETDEWEB)

    Dourougie, C

    1998-07-01

    In the framework of the FMDP (Fissile Materials Disposition Program between the US and Russian, a benchmark was tested. The pin cells contain low enriched uranium (LEU) and mixed oxide fuels (MOX). The calculations are done for a wide range of temperatures and solute boron concentrations, in accidental conditions. (A.L.B.)

  13. Calculus of a reactor VVER-1000 benchmark

    International Nuclear Information System (INIS)

    Dourougie, C.

    1998-01-01

    In the framework of the FMDP (Fissile Materials Disposition Program between the US and Russian, a benchmark was tested. The pin cells contain low enriched uranium (LEU) and mixed oxide fuels (MOX). The calculations are done for a wide range of temperatures and solute boron concentrations, in accidental conditions. (A.L.B.)

  14. 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.

  15. The transport of fuel assemblies. New containers for transport the used nuclear material in Juzbado factory

    International Nuclear Information System (INIS)

    2005-01-01

    Juzbado Manufacturing Facility is designed to be versatile and flexible. It is manufactured different kind of fuel assemblies PWR, BWR and VVER, beginning by the uranium oxide coming from the conversion facilities. The transport of these products (radioactive material fissile) requires the availability of different kind of packages; our models variety is similar to the big manufacturers. It is required a depth knowledge of the licensing process, approvals, manufacturing and handling instruction to be confident. Moreover, the recently changes on the Transport Regulations and the demands for the approval by the Competent Authorities have required the renovation of most of the package designs for the transport of radioactive material fissile worldwide. ENUSA assumed time ago this renovation and it is nowadays in the pick moment of this process. If we also consider the complexity on the management of multimodal international transportations, the Logistic task for the transport of nuclear material associated to the Juzbado factory results in a real changeling area. (Author)

  16. Nuclear reactor for breeding 233U

    International Nuclear Information System (INIS)

    Bohanan, C.S.; Jones, D.H.; Raab, H.F. Jr.; Radkowsky, A.

    1976-01-01

    A light-water-cooled nuclear reactor capable of breeding 233 U for use in a light-water breeder reactor includes physically separated regions containing 235 U fissile material and 238 U fertile material and 232 Th fertile material and 239 Pu fissile material, if available. Preferably the 235 U fissile material and 238 U fertile material are contained in longitudinally movable seed regions and the 239 Pu fissile material and 232 Th fertile material are contained in blanket regions surrounding the seed regions. 1 claim, 5 figures

  17. Sor/89-426, 24 August 1989, transport packaging of radioactive materials regulations, amendment

    International Nuclear Information System (INIS)

    1989-09-01

    These Regulations of 24 September 1983 were amended mainly to clarify the original text and further specify certain requirements. In particular, the definitions of A 1 , A 2 , Fissile Class III package and special Form Radioactive Material have been revoked and replaced by new definitions. Also, a new condition has been added regarding Special Form Radioactive Material. Henceforth, no such material may be transported without a certificate attesting that the material meets the requirements set out in Schedule XII of the Regulations [fr

  18. 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.

  19. 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.)

  20. Experience of air transport of nuclear fuel material in Japan

    International Nuclear Information System (INIS)

    Yamashita, T.; Toguri, D.; Kawasaki, M.

    2004-01-01

    Certified Reference Materials (hereafter called as to CRMs), which are indispensable for Quality Assurance and Material Accountability in nuclear fuel plants, are being provided by overseas suppliers to Japanese nuclear entities as Type A package (non-fissile) through air transport. However, after the criticality accident at JCO in Japan, special law defining nuclear disaster countermeasures (hereafter called as to the LAW) has been newly enforced in June 2000. Thereafter, nuclear fuel materials must meet not only to the existing transport regulations but also to the LAW for its transport

  1. 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

  2. Remote material handling in the Plutonium Immobilization Project. Revision 1

    International Nuclear Information System (INIS)

    Brault, J.R.

    2000-01-01

    With the downsizing of the US and Russian nuclear stockpiles, large quantities of weapons-usable plutonium in the US are being declared excess and will be disposed of by the Department of Energy Fissile Materials Disposition Program. To implement this program, DOE has selected the Savannah River Site (SRS) for the construction and operation of three new facilities: pit disassembly and conversion; mixed oxide fuel fabrication; and plutonium immobilization. The Plutonium Immobilization Project (PIP) will immobilize a portion of the excess plutonium in a hybrid ceramic and glass form containing high level waste for eventual disposal in a geologic repository. The PIP is divided into three distinct operating areas: Plutonium Conversion, First Stage Immobilization, and Second Stage Immobilization. Processing technology for the PIP is being developed jointly by the Lawrence Livermore National Laboratory and Westinghouse Savannah River Company. This paper will discuss development of the automated unpacking and sorting operations in the conversion area, and the automated puck and tray handling operations in the first stage immobilization area. Due to the high radiation levels and toxicity of the materials to be disposed of, the PIP will utilize automated equipment in a contained (glovebox) facility. Most operations involving plutonium-bearing materials will be performed remotely, separating personnel from the radiation source. Source term materials will be removed from the operations during maintenance. Maintenance will then be performed hands on within the containment using glove ports

  3. 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

  4. 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.

  5. 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.

  6. Problems and management of radioactive sources and measures against illicit trafficking of nuclear materials in Bulgaria

    International Nuclear Information System (INIS)

    Strezov, A.

    1998-01-01

    Illicit trafficking of nuclear materials continues to pose a danger to public health and safety and to nuclear non proliferation efforts. The majority of cases so far have involved only small amounts of fissile materials or mainly radioactive sources in Bulgaria. A proper scheme for analysis of seized nuclear materials will be developed based on existing equipment for NDA analysis of nuclear materials supplemented by new system through PHARE project assistance by EU experts. (author)

  7. Material for 258 atom bombs disappeared?

    International Nuclear Information System (INIS)

    Gruemm, H.

    1988-01-01

    In a report published in the news magazine, 'Der Spiegel', it was said that IAEA safeguards obviously had failed, for large amounts of fissile material had disappeared, which could be turned into 258 atomic bombs. The article in this issue of atw by the former Deputy Director General with the IAEA Safeguards Division sketches the background to the assertions made by 'Der Spiegel' and presents an overview of the inspection and verification methods employed by IAEA. (orig./HP) [de

  8. 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

  9. Exemption, exception and other criteria for transport criticality safety

    International Nuclear Information System (INIS)

    Mennerdahl, D.

    2004-01-01

    Many strange concepts, requirements and specifications related to criticality safety are present in the Regulations. Some earlier problems have been corrected but, going back to 1961 and the first edition of the Regulations, it seems as many changes have been to the worse. Fissile material was defined correctly as a material that could consist of or contain fissile nuclides. Materials consisting of pure fissile nuclides don't exist but are important in package designs. 238 Pu was included as a fissile nuclide only as an emergency, because there was no alternative, but this caused some people to think that all nuclides supporting criticality are fissile. Neutron interaction between different (non-identical) packages had to be evaluated, making the transport index or allowable number of packages a credible safety control. That is not true anymore. The 15 gram exception limit for fissile nuclides was combined with a transport mode limit, similar to but more restrictive than the current consignment limit. The confinement system was introduced to help with formulation of a single requirement for safety of the containment system but is becoming something very different. Controls before the first use of a packaging have become controls of the first use of a package, supporting multiple shipments of the same package. The lack of exemption limits for fissile material essentially makes all radioactive materials fissile (all radioactive material contains some fissile atoms). Radioactive material seems to be defined without consideration of the criticality hazard of the material. LSA materials are defined with consideration of criticality, but only relates to quantities in fissile exceptions when other properties can be equally or more important. In July 2004, a number of proposals to IAEA have been submitted by Sweden to improve and expand the criticality safety control of the Regulations. Essential is the introduction of the fissionable nuclide and material concepts in

  10. Exemption, exception and other criteria for transport criticality safety

    Energy Technology Data Exchange (ETDEWEB)

    Mennerdahl, D. [E Mennerdahl Systems, Taeby (Sweden)

    2004-07-01

    Many strange concepts, requirements and specifications related to criticality safety are present in the Regulations. Some earlier problems have been corrected but, going back to 1961 and the first edition of the Regulations, it seems as many changes have been to the worse. Fissile material was defined correctly as a material that could consist of or contain fissile nuclides. Materials consisting of pure fissile nuclides don't exist but are important in package designs. {sup 238}Pu was included as a fissile nuclide only as an emergency, because there was no alternative, but this caused some people to think that all nuclides supporting criticality are fissile. Neutron interaction between different (non-identical) packages had to be evaluated, making the transport index or allowable number of packages a credible safety control. That is not true anymore. The 15 gram exception limit for fissile nuclides was combined with a transport mode limit, similar to but more restrictive than the current consignment limit. The confinement system was introduced to help with formulation of a single requirement for safety of the containment system but is becoming something very different. Controls before the first use of a packaging have become controls of the first use of a package, supporting multiple shipments of the same package. The lack of exemption limits for fissile material essentially makes all radioactive materials fissile (all radioactive material contains some fissile atoms). Radioactive material seems to be defined without consideration of the criticality hazard of the material. LSA materials are defined with consideration of criticality, but only relates to quantities in fissile exceptions when other properties can be equally or more important. In July 2004, a number of proposals to IAEA have been submitted by Sweden to improve and expand the criticality safety control of the Regulations. Essential is the introduction of the fissionable nuclide and material

  11. Institute of Nuclear Materials Management 36. annual meeting: Proceedings. Volume 24

    International Nuclear Information System (INIS)

    Anon.

