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Sample records for reactor fuel characterization

  1. Cermet fuel for fast reactor - Fabrication and characterization

    NASA Astrophysics Data System (ADS)

    Mishra, Sudhir; Kutty, P. S.; Kutty, T. R. G.; Das, Shantanu; Dey, G. K.; Kumar, Arun

    2013-11-01

    (U, Pu)O2 ceramic fuel is the well-established fuel for the fast reactors and (U, Pu, Zr) metallic fuel is the future fuel. Both the fuels have their own merits and demerits. Optimal solution may lie in opting for a fuel which combines the favorable features of both fuel systems. The choice may be the use of cermet fuel which can be either (U, PuO2) or (Enriched U, UO2). In the present study, attempt has been made to fabricate (Natural U, UO2) cermet fuel by powder metallurgy route. Characterization of the fuel has been carried out using dilatometer, differential thermal analyzer, X-ray diffractometer, and Scanning Electron Microscope. The results show a high solidus temperature, high thermal expansion, presence of porosities, etc. in the fuel. The thermal conductivity of the fuel has also been measured. X-ray diffraction study on the fuel compact reveals presence of α U and UO2 phases in the matrix of the fuel.

  2. Reactor Physics Characterization of the HTR Module with UCO Fuel

    SciTech Connect

    Gerhard Strydom

    2011-01-01

    ABSTRACT The HTR Module [1] is a graphite-moderated, helium cooled pebble bed High Temperature Reactor (HTR) design that has been extensively used as a reference template for the former South African and current Chinese HTR [2] programs. This design utilized spherical fuel elements packed into a dynamic pebble bed, consisting of TRISO coated uranium oxide (UO2) fuel kernels with a U-235 enrichment of 7.8% and a Heavy Metal loading of 7 grams per pebble. The main objective of this study is to compare several important reactor physics and core design parameters for the HTR Module and an identical design utilizing UCO fuel kernels. Fuel kernels of this type are currently being tested in the Idaho National Laboratory’s (INL) Advanced Test Reactor (ATR) as part of the larger Next Generation Nuclear Plant (NGNP) project. The PEBBED-THERMIX [3] code, which was developed specifically for the analysis of pebble bed HTRs, was used to compare the coupled neutronic and thermal fluid performance of the two designs.

  3. Characterization of Hanford N Reactor spent fuel and K Basin sludges

    SciTech Connect

    Makenas, B.J.; Omberg, R.P.; Trimble, D.J.; Baker, R.B.

    1996-01-01

    Characterization is in progress for the N Reactor fuel stored in the Hanford K Basins. These activities` support the strategy for removal of fuel from the basins and storage of fuel in a dry condition at an area remote from the Columbia River. This strategy currently consists of placing fuel in a Multi-Canister Overpack (MCO), drying the fuel while it resides in the MCO and conditioning some portion of the fuel to reduce its chemical reactivity. Characterization includes the examination of fuel, canisters, and associated sludge. It consists firstly of in-basin activities such as visual examination, sludge depth measurements, and sampling of gas and liquid in canisters. Secondly characterization ecompasses the examination of samples of fuel and sludge which have been removed from the basins and shipped to laboratories. This paper presents observations made in the basins during the most recent attempts to ship samples from the basins and data obtained in the laboratory hotcells.

  4. Reactor Physics Scoping and Characterization Study on Implementation of TRIGA Fuel in the Advanced Test Reactor

    SciTech Connect

    Jennifer Lyons; Wade R. Marcum; Mark D. DeHart; Sean R. Morrell

    2014-01-01

    The Advanced Test Reactor (ATR), under the Reduced Enrichment for Research and Test Reactors (RERTR) Program and the Global Threat Reduction Initiative (GTRI), is conducting feasibility studies for the conversion of its fuel from a highly enriched uranium (HEU) composition to a low enriched uranium (LEU) composition. These studies have considered a wide variety of LEU plate-type fuels to replace the current HEU fuel. Continuing to investigate potential alternatives to the present HEU fuel form, this study presents a preliminary analysis of TRIGA® fuel within the current ATR fuel envelopes and compares it to the functional requirements delineated by the Naval Reactors Program, which includes: greater than 4.8E+14 fissions/s/g of 235U, a fast to thermal neutron flux ratio that is less than 5% deviation of its current value, a constant cycle power within the corner lobes, and an operational cycle length of 56 days at 120 MW. Other parameters outside those put forth by the Naval Reactors Program which are investigated herein include axial and radial power profiles, effective delayed neutron fraction, and mean neutron generation time.

  5. Safety Evaluation for Packaging for the N Reactor/single pass reactor fuel characterization shipments

    SciTech Connect

    Stevens, P.F.

    1994-10-13

    The purpose of this Safety Evaluation for Packaging (SEP) is to authorize the ChemNuclear CNS 1-13G packaging to ship samples of irradiated fuel elements from the 100 K East and 100 K West basins to the Postirradiation Testing Laboratory (PTL) in support of the spent nuclear fuel characterization effort. It also authorizes the return of the fuel element samples to the 100 K East facility using the same packaging. The CNS 1-13G cask has been-chosen to transport the fuel because it has a Certificate of Compliance (CoC) issued by the US Nuclear Regulatory Commission (NRC) for transporting irradiated oxide and metal fuel in commerce. It is capable of being loaded and offloaded underwater and may be shipped with water in the payload compartment.

  6. Cermet fuel reactors

    SciTech Connect

    Cowan, C.L.; Palmer, R.S.; Van Hoomissen, J.E.; Bhattacharyya, S.K.; Barner, J.O.

    1987-09-01

    Cermet fueled nuclear reactors are attractive candidates for high performance space power systems. The cermet fuel consists of tungsten-urania hexagonal fuel blocks characterized by high strength at elevated temperatures, a high thermal conductivity and resultant high thermal shock resistance. Key features of the cermet fueled reactor design are (1) the ability to achieve very high coolant exit temperatures, and (2) thermal shock resistance during rapid power changes, and (3) two barriers to fission product release - the cermet matrix and the fuel element cladding. Additionally, thre is a potential for achieving a long operating life because of (1) the neutronic insensitivity of the fast-spectrum core to the buildup of fission products and (2) the utilization of a high strength refractory metal matrix and structural materials. These materials also provide resistance against compression forces that potentially might compact and/or reconfigure the core. In addition, the neutronic properties of the refractory materials assure that the reactor remains substantially subcritical under conditions of water immersion. It is concluded that cermet fueled reactors can be utilized to meet the power requirements for a broad range of advanced space applications. 4 refs., 4 figs., 3 tabs.

  7. Characterization of commercially pure aluminum powder for research reactor fuel plates

    SciTech Connect

    Downs, V.D.; Wiencek, T.C.

    1992-11-01

    Aluminum powder is used as the matrix material in the production of uranium aluminide, oxide, and silicide dispersion fuel plates for research and test reactors. variability in the characteristics of the aluminum powder, such as moisture content and particle-size distribution, influences blending and compacting of the aluminum/fuel powder. A detailed study was performed to characterize the physical properties of three aluminum powder lots. An angle-of-shear test was devised to characterize the cohesiveness of the aluminum powder. Flow-rate measurements, apparent density determination, subsieve analysis, surface area measurements, and scanning electron microscopy were also used in the study. It was found that because of the various types of commercially available powders, proper specification of powder variables will ensure the receipt of consistent raw materials. Improved control of the initial powder will reduce the variability of fuel-plate production and will improve overall plate reproducibility. It is recommended that a standard specification be written for the aluminum powder and silicide fuel.

  8. Characterization of commercially pure aluminum powder for research reactor fuel plates

    SciTech Connect

    Downs, V.D. ); Wiencek, T.C. )

    1992-01-01

    Aluminum powder is used as the matrix material in the production of uranium aluminide, oxide, and silicide dispersion fuel plates for research and test reactors. variability in the characteristics of the aluminum powder, such as moisture content and particle-size distribution, influences blending and compacting of the aluminum/fuel powder. A detailed study was performed to characterize the physical properties of three aluminum powder lots. An angle-of-shear test was devised to characterize the cohesiveness of the aluminum powder. Flow-rate measurements, apparent density determination, subsieve analysis, surface area measurements, and scanning electron microscopy were also used in the study. It was found that because of the various types of commercially available powders, proper specification of powder variables will ensure the receipt of consistent raw materials. Improved control of the initial powder will reduce the variability of fuel-plate production and will improve overall plate reproducibility. It is recommended that a standard specification be written for the aluminum powder and silicide fuel.

  9. Pre-conceptual Development and characterization of an extruded graphite composite fuel for the TREAT Reactor

    SciTech Connect

    Luther, Erik; Rooyen, Isabella van; Leckie, Rafael; Papin, Pallas; Nelson, Andrew; Hunter, James

    2015-03-01

    In an effort to explore fuel systems that are more robust under accident scenarios, the DOE-NE has identified the need to resume transient testing. The Transient Reactor Test (TREAT) facility has been identified as the preferred option for the resumption of transient testing of nuclear fuel in the United States. In parallel, NNSA’s Global Threat Reduction Initiative (GTRI) Convert program is exploring the needs to replace the existing highly enriched uranium (HEU) core with low enriched uranium (LEU) core. In order to construct a new LEU core, materials and fabrication processes similar to those used in the initial core fabrication must be identified, developed and characterized. In this research, graphite matrix fuel blocks were extruded and materials properties of were measured. Initially the extrusion process followed the historic route; however, the project was expanded to explore methods to increase the graphite content of the fuel blocks and explore modern resins. Materials properties relevant to fuel performance including density, heat capacity and thermal diffusivity were measured. The relationship between process defects and materials properties will be discussed.

  10. REACTOR FUEL SCAVENGING MEANS

    DOEpatents

    Coffinberry, A.S.

    1962-04-10

    A process for removing fission products from reactor liquid fuel without interfering with the reactor's normal operation or causing a significant change in its fuel composition is described. The process consists of mixing a liquid scavenger alloy composed of about 44 at.% plutoniunm, 33 at.% lanthanum, and 23 at.% nickel or cobalt with a plutonium alloy reactor fuel containing about 3 at.% lanthanum; removing a portion of the fuel and scavenger alloy from the reactor core and replacing it with an equal amount of the fresh scavenger alloy; transferring the portion to a quiescent zone where the scavenger and the plutonium fuel form two distinct liquid layers with the fission products being dissolved in the lanthanum-rich scavenger layer; and the clean plutonium-rich fuel layer being returned to the reactor core. (AEC)

  11. NUCLEAR REACTOR FUEL SYSTEMS

    DOEpatents

    Thamer, B.J.; Bidwell, R.M.; Hammond, R.P.

    1959-09-15

    Homogeneous reactor fuel solutions are reported which provide automatic recombination of radiolytic gases and exhibit large thermal expansion characteristics, thereby providing stability at high temperatures and enabling reactor operation without the necessity of apparatus to recombine gases formed by the radiolytic dissociation of water in the fuel and without the necessity of liquid fuel handling outside the reactor vessel except for recovery processes. The fuels consist of phosphoric acid and water solutions of enriched uranium, wherein the uranium is in either the hexavalent or tetravalent state.

  12. NEUTRONIC REACTOR FUEL COMPOSITION

    DOEpatents

    Thurber, W.C.

    1961-01-10

    Uranium-aluminum alloys in which boron is homogeneously dispersed by adding it as a nickel boride are described. These compositions have particular utility as fuels for neutronic reactors, boron being present as a burnable poison.

  13. NEUTRONIC REACTOR FUEL ELEMENT

    DOEpatents

    Picklesimer, M.L.; Thurber, W.C.

    1961-01-01

    A chemically nonreactive fuel composition for incorporation in aluminum- clad, plate type fuel elements for neutronic reactors is described. The composition comprises a mixture of aluminum and uranium carbide particles, the uranium carbide particles containing at least 80 wt.% UC/sub 2/.

  14. NUCLEAR REACTOR FUEL ELEMENT

    DOEpatents

    Currier, E.L. Jr.; Nicklas, J.H.

    1963-06-11

    A fuel plate is designed for incorporation into control rods of the type utilized in high-flux test reactors. The fuel plate is designed so that the portion nearest the poison section of the control rod contains about one-half as much fissionable material as in the rest of the plate, thereby eliminating dangerous flux peaking in that portion. (AEC)

  15. NEUTRONIC REACTOR FUEL ELEMENT

    DOEpatents

    Horning, W.A.; Lanning, D.D.; Donahue, D.J.

    1959-10-01

    A fuel slug for a reactor which acts as a safety device is described. The fuel slug is an aluminum tube with a foil lining the inside surface of the tube, the foil being fabricated of uranium in a lead matrix.

  16. Hybrid reactors. [Fuel cycle

    SciTech Connect

    Moir, R.W.

    1980-09-09

    The rationale for hybrid fusion-fission reactors is the production of fissile fuel for fission reactors. A new class of reactor, the fission-suppressed hybrid promises unusually good safety features as well as the ability to support 25 light-water reactors of the same nuclear power rating, or even more high-conversion-ratio reactors such as the heavy-water type. One 4000-MW nuclear hybrid can produce 7200 kg of /sup 233/U per year. To obtain good economics, injector efficiency times plasma gain (eta/sub i/Q) should be greater than 2, the wall load should be greater than 1 MW.m/sup -2/, and the hybrid should cost less than 6 times the cost of a light-water reactor. Introduction rates for the fission-suppressed hybrid are usually rapid.

  17. NUCLEAR REACTOR FUEL ELEMENT

    DOEpatents

    Wheelock, C.W.; Baumeister, E.B.

    1961-09-01

    A reactor fuel element utilizing fissionable fuel materials in plate form is described. This fuel element consists of bundles of fuel-bearing plates. The bundles are stacked inside of a tube which forms the shell of the fuel element. The plates each have longitudinal fins running parallel to the direction of coolant flow, and interspersed among and parallel to the fins are ribs which position the plates relative to each other and to the fuel element shell. The plate bundles are held together by thin bands or wires. The ex tended surface increases the heat transfer capabilities of a fuel element by a factor of 3 or more over those of a simple flat plate.

  18. FUEL ASSAY REACTOR

    DOEpatents

    Spinrad, B.I.; Sandmeier, H.A.; Martens, F.H.

    1962-12-25

    A reactor having maximum sensitivity to perturbations is described comprising a core consisting of a horizontally disposed, rectangular, annular fuel zone containing enriched uranium dioxide dispersed in graphite, the concentration of uranium dioxide increasing from the outside to the inside of the fuel zone, an internal reflector of graphite containing an axial test opening disposed within the fuel zone, an external graphite reflector, means for changing the neutron spectrum in the test opening, and means for measuring perturbations in the neutron flux caused by the introduction of different fuel elements into the test opening. (AEC)

  19. NEUTRONIC REACTOR FUEL PUMP

    DOEpatents

    Cobb, W.G.

    1959-06-01

    A reactor fuel pump is described which offers long life, low susceptibility to radiation damage, and gaseous fission product removal. An inert-gas lubricated bearing supports a journal on one end of the drive shsft. The other end has an impeller and expansion chamber which effect pumping and gas- liquid separation. (T.R.H.)

  20. NEUTRONIC REACTOR FUEL ELEMENT

    DOEpatents

    Stacy, J.T.

    1958-12-01

    A reactor fuel element having a core of molybdenum-uranium alloy jacketed in stainless steel is described. A barrier layer of tungsten, tantalum, molybdenum, columbium, or silver is interposed between the core and jacket to prevent formation of a low melting eutectic between uranium and the varlous alloy constituents of the stainless steel.

  1. JACKETED REACTOR FUEL ELEMENT

    DOEpatents

    Smith, K.F.; Van Thyne, R.J.

    1958-12-01

    A fuel element is described for fast reactors comprised of a core of uranium metal containing material and a jacket around the core, the jacket consisting of from 2.5 to 15 percent of titanium, from 1 to 5 percent of niobium, and from 80 to 96.5 percent of vanadium.

  2. Covariance Applications in Criticality Safety, Light Water Reactor Analysis, and Spent Fuel Characterization

    NASA Astrophysics Data System (ADS)

    Williams, M. L.; Wiarda, D.; Ilas, G.; Marshall, W. J.; Rearden, B. T.

    2015-01-01

    A new covariance data library based on ENDF/B-VII.1 was recently processed for the SCALE nuclear analysis code system. The multigroup covariance data are discussed here, along with testing and application results for critical benchmark experiments. The cross section covariance library, along with covariances for fission product yields and decay data, is used to compute uncertainties in the decay heat produced by a burned reactor fuel assembly.

  3. Covariance Applications in Criticality Safety, Light Water Reactor Analysis, and Spent Fuel Characterization

    DOE PAGESBeta

    Williams, M. L.; Wiarda, D.; Ilas, G.; Marshall, W. J.; Rearden, B. T.

    2014-06-15

    Recently, we processed a new covariance data library based on ENDF/B-VII.1 for the SCALE nuclear analysis code system. The multigroup covariance data are discussed here, along with testing and application results for critical benchmark experiments. Moreover, the cross section covariance library, along with covariances for fission product yields and decay data, is used to compute uncertainties in the decay heat produced by a burned reactor fuel assembly.

  4. Covariance Applications in Criticality Safety, Light Water Reactor Analysis, and Spent Fuel Characterization

    SciTech Connect

    Williams, M.L. Wiarda, D.; Ilas, G.; Marshall, W.J.; Rearden, B.T.

    2015-01-15

    A new covariance data library based on ENDF/B-VII.1 was recently processed for the SCALE nuclear analysis code system. The multigroup covariance data are discussed here, along with testing and application results for critical benchmark experiments. The cross section covariance library, along with covariances for fission product yields and decay data, is used to compute uncertainties in the decay heat produced by a burned reactor fuel assembly.

  5. COMPARTMENTED REACTOR FUEL ELEMENT

    DOEpatents

    Cain, F.M. Jr.

    1962-09-11

    A method of making a nuclear reactor fuel element of the elongated red type is given wherein the fissionable fuel material is enclosed within a tubular metal cladding. The method comprises coating the metal cladding tube on its inside wall with a brazing alloy, inserting groups of cylindrical pellets of fissionable fuel material into the tube with spacing members between adjacent groups of pellets, sealing the ends of the tubes to leave a void space therewithin, heating the tube and its contents to an elevated temperature to melt the brazing alloy and to expand the pellets to their maximum dimensions under predetermined operating conditions thereby automatically positioning the spacing members along the tube, and finally cooling the tube to room temperature whereby the spacing disks become permanently fixed at their edges in the brazing alloy and define a hermetically sealed compartment for each fl group of fuel pellets. Upon cooling, the pellets contract thus leaving a space to accommodate thermal expansion of the pellets when in use in a reactor. The spacing members also provide lateral support for the tubular cladding to prevent collapse thereof when subjected to a reactor environment. (AEC)

  6. Nuclear reactor fuel element

    DOEpatents

    Johnson, Carl E.; Crouthamel, Carl E.

    1980-01-01

    A nuclear reactor fuel element is described which has an outer cladding, a central core of fissionable or mixed fissionable and fertile fuel material and a layer of oxygen gettering material on the inner surface of the cladding. The gettering material reacts with oxygen released by the fissionable material during irradiation of the core thereby preventing the oxygen from reacting with and corroding the cladding. Also described is an improved method for coating the inner surface of the cladding with a layer of gettering material.

  7. Fossil fuel furnace reactor

    DOEpatents

    Parkinson, William J.

    1987-01-01

    A fossil fuel furnace reactor is provided for simulating a continuous processing plant with a batch reactor. An internal reaction vessel contains a batch of shale oil, with the vessel having a relatively thin wall thickness for a heat transfer rate effective to simulate a process temperature history in the selected continuous processing plant. A heater jacket is disposed about the reactor vessel and defines a number of independent controllable temperature zones axially spaced along the reaction vessel. Each temperature zone can be energized to simulate a time-temperature history of process material through the continuous plant. A pressure vessel contains both the heater jacket and the reaction vessel at an operating pressure functionally selected to simulate the continuous processing plant. The process yield from the oil shale may be used as feedback information to software simulating operation of the continuous plant to provide operating parameters, i.e., temperature profiles, ambient atmosphere, operating pressure, material feed rates, etc., for simulation in the batch reactor.

  8. NUCLEAR REACTOR FUEL-BREEDER FUEL ELEMENT

    DOEpatents

    Currier, E.L. Jr.; Nicklas, J.H.

    1962-08-14

    A fuel-breeder fuel element was developed for a nuclear reactor wherein discrete particles of fissionable material are dispersed in a matrix of fertile breeder material. The fuel element combines the advantages of a dispersion type and a breeder-type. (AEC)

  9. REACTOR FUEL ELEMENTS TESTING CONTAINER

    DOEpatents

    Whitham, G.K.; Smith, R.R.

    1963-01-15

    This patent shows a method for detecting leaks in jacketed fuel elements. The element is placed in a sealed tank within a nuclear reactor, and, while the reactor operates, the element is sparged with gas. The gas is then led outside the reactor and monitored for radioactive Xe or Kr. (AEC)

  10. Fabrication and Pre-irradiation Characterization of a Minor Actinide and Rare Earth Containing Fast Reactor Fuel Experiment for Irradiation in the Advanced Test Reactor

    SciTech Connect

    Timothy A. Hyde

    2012-06-01

    The United States Department of Energy, seeks to develop and demonstrate the technologies needed to transmute the long-lived transuranic actinide isotopes contained in spent nuclear fuel into shorter lived fission products, thereby decreasing the volume of material requiring disposal and reducing the long-term radiotoxicity and heat load of high-level waste sent to a geologic repository. This transmutation of the long lived actinides plutonium, neptunium, americium and curium can be accomplished by first separating them from spent Light Water Reactor fuel using a pyro-metalurgical process, then reprocessing them into new fuel with fresh uranium additions, and then transmuted to short lived nuclides in a liquid metal cooled fast reactor. An important component of the technology is developing actinide-bearing fuel forms containing plutonium, neptunium, americium and curium isotopes that meet the stringent requirements of reactor fuels and materials.

  11. FUEL ELEMENTS FOR NEUTRONIC REACTORS

    DOEpatents

    Foote, F.G.; Jette, E.R.

    1963-05-01

    A fuel element for a nuclear reactor is described that consists of a jacket containing a unitary core of fissionable material and a filling of a metal of the group consisting of sodium and sodium-potassium alloys. (AEC)

  12. MOLTEN FLUORIDE NUCLEAR REACTOR FUEL

    DOEpatents

    Barton, C.J.; Grimes, W.R.

    1960-01-01

    Molten-salt reactor fuel compositions consisting of mixtures of fluoride salts are reported. In its broadest form, the composition contains an alkali fluoride such as sodium fluoride, zirconium tetrafluoride, and a uranium fluoride, the latter being the tetrafluoride or trifluoride or a mixture of the two. An outstanding property of these fuel compositions is a high coeffieient of thermal expansion which provides a negative temperature coefficient of reactivity in reactors in which they are used.

  13. FUSED REACTOR FUELS

    DOEpatents

    Mayer, S.W.

    1962-11-13

    This invention relates to a nuciear reactor fuel composition comprising (1) from about 0.01 to about 50 wt.% based on the total weight of said composition of at least one element selected from the class consisting of uranium, thorium, and plutonium, wherein said eiement is present in the form of at least one component selected from the class consisting of oxides, halides, and salts of oxygenated anions, with components comprising (2) at least one member selected from the class consisting of (a) sulfur, wherein the sulfur is in the form of at least one entity selected irom the class consisting of oxides of sulfur, metal sulfates, metal sulfites, metal halosulfonates, and acids of sulfur, (b) halogen, wherein said halogen is in the form of at least one compound selected from the class of metal halides, metal halosulfonates, and metal halophosphates, (c) phosphorus, wherein said phosphorus is in the form of at least one constituent selected from the class consisting of oxides of phosphorus, metal phosphates, metal phosphites, and metal halophosphates, (d) at least one oxide of a member selected from the class consisting of a metal and a metalloid wherein said oxide is free from an oxide of said element in (1); wherein the amount of at least one member selected from the class consisting of halogen and sulfur is at least about one at.% based on the amount of the sum of said sulfur, halogen, and phosphorus atom in said composition; and wherein the amount of said 2(a), 2(b) and 2(c) components in said composition which are free from said elements of uranium, thorium, arid plutonium, is at least about 60 wt.% based on the combined weight of the components of said composition which are free from said elements of uranium, thorium, and plutonium. (AEC)

  14. Characterization of bacterial and archaeal communities in air-cathode microbial fuel cells, open circuit and sealed-off reactors.