    1995-01-01

    The sessions at this meeting covered the following topics: international safeguards; physical protection; material control and accountability (MC and A) training; waste overview; nonproliferation and arms control -- US/former Soviet Union; MC and A/physical protection -- SNM storage; spent fuel storage; physical protection -- intrusion sensors and access delay; MC and A -- gamma ray and holdup measurements; MC and A -- neutron based measurements; MC and A/nonproliferation and arms control -- fissile materials disposition; low level waste; packaging and transportation; international safeguards and nonproliferation and arms control -- IAEA verification in the US; international safeguards and physical protection -- video systems; poster and demonstrations; international safeguards and physical protection -- T.I.D.'s, tags and seals; high level wastes and mixed waste; MC and A -- bulk measurements; MC and A and physical protection -- policy; nonproliferation and arms control -- detection and monitoring techniques; nonproliferation and arms control -- US-Russian lab-to-lab safeguards cooperation; MC and A -- information systems; physical protection and MC and A -- vulnerability assessments; waste management -- measurement; international safeguards -- integrated and remote monitoring systems; MC and A -- material control and accountability and indicators; nonproliferation and arms control -- treaties and analysis; nonproliferation and arms control -- plutonium/HEU nonproliferation; MC and A -- measurement control and verification; and international safeguards and nonproliferation and arms control -- environmental monitoring and export control. Separate abstracts were prepared for some of the papers in this volume

  12. 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...

  13. Apparatus and method for identification of matrix materials in which transuranic elements are embedded using thermal neutron capture gamma-ray emission

    Science.gov (United States)

    Close, D.A.; Franks, L.A.; Kocimski, S.M.

    1984-08-16

    An invention is described that enables the quantitative simultaneous identification of the matrix materials in which fertile and fissile nuclides are embedded to be made along with the quantitative assay of the fertile and fissile materials. The invention also enables corrections for any absorption of neutrons by the matrix materials and by the measurement apparatus by the measurement of the prompt and delayed neutron flux emerging from a sample after the sample is interrogated by simultaneously applied neutrons and gamma radiation. High energy electrons are directed at a first target to produce gamma radiation. A second target receives the resulting pulsed gamma radiation and produces neutrons from the interaction with the gamma radiation. These neutrons are slowed by a moderator surrounding the sample and bathe the sample uniformly, generating second gamma radiation in the interaction. The gamma radiation is then resolved and quantitatively detected, providing a spectroscopic signature of the constituent elements contained in the matrix and in the materials within the vicinity of the sample. (LEW)

  14. 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

  15. Hot cell works and related irradiation tests in fission reactor for development of new materials for nuclear application

    International Nuclear Information System (INIS)

    Shikama, Tatsuo

    1999-01-01

    Present status of research works in Oarai Branch, Institute for Materials Research, Tohoku University, utilizing Japan Materials Testing Reactor and related hot cells will be described.Topics are mainly related with nuclear materials studies, excluding fissile materials, which is mainly aiming for development of materials for advanced nuclear systems such as a nuclear fusion reactor. Conflict between traditional and routined procedures and new demands will be described and future perspective is discussed. (author)

  16. Effects of various moderators on the critical mass of plutonium

    International Nuclear Information System (INIS)

    Doherty, A.L.

    1986-01-01

    The fissile material storage tanks in the Hanford Plutonium Critical Mass Laboratory (CML) in Richland, Washington, are presently being upgraded. During the design and planning phase of this modification, criticality analysis was necessary to compare potential moderator/absorber materials used as isolators between tanks. A parameter study was performed to assist in determining the appropriate moderator material to be used in the plutonium nitrate storage tank system in the mix room at the CML. Four moderator/absorber materials were identified as providing adequate isolation between the tanks

  17. 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

  18. 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.

  19. Dossier: transport of radioactive materials

    International Nuclear Information System (INIS)

    Mignon, H.; Brachet, Y.; Turquet de Beauregard, G.; Mauny, G.; Robine, F.; Plantet, F.; Pestel Lefevre, O.; Hennenhofer, G.; Bonnemains, J.

    1997-01-01

    This dossier is entirely devoted to the transportation of radioactive and fissile materials of civil use. It comprises 9 papers dealing with: the organization of the control of the radioactive materials transport safety (safety and security aspects, safety regulations, safety analysis and inspection, emergency plans, public information), the technical aspects of the regulation concerning the transport of radioactive materials (elaboration of regulations and IAEA recommendations, risk assessments, defense in depth philosophy and containers, future IAEA recommendations, expertise-research interaction), the qualification of containers (regulations, test facilities), the Transnucleaire company (presentation, activity, containers for spent fuels), the packages of radioactive sources for medical use (flux, qualification, safety and transport), an example of accident during radioactive materials transportation: the Apach train derailment (February 4, 1997), the sea transport of radioactive materials (international maritime organization (OMI), international maritime dangerous goods (IMDG) code, irradiated nuclear fuel (INF) safety rules), the transport of radioactive materials in Germany, and the point of view from an external observer. (J.S.)

  20. Induced fission track distribution from highly radioactive particles in fallout materials

    International Nuclear Information System (INIS)

    Hashimoto, Tetsuo; Okada, Tatemichi

    1987-01-01

    Some highly radioactive fallout particles (GPs) from the 19th Chinese nuclear detonation were followed to the neutron irradiation in a reactor after sandwiched with mica detectors. The interesting star-like fission track patterns were revealed on the etched surface of the mica detectors. The simple chemical separation procedure for the GPs was applied for the separation of U and Pu as fissile elements and the both resultant fractions were examined with the similar high sensitive fission tracking detection. Subsequently, a representative track pattern from a black spherical particle was subjected to the determination of fissile nuclide content; comparing the total fission events evaluated on the basis of the numerical calculation of track densities with the total thermal neutron fluence. The results implied that the uranium is responsible for the main fissile nuclide remaining within a particle as unfissioned fractions and should be certainly enriched with respect to U-235 within such small fallout particles. This sophisticated method was also applied to determine the dead GPs, which have been highly radioactive particles just after the detonations, in the rain and snow-residual materials. Many induced star-like fission tracks verified certainly that there remains a lot of dead particles in the atmosheric environment till nowadays. (author)

  1. 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)

  2. Fission meter and neutron detection using poisson distribution comparison

    Science.gov (United States)

    Rowland, Mark S; Snyderman, Neal J

    2014-11-18

    A neutron detector system and method 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 into information that a first responder can use to discriminate materials. The system comprises counting neutrons from the unknown source and detecting excess grouped neutrons to identify fission in the unknown source. Comparison of the observed neutron count distribution with a Poisson distribution is performed to distinguish fissile material from non-fissile material.

  3. 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.

  4. 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.

  5. Safety aspects related to burnup increase and mixed oxide fuel

    International Nuclear Information System (INIS)

    Thomas, W.

    1992-01-01

    The dominant factor presently limiting the fuel burnup is the response of the cladding hulls. To maintain the excellent record of very low fuel failure rates for increased burnups further technical development is underway and necessary. In the nuclear fuel cycle increased burnups lead to a remarkable reduction of spent fuel arisings and corresponding economic savings. Thermal recycling of plutonium presently provides an opportunity to reduce the rising accumulation of plutunium in a situation where there is no demand for this fissile material in Fast Breeder Reactors. (orig.) [de

  6. Economics of the specification 6M safety re-evaluation and regulatory requirements

    International Nuclear Information System (INIS)

    Hopper, C.M.

    1985-01-01

    The objective of this work was to examine the potential economic impact of the DOT Specification 6M criticality safety re-evaluation and regulatory requirements. The examination was based upon comparative analyses of current authorized fissile material load limits for the 6M, current Federal regulations (and interpretations) limiting the contents of Type B fissile material packages, limiting aggregates of fissile material packages, and recent proposed fissile material mass limits derived from specialized criticality safety analyses of the 6M package. The work examines influences on cost in transportation, handling, and storage of fissile materials. Depending upon facility throughput requirements (and assumed incremental costs of fissile material packaging, storage, and transport), operating, facility storage capacity, and transportation costs can be reduced significantly. As an example of the pricing algorithm application based upon reasonable cost influences, the magnitude of the first year cost reductions could extend beyond four times the cost of the packaging nuclear criticality safety re-evaluation. 1 tab

  7. Mixing ratio sensor for alcohol mixed fuel

    Energy Technology Data Exchange (ETDEWEB)

    Miyata, Shigeru; Matsubara, Yoshihiro

    1987-08-24

    In order to improve the combustion efficiency of an internal combustion engine using gasoline-alcohol mixed fuel and to reduce harmful substance in its exhaust gas, it is necessary to control strictly the air-fuel ratio to be supplied and the ignition timing. In order to detect the mixing ratio of the mixed fuel, a mixing ratio sensor has so far been proposed to detect the above mixing ratio by casting a ray of light to the mixed fuel and utilizing a change of critical angle associated with the change of the composition of the fluid of the mixed fuel. However, because of the arrangement of its transparent substance in the fuel passage with the sealing material in between, this sensor invited the leakage of the fluid due to deterioration of the sealing material, etc. and its cost became high because of too many parts to be assembled. In view of the above, in order to reduce the number of parts, to lower the cost of parts and the assembling cost and to secure no fluid leakage from the fuel passage, this invention formed the above fuel passage and the above transparent substance both concerning the above mixing ratio sensor in an integrated manner using light transmitting resin. (3 figs)

  8. 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

  9. 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.

  10. 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

  11. Proliferation risks; Proliferatierisico's

    Energy Technology Data Exchange (ETDEWEB)

    Carchon, R

    1998-09-01

    The report gives an overview of different aspects related to safeguards of fissile materials. Existing treaties including the Non-Proliferation Treaty, and the Tlatelolco and the Rarotonga Treaties are discussed. An overview of safeguards systems for the control of fissile materials as well as the role of various authorities is given. An overall overview of proliferation risks, the physical protection of fissile materials and the trade in fissile materials is given. Finally, the status in problem countries and de facto nuclear weapon states is discussed.

  12. Proliferation risks

    International Nuclear Information System (INIS)

    Carchon, R.