    PubMed

    Shehab, Noura; Li, Dong; Amy, Gary L; Logan, Bruce E; Saikaly, Pascal E

    2013-11-01

    A large percentage of organic fuel consumed in a microbial fuel cell (MFC) is lost as a result of oxygen transfer through the cathode. In order to understand how this oxygen transfer affects the microbial community structure, reactors were operated in duplicate using three configurations: closed circuit (CC; with current generation), open circuit (OC; no current generation), and sealed off cathodes (SO; no current, with a solid plate placed across the cathode). Most (98 %) of the chemical oxygen demand (COD) was removed during power production in the CC reactor (maximum of 640 ± 10 mW/m(2)), with a low percent of substrate converted to current (coulombic efficiency of 26.5 ± 2.1 %). Sealing the cathode reduced COD removal to 7 %, but with an open cathode, there was nearly as much COD removal by the OC reactor (94.5 %) as the CC reactor. Oxygen transfer into the reactor substantially affected the composition of the microbial communities. Based on analysis of the biofilms using 16S rRNA gene pyrosequencing, microbes most similar to Geobacter were predominant on the anodes in the CC MFC (72 % of sequences), but the most abundant bacteria were Azoarcus (42 to 47 %) in the OC reactor, and Dechloromonas (17 %) in the SO reactor. Hydrogenotrophic methanogens were most predominant, with sequences most similar to Methanobacterium in the CC and SO reactor, and Methanocorpusculum in the OC reactors. These results show that oxygen leakage through the cathode substantially alters the bacterial anode communities, and that hydrogenotrophic methanogens predominate despite high concentrations of acetate. The predominant methanogens in the CC reactor most closely resembled those in the SO reactor, demonstrating that oxygen leakage alters methanogenic as well as general bacterial communities.

  15. FUEL ELEMENT FOR NUCLEAR REACTORS

    DOEpatents

    Bassett, C.H.

    1961-05-16

    A fuel element particularly adapted for use in nuclear reactors of high power density is offered. It has fissionable fuel pellet segments mounted in a tubular housing and defining a central passage in the fuel element. A burnable poison element extends through the central passage, which is designed to contain more poison material at the median portion than at the end portions thereby providing a more uniform hurnup and longer reactivity life.

  16. Gaseous fuel nuclear reactor research

    NASA Technical Reports Server (NTRS)

    Schwenk, F. C.; Thom, K.

    1975-01-01

    Gaseous-fuel nuclear reactors are described; their distinguishing feature is the use of fissile fuels in a gaseous or plasma state, thereby breaking the barrier of temperature imposed by solid-fuel elements. This property creates a reactor heat source that may be able to heat the propellant of a rocket engine to 10,000 or 20,000 K. At this temperature level, gas-core reactors would provide the breakthrough in propulsion needed to open the entire solar system to manned and unmanned spacecraft. The possibility of fuel recycling makes possible efficiencies of up to 65% and nuclear safety at reduced cost, as well as high-thrust propulsion capabilities with specific impulse up to 5000 sec.

  17. FUEL ELEMENT FOR NUCLEAR REACTORS

    DOEpatents

    Bassett, C.H.

    1961-11-21

    A fuel element is designed which is particularly adapted for reactors of high power density used to generate steam for the production of electricity. The fuel element consists of inner and outer concentric tubes forming an annular chamber within which is contained fissionable fuel pellet segments, wedge members interposed between the fuel segments, and a spring which, acting with wedge members, urges said fuel pellets radially into contact against the inner surface of the outer tube. The wedge members may be a fertile material convertible into fissionable fuel material by absorbing neutrons emitted from the fissionable fuel pellet segments. The costly grinding of cylindrical fuel pellets to close tolerances for snug engagement is reduced because the need to finish the exact size is eliminated. (AEC)

  18. Spent fuel assembly hardware: Characterization and 10 CFR 61 classification for waste disposal: Volume 3, Calculated activity profiles of spent nuclear fuel assembly hardware for boiling water reactors

    SciTech Connect

    Short, S.M.; Luksic, A.T.; Schutz, M.E.

    1989-06-01

    Consolidation of spent fuel is under active consideration as the US Department of Energy plans to dispose of spent fuel as required by the Nuclear Waste Policy Act of 1982. During consolidation, the fuel pins are removed from an intact fuel assembly and repackaged into a more compact configuration. After repackaging, approximately 30 kg of residual spent fuel assembly hardware per assembly that is also radioactive and required disposal. Understanding the nature of this secondary waste stream is critical to designing a system that will properly handle, package, store, and dispose of the waste. This report presents a methodology for estimating the radionuclide inventory in irradiated spent fuel hardware. Ratios are developed that allow the use of ORIGEN2 computer code calculations to be applied to regions that are outside the fueled region. The ratios are based on the analysis of samples of irradiated hardware from spent fuel assemblies. The results of this research are presented in three volumes. In Volume 1, the development of scaling factors that can be used with ORIGEN2 calculations to estimate activation of spent fuel assembly hardware is documented. The results from laboratory analysis of irradiated spent-fuel hardware samples are also presented in Volume 1. In Volume 2 and 3, the calculated flux profiles of spent nuclear fuel assemblies are presented for pressurized water reactors and boiling water reactors, respectively. The results presented in Volumes 2 and 3 were used to develop the scaling factors documented in Volume 1.

  19. Spent fuel assembly hardware: Characterization and 10 CFR 61 classification for waste disposal: Volume 2, Calculated activity profiles of spent nuclear fuel assembly hardware for pressurized water reactors

    SciTech Connect

    Short, S.M.; Luksic, A.T.; Lotz, T.L.; Schutz, M.E.

    1989-06-01

    Consolidation of spent fuel is under active consideration as the US Department of Energy plans to dispose of spent fuel as required by the Nuclear Waste Policy Act of 1982. During consolidation, the fuel pins are removed from an intact fuel assembly and repackaged into a more compact configuration. After repackaging, approximately 30 kg of residual spent fuel assembly hardware per assembly remains that is also radioactive and requires disposal. Understanding the nature of this secondary waste stream is critical to designing a system that will properly handle, package, store, and dispose of the waste. This report present a methodology for estimating the radionuclide inventory in irradiated spent fuel hardware. Ratios are developed that allow the use of ORIGEN2 computer code calculations to be applied to regions that are outside the fueled region. The ratios are based on the analysis of samples of irradiated hardware from spent fuel assemblies. The results of this research are presented in three volumes. In Volume 1, the development of scaling factors that can be used with ORIGEN2 calculations to estimate activation of spent fuel assembly hardware is documented. The results from Laboratory analysis of irradiated spent-fuel hardware samples are also presented in Volume 1. In Volumes 2 and 3, the calculated flux profiles of spent nuclear fuel assemblies are presented for pressurized water reactors and boiling water reactors, respectively. The results presented in Volumes 2 and 3 were used to develop the scaling factors documented in Volume 1.

  20. FUEL ELEMENT FOR NUCLEAR REACTORS

    DOEpatents

    Bassett, C.H.

    1961-05-01

    A nuclear reactor fuel element comprising high density ceramic fissionable material enclosed in a tubular cladding of corrosion-resistant material is described. The fissionable material is in the form of segments of a tube which have cooperating tapered interfaces which produce outward radial displacement when the segments are urged axially together. A resilient means is provided within the tubular housing to constantly urge the fuel segments axially. This design maintains the fuel material in tight contacting engagement against the inner surface of the outer cladding tube to eliminate any gap therebetween which may be caused by differential thermal expansion between the fuel material and the material of the tube.

  1. FUEL ELEMENTS FOR NUCLEAR REACTORS

    DOEpatents

    Blainey, A.; Lloyd, H.

    1961-07-11

    A method of sheathing a tubular fuel element for a nuclear reactor is described. A low melting metal core member is centered in a die, a layer of a powdered sheathing substance is placed on the bottom of the die, the tubular fuel element is inserted in the die, the space between the tubular fuel element and the die walls and core member is filled with the same powdered sheathing substance, a layer of the same substance is placed over the fissile material, and the charge within the die is subjected to pressure in the direction of the axis of the fuel element at the sintering temperature of the protective substance.

  2. NEUTRONIC REACTOR FUEL ELEMENT

    DOEpatents

    Shackleford, M.H.

    1958-12-16

    A fuel element possessing good stability and heat conducting properties is described. The fuel element comprises an outer tube formed of material selected from the group consisting of stainhess steel, V, Ti. Mo. or Zr, a fuel tube concentrically fitting within the outer tube and containing an oxide of an isotope selected from the group consisting of U/sup 235/, U/sup 233/, and Pu/sup 239/, and a hollow, porous core concentrically fitting within the fuel tube and formed of an oxide of an element selected from the group consisting of Mg, Be, and Zr.

  3. NEUTRONIC REACTOR FUEL ELEMENT

    DOEpatents

    Gurinsky, D.H.; Powell, R.W.; Fox, M.

    1959-11-24

    A nuclear fuel element comprising a plurality of nuclear fuel bearing strips is presented. The strips are folded along their longitudinal axes to an angle of about 60 deg and are secured at each end by ferrule to form an elongated assembly suitable for occupying a cylindrical coolant channel.

  4. Criticality of spent reactor fuel

    SciTech Connect

    Harris, D.R.

    1987-01-01

    The storage capacity of spent reactor fuel pools can be greatly increased by consolidation. In this process, the fuel rods are removed from reactor fuel assemblies and are stored in close-packed arrays in a canister or skeleton. An earlier study examined criticality consideration for consolidation of Westinghouse fuel, assumed to be fresh, in canisters at the Millstone-2 spent-fuel pool and in the General Electric IF-300 shipping cask. The conclusions were that the fuel rods in the canister are so deficient in water that they are adequately subcritical, both in normal and in off-normal conditions. One potential accident, the water spill event, remained unresolved in the earlier study. A methodology is developed here for spent-fuel criticality and is applied to the water spill event. The methodology utilizes LEOPARD to compute few-group cross sections for the diffusion code PDQ7, which then is used to compute reactivity. These codes give results for fresh fuel that are in good agreement with KENO IV-NITAWL Monte Carlo results, which themselves are in good agreement with continuous energy Monte Carlo calculations. These methodologies are in reasonable agreement with critical measurements for undepleted fuel.

  5. FUEL ELEMENT FOR NUCLEAR REACTORS

    DOEpatents

    Dickson, J.J.

    1963-09-24

    A method is described whereby fuel tubes or pins are cut, loaded with fuel pellets and a heat transfer medium, sealed at each end with slotted fittings, and assembled into a rectangular tube bundle to form a fuel element. The tubes comprising the fuel element are laterally connected between their ends by clips and tabs to form a linear group of spaced parallel tubes, which receive their vertical support by resting on a grid. The advantages of this method are that it permits elimination of structural material (e.g., fuel-element cans) within the reactor core, and removal of at least one fuel pin from an element and replacement thereof so that a burnable poison may be utilized during the core lifetime. (AEC)

  6. Nuclear reactor composite fuel assembly

    DOEpatents

    Burgess, Donn M.; Marr, Duane R.; Cappiello, Michael W.; Omberg, Ronald P.

    1980-01-01

    A core and composite fuel assembly for a liquid-cooled breeder nuclear reactor including a plurality of elongated coextending driver and breeder fuel elements arranged to form a generally polygonal bundle within a thin-walled duct. The breeder elements are larger in cross section than the driver elements, and each breeder element is laterally bounded by a number of the driver elements. Each driver element further includes structure for spacing the driver elements from adjacent fuel elements and, where adjacent, the thin-walled duct. A core made up of the fuel elements can advantageously include fissile fuel of only one enrichment, while varying the effective enrichment of any given assembly or core region, merely by varying the relative number and size of the driver and breeder elements.