    1998-09-01

    The report gives an overview of different aspects related to safeguards of fissile materials. Existing treaties including the Non-Proliferation Treaty, and the Tlatelolco and the Rarotonga Treaties are discussed. An overview of safeguards systems for the control of fissile materials as well as the role of various authorities is given. An overall overview of proliferation risks, the physical protection of fissile materials and the trade in fissile materials is given. Finally, the status in problem countries and de facto nuclear weapon states is discussed

  13. 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)

  14. Doses to railroad workers from shipments of radioactive materials

    International Nuclear Information System (INIS)

    Fields, D.E.; Cottrell, W.D.

    1988-01-01

    Fissile and high-level radioactive wastes are currently transported over long distances by truck and by rail transportation systems. The primary form of fissile material is spent reactor fuel. Transportation operations within DOE are controlled through the Transportation Operations and Management System. DOE projected increases in the rate of shipments have generated concern by railroad companies that railroad workers may be exposed to levels of radiation sufficiently high that a radiation protection program may need to be implemented. To address railroad company concerns, the Health and Safety Research Division at Oak Ridge National Laboratory has estimated doses to railroad workers for two exposure scenarios that were constructed using worker activity data obtained from CSX Transportation for crew and maintenance workers. This characterization of railroad worker activity patterns includes a quantitative evaluation of the duration and rate of exposure. These duration and exposure rate values were evaluated using each of three exposure rate vs. distance models to generate exposure estimates. 14 refs., 1 tab

  15. 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

  16. Hybrid Gama Emission Tomography (HGET): FY16 Annual Report

    Energy Technology Data Exchange (ETDEWEB)

    Miller, Erin A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Smith, Leon E. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Wittman, Richard S. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Campbell, Luke W. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Deshmukh, Nikhil S. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Zalavadia, Mital A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Batie, Margo A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Mozin, Vladimir V. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2017-02-01

    Current International Atomic Energy Agency (IAEA) methodologies for the verification of fresh low-enriched uranium (LEU) and mixed oxide (MOX) fuel assemblies are volume-averaging methods that lack sensitivity to individual pins. Further, as fresh fuel assemblies become more and more complex (e.g., heavy gadolinium loading, high degrees of axial and radial variation in fissile concentration), the accuracy of current IAEA instruments degrades and measurement time increases. Particularly in light of the fact that no special tooling is required to remove individual pins from modern fuel assemblies, the IAEA needs new capabilities for the verification of unirradiated (i.e., fresh LEU and MOX) assemblies to ensure that fissile material has not been diverted. Passive gamma emission tomography has demonstrated potential to provide pin-level verification of spent fuel, but gamma-ray emission rates from unirradiated fuel emissions are significantly lower, precluding purely passive tomography methods. The work presented here introduces the concept of Hybrid Gamma Emission Tomography (HGET) for verification of unirradiated fuels, in which a neutron source is used to actively interrogate the fuel assembly and the resulting gamma-ray emissions are imaged using tomographic methods to provide pin-level verification of fissile material concentration.

  17. Consideration of Nuclear Criticality When Directly Disposing Highly Enriched Spent Nuclear Fuel in Unsaturated Tuff - II: Geochemical Constraints

    International Nuclear Information System (INIS)

    Rechard, Rob P.; Sanchez, Lawrence C.; Trellue, Holly R.

    2003-01-01

    This article presents several reasonable cases in which four mechanisms - dissolution, physical mixing, adsorption, and precipitation (either chemical change or evaporation) - might concentrate fissile material in and around a disposal container for radioactive waste at the proposed repository at Yucca Mountain, Nevada. The possible masses, concentrations, and volume are then compared to criticality limits. The cases examined evaluate the geologic barrier role in preventing criticality since engineered options for preventing criticality (e.g., boron or gadolinium neutron absorber in the disposal container) are not considered. The solid concentrations able to form in the natural environment are insufficient for criticality to occur because (a) solutions of 235 U and 239 Pu are clearly not critical; (b) physical mixing of fissile material with the entire potential iron oxide (as goethite - FeOOH) in a waste package is not critical; (c) the adsorption of 239 Pu on consolidated iron oxide in a waste package is not critical; (d) the adsorption of 235 U on consolidated iron oxide in a waste package is not critical when accounting for reduced adsorption because of carbonates at high pH; (e) the filtration of iron oxide colloids, containing fissile material, by the thin invert material is not critical; (f) insufficient retention through precipitation of 235 U or 239 Pu occurs in the invert; (g) adsorption of 235 U and 239 Pu on devitrified or clinoptolite-rich tuff below the repository is not critical; (h) the average precipitation/adsorption of 235 U as uranyl silicates in the tuff is not critical by analogy with calcite deposition in lithophysae at Yucca Mountain; and (i) precipitation/adsorption (caused by cyclic drying) as uranyl silicates on fracture surfaces of the tuff is not critical by analogy with the oxidation of UO 2 , migration of U VI , and precipitation in fractures at the Nopal I ore deposit in Mexico

  18. 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.

  19. Tank 40 final sludge batch 9 chemical and fissile radionuclide characterization results

    Energy Technology Data Exchange (ETDEWEB)

    Bannochie, C. J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Kubilius, W. P. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Pareizs, J. M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-06-26

    A sample of Sludge Batch (SB) 9 was pulled from Tank 40 in order to obtain radionuclide inventory analyses necessary for compliance with the Waste Acceptance Product Specifications (WAPS)i. The SB9 WAPS sample was also analyzed for chemical composition, including noble metals, and fissile constituents, and these results are reported here. These analyses along with the WAPS radionuclide analyses will help define the composition of the sludge in Tank 40 that is fed to the Defense Waste Processing Facility (DWPF) as SB9. At the Savannah River National Laboratory (SRNL), the 3-L Tank 40 SB9 sample was transferred from the shipping container into a 4-L high density polyethylene bottle and solids were allowed to settle. Supernate was then siphoned off and circulated through the shipping container to complete the transfer of the sample. Following thorough mixing of the 3-L sample, a 547 g sub-sample was removed. This sub-sample was then utilized for all subsequent slurry sample preparations. Eight separate aliquots of the slurry were digested, four with HNO3/HCl (aqua regiaii) in sealed Teflon® vessels and four with NaOH/Na2O2 (alkali or peroxide fusioniii) using Zr crucibles. Three Analytical Reference Glass – 1iv (ARG-1) standards were digested along with a blank for each preparation. Each aqua regia digestion and blank was diluted to 1:100 with deionized water and submitted to Analytical Development (AD) for inductively coupled plasma – atomic emission spectroscopy (ICP-AES) analysis, inductively coupled plasma – mass spectrometry (ICP-MS) analysis, atomic absorption spectroscopy (AA) for As and Se, and cold vapor atomic absorption spectroscopy (CV-AA) for Hg. Equivalent dilutions of the alkali fusion digestions and blank were submitted to AD for ICP-AES analysis. Tank 40 SB9 supernate was collected from a mixed slurry sample in the SRNL Shielded Cells and submitted to AD for ICP-AES, ion chromatography (IC), total base/free OH-/other base, total inorganic

  20. 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.

  1. 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

  2. The safety of radioactive materials transport; La surete des transports de matieres radioactives

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2007-07-01

    The rule of the radioactive materials transport contains two different objectives: the safety, or physical protection, consists in preventing the losses, the disappearances, the thefts and the diversions of the nuclear materials (useful materials for weapons); the high civil servant of defence near the Minister of Economy, Finance and Industry is the responsible authority; the safety consists in mastering the risks of irradiation, contamination and criticality presented by the radioactive and fissile materials transport, in order that man and environment do not undergo the nuisances. The control of the safety is within the competence of the Asn. (N.C.)

  3. 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

  4. Destructive and nondestructive methods for controlling nuclear materials for the purpose of safeguards in the CSSR

    International Nuclear Information System (INIS)

    Krivanek, M.; Krtil, J.; Moravec, J.; Pacak, P.; Sus, F.

    1977-01-01

    Central Control Laboratory (CCL) of the Nuclear Research Institute was charged with the control of nuclear materials in CSSR within the framework of the safeguards system. The CCL has been directed by the Department of nuclear safety and safeguards of CAEC according to a long-term plan, elaborated for controlling nuclear material in CSSR. The CCL has mainly been performing independent, rapid, accurate, and reliable analyses of nuclear materials, using destructive as well as non-destructive methods; the analyses of samples taken in MBA's in CSSR are mentioned, concerning the determinations of U, Pu, and Th contents, isotopic compositions of U and Pu, and burn up. The results of the analyses have served for the material and isotopic balances of fissile materials and the control of fuel reprocessing under laboratory conditions. The methods for sampling and sample transport as well as sample treatment before the analysis are described. The experience is given, obtained at CCL during a routine application of chemical methods for highly precise determinations of U, Pu, and Th (titration-based methods), mass-spectrometric determinations of U and Pu (isotopic composition, IDA using 233 U and 242 Pu), and burn-up determinations based on radioactive fissile products (Cs, Ru, Ce) and stable Nd isotopes. Some non-destructive methods for controlling nuclear materials (passive gamma-spectrometry) are discussed

  5. Gallium-cladding compatibility testing plan: Phase 3: Test plan for centrally heated surrogate rodlet test. Revision 2

    International Nuclear Information System (INIS)

    Morris, R.N.; Baldwin, C.A.; Wilson, D.F.

    1998-07-01

    The Fissile Materials Disposition Program (FMDP) is investigating the use of weapons grade plutonium in mixed oxide (MOX) fuel for light-water reactors (LWR). Commercial MOX fuel has been successfully used in overseas reactors for many years; however, weapons derived fuel may differ from the previous commercial fuels because of small amounts of gallium impurities. A concern presently exists that the gallium may migrate out of the fuel, react with and weaken the clad, and thereby promote loss of fuel pin integrity. Phases 1 and 2 of the gallium task are presently underway to investigate the types of reactions that occur between gallium and clad materials. This is a Level-2 document as defined in the Fissile Materials Disposition Program Light-Water Reactor Mixed-Oxide Fuel Irradiation Test Project Plan. This Plan summarizes the projected Phase 3 Gallium-Cladding compatibility heating test and the follow-on post test examination (PTE). This work will be performed using centrally-heated surrogate pellets, to avoid unnecessary complexities and costs associated with working with plutonium and an irradiation environment. Two sets of rodlets containing pellets prepared by two different methods will be heated. Both sets will have an initial bulk gallium content of approximately 10 ppm. The major emphasis of the PTE task will be to examine the material interactions, particularly indications of gallium transport from the pellets to the clad

  6. Report on the Trilateral Initiative. IAEA verification of weapon-origin material in the Russian Federation and the United States

    International Nuclear Information System (INIS)

    Shea, Thomas E.

    2001-01-01

    Just over five years ago, the Trilateral Initiative was launched to investigate the technical, legal and financial issues associated with IAEA verification of weapon-origin fissile material in the Russian Federation and the United States. Since then, the Joint Working Group has developed concepts and equipment suitable for such a verification mission, anticipating that the States would submit classified forms of fissile material to IAEA verification under new agreements developed for this purpose. This article summarizes the accomplishments to date and identifies the future steps foreseen under the Trilateral Initiative. As there is no legal commitment on the Parties to this Initiative as yet, the issues considered are still changing. Since it was launched, the Initiative has been given a sense of importance and weight, raising the expectations of the international community. The Final Document of the 2000 Conference on the Treaty on the Non-Proliferation of Nuclear Weapons (NPT), for example, under the review of Article VI of the Treaty, includes the statement to 'complete and implement the Trilateral Initiative'. It was launched following independent statements by the President of the United States beginning in 1993, and by the President of the Russian Federation in 1996. It is an Initiative between the IAEA, the Russian Federation and the United States that is in the context of Article VI of the NPT. The intention is to examine the technical, legal and financial issues associated with IAEA verification of weapon origin and other fissile material released from defense programmes in those two countries

  7. Fuel cycle parameters for strategy studies

    International Nuclear Information System (INIS)

    Archinoff, G.H.

    1979-05-01

    This report summarizes seven fuel cycle parameters (efficiency, specific power, burnup, equilibrium net fissile feed, equilibrium net fissile surplus, first charge fissile content, and whether or not fuel reprocessing is required) to be used in long-term strategy analyses of fuel cycles based on natural UO 2 , low enriched uranium, mixed oxides, plutonium topped thorium, uranium topped thorium, and the fast breeder oxide cycle. (LL)

  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. Delayed Gamma-Ray Spectroscopy for Non-Destructive Assay of Nuclear Materials

    International Nuclear Information System (INIS)

    Ludewigt, Bernhard; Mozin, Vladimir; Campbell, Luke; Favalli, Andrea; Hunt, Alan W.; Reedy, Edward T.E.; Seipel, Heather

    2015-01-01

    High-energy, beta-delayed gamma-ray spectroscopy is a potential, non-destructive assay techniques for the independent verification of declared quantities of special nuclear materials at key stages of the fuel cycle and for directly assaying nuclear material inventories for spent fuel handling, interim storage, reprocessing facilities, repository sites, and final disposal. Other potential applications include determination of MOX fuel composition, characterization of nuclear waste packages, and challenges in homeland security and arms control verification. Experimental measurements were performed to evaluate fission fragment yields, to test methods for determining isotopic fractions, and to benchmark the modeling code package. Experimental measurement campaigns were carried out at the IAC using a photo-neutron source and at OSU using a thermal neutron beam from the TRIGA reactor to characterize the emission of high-energy delayed gamma rays from 235 U, 239 Pu, and 241 Pu targets following neutron induced fission. Data were collected for pure and combined targets for several irradiation/spectroscopy cycle times ranging from 10/10 seconds to 15/30 minutes.The delayed gamma-ray signature of 241 Pu, a significant fissile constituent in spent fuel, was measured and compared to 239 Pu. The 241 Pu/ 239 Pu ratios varied between 0.5 and 1.2 for ten prominent lines in the 2700-3600 keV energy range. Such significant differences in relative peak intensities make it possible to determine relative fractions of these isotopes in a mixed sample. A method for determining fission product yields by fitting the energy and time dependence of the delayed gamma-ray emission was developed and demonstrated on a limited 235 U data set. De-convolution methods for determining fissile fractions were developed and tested on the experimental data. The use of high count-rate LaBr 3 detectors was investigated as a potential alternative to HPGe detectors. Modeling capabilities were added to an

  10. A neutron booster for spallation sources--application to accelerator driven systems and isotope production

    CERN Document Server

    Galy, J; Van Dam, H; Valko, J

    2002-01-01

    One can design a critical system with fissile material in the form of a thin layer on the inner surface of a cylindrical neutron moderator such as graphite or beryllium. Recently, we have investigated the properties of critical and near critical systems based on the use of thin actinide layers of uranium, plutonium and americium. The thickness of the required fissile layer depends on the type of fissile material, its concentration in the layer and on the geometrical arrangement, but is typically in the mu m-mm range. The resulting total mass of fissile material can be as low as 100 g. Thin fissile layers have a variety of applications in nuclear technology--for example in the design neutron amplifiers for medical applications and 'fast' islands in thermal reactors for waste incineration. In the present paper, we investigate the properties of a neutron booster unit for spallation sources and isotope production. In those applications a layer of fissile material surrounds the spallation source. Such a module cou...

  11. 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.)

  12. 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...

  13. 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

  14. 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

  15. 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.

  16. 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

  17. Canadian accelerator breeder system development

    International Nuclear Information System (INIS)

    Schriber, S.O.

    1982-11-01

    A shortage of fissile material at a reasonable price is expected to occur in the early part of the twenty-first century. Converting fertile material to fissile material by electronuclar methods is an option that can extend th world's resources of fissionable material, supplying fuel for nuclear power stations. This paper presents the rationale for electronuclear breeders and describes the Canadian development program for an accelerator breeder facility that could produce 1 Mg of fissile material per year

  18. Evolutionary/advanced light water reactor data report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-02-09

    The US DOE Office of Fissile Material Disposition is examining options for placing fissile materials that were produced for fabrication of weapons, and now are deemed to be surplus, into a condition that is substantially irreversible and makes its use in weapons inherently more difficult. The principal fissile materials subject to this disposition activity are plutonium and uranium containing substantial fractions of plutonium-239 uranium-235. The data in this report, prepared as technical input to the fissile material disposition Programmatic Environmental Impact Statement (PEIS) deal only with the disposition of plutonium that contains well over 80% plutonium-239. In fact, the data were developed on the basis of weapon-grade plutonium which contains, typically, 93.6% plutonium-239 and 5.9% plutonium-240 as the principal isotopes. One of the options for disposition of weapon-grade plutonium being considered is the power reactor alternative. Plutonium would be fabricated into mixed oxide (MOX) fuel and fissioned (``burned``) in a reactor to produce electric power. The MOX fuel will contain dioxides of uranium and plutonium with less than 7% weapon-grade plutonium and uranium that has about 0.2% uranium-235. The disposition mission could, for example, be carried out in existing power reactors, of which there are over 100 in the United States. Alternatively, new LWRs could be constructed especially for disposition of plutonium. These would be of the latest US design(s) incorporating numerous design simplifications and safety enhancements. These ``evolutionary`` or ``advanced`` designs would offer not only technological advances, but also flexibility in siting and the option of either government or private (e.g., utility) ownership. The new reactor designs can accommodate somewhat higher plutonium throughputs. This data report deals solely with the ``evolutionary`` LWR alternative.

  19. Evolutionary/advanced light water reactor data report

    International Nuclear Information System (INIS)

    1996-01-01

    The US DOE Office of Fissile Material Disposition is examining options for placing fissile materials that were produced for fabrication of weapons, and now are deemed to be surplus, into a condition that is substantially irreversible and makes its use in weapons inherently more difficult. The principal fissile materials subject to this disposition activity are plutonium and uranium containing substantial fractions of plutonium-239 uranium-235. The data in this report, prepared as technical input to the fissile material disposition Programmatic Environmental Impact Statement (PEIS) deal only with the disposition of plutonium that contains well over 80% plutonium-239. In fact, the data were developed on the basis of weapon-grade plutonium which contains, typically, 93.6% plutonium-239 and 5.9% plutonium-240 as the principal isotopes. One of the options for disposition of weapon-grade plutonium being considered is the power reactor alternative. Plutonium would be fabricated into mixed oxide (MOX) fuel and fissioned (''burned'') in a reactor to produce electric power. The MOX fuel will contain dioxides of uranium and plutonium with less than 7% weapon-grade plutonium and uranium that has about 0.2% uranium-235. The disposition mission could, for example, be carried out in existing power reactors, of which there are over 100 in the United States. Alternatively, new LWRs could be constructed especially for disposition of plutonium. These would be of the latest US design(s) incorporating numerous design simplifications and safety enhancements. These ''evolutionary'' or ''advanced'' designs would offer not only technological advances, but also flexibility in siting and the option of either government or private (e.g., utility) ownership. The new reactor designs can accommodate somewhat higher plutonium throughputs. This data report deals solely with the ''evolutionary'' LWR alternative

  20. 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

  1. Non-Destructive Testing for Control of Radioactive Waste Package

    Science.gov (United States)

    Plumeri, S.; Carrel, F.

    2015-10-01

    Characterization and control of radioactive waste packages are important issues in the management of a radioactive waste repository. Therefore, Andra performs quality control inspection on radwaste package before disposal to ensure the compliance of the radwast characteristics with Andra waste disposal specifications and to check the consistency between Andra measurements results and producer declared properties. Objectives of this quality control are: assessment and improvement of producer radwaste packages quality mastery, guarantee of the radwaste disposal safety, maintain of the public confidence. To control radiological characteristics of radwaste package, non-destructive passive methods (gamma spectrometry and neutrons counting) are commonly used. These passive methods may not be sufficient, for instance to control the mass of fissile material contained inside radwaste package. This is particularly true for large concrete hull of heterogeneous radwaste containing several actinides mixed with fission products like 137Cs. Non-destructive active methods, like measurement of photofission delayed neutrons, allow to quantify the global mass of actinides and is a promising method to quantify mass of fissile material. Andra has performed different non-destructive measurements on concrete intermediate-level short lived nuclear waste (ILW-SL) package to control its nuclear material content. These tests have allowed Andra to have a first evaluation of the performance of photofission delayed neutron measurement and to identify development needed to have a reliable method, especially for fissile material mass control in intermediate-level long lived waste package.