  7. Electron Microscopy Characterization of an As-Fabricated Research Reactor Fuel Plate Comprised of U-7Mo Particles Dispersed in an Al-2Si Alloy Matrix

    SciTech Connect

    Dennis D. Keiser, Jr.; J. Gan; J. F. Jue; B. D. Miller

    2010-11-01

    To understand the microstructural development of nuclear fuel plates during irradiation, it is imperative to know the microstructure of a fuel plate after all the fabrication steps have been completed and before it is inserted into the reactor. To this end, a U–7 wt.% Mo alloy research reactor dispersion fuel plate with Al–2 wt.% Si matrix was destructively examined using scanning and transmission electron microscopy to characterize the developed microstructure after fabrication. Of particular interest for this study was how the Si that was added to the fuel matrix partitioned between the various fuel plate phases during fabrication. Si was added to the matrix so that the microstructure that developed during fuel fabrication would exhibit good irradiation behavior. SEM analysis was used to identify the representative microstructure, the compositions of the various phases, and the partitioning behavior of the fuel and matrix constituents. TEM analysis was employed to definitively identify the phases in the U–7Mo alloy and the phases that formed due to diffusional interactions between the fuel particles and matrix during fuel plate fabrication. The TEM results are the first reported for an as-fabricated U–7 wt.% Mo dispersion fuel plate with an Al alloy matrix. SEM results showed that a significant portion of the original ?-(U–Mo) fuel particles had transformed to a lamellar microstructure, comprised of a-U and either ? or ?' phases, and the fuel/matrix interaction layers were enriched in Si. TEM analysis identified an ordered fcc (U–Mo)(Al–Si)3 type of phase, which formed at the decomposed U–7Mo/matrix interface and extended into the lamellar microstructure. Some regions of the U–7Mo particles retained the single-phase ?-(U–Mo). Small precipitate phases were observed in the fuel meat matrix that contained Fe, Al, and Si. The Si that is added to the matrix of a U–Mo dispersion fuel plate to improve irradiation performance appears to result in

  8. Corrosion of spent Advanced Test Reactor fuel

    SciTech Connect

    Lundberg, L.B.; Croson, M.L.

    1994-11-01

    The results of a study of the condition of spent nuclear fuel elements from the Advanced Test Reactor (ATR) currently being stored underwater at the Idaho National Engineering Laboratory (INEL) are presented. This study was motivated by a need to estimate the corrosion behavior of dried, spent ATR fuel elements during dry storage for periods up to 50 years. The study indicated that the condition of spent ATR fuel elements currently stored underwater at the INEL is not very well known. Based on the limited data and observed corrosion behavior in the reactor and in underwater storage, it was concluded that many of the fuel elements currently stored under water in the facility called ICPP-603 FSF are in a degraded condition, and it is probable that many have breached cladding. The anticipated dehydration behavior of corroded spent ATR fuel elements was also studied, and a list of issues to be addressed by fuel element characterization before and after forced drying of the fuel elements and during dry storage is presented.

  9. NUCLEAR REACTOR FUEL ELEMENT ASSEMBLY

    DOEpatents

    Stengel, F.G.

    1963-12-24

    A method of fabricating nuclear reactor fuel element assemblies having a plurality of longitudinally extending flat fuel elements in spaced parallel relation to each other to form channels is presented. One side of a flat side plate is held contiguous to the ends of the elements and a welding means is passed along the other side of the platertransverse to the direction of the longitudinal extension of the elements. The setting and speed of travel of the welding means is set to cause penetration of the side plate with welds at bridge the gap in each channel between adjacent fuel elements with a weld-through bubble of predetermined size. The fabrication of a high strength, dependable fuel element is provided, and the reduction of distortion and high production costs are facilitated by this method. (AEC)

  10. Fuel handling apparatus for a nuclear reactor

    DOEpatents

    Hawke, Basil C.

    1987-01-01

    Fuel handling apparatus for transporting fuel elements into and out of a nuclear reactor and transporting them within the reactor vessel extends through a penetration in the side of the reactor vessel. A lateral transport device carries the fuel elements laterally within the vessel and through the opening in the side of the vessel, and a reversible lifting device raises and lowers the fuel elements. In the preferred embodiment, the lifting device is supported by a pair of pivot arms.

  11. Fast Reactor Fuel Type and Reactor Safety Performance

    SciTech Connect

    R. Wigeland; J. Cahalan

    2009-09-01

    Fast Reactor Fuel Type and Reactor Safety Performance R. Wigeland , Idaho National Laboratory J. Cahalan, Argonne National Laboratory The sodium-cooled fast neutron reactor is currently being evaluated for the efficient transmutation of the highly-hazardous, long-lived, transuranic elements that are present in spent nuclear fuel. One of the fundamental choices that will be made is the selection of the fuel type for the fast reactor, whether oxide, metal, carbide, nitride, etc. It is likely that a decision on the fuel type will need to be made before many of the related technologies and facilities can be selected, from fuel fabrication to spent fuel reprocessing. A decision on fuel type should consider all impacts on the fast reactor system, including safety. Past work has demonstrated that the choice of fuel type may have a significant impact on the severity of consequences arising from accidents, especially for severe accidents of low probability. In this paper, the response of sodium-cooled fast reactors is discussed for both oxide and metal fuel types, highlighting the similarities and differences in reactor response and accident consequences. Any fast reactor facility must be designed to be able to successfully prevent, mitigate, or accommodate all consequences of potential events, including accidents. This is typically accomplished by using multiple barriers to the release of radiation, including the cladding on the fuel, the intact primary cooling system, and most visibly the reactor containment building. More recently, this has also included the use of ‘inherent safety’ concepts to reduce or eliminate the potential for serious damage in some cases. Past experience with oxide and metal fuel has demonstrated that both fuel types are suitable for use as fuel in a sodium-cooled fast reactor. However, safety analyses for these two fuel types have also shown that there can be substantial differences in accident consequences due to the neutronic and

  12. Stationary Liquid Fuel Fast Reactor

    SciTech Connect

    Yang, Won Sik; Grandy, Andrew; Boroski, Andrew; Krajtl, Lubomir; Johnson, Terry

    2015-09-30

    For effective burning of hazardous transuranic (TRU) elements of used nuclear fuel, a transformational advanced reactor concept named SLFFR (Stationary Liquid Fuel Fast Reactor) was proposed based on stationary molten metallic fuel. The fuel enters the reactor vessel in a solid form, and then it is heated to molten temperature in a small melting heater. The fuel is contained within a closed, thick container with penetrating coolant channels, and thus it is not mixed with coolant nor flow through the primary heat transfer circuit. The makeup fuel is semi- continuously added to the system, and thus a very small excess reactivity is required. Gaseous fission products are also removed continuously, and a fraction of the fuel is periodically drawn off from the fuel container to a processing facility where non-gaseous mixed fission products and other impurities are removed and then the cleaned fuel is recycled into the fuel container. A reference core design and a preliminary plant system design of a 1000 MWt TRU- burning SLFFR concept were developed using TRU-Ce-Co fuel, Ta-10W fuel container, and sodium coolant. Conservative design approaches were adopted to stay within the current material performance database. Detailed neutronics and thermal-fluidic analyses were performed to develop a reference core design. Region-dependent 33-group cross sections were generated based on the ENDF/B-VII.0 data using the MC2-3 code. Core and fuel cycle analyses were performed in theta-r-z geometries using the DIF3D and REBUS-3 codes. Reactivity coefficients and kinetics parameters were calculated using the VARI3D perturbation theory code. Thermo-fluidic analyses were performed using the ANSYS FLUENT computational fluid dynamics (CFD) code. Figure 0.1 shows a schematic radial layout of the reference 1000 MWt SLFFR core, and Table 0.1 summarizes the main design parameters of SLFFR-1000 loop plant. The fuel container is a 2.5 cm thick cylinder with an inner radius of 87.5 cm. The fuel

  13. Microstructure Characterization of RERTR Fuel

    SciTech Connect

    J. Gan; B. D. Miller; D. D. Keiser; T. R. Allen; D. M. Wachs

    2008-09-01

    A variety of phases have the potential to develop in the irradiated fuels for the reduced enrichment research test reactor (RERTR) program. To study the radiation stability of these potential phases, three depleted uranium alloys were cast. The phases of interest were identified including U(Si,Al)3, (U,Mo)(Si,Al)3, UMo2Al20, UAl4, and U6Mo4Al43. These alloys were irradiated with 2.6 MeV protons at 200ºC up to 3.0 dpa. The microstructure is characterized using SEM and TEM. Microstructural characterization for an archive dispersion fuel plate (U-7Mo fuel particles in Al-2%Si cladding) was also carried out. TEM sample preparation for the irradiated dispersion fuel has been developed.

  14. Proliferation Resistant Nuclear Reactor Fuel

    SciTech Connect

    Gray, L W; Moody, K J; Bradley, K S; Lorenzana, H E

    2011-02-18

    Global appetite for fission power is projected to grow dramatically this century, and for good reason. Despite considerable research to identify new sources of energy, fission remains the most plentiful and practical alternative to fossil fuels. The environmental challenges of fossil fuel have made the fission power option increasingly attractive, particularly as we are forced to rely on reserves in ecologically fragile or politically unstable corners of the globe. Caught between a globally eroding fossil fuel reserve as well as the uncertainty and considerable costs in the development of fusion power, most of the world will most likely come to rely on fission power for at least the remainder of the 21st century. Despite inevitable growth, fission power faces enduring challenges in sustainability and security. One of fission power's greatest hurdles to universal acceptance is the risk of potential misuse for nefarious purposes of fissionable byproducts in spent fuel, such as plutonium. With this issue in mind, we have discussed intrinsic concepts in this report that are motivated by the premise that the utility, desirability, and applicability of nuclear materials can be reduced. In a general sense, the intrinsic solutions aim to reduce or eliminate the quantity of existing weapons usable material; avoid production of new weapons-usable material through enrichment, breeding, extraction; or employ engineering solutions to make the fuel cycle less useful or more difficult for producing weapons-usable material. By their nature, these schemes require modifications to existing fuel cycles. As such, the concomitants of these modifications require engagement from the nuclear reactor and fuel-design community to fully assess their effects. Unfortunately, active pursuit of any scheme that could further complicate the spread of domestic nuclear power will probably be understandably unpopular. Nevertheless, the nonproliferation and counterterrorism issues are paramount, and

  15. EXPERIMENTAL LIQUID METAL FUEL REACTOR

    DOEpatents

    Happell, J.J.; Thomas, G.R.; Denise, R.P.; Bunts, J.L. Jr.

    1962-01-23

    A liquid metal fuel nuclear fission reactor is designed in which the fissionable material is dissolved or suspended in a liquid metal moderator and coolant. The liquid suspension flows into a chamber in which a critical amount of fissionable material is obtained. The fluid leaves the chamber and the heat of fission is extracted for power or other utilization. The improvement is in the support arrangement for a segrnented graphite core to permit dif ferential thermal expansion, effective sealing between main and blanket liquid metal flows, and avoidance of excessive stress development in the graphite segments. (AEC)

  16. Corrosion Minimization for Research Reactor Fuel

    SciTech Connect

    Eric Shaber; Gerard Hofman

    2005-06-01

    Existing university research reactors are being converted to use low-enriched uranium fue to eliminate the use of highly-enriched uranium. These conversions require increases in fuel loading that will result in the use of elements with more fuel plates, resulting in a net decrease in the water annulus between fuel plates. The proposed decrease in the water annulus raises questions about the requirements and stability of the surface hydroxide on the aluminum fuel cladding and the potential for runaway corrosion resulting in fuel over-temperature incidents. The Nuclear Regulatory Commission (NRC), as regulator for these university reactors, must ensure that proposed fuel modifications will not result in any increased risk or hazard to the reactor operators or the public. This document reviews the characteristics and behavior of aluminum hydroxides, analyzes the drivers for fuel plate corrosion, reviews relevant historical incidents, and provides recommendations on fuel design, surface treatment, and reactor operational practices to avoid corrosion issues.

  17. Coprocessing of thermal reactor fuels

    SciTech Connect

    Ballard, W.W. Jr.

    1985-01-01

    The Nuclear Power Development Division (NPD) under the Assistant Secretary for Energy Technology in the Department of Energy (DOE) is responsible for examining alternative nuclear reactor fuel recycle systems which have potential for reducing the risk of proliferation of nuclear weapons. NPD is administering a base technology program of research and development and design studies which will provide a sound technical foundation for evaluating the nonproliferation potential and commercial feasibility of these alternatives. The Savannah River Laboratory (SRL) has been assigned as the technical lead for those activities associated with the processing of thermal reactor fuel. In order to systematically identify technical requirements and design solutions, SRL periodically updates a Design Integration Study (DIS). The reference process being incorporated into the current DIS is coprocessing uranium and plutonium in a manner whereby pure plutonium is never available in a separate stream. As with other processes, coprocessing doesn't offer a technical fix for preventing proliferators. A flowsheet for this reference process is described with particular emphasis on technical issues and proliferation resistance advantages of coprocessing over conventional Purex processing.

  18. Fuel elements of research reactor CM

    SciTech Connect

    Kozlov, A.V.; Morozov, A.V.; Vatulin, A.V.; Ershov, S.A.