  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. A Strategy for Quantifying Radioactive Material in a Low-Level Waste Incineration Facility

    International Nuclear Information System (INIS)

    Hochel, R.C.

    1997-03-01

    One of the methods proposed by the U.S. Department of Energy (DOE) for the volume reduction and stabilization of a variety of low-level radioactive wastes (LLW) is incineration. Many commercial incinerators are in operation treating both non-hazardous and hazardous wastes. These can obtain volume reductions factors of 50 or more for certain wastes, and produce a waste (ash) that can be easily stabilized if necessary by vitrification or cementation. However, there are few incinerators designed to accommodate radioactive wastes. One has been recently built at the Savannah River Site (SRS) near Aiken, SC and is burning non-radioactive hazardous waste and radioactive wastes in successive campaigns. The SRS Consolidated Incineration Facility (CIF) is RCRA permitted as a Low Chemical Hazard, Radiological facility as defined by DOE criteria (Ref. 1). Accordingly, the CIF must operate within specified chemical, radionuclide, and fissile material inventory limits (Ref. 2). The radionuclide and fissile material limits are unique to radiological or nuclear facilities, and require special measurement and removal strategies to assure compliance, and the CIF may be required to shut down periodically in order to clean out the radionuclide inventory which builds up in various parts of the facility

  4. 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

  5. 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

  6. 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

  7. Uranium-233 waste definition: Disposal options, safeguards, criticality control, and arms control

    International Nuclear Information System (INIS)

    Forsberg, C.W.; Storch, S.N.; Lewis, L.C.

    1998-01-01

    The US investigated the use of 233 U for weapons, reactors, and other purposes from the 1950s into the 1970s. Based on the results of these investigations, it was decided not to use 233 U on a large scale. Most of the 233 U-containing materials were placed in long-term storage. At the end of the cold war, the US initiated, as part of its arms control policies, a disposition program for excess fissile materials. Other programs were accelerated for disposal of radioactive wastes placed in storage during the cold war. Last, potential safety issues were identified related to the storage of some 233 U-containing materials. Because of these changes, significant activities associated with 233 U-containing materials are expected. This report is one of a series of reports to provide the technical bases for future decisions on how to manage this material. A basis for defining when 233 U-containing materials can be managed as waste and when they must be managed as concentrated fissile materials has been developed. The requirements for storage, transport, and disposal of radioactive wastes are significantly different than those for fissile materials. Because of these differences, it is important to classify material in its appropriate category. The establishment of a definition of what is waste and what is fissile material will provide the guidance for appropriate management of these materials. Wastes are defined in this report as materials containing sufficiently small masses or low concentrations of fissile materials such that they can be managed as typical radioactive waste. Concentrated fissile materials are defined herein as materials containing sufficient fissile content such as to warrant special handling to address nuclear criticality, safeguards, and arms control concerns

  8. Tour of Los Alamos Safeguards R and D laboratories: demonstration and use of NDA instruments and material control and accounting simulation

    International Nuclear Information System (INIS)

    Anon.

    1984-01-01

    A description is presented of the nondestructive assay techniques and instrumentation for measuring the fissile content of fuel assemblies and fuel components. The course participants had a hands-on tour of this instrumentation and material accounting and control systems at Los Alamos National Laboratory

  9. 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

  10. High pressure freon decontamination of remote equipment

    International Nuclear Information System (INIS)

    Wilson, C.E.

    1987-01-01

    A series of decontamination tests using high pressure FREON 113 was conducted in the 200 Area of the Hanford site. The intent of these tests was to evaluate the effectiveness of FREON 113 in decontamination of manipulator components, tools, and equipment items contaminated with mixed fission products. The test results indicated that high pressure FREON 113 is very effective in removing fissile material from a variety of objects and can reduce both the quantity and the volume of the radioactive waste material presently being buried

  11. 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

  12. Plutonium characterisation with prompt high energy gamma-rays from (n,gamma) reactions for nuclear warhead dismantlement verification

    Energy Technology Data Exchange (ETDEWEB)

    Postelt, Frederik; Gerald, Kirchner [Carl Friedrich von Weizsaecker-Centre for Science and Peace Research, Hamburg (Germany)

    2015-07-01

    Measurements of neutron induced gammas allow the characterisation of fissile material (i.e. plutonium and uranium), despite self- and additional shielding. Most prompt gamma-rays from radiative neutron capture reactions in fissile material have energies between 3 and 6.5 MeV. Such high energy photons have a high penetrability and therefore minimise shielding and self-absorption effects. They are also isotope specific and therefore well suited to determine the isotopic composition of fissile material. As they are non-destructive, their application in dismantlement verification is desirable. Disadvantages are low detector efficiencies at high gamma energies, as well as a high background of gammas which result from induced fission reactions in the fissile material, as well as delayed gammas from both, (n,f) and(n,gamma) reactions. In this talk, simulations of (n,gamma) measurements and their implications are presented. Their potential for characterising fissile material is assessed and open questions are addressed.

  13. 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

  14. Nuclear fission fragment excitation of electronic transition laser media

    International Nuclear Information System (INIS)

    Lorents, D.C.; McCusker, M.V.; Rhodes, C.K.

    1976-01-01

    The properties of high energy electronic transition lasers excited by fission fragments are expanded. Specific characteristics of the media including density, excitation rates, wavelength, kinetics, fissile material, scale size, and medium uniformity are assessed. The use of epithermal neutrons, homogeneously mixed fissile material, and special high cross section nuclear isotopes to optimize coupling of the energy to the medium are shown to be important considerations maximizing the scale size, energy deposition, and medium uniformity. A performance limit point of approximately 1000 J/l in approximately 100 μs pulses is established for a large class of systems operating in the near ultraviolet and visible spectral regions. It is demonstrated that e-beam excitation can be used to simulate nuclear pumping conditions to facilitate the search for candidate media. Experimental data for the kinetics of a XeF* laser operating in Ar/Xe/F 2 /UF 6 mixtures are given. These reactor-pumped systems are suitable for scaling to volumes on the order of (meters) 3

  15. 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)

  16. Process for the fabrication of a nuclear fuel

    International Nuclear Information System (INIS)

    Hirose, Yasuo.

    1970-01-01

    Herein disclosed is a process for fabricating a nuclear fuel incorporating either uranium or plutonium. A pellet-like substrate consisting of a packed powder ceramic fuel such as uranium or plutonium is prepared with the horizontal surface of the body provided with a masking. Next, after impregnating the substrate voids with a solution consisting of a fissile material or mixture of fissile material and poison, the solvent is removed by a chemical deposition process which causes the impregnated material to migrate through capillary action toward the vicinity of the fuel body surface. Sintering and pyrolysis of the deposited material and masking are subsequently carried out to yield a fuel body having adjacent to its surface an intensely concentrated layer of either fissile material or a mixture of fissile material and poison. (Owens, K.J.)

  17. Nuclear materials facility safety initiative

    International Nuclear Information System (INIS)

    Peddicord, K.L.; Nelson, P.; Roundhill, M.; Jardine, L.J.; Lazarev, L.; Moshkov, M.; Khromov, V.V.; Kruchkov, E.; Bolyatko, V.; Kazanskij, Yu.; Vorobeva, I.; Lash, T.R.; Newton, D.; Harris, B.

    2000-01-01

    Safety in any facility in the nuclear fuel cycle is a fundamental goal. However, it is recognized that, for example, should an accident occur in either the U.S. or Russia, the results could seriously delay joint activities to store and disposition weapons fissile materials in both countries. To address this, plans are underway jointly to develop a nuclear materials facility safety initiative. The focus of the initiative would be to share expertise which would lead in improvements in safety and safe practices in the nuclear fuel cycle.The program has two components. The first is a lab-to-lab initiative. The second involves university-to-university collaboration.The lab-to-lab and university-to-university programs will contribute to increased safety in facilities dealing with nuclear materials and related processes. These programs will support important bilateral initiatives, develop the next generation of scientists and engineers which will deal with these challenges, and foster the development of a safety culture

  18. Preparation and development of new Pu spike isotopic reference materials at IRMM

    Energy Technology Data Exchange (ETDEWEB)

    Jakopic, Rozle; Bauwens, Jeroen; Richter, Stephan; Sturm, Monika; Verbruggen, Andre; Wellum, Roger; Eykens, Roger; Kehoe, Frances; Kuehn, Heinz; Aregbe, Yetunde [Institute for Reference Materials and Measurements (IRMM) Joint Research Centre, European Commission, Geel, (Belgium)

    2011-12-15

    spikes are applied to measure the uranium and plutonium content of dissolved fuel solutions using IDMS. They are prepared by IRMM to fulfil the existing requirements for reliable spike IRMs in fissile material control from European Safeguards authorities and customers from industry. IRMM-046b, a mixed uranium-plutonium spike IRM of highly enriched {sup 233}U and {sup 242}Pu that dates from 1995, was re-certified for isotope amount content and isotopic composition, each with considerably smaller combined uncertainties. IRMM-046c, a new mixed uranium-plutonium spike, and IRMM-049d, a highly enriched {sup 242}Pu spike, have been prepared. IRMM-049d was prepared from the same stock solution as its predecessor IRMM-049c, dating from 1996, but the new 242Pu spike has certified values with smaller combined uncertainties. The traceability of the certified values to the SI is established through an unbroken chain of comparisons, all having stated uncertainties. IRMM is also co-operating with the Institute for Transuranium Elements (EC-JRC-ITU) in a feasibility study on the development of Pu reference materials for 'age dating' in nuclear forensics. In the course of this work, the reference materials NBS SRM 946, 947 and 948 (NBL CRM 136, 137 and 138) will be investigated among others.