    2013-07-01

    In 1961 the CM research reactor was commissioned at the Research Institute of Atomic Reactors (Dimitrovgrad, Russia), it was intended to carry on investigations and the production of transuranium nuclides. The reactor is of a tank type. Original fuel assembly contained plate fuels that were spaced with vanes and corrugated bands. Nickel was used as a cladding material, fuel meat was produced from UO{sub 2} + electrolytic nickel composition. Fuel plates have been replaced by self-spacing cross-shaped dispersion fuels clad in stainless steel. In 2005 the reactor was updated. The purpose of this updating was to increase the quantity of irradiation channels in the reactor core and to improve the neutron balance. The updating was implemented at the expense of 20 % reduction in the quantity of fuel elements in the core which released a space for extra channels and decreased the mass of structural materials in the core. The updated reactor is loaded with modified standard fuel elements with 20 % higher uranium masses. At the same time stainless steel in fuel assembly shrouds was substituted by zirconium alloy. Today in progress are investigations and work to promote the second stage of reactor updating that involve developments of cross-shaped fuel elements having low neutron absorption matrix materials. This article gives an historical account of the design and main technical changes that occurred for the CM reactor since its commissioning.

  19. Fuels characterization studies. [jet fuels

    NASA Technical Reports Server (NTRS)

    Seng, G. T.; Antoine, A. C.; Flores, F. J.

    1980-01-01

    Current analytical techniques used in the characterization of broadened properties fuels are briefly described. Included are liquid chromatography, gas chromatography, and nuclear magnetic resonance spectroscopy. High performance liquid chromatographic ground-type methods development is being approached from several directions, including aromatic fraction standards development and the elimination of standards through removal or partial removal of the alkene and aromatic fractions or through the use of whole fuel refractive index values. More sensitive methods for alkene determinations using an ultraviolet-visible detector are also being pursued. Some of the more successful gas chromatographic physical property determinations for petroleum derived fuels are the distillation curve (simulated distillation), heat of combustion, hydrogen content, API gravity, viscosity, flash point, and (to a lesser extent) freezing point.

  20. K Basin spent nuclear fuel characterization

    SciTech Connect

    LAWRENCE, L.A.

    1999-02-10

    The results of the characterization efforts completed for the N Reactor fuel stored in the Hanford K Basins were Collected and summarized in this single referencable document. This summary provides a ''road map'' for what was done and the results obtained for the fuel characterization program initiated in 1994 and scheduled for completion in 1999 with the fuel oxidation rate measurement under moist inert atmospheres.

  1. Research reactor de-fueling and fuel shipment

    SciTech Connect

    Ice, R.D.; Jawdeh, E.; Strydom, J.

    1998-08-01

    Planning for the Georgia Institute of Technology Research Reactor operations during the 1996 Summer Olympic Games began in early 1995. Before any details could be outlined, several preliminary administrative decisions had to be agreed upon by state, city, and university officials. The two major administrative decisions involving the reactor were (1) the security level and requirements and (2) the fuel status of the reactor. The Georgia Tech Research Reactor (GTRR) was a heavy-water moderated and cooled reactor, fueled with high-enriched uranium. The reactor was first licensed in 1964 with an engineered lifetime of thirty years. The reactor was intended for use in research applications and as a teaching facility for nuclear engineering students and reactor operators. Approximately one year prior to the olympics, the Georgia Tech administration decided that the GTRR fuel would be removed. In addition, a heightened, beyond regulatory requirements, security system was to be implemented. This report describes the scheduling, operations, and procedures.

  2. Plasma-gun fueling for tokamak reactors

    SciTech Connect

    Ehst, D.A.

    1980-11-01

    In light of the uncertain extrapolation of gas puffing for reactor fueling and certain limitations to pellet injection, the snowplow plasma gun has been studied as a fueling device. Based on current understanding of gun and plasma behavior a design is proposed, and its performance is predicted in a tokamak reactor environment.

  3. DECONTAMINATION OF NEUTRON-IRRADIATED REACTOR FUEL

    DOEpatents

    Buyers, A.G.; Rosen, F.D.; Motta, E.E.

    1959-12-22

    A pyrometallurgical method of decontaminating neutronirradiated reactor fuel is presented. In accordance with the invention, neutron-irradiated reactor fuel may be decontaminated by countercurrently contacting the fuel with a bed of alkali and alkaine fluorides under an inert gas atmosphere and inductively melting the fuel and tracking the resulting descending molten fuel with induction heating as it passes through the bed. By this method, a large, continually fresh surface of salt is exposed to the descending molten fuel which enhances the efficiency of the scrubbing operation.

  4. United States Domestic Research Reactor Infrastrucutre TRIGA Reactor Fuel Support

    SciTech Connect

    Douglas Morrell

    2011-03-01

    The United State Domestic Research Reactor Infrastructure Program at the Idaho National Laboratory manages and provides project management, technical, quality engineering, quality inspection and nuclear material support for the United States Department of Energy sponsored University Reactor Fuels Program. This program provides fresh, unirradiated nuclear fuel to Domestic University Research Reactor Facilities and is responsible for the return of the DOE-owned, irradiated nuclear fuel over the life of the program. This presentation will introduce the program management team, the universities supported by the program, the status of the program and focus on the return process of irradiated nuclear fuel for long term storage at DOE managed receipt facilities. It will include lessons learned from research reactor facilities that have successfully shipped spent fuel elements to DOE receipt facilities.

  5. Development and numerical/experimental characterization of a lab-scale flat flame reactor allowing the analysis of pulverized solid fuel devolatilization and oxidation at high heating rates.

    PubMed

    Lemaire, R; Menanteau, S

    2016-01-01

    This paper deals with the thorough characterization of a new experimental test bench designed to study the devolatilization and oxidation of pulverized fuel particles in a wide range of operating conditions. This lab-scale facility is composed of a fuel feeding system, the functioning of which has been optimized by computational fluid dynamics. It allows delivering a constant and time-independent mass flow rate of fuel particles which are pneumatically transported to the central injector of a hybrid McKenna burner using a carrier gas stream that can be inert or oxidant depending on the targeted application. A premixed propane/air laminar flat flame stabilized on the porous part of the burner is used to generate the hot gases insuring the heating of the central coal/carrier-gas jet with a thermal gradient similar to those found in industrial combustors (>10(5) K/s). In the present work, results issued from numerical simulations performed a priori to characterize the velocity and temperature fields in the reaction chamber have been analyzed and confronted with experimental measurements carried out by coupling particle image velocimetry, thermocouple and two-color pyrometry measurements so as to validate the order of magnitude of the heating rate delivered by such a new test bench. Finally, the main features of the flat flame reactor we developed have been discussed with respect to those of another laboratory-scale system designed to study coal devolatilization at a high heating rate. PMID:26827350

  6. Development and numerical/experimental characterization of a lab-scale flat flame reactor allowing the analysis of pulverized solid fuel devolatilization and oxidation at high heating rates

    NASA Astrophysics Data System (ADS)

    Lemaire, R.; Menanteau, S.

    2016-01-01

    This paper deals with the thorough characterization of a new experimental test bench designed to study the devolatilization and oxidation of pulverized fuel particles in a wide range of operating conditions. This lab-scale facility is composed of a fuel feeding system, the functioning of which has been optimized by computational fluid dynamics. It allows delivering a constant and time-independent mass flow rate of fuel particles which are pneumatically transported to the central injector of a hybrid McKenna burner using a carrier gas stream that can be inert or oxidant depending on the targeted application. A premixed propane/air laminar flat flame stabilized on the porous part of the burner is used to generate the hot gases insuring the heating of the central coal/carrier-gas jet with a thermal gradient similar to those found in industrial combustors (>105 K/s). In the present work, results issued from numerical simulations performed a priori to characterize the velocity and temperature fields in the reaction chamber have been analyzed and confronted with experimental measurements carried out by coupling particle image velocimetry, thermocouple and two-color pyrometry measurements so as to validate the order of magnitude of the heating rate delivered by such a new test bench. Finally, the main features of the flat flame reactor we developed have been discussed with respect to those of another laboratory-scale system designed to study coal devolatilization at a high heating rate.

  7. Gaseous fuel reactors for power systems

    NASA Technical Reports Server (NTRS)

    Kendall, J. S.; Rodgers, R. J.

    1977-01-01

    Gaseous-fuel nuclear reactors have significant advantages as energy sources for closed-cycle power systems. The advantages arise from the removal of temperature limits associated with conventional reactor fuel elements, the wide variety of methods of extracting energy from fissioning gases, and inherent low fissile and fission product in-core inventory due to continuous fuel reprocessing. Example power cycles and their general performance characteristics are discussed. Efficiencies of gaseous fuel reactor systems are shown to be high with resulting minimal environmental effects. A technical overview of the NASA-funded research program in gaseous fuel reactors is described and results of recent tests of uranium hexafluoride (UF6)-fueled critical assemblies are presented.

  8. Microstructural Characterization of Cast Metallic Transmutation Fuels

    SciTech Connect

    J. I. Cole; D. D. Keiser; J. R. Kennedy

    2007-09-01

    As part of the Global Nuclear Energy Partnership (GNEP) and the Advanced Fuel Cycle Initiative (AFCI), the US Department of Energy (DOE) is participating in an international collaboration to irradiate prototypic actinide-bearing transmutation fuels in the French Phenix fast reactor (FUTURIX-FTA experiment). The INL has contributed to this experiment by fabricating and characterizing two compositions of metallic fuel; a non-fertile 48Pu-12Am-40Zr fuel and a low-fertile 35U-29Pu-4Am-2Np-30Zr fuel for insertion into the reactor. This paper highlights results of the microstructural analysis of these cast fuels, which were reasonably homogeneous in nature, but had several distinct phase constituents. Spatial variations in composition appeared to be more pronounced in the low-fertile fuel when compared to the non-fertile fuel.

  9. FUEL-BREEDER FUEL ELEMENT FOR NUCLEAR REACTOR

    DOEpatents

    Abbott, W.E.; Balent, R.

    1958-09-16

    A fuel element design to facilitate breeding reactor fuel is described. The fuel element is comprised of a coatainer, a central core of fertile material in the container, a first bonding material surrounding the core, a sheet of fissionable material immediately surrounding the first bonding material, and a second bonding material surrounding the fissionable material and being in coniact with said container.

  10. NEUTRONIC REACTOR FUEL ELEMENT AND CORE SYSTEM

    DOEpatents

    Moore, W.T.

    1958-09-01

    This patent relates to neutronic reactors and in particular to an improved fuel element and a novel reactor core system for facilitating removal of contaminating fission products, as they are fermed, from association with the flssionable fuel, so as to mitigate the interferent effects of such fission products during reactor operation. The fuel elements are comprised of tubular members impervious to fluid and contatning on their interior surfaces a thin layer of fissionable material providing a central void. The core structure is comprised of a plurality of the tubular fuel elements arranged in parallel and a closed manifold connected to their ends. In the reactor the core structure is dispersed in a water moderator and coolant within a pressure vessel, and a means connected to said manifuld is provided for withdrawing and disposing of mobile fission product contamination from the interior of the feel tubes and manifold.

  11. Post irradiation examination of thermal reactor fuels

    NASA Astrophysics Data System (ADS)

    Sah, D. N.; Viswanathan, U. K.; Ramadasan, E.; Unnikrishnan, K.; Anantharaman, S.

    2008-12-01

    The post irradiation examination (PIE) facility at the Bhabha Atomic Research Centre (BARC) has been in operation for more than three decades. Over these years this facility has been utilized for examination of experimental fuel pins and fuels from commercial power reactors operating in India. In a program to assess the performance of (U,Pu)O 2 MOX fuel prior to its introduction in commercial reactors, three experimental MOX fuel clusters irradiated in the pressurized water loop (PWL) of CIRUS up to burnup of 16 000 MWd/tU were examined. Fission gas release from these pins was measured by puncture test. Some of these fuel pins in the cluster contained controlled porosity pellets, low temperature sintered (LTS) pellets, large grain size pellets and annular pellets. PIE has also been carried out on natural UO 2 fuel bundles from Indian PHWRs, which included two high burnup (˜15 000 MWd/tU) bundles. Salient investigations carried out consisted of visual examination, leak testing, axial gamma scanning, fission gas analysis, microstructural examination of fuel and cladding, β, γ autoradiography of the fuel cross-section and fuel central temperature estimation from restructuring. A ThO 2 fuel bundle irradiated in Kakrapar Atomic Power Station (KAPS) up to a nominal fuel burnup of ˜11 000 MWd/tTh was also examined to evaluate its in-pile performance. The performance of the BWR fuel pins of Tarapur Atomic Power Stations (TAPS) was earlier assessed by carrying out PIE on 18 fuel elements selected from eight fuel assemblies irradiated in the two reactors. The burnup of these fuel elements varied from 5000 to 29 000 MWd/tU. This paper provides a brief review of some of the fuels examined and the results obtained on the performance of natural UO 2, enriched UO 2, MOX, and ThO 2 fuels.

  12. NON-CORROSIVE REACTOR FUEL SYSTEM

    DOEpatents

    Herrick, C.C.

    1962-08-14

    A non-corrosive nuclear reactor fuel system was developed utilizing a molten plutonium-- iron alloy fuel having about 2 at.% carbon and contained in a tantalum vessel. This carbon reacts with the interior surface of the tantalum vessel to form a plutonium resistant self-healing tantalum carbide film. (AEC)

  13. Nuclear reactor with fuel pin bracing grid

    SciTech Connect

    Jolly, R.

    1980-01-01

    A fuel pin bracing grid for a nuclear fuel sub-assembly comprises a honeycomb array of unit cells formed from discrete strips. The cells are hexagonal, three alternate sides having windows and the remaining sides have linear groups of three embossments to provide guide pads for fuel pins. The openings provide a measure of compliancy for the grid to facilitate insertion and withdrawal of the pins. A fuel sub-assembly for a liquid metal cooled fast breeder nuclear reactor has a central fuel section with end extensions, the fuel section comprising a bundle of fuel pins in a hexagonal wrapper the pins being braced by a series of grids according to the invention. Reprocessing of the fuel is facilitated because the pins are withdrawable collectively from the compliant grids and wrapper combination merely by cutting an end extension from the wrapper. 4 claims.

  14. History of fast reactor fuel development

    NASA Astrophysics Data System (ADS)

    Kittel, J. H.; Frost, B. R. T.; Mustelier, J. P.; Bagley, K. Q.; Crittenden, G. C.; Van Dievoet, J.

    1993-09-01

    The first fast breeder reactors, constructed in the 1945-1960 time period, used metallic fuels composed of uranium, plutonium, or their alloys. They were chosen because most existing reactor operating experience had been obtained on metallic fuels and because they provided the highest breeding ratios. Difficulties in obtaining adequate dimensional stability in metallic fuel elements under conditions of high fuel burnup led in the 1960s to the virtual worldwide choice of ceramic fuels. Although ceramic fuels provide lower breeding performance, this objective is no longer an important consideration in most national programs. Mixed uranium and plutonium dioxide became the ceramic fuel that has received the widest use. The more advanced ceramic fuels, mixed uranium and plutonium carbides and nitrides, continue under development. More recently, metal fuel elements of improved design have joined ceramic fuels in achieving goal burnups of 15 to 20 percent. Low-swelling fuel cladding alloys have also been continuously developed to deal with the unexpected problem of void formation in stainless steels subjected to fast neutron irradiation, a phenomenon first observed in the 1960s.

  15. Fuel assembly for nuclear reactors

    DOEpatents

    Creagan, Robert J.; Frisch, Erling

    1977-01-01

    A new and improved fuel assembly is formed to minimize the amount of parasitic structural material wherein a plurality of hollow tubular members are juxtaposed to the fuel elements of the assembly. The tubular members may serve as guide tubes for control elements and are secured to a number of longitudinally spaced grid members along the fuel assembly. The grid members include means thereon engaging each of the fuel elements to laterally position the fuel elements in a predetermined array. Openings in the bottom of each hollow member serve as a shock absorber to cushion shock transmitted to the structure when the control elements are rapidly inserted in their corresponding tubular members.

  16. (UA1 reactor fuels safety and performance)

    SciTech Connect

    Taleyarkhan, R.P.

    1990-07-13

    The traveler visited several reactor and hot cell experimental facilities connected with JAERI at the Oarai and Tokai establishments. Uranium silicide fission product release experimental data and related acquisition systems were discussed. A presentation was made by the traveler on analysis and modeling of fission product release from UAl reactor fuels. Data obtained by JAERI thus far were offered to the traveler for Oak Ridge National Laboratory (ORNL) review and analysis. This data confirmed key aspects of ORNL theoretical model predictions and will be useful for Advanced Neutron Source (ANS) design. The Oarai establishment expressed their interest and willingness to pursue ORNL/JAERI cooperative efforts in understanding volatile fission product release behavior from silicide fuels. The traveler also presented a perspective overview on ORNL severe accident analysis technology and identified areas for cooperation in JAERI's forthcoming transient testing program. JAERI staff presented plans for evaluating silicide fuel performance under transient reactivity insertion accident conditions in the Nuclear Safety Research Reactor (NSRR) facility. A surprise announcement was made concerning JAERI's most recent initiative relating to the construction of a safety demonstration reactor (SDR) at the Tokai site. The purpose of this reactor facility would be to demonstrate operational safety of both Pressurized Water Reactors (PWRs) and Boiling Water Reactors (BWRs) in support of Japan's nuclear power industry.

  17. Evolution of Particle Bed Reactor Fuel

    NASA Astrophysics Data System (ADS)

    Jensen, Russell R.; Evans, Robert S.; Husser, Dewayne L.; Kerr, John M.

    1994-07-01

    To realize the potential performance advantages inherent in a particle bed reactor (PBR) for nuclear thermal propulsion (NTP) applications, high performance particle fuel is required. This fuel must operate safely and without failure at high temperature in high pressure, flowing hydrogen propellant. The mixed mean outlet temperature of the propellant is an important characteristic of PBR performance. This temperature is also a critical parameter for fuel particle design because it dictates the required maximum fuel operating temperature. In this paper, the evolution in PBR fuel form to achieve higher operating temperatures is discussed and the potential thermal performance of the different fuel types is evaluated. It is shown that the optimum fuel type for operation under the demanding conditions in a PBR is a coated, solid carbide particle.

  18. Microstructural Characterization of a Mg Matrix U-Mo Dispersion Fuel Plate Irradiated in the Advanced Test Reactor to High Fission Density: SEM Results

    NASA Astrophysics Data System (ADS)

    Keiser, Dennis D.; Jue, Jan-Fong; Miller, Brandon D.; Gan, Jian; Robinson, Adam B.; Medvedev, Pavel G.; Madden, James W.; Moore, Glenn A.

    2016-06-01

    Low-enriched (U-235 <20 pct) U-Mo dispersion fuel is being developed for use in research and test reactors. In most cases, fuel plates with Al or Al-Si alloy matrices have been tested in the Advanced Test Reactor to support this development. In addition, fuel plates with Mg as the matrix have also been tested. The benefit of using Mg as the matrix is that it potentially will not chemically interact with the U-Mo fuel particles during fabrication or irradiation, whereas with Al and Al-Si alloys such interactions will occur. Fuel plate R9R010 is a Mg matrix fuel plate that was aggressively irradiated in ATR. This fuel plate was irradiated as part of the RERTR-8 experiment at high temperature, high fission rate, and high power, up to high fission density. This paper describes the results of the scanning electron microscopy (SEM) analysis of an irradiated fuel plate using polished samples and those produced with a focused ion beam. A follow-up paper will discuss the results of transmission electron microscopy (TEM) analysis. Using SEM, it was observed that even at very aggressive irradiation conditions, negligible chemical interaction occurred between the irradiated U-7Mo fuel particles and Mg matrix; no interconnection of fission gas bubbles from fuel particle to fuel particle was observed; the interconnected fission gas bubbles that were observed in the irradiated U-7Mo particles resulted in some transport of solid fission products to the U-7Mo/Mg interface; the presence of microstructural pathways in some U-9.1 Mo particles that could allow for transport of fission gases did not result in the apparent presence of large porosity at the U-7Mo/Mg interface; and, the Mg-Al interaction layers that were present at the Mg matrix/Al 6061 cladding interface exhibited good radiation stability, i.e. no large pores.

  19. Gaseous fuel reactor systems for aerospace applications

    NASA Technical Reports Server (NTRS)

    Thom, K.; Schwenk, F. C.

    1977-01-01

    Research on the gaseous fuel nuclear rocket concept continues under the programs of the U.S. National Aeronautics and Space Administration (NASA) Office for Aeronautics and Space Technology and now includes work related to power applications in space and on earth. In a cavity reactor test series, initial experiments confirmed the low critical mass determined from reactor physics calculations. Recent work with flowing UF6 fuel indicates stable operation at increased power levels. Preliminary design and experimental verification of test hardware for high-temperature experiments have been accomplished. Research on energy extraction from fissioning gases has resulted in lasers energized by fission fragments. Combined experimental results and studies indicate that gaseous-fuel reactor systems have significant potential for providing nuclear fission power in space and on earth.

  20. CHARACTERIZATION OF RADIOACTIVITY IN THE REACTOR VESSEL OF THE HEAVY WATER COMPONENT TEST REACTOR

    SciTech Connect

    Vinson, Dennis

    2010-06-01

    The Heavy Water Component Test Reactor (HWCTR) facility is a pressurized heavy water reactor that was used to test candidate fuel designs for heavy water power reactors. The reactor operated at nominal power of 50 MW{sub th}. The reactor coolant loop operated at 1200 psig and 250 C. Two isolated test loop were designed into the reactor to provide special test conditions. Fig. 1 shows a cut-away view of the reactor. The two loops are contained in four inch diameter stainless steel piping. The HWCTR was operated for only a short duration, from March 1962 to December 1964 in order to test the viability of test fuel elements and other reactor components for use in a heavy water power reactor. The reactor achieved 13,882 MWd of total power while testing 36 different fuel assemblies. In the course of operation, HWCTR experienced the cladding failures of 10 separate test fuel assemblies. In each case, the cladding was breached with some release of fuel core material into the isolated test loop, causing fission product and actinide contamination in the main coolant loop and the liquid and boiling test loops. Despite the contribution of the contamination from the failed fuel, the primary source of radioactivity in the HWCTR vessel and internals is the activation products in the thermal shields, and to a lesser degree, activation products in the reactor vessel walls and liner. A detailed facility characterization report of the HWCTR facility was completed in 1996. Many of the inputs and assumptions in the 1996 characterization report were derived from the HWCTR decommissioning plan published in 1975. The current paper provides an updated assessment of the radioisotopic characteristics of the HWCTR vessel and internals to support decommissioning activities on the facility.

  1. Characterization plan for Hanford spent nuclear fuel

    SciTech Connect

    Abrefah, J.; Thornton, T.A.; Thomas, L.E.; Berting, F.M.; Marschman, S.C.

    1994-12-01

    Reprocessing of spent nuclear fuel (SNF) at the Hanford Site Plutonium-Uranium Extraction Plant (PUREX) was terminated in 1972. Since that time a significant quantity of N Reactor and Single-Pass Reactor SNF has been stored in the 100 Area K-East (KE) and K-West (KW) reactor basins. Approximately 80% of all US Department of Energy (DOE)-owned SNF resides at Hanford, the largest portion of which is in the water-filled KE and KW reactor basins. The basins were not designed for long-term storage of the SNF and it has become a priority to move the SNF to a more suitable location. As part of the project plan, SNF inventories will be chemically and physically characterized to provide information that will be used to resolve safety and technical issues for development of an environmentally benign and efficient extended interim storage and final disposition strategy for this defense production-reactor SNF.

  2. Optimally moderated nuclear fission reactor and fuel source therefor

    DOEpatents

    Ougouag, Abderrafi M.; Terry, William K.; Gougar, Hans D.

    2008-07-22

    An improved nuclear fission reactor of the continuous fueling type involves determining an asymptotic equilibrium state for the nuclear fission reactor and providing the reactor with a moderator-to-fuel ratio that is optimally moderated for the asymptotic equilibrium state of the nuclear fission reactor; the fuel-to-moderator ratio allowing the nuclear fission reactor to be substantially continuously operated in an optimally moderated state.

  3. MOLTEN PLUTONIUM FUELED FAST BREEDER REACTOR

    DOEpatents

    Kiehn, R.M.; King, L.D.P.; Peterson, R.E.; Swickard, E.O. Jr.

    1962-06-26

    A description is given of a nuclear fast reactor fueled with molten plutonium containing about 20 kg of plutonium in a tantalum container, cooled by circulating liquid sodium at about 600 to 650 deg C, having a large negative temperature coefficient of reactivity, and control rods and movable reflector for criticality control. (AEC)

  4. PLUTONIUM METALLIC FUELS FOR FAST REACTORS

    SciTech Connect

    STAN, MARIUS; HECKER, SIEGFRIED S.