  19. 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)

  20. 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)

  1. K east encapsulation packager modifications

    International Nuclear Information System (INIS)

    Jensen, M.A.

    1994-01-01

    This Supporting Document analyzes a proposal for reducing the under-packager volume to decrease the amount of fissile material that could accumulate there. The analysis shows that restricting the under packager volume to no more than 4080 in 3 will assure that if accumulated fissile material beneath the packager is added to the worst-case mass of fissile material in the discharge chute, a k eff of 0.98 will not be exceeded

  2. Methods and apparatuses for the development of microstructured nuclear fuels

    Science.gov (United States)

    Jarvinen, Gordon D [Los Alamos, NM; Carroll, David W [Los Alamos, NM; Devlin, David J [Santa Fe, NM

    2009-04-21

    Microstructured nuclear fuel adapted for nuclear power system use includes fissile material structures of micrometer-scale dimension dispersed in a matrix material. In one method of production, fissile material particles are processed in a chemical vapor deposition (CVD) fluidized-bed reactor including a gas inlet for providing controlled gas flow into a particle coating chamber, a lower bed hot zone region to contain powder, and an upper bed region to enable powder expansion. At least one pneumatic or electric vibrator is operationally coupled to the particle coating chamber for causing vibration of the particle coater to promote uniform powder coating within the particle coater during fuel processing. An exhaust associated with the particle coating chamber and can provide a port for placement and removal of particles and powder. During use of the fuel in a nuclear power reactor, fission products escape from the fissile material structures and come to rest in the matrix material. After a period of use in a nuclear power reactor and subsequent cooling, separation of the fissile material from the matrix containing the embedded fission products will provide an efficient partitioning of the bulk of the fissile material from the fission products. The fissile material can be reused by incorporating it into new microstructured fuel. The fission products and matrix material can be incorporated into a waste form for disposal or processed to separate valuable components from the fission products mixture.

  3. Neutron method for NDA in the Sapphire Project

    International Nuclear Information System (INIS)

    Lewis, K.D.

    1995-01-01

    The implementation of Project Sapphire, the top-secret mission to the Republic of Kazakhstan to recover weapons-grade nuclear materials, consisted of four major elements: (1) repacking of fissile material from Kazakh containers into suitable U.S. containers; (2) nondestructive analyses (NDA) to quantify the 235 U content of each container for nuclear criticality safety and compliance purposes; (3) packaging of the fissile material containers into 6M/2R drums, which are internationally approved for shipping fissile material; and (4) shipping or transport of the recovered fissile material to the United States. This paper discusses the development and application of a passive neutron counting technique used in the NDA phase of the Sapphire operations to analyze uranium/beryllium (U/Be) alloys and compounds for 235 U content

  4. A neutron method for NDA analysis in the SAPPHIRE Project

    International Nuclear Information System (INIS)

    Lewis, K.D.

    1995-01-01

    The implementation of Project SAPPHIRE, the top secret mission to the Republic of Kazakhstan to recover weapons grade nuclear materials, consisted of four major elements: (1) the re-packing of fissile material from Kazakh containers into suitable US containers; (2) nondestructive analyses (NDA) to quantify the U-235 content of each container for Nuclear Criticality Safety and compliance purposes; (3) the packaging of the fissile material containers into 6M/2R drums, which are internationally approved for shipping fissile material; and (4) the shipping or transport of the recovered fissile material to the United States. This paper discusses the development and application of a passive neutron counting technique used in the NDA phase of SAPPHIRE operations to analyze uranium/beryllium (U/Be) alloys and compounds for U-235 content

  5. 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

  6. Quality assurance for packaging of radioactive and hazardous materials

    International Nuclear Information System (INIS)

    Gustafson, L.D.

    1986-01-01

    The Department of Energy (DOE) has required for many years that quality assurance programs be established and implemented for the packaging of radioactive and hazardous materials. This paper identifies various requirement principles and related actions involved in establishing effective quality assurance for packaging of radioactive and hazardous materials. A primary purpose of these quality assurance program activities is to provide assurance that the packaging and transportation of hazardous materials, which includes radioactive and fissile materials, are in conformance with appropriate governmental regulations. Applicable regulations include those issued by the Nuclear Regulatory Commission (NRC), the Department of Transportation (DOT), and the Environmental Protection Agency (EPA). DOE Order 5700.6A establishes that quality assurance requirements are to be applied in accordance with national consensus standards where suitable ones are available. In the nuclear area, ANSI/ASME NQA-1 is the preferred standard

  7. Neutron data error estimate of criticality calculations for lattice in shielding containers with metal fissionable materials

    International Nuclear Information System (INIS)

    Vasil'ev, A.P.; Krepkij, A.S.; Lukin, A.V.; Mikhal'kova, A.G.; Orlov, A.I.; Perezhogin, V.D.; Samojlova, L.Yu.; Sokolov, Yu.A.; Terekhin, V.A.; Chernukhin, Yu.I.

    1991-01-01

    Critical mass experiments were performed using assemblies which simulated one-dimensional lattice consisting of shielding containers with metal fissile materials. Calculations of the criticality of the above assemblies were carried out using the KLAN program with the BAS neutron constants. Errors in the calculations of the criticality for one-, two-, and three-dimensional lattices are estimated. 3 refs.; 1 tab

  8. Light water breeder reactor using a uranium-plutonium cycle

    International Nuclear Information System (INIS)

    Radkowsky, A.; Chen, R.

    1990-01-01

    This patent describes a light water receptor (LWR) for breeding fissile material using a uranium-plutonium cycle. It comprises: a prebreeder section having plutonium fuel containing a Pu-241 component, the prebreeder section being operable to produce enriched plutonium having an increased Pu-241 component; and a breeder section for receiving the enriched plutonium from the prebreeder section, the breeder section being operable for breeding fissile material from the enriched plutonium fuel. This patent describes a method of operating a light water nuclear reactor (LWR) for breeding fissile material using a uranium-plutonium cycle. It comprises: operating the prebreeder to produce enriched plutonium fuel having an increased Pu-241 component; fueling a breeder section with the enriched plutonium fuel to breed the fissile material

  9. Standard test method for non-destructive assay of nuclear material in waste by passive and active neutron counting using a differential Die-away system

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2009-01-01

    1.1 This test method covers a system that performs nondestructive assay (NDA) of uranium or plutonium, or both, using the active, differential die-away technique (DDT), and passive neutron coincidence counting. Results from the active and passive measurements are combined to determine the total amount of fissile and spontaneously-fissioning material in drums of scrap or waste. Corrections are made to the measurements for the effects of neutron moderation and absorption, assuming that the effects are averaged over the volume of the drum and that no significant lumps of nuclear material are present. These systems are most widely used to assay low-level and transuranic waste, but may also be used for the measurement of scrap materials. The examples given within this test method are specific to the second-generation Los Alamos National Laboratory (LANL) passive-active neutron assay system. 1.1.1 In the active mode, the system measures fissile isotopes such as 235U and 239Pu. The neutrons from a pulsed, 14-MeV ne...

  10. Metal Poisons for Criticality in Waste Streams

    International Nuclear Information System (INIS)

    Williamson, T.G.; Goslen, A.Q.

    1996-01-01

    Many of the wastes from processing fissile materials contain metals which may serve as nuclear criticality poisons. It would be advantageous to the criticality evaluation of these wastes to demonstrate that the poisons remain with the fissile materials and to demonstrate an always safe poison-to-fissile ratio. The first task, demonstrating that the materials stay together, is the job of the chemist, the second, calculating an always safe ratio, is an object of this paper

  11. 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...

  12. 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

  13. 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.

  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. Core of a fast neutron nuclear reactor

    International Nuclear Information System (INIS)

    Giacometti, Christian; Mougniot, J.-C.; Ravier, Jean.

    1974-01-01

    The fast neutron nuclear reactor described includes an internal area in fissile material completely enclosed in an area of fertile material forming the outside blanket. The internal fissile area is provided with housings exclusively filled with fertile material forming one or more inside blankets. In this core the internal blankets are shaped like rings vertically separating superimposed rings of fissile material. The blanket of material nearest to the periphery is circumscribed externally by a contour having an indented shape on its straight section so as to increase the contact area between this blanket and the external blanket [fr

  16. Managing a mixed waste program

    International Nuclear Information System (INIS)

    Koch, J.D.

    1994-01-01

    IT Corporation operates an analytical laboratory in St. Louis capable of analyzing environmental samples that are contaminated with both chemical and radioactive materials. Wastes generated during these analyses are hazardous in nature; some are listed wastes others exhibit characteristic hazards. When the original samples contain significant quantities of radioactive material, the waste must be treated as a mixed waste. A plan was written to document the waste management program describing the management of hazardous, radioactive and mixed wastes. This presentation summarizes the methods employed by the St. Louis facility to reduce personnel exposures to the hazardous materials, minimize the volume of mixed waste and treat the materials prior to disposal. The procedures that are used and the effectiveness of each procedure will also be discussed. Some of the lessons that have been learned while dealing with mixed wastes will be presented as well as the solutions that were applied. This program has been effective in reducing the volume of mixed waste that is generated. The management program also serves as a method to manage the costs of the waste disposal program by effectively segregating the different wastes that are generated

  17. 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.

  18. Applying RFID technology in nuclear materials management

    International Nuclear Information System (INIS)

    Tsai, H.; Chen, K.; Liu, Y.; Norair, J.P.; Bellamy, S.; Shuler, J.