    2007-02-07

    Early interest in metallic plutonium fuels for fast reactors led to much research on plutonium alloy systems including binary solid solutions with the addition of aluminum, gallium, or zirconium and low-melting eutectic alloys with iron and nickel or cobalt. There was also interest in ternaries of these elements with plutonium and cerium. The solid solution and eutectic alloys have most unusual properties, including negative thermal expansion in some solid-solution alloys and the highest viscosity known for liquid metals in the Pu-Fe system. Although metallic fuels have many potential advantages over ceramic fuels, the early attempts were unsuccessful because these fuels suffered from high swelling rates during burn up and high smearing densities. The liquid metal fuels experienced excessive corrosion. Subsequent work on higher-melting U-PuZr metallic fuels was much more promising. In light of the recent rebirth of interest in fast reactors, we review some of the key properties of the early fuels and discuss the challenges presented by the ternary alloys.

  5. FUEL ELEMENT FOR NUCLEAR REACTOR

    DOEpatents

    Carney, K.G. Jr.

    1959-07-14

    A nuclear fuel element comprising a large number og wafers of fissionable material and a protective jacket having compartments holding these wafers is described. The compartments of the jacket aid the removal of heat from the wafers, keep the wafers or fragments thereof from migrating in the jacket, and permit the escape of gaseous fission products.

  6. PYROCHEMICAL DECONTAMINATION METHOD FOR REACTOR FUEL

    DOEpatents

    Buyers, A.G.

    1959-06-30

    A pyro-chemical method is presented for decontaminating neutron irradiated uranium and separating plutonium therefrom by contact in the molten state with a metal chloride salt. Uranium trichloride and uranium tetrachloride either alone or in admixture with alkaline metal and alkaline eanth metal fluorides under specified temperature and specified phase ratio conditions extract substantially all of the uranium from the irradiated uranium fuel together with certain fission products. The phases are then separated leaving purified uranium metal. The uranium and plutonium in the salt phase can be reduced to forin a highly decontaminated uraniumplutonium alloy. The present method possesses advantages for economically decontaminating irradiated nuclear fuel elements since irradiated fuel may be proccessed immediately after withdrawal from the reactor and the uranium need not be dissolved and later reduced to the metallic form. Accordingly, the uranium may be economically refabricated and reinserted into the reactor.

  7. NEUTRONIC REACTOR AND FUEL ELEMENT THEREFOR

    DOEpatents

    Szilard, L.; Young, G.J.

    1958-03-01

    This patent relates to a reactor design of the type which employs solid fuel elements disposed in channels within the moderator through which channels and around the fuel elements is conveyed a coolant fiuid. The coolant channels are comprised of aluminum tubes extending through a solid moderator such as graphite and the fuel elements are comprised of an elongated solid body of natural uranium jacketed in an aluminum jacket with the ends thereof closed by aluminum caps of substantially greater thickness than the jacket was and in good thermal contact with the fuel material to facilitate the conduction of heat from the central portion of said ends to the coolant surrounding the fuel element to prevent overheating of said central portion.

  8. Sintering and characterization of ZrN and (Dy,Zr)N as surrogate materials for fast reactor nitride fuel

    NASA Astrophysics Data System (ADS)

    Pukari, Merja; Takano, Masahide

    2014-01-01

    Pellets of inert matrix material ZrN, and surrogate nitride fuel material (Dy0.4Zr0.6)N, are fabricated for the purpose of investigating the origin and the effect of carbon and oxygen impurity concentrations. Oxygen concentrations of up to 1.2 wt% are deliberately introduced into the materials with two separate methods. The achievable pellet densities of these materials, as a function of O content, sintering temperature and dimensional powder properties are determined. O dissolved into (Dy,Zr)N increases the achievable densities to a larger extent than if dissolved into ZrN. The segregation of O-rich phases in ZrN indicates a low O solubility in the material. Oxygen pick-up during the fabrication of the product as well as its exposure to air is demonstrated. The quality of the materials is monitored by the systematic analysis of O, N and C contents throughout the fabrication and sintering processes, supported by XRD and SEM analyses.

  9. FUEL COMPOSITION FOR NUCLEAR REACTORS

    DOEpatents

    Andersen, J.C.

    1963-08-01

    A process for making refractory nuclear fuel elements involves heating uranium and silicon powders in an inert atmosphere to 1600 to 1800 deg C to form USi/sub 3/; adding silicon carbide, carbon, 15% by weight of nickel and aluminum, and possibly also molybdenum and silicon powders; shaping the mixture; and heating to 1700 to 2050 deg C again in an inert atmosphere. Information on obtaining specific compositions is included. (AEC)

  10. Natural fueling of a tokamak fusion reactor

    NASA Astrophysics Data System (ADS)

    Wan, Weigang; Parker, Scott E.; Chen, Yang; Perkins, Francis W.

    2010-04-01

    A natural fueling mechanism that helps to maintain the main core deuterium and tritium (DT) density profiles in a tokamak fusion reactor is presented. In H-mode plasmas dominated by ion-temperature gradient (ITG) driven turbulence, cold DT ions near the edge will naturally pinch radially inward toward the core. This mechanism is due to the quasineutral heat flux dominated nature of ITG turbulence and still applies when trapped and passing kinetic electron effects are included. Fueling using shallow pellet injection is augmented by an inward pinch of cold DT fuel. The natural fueling mechanism is investigated using the gyrokinetic turbulence code GEM [Y. Chen and S. E. Parker, J. Comput. Phys. 220, 839 (2007)] and is analyzed using quasilinear theory. Profiles similar to those used for conservative International Thermonuclear Experimental Reactor [R. Aymar et al., Nucl. Fusion 41, 1301 (2001)] transport modeling that have a completely flat density profile are examined and it is found that natural fueling actually reduces the linear growth rate and energy transport. Natural fueling requires a two-component plasma and ion-ion and charge-exchange collisions set limits on this favorable effect.

  11. Performance of boiling water reactor fuel lead test assemblies to 35 MWd/kg U

    SciTech Connect

    Rowland, T.C.; Ikemoto, R.N.; Gehl, S.

    1986-01-01

    This joint Electric Power Research Institute/General Electric (EPRI/GE) fuel performance program involved thorough preirradiation characterization of fuel used in lead test assemblies (LTAs), detailed surveillance of their operation, and interim site examinations of the assemblies during reactor outages. The program originally included four GE-5 LTAs operating in the Peach Bottom-2 (PB-2) reactor. The program was later modified to include the pressurized fuel rod test assembly in the Peach Bottom-3 (PB-3) reactor. The program modification also included extending the operation of the PB-2 and PB-3 LTA fuel beyond normal discharge exposures. Results are summarized in the paper.

  12. Safeguards instrument to monitor spent reactor fuel

    NASA Astrophysics Data System (ADS)

    Nicholson, N.; Dowdy, E. J.; Holt, D. M.; Stump, C.

    1981-10-01

    A hand held instrument for monitoring irradiated nuclear fuel inventories located in water filled storage ponds has been developed. This instrument provides sufficient precise qualitative and quantitative information to be useful as a confirmatory technique to International Atomic Energy Agency inspectors, and is believed to be of potential use to nuclear fuel managers and to operators of spent fuel storage facilities, both at reactor and away from reactor, and to operators of nuclear fuel reprocessing plants. Because the Cerenkov radiation glow can barely be seen by the unaided eye under darkened conditions, a night vision device is incorporated to aid the operator in locating the fuel assembly to be measured. Beam splitting optics placed in front of the image intensifier and a preset aperture select a predetermined portion of the observed scene for measurement of the light intensity using a photomultiplier (PM) tube and digital readout. The PM tube gain is adjusted by use of an internal optical reference source, providing long term repeatability and instrument to instrument consistency. Interchangeable lenses accommodate various viewing and measuring conditions.

  13. JACKETED FUEL ELEMENTS FOR GRAPHITE MODERATED REACTORS

    DOEpatents

    Szilard, L.; Wigner, E.P.; Creutz, E.C.

    1959-05-12

    Fuel elements for a heterogeneous, fluid cooled, graphite moderated reactor are described. The fuel elements are comprised of a body of natural uranium hermetically sealed in a jacket of corrosion resistant material. The jacket, which may be aluminum or some other material which is non-fissionable and of a type having a low neutron capture cross-section, acts as a barrier between the fissioning isotope and the coolant or moderator or both. The jacket minimizes the tendency of the moderator and coolant to become radioactive and/or contaminated by fission fragments from the fissioning isotope.

  14. Nuclear reactor fuel rod attachment system

    DOEpatents

    Not Available

    1980-09-17

    A reusable system is described for removably attaching a nuclear reactor fuel rod to a support member. A locking cap is secured to the fuel rod and a locking strip is fastened to the support member. The locking cap has two opposing fingers shaped to form a socket having a body portion. The locking strip has an extension shaped to rigidly attach to the socket's body portion. The locking cap's fingers are resiliently deflectable. For attachment, the locking cap is longitudinally pushed onto the locking strip causing the extension to temporarily deflect open the fingers to engage the socket's body portion. For removal, the process is reversed.

  15. A compact breed and burn fast reactor using spent nuclear fuel blanket

    SciTech Connect

    Hartanto, D.; Kim, Y.

    2012-07-01

    A long-life breed-and-burn (B and B) type fast reactor has been investigated from the neutronics points of view. The B and B reactor has the capability to breed the fissile fuels and use the bred fuel in situ in the same reactor. In this work, feasibility of a compact sodium-cooled B and B fast reactor using spent nuclear fuel as blanket material has been studied. In order to derive a compact B and B fast reactor, a tight fuel lattice and relatively large fuel pin are used to achieve high fuel volume fraction. The core is initially loaded with an LEU (Low Enriched Uranium) fuel and a metallic fuel is used in the core. The Monte Carlo depletion has been performed for the core to see the long-term behavior of the B and B reactor. Several important parameters such as reactivity coefficients, delayed neutron fraction, prompt neutron generation lifetime, fission power, and fast neutron fluence, are analyzed through Monte Carlo reactor analysis. Evolution of the core fuel composition is also analyzed as a function of burnup. Although the long-life small B and B fast reactor is found to be feasible from the neutronics point of view, it is characterized to have several challenging technical issues including a very high fast neutron fluence of the structural materials. (authors)

  16. Fuel handling system for a nuclear reactor

    DOEpatents

    Saiveau, James G.; Kann, William J.; Burelbach, James P.

    1986-01-01

    A pool type nuclear fission reactor has a core, with a plurality of core elements and a redan which confines coolant as a hot pool at a first end of the core separated from a cold pool at a second end of the core by the redan. A fuel handling system for use with such reactors comprises a core element storage basket located outside of the redan in the cold pool. An access passage is formed in the redan with a gate for opening and closing the passage to maintain the temperature differential between the hot pool and the cold pool. A mechanism is provided for opening and closing the gate. A lifting arm is also provided for manipulating the fuel core elements through the access passage between the storage basket and the core when the redan gate is open.

  17. Fuel handling system for a nuclear reactor

    DOEpatents

    Saiveau, James G.; Kann, William J.; Burelbach, James P.

    1986-12-02

    A pool type nuclear fission reactor has a core, with a plurality of core elements and a redan which confines coolant as a hot pool at a first end of the core separated from a cold pool at a second end of the core by the redan. A fuel handling system for use with such reactors comprises a core element storage basket located outside of the redan in the cold pool. An access passage is formed in the redan with a gate for opening and closing the passage to maintain the temperature differential between the hot pool and the cold pool. A mechanism is provided for opening and closing the gate. A lifting arm is also provided for manipulating the fuel core elements through the access passage between the storage basket and the core when the redan gate is open.

  18. NUCLEAR REACTOR AND THERMIONIC FUEL ELEMENT THEREFOR

    DOEpatents

    Rasor, N.S.; Hirsch, R.L.

    1963-12-01

    The patent relates to the direct conversion of fission heat to electricity by use of thermionic plasma diodes having fissionable material cathodes, said diodes arranged to form a critical mass in a nuclear reactor. The patent describes a fuel element comprising a plurality of diodes each having a fissionable material cathode, an anode around said cathode, and an ionizable gas therebetween. Provision is made for flowing the gas and current serially through the diodes. (AEC)

  19. Proliferation resistance of small modular reactors fuels

    SciTech Connect

    Polidoro, F.; Parozzi, F.; Fassnacht, F.; Kuett, M.; Englert, M.