    2008-01-01

    The Packaging Certification Program (PCP) of US Department of Energy (DOE) Environmental Management (EM), Office of Safety Management and Operations (EM-60), has developed a radio frequency identification (RFID) system for the management of nuclear materials. Argonne National Laboratory, a PCP supporting laboratory, and Savi Technology, a Lockheed Martin Company, are collaborating in the development of the RFID system, a process that involves hardware modification (form factor, seal sensor and batteries), software development and irradiation experiments. Savannah River National Laboratory and Argonne will soon field test the active RFID system on Model 9975 drums, which are used for storage and transportation of fissile and radioactive materials. Potential benefits of the RFID system are enhanced safety and security, reduced need for manned surveillance, real time access of status and history data, and overall cost effectiveness

  19. Supplying the six. [Supplies of nuclear fuels and ores to the European Community

    Energy Technology Data Exchange (ETDEWEB)

    Oboussier, F

    1975-07-01

    Under the Euratom Treaty, the European Community must ensure that all users in the Community receive a regular and equitable supply of ores and nuclear fuels. Supply to users in the Community of ores, source materials, and special fissile materials is based on the principle of equal access of the users to the supply sources. To ensure such equal access, the Treaty prohibits all practices designed to secure a privileged position for certain users. In addition, an agency has been set up with two essential rights--that of an option on all ores, source materials, and special fissile materials produced in the territories of the Member States; and the exclusive right to conclude all contracts relating to the supply of ores, source materials, and special fissile materials coming from inside the Community or from outside. Dealings of the Agency with outside agencies, especially the former US AEC, are described. The uranium market and its economics and the availability of special fissile materials are summarized. (MCW)

  20. Delayed Gamma-Ray Spectroscopy for Non-Destructive Assay of Nuclear Materials

    Energy Technology Data Exchange (ETDEWEB)

    Ludewigt, Bernhard [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Mozin, Vladimir [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Campbell, Luke [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Favalli, Andrea [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Hunt, Alan W. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Reedy, Edward T.E. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Seipel, Heather [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-06-01

    High-­energy, beta-delayed gamma-­ray spectroscopy is a potential, non-­destructive assay techniques for the independent verification of declared quantities of special nuclear materials at key stages of the fuel cycle and for directly assaying nuclear material inventories for spent fuel handling, interim storage, reprocessing facilities, repository sites, and final disposal. Other potential applications include determination of MOX fuel composition, characterization of nuclear waste packages, and challenges in homeland security and arms control verification. Experimental measurements were performed to evaluate fission fragment yields, to test methods for determining isotopic fractions, and to benchmark the modeling code package. Experimental measurement campaigns were carried out at the IAC using a photo-­neutron source and at OSU using a thermal neutron beam from the TRIGA reactor to characterize the emission of high-­energy delayed gamma rays from 235U, 239Pu, and 241Pu targets following neutron induced fission. Data were collected for pure and combined targets for several irradiation/spectroscopy cycle times ranging from 10/10 seconds to 15/30 minutes.The delayed gamma-ray signature of 241Pu, a significant fissile constituent in spent fuel, was measured and compared to 239Pu. The 241Pu/239Pu ratios varied between 0.5 and 1.2 for ten prominent lines in the 2700-­3600 keV energy range. Such significant differences in relative peak intensities make it possible to determine relative fractions of these isotopes in a mixed sample. A method for determining fission product yields by fitting the energy and time dependence of the delayed gamma-­ray emission was developed and demonstrated on a limited 235U data set. De-­convolution methods for determining fissile fractions were developed and tested on the experimental data. The use of high count-­rate LaBr3 detectors

  1. Mixed-linker zeolitic imidazolate framework mixed-matrix membranes for aggressive CO2 separation from natural gas

    KAUST Repository

    Thompson, Joshua A.

    2014-07-01

    Zeolitic imidazolate framework (ZIF) materials are a promising subclass of metal-organic frameworks (MOF) for gas separations. However, due to the deleterious effects of gate-opening phenomena associated with organic linker rotation near the limiting pore apertures of ZIFs, there have been few demonstrations of improved gas separation properties over pure polymer membranes when utilizing ZIF materials in composite membranes for CO2-based gas separations. Here, we report a study of composite ZIF/polymer membranes, containing mixed-linker ZIF materials with ZIF-8 crystal topologies but composed of different organic linker compositions. Characterization of the mixed-linker ZIFs shows that the mixed linker approach offers control over the porosity and pore size distribution of the materials, as determined from nitrogen physisorption and Horváth-Kawazoe analysis. Single gas permeation measurements on mixed-matrix membranes reveal that inclusion of mixed-linker ZIFs yields membranes with better ideal CO2/CH4 selectivity than membranes containing ZIF-8. This improvement is shown to likely occur from enhancement in the diffusion selectivity of the membranes associated with controlling the pore size distribution of the ZIF filler. Mixed-gas permeation experiments show that membranes with mixed-linker ZIFs display an effective plasticization resistance that is not typical of the pure polymeric matrix. Overall, we demonstrate that mixed-linker ZIFs can improve the gas separation properties in composite membranes and may be applicable to aggressive CO2 concentrations in natural gas feeds. © 2013 Elsevier Inc. All rights reserved.

  2. Conversion of highly enriched uranium in thorium-232 based oxide fuel for light water reactors: MOX-T fuel

    Energy Technology Data Exchange (ETDEWEB)

    Vapirev, E I; Jordanov, T; Christoskov, I [Sofia Univ. (Bulgaria). Fizicheski Fakultet

    1994-12-31

    The idea of conversion of highly enriched uranium (HEU) from warheads without mixing it with natural uranium as well as the utilization of plutonium as fuel component is discussed. A nuclear fuel which is a mixture of 4% {sup 235}U (HEU) as a fissile isotope and 96 % {sup 232}Th (ThO{sub 2}) as a non-fissile isotope in a mixed oxide with thorium fuel is proposed. It is assumed that plutonium can also be used in the proposed fuel in a mixture with {sup 235}U. The following advantages of the use of HEU in LWRs in mixed {sup 235}U - Th fuel are pointed out: (1) No generation of long-living plutonium and americium isotopes (in case of reprocessing the high level radioactive wastes will contain only fission fragments and uranium); (2) The high conversion ratio of Th extends the expected burnup by approximately 1/3 without higher initial enrichment (the same initial enrichment simplifies the problem for compensation of the excess reactivity in the beginning with burnable poison and boric acid); (3) The high conversion ratio of Th allows the fuel utilization with less initial enrichment (by approx. 1/3) for the same burnup; thus less excess reactivity has to be compensated after reloading; in case of fuel reprocessing all fissile materials ({sup 235}U + {sup 233}U) could be chemically extracted. Irrespectively to the optimistic expectations outlined, further work including data on optimal loading and reloading schemes, theoretical calculations of thermal properties of {sup 235}U + Th fuel rods, manufacturing of several test fuel assemblies and investigations of their operational behaviour in a reactor core is still needed. 1 fig., 7 refs.

  3. 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

  4. 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)

  5. Explanatory material for the IAEA regulations for the safe transport of radioactive material (1985 edition). 2. ed

    International Nuclear Information System (INIS)

    1987-01-01

    This document pertains to Safety Series No. 7 of the IAEA, which is to explain the provisions of the IAEA Safety Series No. 6 in order to help comprehension of the regulatory standards and to promote compliance, public acceptance and further development of the Regulations. The document also reflects corrections and changes implemented by the 1986 Supplement to the Regulations for the Safe Transport of Radioactive Material. The intent of the document is to show why certain provisions of Safety Series No. 6 exist, why they are so formed (including any relevant history) and the rationale behind the provisions. Definitions are presented, basic principles established, activity and fissile material limits as well as computational techniques are presented. The detailed requirements (the latter sections are built on this information) concern: shipping and storage, material packagings and packages which govern design. Test requirements are provided. Approval and administrative requirements are stated. Heavy emphasis is placed on providing safety through design. It contains the cornerstone of the basic requirements for packagings, packages and material-related aspects.

  6. Addendum 6 to CSAR 79-038 out-of-hood plutonium storage (burial box)

    International Nuclear Information System (INIS)

    Chiao, T.

    1995-01-01

    The Addendum considered an increase in the limit of fissile material in a stacked container array to 500 grams. In other words, the sum of fissile material in an array of containers is limited to 500 grams, regardless of whether the containers are stacked or not. The results of this evaluation indicates that with the modification of the fissile limits described, the system of a container array will stay sub-critical

  7. Computational and Experimental Investigations of the Coolant Flow in the Cassette Fissile Core of a KLT-40S Reactor

    Science.gov (United States)

    Dmitriev, S. M.; Varentsov, A. V.; Dobrov, A. A.; Doronkov, D. V.; Pronin, A. N.; Sorokin, V. D.; Khrobostov, A. E.

    2017-07-01

    Results of experimental investigations of the local hydrodynamic and mass-exchange characteristics of a coolant flowing through the cells in the characteristic zones of a fuel assembly of a KLT-40S reactor plant downstream of a plate-type spacer grid by the method of diffusion of a gas tracer in the coolant flow with measurement of its velocity by a five-channel pneumometric probe are presented. An analysis of the concentration distribution of the tracer in the coolant flow downstream of a plate-type spacer grid in the fuel assembly of the KLT-40S reactor plant and its velocity field made it possible to obtain a detailed pattern of this flow and to determine its main mechanisms and features. Results of measurement of the hydraulic-resistance coefficient of a plate-type spacer grid depending on the Reynolds number are presented. On the basis of the experimental data obtained, recommendations for improvement of the method of calculating the flow rate of a coolant in the cells of the fissile core of a KLT-40S reactor were developed. The results of investigations of the local hydrodynamic and mass-exchange characteristics of the coolant flow in the fuel assembly of the KLT-40S reactor plant were accepted for estimating the thermal and technical reliability of the fissile cores of KLT-40S reactors and were included in the database for verification of computational hydrodynamics programs (CFD codes).

  8. Fissile fuel production and usage of thermal reactor waste fueled with UO2 by means of hybrid reactor system

    International Nuclear Information System (INIS)

    Ipek, O.

    1997-01-01

    The use of Fast Breeder Reactors to produce fissile fuel from nuclear waste and the operation of these reactors with a new neutron source are becoming today' topic. In the thermonuclear reactors, it is possible to use 2.45-14.1 MeV - neutrons which can be obtained by D-T, D-D Semicatalyzed (D-D) and other fusion reactions. To be able to do these, Hybrid Reactor System, which still has experimental and theoretical studies, have to be taken into consideration.In this study, neutronic analysis of hybrid blanket with grafit reflector, is performed. D-T driven fusion reaction is surrounded by UO 2 fuel layer and the production of ''2''3''9Pu fissile fuel from waste ''2''3''8U is analyzed. It is also compared to the other possible fusion reactions. The results show that 815.8 kg/year ''2''3''8Pu with D-T reaction and 1431.6 kg/year ''2''3''8Pu with semicatalyzed (D-D) reaction can be produced for 1000 MW fusion power. This means production of 2.8/ year and 4.94/ year LWR respectively. In addition, 1000 MW fusion flower is is multiplicated to 3415 MW and 4274 MW for D-T and semicatalyzed (D-D) reactions respectively. The system works subcritical and these values are 0.4115 and 0.312 in order. The calculations, ANISN-ORNL code, S 16 -P 3 approach and DLC36 data library are used

  9. Gallium-cladding compatibility testing plan. Phases 1 and 2: Test plan for gallium corrosion tests. Revision 2

    International Nuclear Information System (INIS)

    Wilson, D.F.; Morris, R.N.

    1998-05-01

    This test plan is a Level-2 document as defined in the Fissile Materials Disposition Program Light-Water-Reactor Mixed-Oxide Fuel Irradiation Test Project Plan. The plan summarizes and updates the projected Phases 1 and 2 Gallium-Cladding compatibility corrosion testing and the following post-test examination. This work will characterize the reactions and changes, if any, in mechanical properties that occur between Zircaloy clad and gallium or gallium oxide in the temperature range 30--700 C

  10. Minimization of the fission product waste by using thorium based fuel instead of uranium dioxide

    Energy Technology Data Exchange (ETDEWEB)

    Galahom, A. Abdelghafar, E-mail: Agalahom@yahoo.com

    2017-04-01

    This research discusses the neutronic characteristics of VVER-1200 assembly fueled with five different fuel types based on thorium. These types of fuel based on mixing thorium as a fertile material with different fissile materials. The neutronic characteristics of these fuels are investigated by comparing their neutronic characteristics with the conventional uranium dioxide fuel using the MCNPX code. The objective of this study is to reduce the production of long-lived actinides, get rid of plutonium component and to improve the fuel cycle economy while maintaining acceptable values of the neutronic safety parameters such as moderator temperature coefficient, Doppler coefficient and effective delayed neutrons (β). The thorium based fuel has a more negative Doppler coefficient than uranium dioxide fuel. The moderator temperature coefficient (MTC) has been calculated for the different proposed fuels. Also, the fissile inventory ratio has been calculated at different burnup step. The use of Th-232 as a fertile material instead of U-238 in a nuclear fuel is the most promising fuel in VVER-1200 as it is the ideal solution to avoid the production of more plutonium components and long-lived minor actinides. The reactor grade plutonium accumulated in light water reactor with burnup can be recycled by mixing it with Th-232 to fuel the VVER-1200 assembly. The concentrations of Xe-135 and Sm-151 have been investigated, due to their high thermal neutron absorption cross section.

  11. Detecting nuclear warheads

    International Nuclear Information System (INIS)

    Fetter, S.; Frolov, V.A.; Prilutsky, O.F.; Rodionov, S.N.; Sagdeev, R.Z.; Miller, M.

    1992-01-01

    To the best of our knowledge, all nuclear weapons contain at least several kilograms of fissile material - material that can sustain a chain reaction. Such material provides the energy for fission explosives such as those that destroyed Hiroshima and Nagasaki; it is also used in the fission trigger modern thermonuclear weapons. The two fissile materials used in US and Soviet warheads are weapon grade uranium (WgU) and weapon-grade plutonium (WgPu). Fissile materials are radioactive; they are very dense and absorb certain radiations very well; and they can be fissioned. This paper reports on the two basic ways to detect fissile material: passive detection of the radiation emitted by its radioactive decay, or active detection involving either radiographing (x-raying) an object with neutrons or high-energy photons and detecting particles emitted by the resulting induced fissions. Passive detection is the preferred technique for verification purposes because of its simplicity and safety

  12. Calculational assessment of critical experiments with mixed-oxide fuel pin arrays moderated by organic solution

    International Nuclear Information System (INIS)

    Smolen, G.R.; Funabashi, H.

    1987-01-01

    Critical experiments have been conducted with organically moderated mixed-oxide (MOX) fuel pin assemblies at the Pacific Northwest Lab. Critical Mass Lab. These experiments are part of a joint exchange program between the US Dept. of Energy and the Power Reactor and Nuclear Fuel Development Corp. of Japan in the area of criticality data development. The purpose of these experiments is to benchmark computer codes and cross-section libraries and to assess the reactivity difference between systems moderated by water and those moderated by an organic solution. Past studies have indicated that some organic mixtures may be better moderators than water. This topic is of particular importance to the criticality safety of fuel processing plants where fissile material is dissolved in organic solutions during the solvent extraction process. In the past, it has been assumed that the codes and libraries benchmarked with water-moderated experiments were adequate when performing design and licensing studies of organically moderated systems. Calculations presented in this paper indicated that the Scale code system and the 27-energy-group cross-section library accurately compute k/sub eff/ for organically moderated MOX fuel pin assemblies. Furthermore, the reactivity of an organic solution with a 32 vol % TBP/68 vol% NPH mixture in a heterogeneous configuration is the same, for practical purposes, as water

  13. Calculational assessment of critical experiments with mixed oxide fuel pin arrays moderated by organic solution

    International Nuclear Information System (INIS)

    Smolen, G.R.

    1987-01-01

    Critical experiments have been conducted with organic-moderated mixed oxide (MOX) fuel pin assemblies at the Pacific Northwest Laboratory (PNL) Critical Mass Laboratory (CML). These experiments are part of a joint exchange program between the United States Department of Energy (USDOE) and the Power Reactor and Nuclear Fuel Development Corporation (PNC) of Japan in the area of criticality data development. The purpose of these experiments is to benchmark computer codes and cross-section libraries and to assess the reactivity difference between systems moderated by water and those moderated by an organic solution. Past studies have indicated that some organic mixtures may be better moderators than water. This topic is of particular importance to the criticality safety of fuel processing plants where fissile material is dissolved in organic solutions during the solvent extraction process. In the past, it has been assumed that the codes and libraries benchmarked with water-moderated experiments were adequate when performing design and licensing studies of organic-moderated systems. Calculations presented in this paper indicated that the SCALE code system and the 27-energy-group cross-section accurately compute k-effectives for organic moderated MOX fuel-pin assemblies. Furthermore, the reactivity of an organic solution with a 32-vol-% TBP/68-vol-% NPH mixture in a heterogeneous configuration is the same, for practical purposes, as water. 5 refs

  14. 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

  15. Composite material dosimeters

    Science.gov (United States)

    Miller, Steven D.

    1996-01-01

    The present invention is a composite material containing a mix of dosimeter material powder and a polymer powder wherein the polymer is transparent to the photon emission of the dosimeter material powder. By mixing dosimeter material powder with polymer powder, less dosimeter material is needed compared to a monolithic dosimeter material chip. Interrogation is done with excitation by visible light.

  16. Safety analysis report for the Neutron Multiplier Facility, 329 Building

    International Nuclear Information System (INIS)

    Rieck, H.G.

    1978-09-01

    Neutron multiplication is a process wherein the flux of a neutron source such as 252 Cf is enhanced by fission reactions that occur in a subcritical assemblage of fissile material. The multiplication factor of the device depends upon the consequences of neutron reactions with matter and is independent of the initial number of neutrons present. Safe utilization of such a device demands that the fissile material assemblage be maintained in a subcritical state throughout all normal and credibly abnormal conditions. Examples of things that can alter the multiplication factor (and degree of subcriticality) are temperature fluctuations, changes in moderator material such as voiding or composition, addition of fissile materials, and change in assembly configuration. The Neutron Multiplier Facility (NMF) utilizes a multiplier- 252 Cf assembly to produce neutrons for activation analysis of organic and inorganic environmental samples and for on-line mass spectrometry analysis of fission products which diffuse from a stationary fissile target (less than or equal to 4 g fissile material) located in the Neutron Multiplier. The NMF annex to the 329 Building provides close proximity to related counting equipment, and delay between sample irradiation and counting is minimized

  17. 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

  18. FMDP reactor alternative summary report. Volume 1 - existing LWR alternative

    International Nuclear Information System (INIS)

    Greene, S.R.; Bevard, B.B.

    1996-01-01

    Significant quantities of weapons-usable fissile materials [primarily plutonium and highly enriched uranium (HEU)] are becoming surplus to national defense needs in both the United States 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 (ES ampersand H) consequences if surplus fissile materials are not properly managed. This document summarizes the results of analysis concerned with existing light water reactor plutonium disposition alternatives

  19. FMDP reactor alternative summary report. Volume 1 - existing LWR alternative

    Energy Technology Data Exchange (ETDEWEB)

    Greene, S.R.; Bevard, B.B. [and others

    1996-10-07

    Significant quantities of weapons-usable fissile materials [primarily plutonium and highly enriched uranium (HEU)] are becoming surplus to national defense needs in both the United States 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 (ES&H) consequences if surplus fissile materials are not properly managed. This document summarizes the results of analysis concerned with existing light water reactor plutonium disposition alternatives.

  20. effects of mixed of mixed of mixed alkaline earth oxides in potash

    African Journals Online (AJOL)

    eobe

    Si, P) are network formers, and that materials whose. Nigerian ... made by mixing sand (SiO2), potassium carbonates. (K2Co3) .... The edges of the glass were grounded using ..... surface energies of minerals; theoritical estimate for oxides ...