    2013-07-01

    In this paper the proliferation resistance of different types of Small Modular Reactors (SMRs) has been examined and classified with criteria available in the literature. In the first part of the study, the level of proliferation attractiveness of traditional low-enriched UO{sub 2} and MOX fuels to be used in SMRs based on pressurized water technology has been analyzed. On the basis of numerical simulations both cores show significant proliferation risks. Although the MOX core is less proliferation prone in comparison to the UO{sub 2} core, it still can be highly attractive for diversion or undeclared production of nuclear material. In the second part of the paper, calculations to assess the proliferation attractiveness of fuel in typical small sodium cooled fast reactor show that proliferation risks from spent fuel cannot be neglected. The core contains a highly attractive plutonium composition during the whole life cycle. Despite some aspects of the design like the sealed core that enables easy detection of unauthorized withdrawal of fissile material and enhances proliferation resistance, in case of open Non-Proliferation Treaty break-out, weapon-grade plutonium in sufficient quantities could be extracted from the reactor core.

  20. Spring element for holding down nuclear reactor fuel assembly

    SciTech Connect

    Steinke, A.

    1981-07-14

    Spring element is described for holding down and bracing a fuel assembly against a hold-down plate upwardly limiting the reactor core of a nuclear reactor. Includes a spring-loaded rod-shaped member separately formed independently of the fuel assembly and being slidable axially and form-lockingly into the fuel assembly.

  1. Advanced ceramic cladding for water reactor fuel

    SciTech Connect

    Feinroth, H.

    2000-07-01

    Under the US Department of Energy's Nuclear Energy Research Initiatives (NERI) program, continuous fiber ceramic composites (CFCCs) are being developed as cladding for water reactor fuel elements. The purpose is to substantially increase the passive safety of water reactors. A development effort was initiated in 1991 to fabricate CFCC-clad tubes using commercially available fibers and a sol-gel process developed by McDermott Technologies. Two small-diameter CFCC tubes were fabricated using pure alumina and alumina-zirconia fibers in an alumina matrix. Densities of {approximately}60% of theoretical were achieved. Higher densities are required to guarantee fission gas containment. This NERI work has just begun, and only preliminary results are presented herein. Should the work prove successful, further development is required to evaluate CFCC cladding and performance, including in-pile tests containing fuel and exploring a marriage of CFCC cladding materials with suitable advanced fuel and core designs. The possibility of much higher temperature core designs, possibly cooled with supercritical water, and achievement of plant efficiencies {ge}50% would be examined.

  2. Spent nuclear fuel discharges from U.S. reactors 1994

    SciTech Connect

    1996-02-01

    Spent Nuclear Fuel Discharges from US Reactors 1994 provides current statistical data on fuel assemblies irradiated at commercial nuclear reactors operating in the US. This year`s report provides data on the current inventories and storage capacities at these reactors. Detailed statistics on the data are presented in four chapters that highlight 1994 spent fuel discharges, storage capacities and inventories, canister and nonfuel component data, and assembly characteristics. Five appendices, a glossary, and bibliography are also included. 10 figs., 34 tabs.

  3. Solar Thermal Reactor Materials Characterization

    SciTech Connect

    Lichty, P. R.; Scott, A. M.; Perkins, C. M.; Bingham, C.; Weimer, A. W.

    2008-03-01

    Current research into hydrogen production through high temperature metal oxide water splitting cycles has created a need for robust high temperature materials. Such cycles are further enhanced by the use of concentrated solar energy as a power source. However, samples subjected to concentrated solar radiation exhibited lifetimes much shorter than expected. Characterization of the power and flux distributions representative of the High Flux Solar Furnace(HFSF) at the National Renewable Energy Laboratory(NREL) were compared to ray trace modeling of the facility. In addition, samples of candidate reactor materials were thermally cycled at the HFSF and tensile failure testing was performed to quantify material degradation. Thermal cycling tests have been completed on super alloy Haynes 214 samples and results indicate that maximum temperature plays a significant role in reduction of strength. The number of cycles was too small to establish long term failure trends for this material due to the high ductility of the material.

  4. Nuclear reactor fuel rod attachment system

    DOEpatents

    Christiansen, David W.

    1982-01-01

    A reusable system for removably attaching a nuclear reactor fuel rod (12) to a support member (14). A locking cap (22) is secured to the fuel rod (12) and a locking strip (24) is fastened to the support member (14). The locking cap (22) has two opposing fingers (24a and 24b) shaped to form a socket having a body portion (26). The locking strip has an extension (36) shaped to rigidly attach to the socket's body portion (26). The locking cap's fingers are resiliently deflectable. For attachment, the locking cap (22) is longitudinally pushed onto the locking strip (24) causing the extension (36) to temporarily deflect open the fingers (24a and 24b) to engage the socket's body portion (26). For removal, the process is reversed.

  5. Fuel Cycle Performance of Thermal Spectrum Small Modular Reactors

    SciTech Connect

    Worrall, Andrew; Todosow, Michael

    2016-01-01

    Small modular reactors may offer potential benefits, such as enhanced operational flexibility. However, it is vital to understand the holistic impact of small modular reactors on the nuclear fuel cycle and fuel cycle performance. The focus of this paper is on the fuel cycle impacts of light water small modular reactors in a once-through fuel cycle with low-enriched uranium fuel. A key objective of this paper is to describe preliminary reactor core physics and fuel cycle analyses conducted in support of the U.S. Department of Energy Office of Nuclear Energy Fuel Cycle Options Campaign. Challenges with small modular reactors include: increased neutron leakage, fewer assemblies in the core (and therefore fewer degrees of freedom in the core design), complex enrichment and burnable absorber loadings, full power operation with inserted control rods, the potential for frequent load-following operation, and shortened core height. Each of these will impact the achievable discharge burn-up in the reactor and the fuel cycle performance. This paper summarizes the results of an expert elicitation focused on developing a list of the factors relevant to small modular reactor fuel, core, and operation that will impact fuel cycle performance. Preliminary scoping analyses were performed using a regulatory-grade reactor core simulator. The hypothetical light water small modular reactor considered in these preliminary scoping studies is a cartridge type one-batch core with 4.9% enrichment. Some core parameters, such as the size of the reactor and general assembly layout, are similar to an example small modular reactor concept from industry. The high-level issues identified and preliminary scoping calculations in this paper are intended to inform on potential fuel cycle impacts of one-batch thermal spectrum SMRs. In particular, this paper highlights the impact of increased neutron leakage and reduced number of batches on the achievable burn-up of the reactor. Fuel cycle performance

  6. Integral reactor system and method for fuel cells

    DOEpatents

    Fernandes, Neil Edward; Brown, Michael S; Cheekatamarla, Praveen; Deng, Thomas; Dimitrakopoulos, James; Litka, Anthony F

    2013-11-19

    A reactor system is integrated internally within an anode-side cavity of a fuel cell. The reactor system is configured to convert hydrocarbons to smaller species while mitigating the lower production of solid carbon. The reactor system may incorporate one or more of a pre-reforming section, an anode exhaust gas recirculation device, and a reforming section.

  7. Laser-based characterization of nuclear fuel plates

    SciTech Connect

    Smith, James A.; Cottle, Dave L.; Rabin, Barry H.

    2014-02-18

    Ensuring the integrity of fuel-clad and clad-clad bonding in nuclear fuels is important for safe reactor operation and assessment of fuel performance, yet the measurement of bond strengths in actual fuels has proved challenging. The laser shockwave technique (LST) originally developed to characterize structural adhesion in composites is being employed to characterize interface strength in a new type of plate fuel being developed at Idaho National Laboratory (INL). LST is a non-contact method that uses lasers for the generation and detection of large-amplitude acoustic waves and is well suited for application to both fresh and irradiated nuclear-fuel plates. This paper will report on initial characterization results obtained from fresh fuel plates manufactured by different processes, including hot isostatic pressing, friction stir welding, and hot rolling.

  8. Laser-Based Characterization of Nuclear Fuel Plates

    SciTech Connect

    James A. Smith; David L. Cottle; Barry H. Rabin

    2013-07-01

    Ensuring the integrity of fuel-clad and clad-clad bonding in nuclear fuels is important for safe reactor operation and assessment of fuel performance, yet the measurement of bond strengths in actual fuels has proved challenging. The laser shockwave technique (LST) originally developed to characterize structural adhesion in composites is being employed to characterize interface strength in a new type of plate fuel being developed at Idaho National Laboratory (INL). LST is a non-contact method that uses lasers for the generation and detection of large-amplitude acoustic waves and is well suited for application to both fresh and irradiated nuclear-fuel plates. This paper will report on initial characterization results obtained from fresh fuel plates manufactured by different processes, including hot isostatic pressing, friction stir welding, and hot rolling.

  9. Laser-based characterization of nuclear fuel plates

    NASA Astrophysics Data System (ADS)

    Smith, James A.; Cottle, Dave L.; Rabin, Barry H.

    2014-02-01

    Ensuring the integrity of fuel-clad and clad-clad bonding in nuclear fuels is important for safe reactor operation and assessment of fuel performance, yet the measurement of bond strengths in actual fuels has proved challenging. The laser shockwave technique (LST) originally developed to characterize structural adhesion in composites is being employed to characterize interface strength in a new type of plate fuel being developed at Idaho National Laboratory (INL). LST is a non-contact method that uses lasers for the generation and detection of large-amplitude acoustic waves and is well suited for application to both fresh and irradiated nuclear-fuel plates. This paper will report on initial characterization results obtained from fresh fuel plates manufactured by different processes, including hot isostatic pressing, friction stir welding, and hot rolling.

  10. Characteristics and Dose Levels for Spent Reactor Fuels

    SciTech Connect

    Coates, Cameron W

    2007-01-01

    Current guidance considers highly radioactive special nuclear materials to be those materials that, unshielded, emit a radiation dose [rate] measured at 1 m which exceeds 100 rem/h. Smaller, less massive fuel assemblies from research reactors can present a challenge from the point of view of self protection because of their size (lower dose, easier to handle) and the desirability of higher enrichments; however, a follow-on study to cross-compare dose trends of research reactors and power reactors was deemed useful to confirm/verify these trends. This paper summarizes the characteristics and dose levels of spent reactor fuels for both research reactors and power reactors and extends previous studies aimed at quantifying expected dose rates from research reactor fuels worldwide.

  11. Safeguards operations in the integral fast reactor fuel cycle

    SciTech Connect

    Goff, K.M.; Benedict, R.W.; Brumbach, S.B.; Dickerman, C.E.; Tompot, R.W.

    1994-08-01

    Argonne National Laboratory is currently demonstrating the fuel cycle for the Integral Fast Reactor (IFR), an advanced reactor concept that takes advantage of the properties of metallic fuel and liquid metal cooling to offer significant improvements in reactor safety, operation, fuel-cycle economics, environmental protection, and safeguards. The IFR fuel cycle employs a pyrometallurgical process using molten salts and liquid metals to recover actinides from spent fuel. The safeguards aspects of the fuel cycle demonstration must be approved by the United States Department of Energy, but a further goal of the program is to develop a safeguards system that could gain acceptance from the Nuclear Regulatory Commission and International Atomic Energy Agency. This fuel cycle is described with emphasis on aspects that differ from aqueous reprocessing and on its improved safeguardability due to decreased attractiveness and diversion potential of all process streams, including the fuel product.

  12. Reactor-specific spent fuel discharge projections, 1987-2020

    SciTech Connect

    Walling, R.C.; Heeb, C.M.; Purcell, W.L.

    1988-03-01

    The creation of five reactor-specific spent fuel data bases that contain information on the projected amounts of spent fuel to be discharged from U.S. commercial nuclear reactors through the year 2020 is described. The data bases contain detailed spent fuel information from existing, planned, and projected pressurized water reactors (PWR) and boiling water eactors (BWR), and one existing high temperature gas reactor (HTGR). The projections are based on individual reactor information supplied by the U.S. reactor owners. The basic information is adjusted to conform to Energy Information Administration (EIA) forecasts for nuclear installed capacity, generation, and spent fuel discharged. The EIA cases considered are: No New Orders (assumes increasing burnup), No New Orders with No Increased Burnup, Upper Reference (assumes increasing burnup), Upper Reference with No Increased Burnup, and Lower Reference (assumes increasing burnup). Detailed, by-reactor tables are provided for annual discharged amounts of spent fuel, for storage requirements assuming maximum at-reactor storage, and for storage requirements assuming maximum at-reactor storage plus intra-utility transshipment of spent fuel. 8 refs., 8 figs., 10 tabs.

  13. Reactor-specific spent fuel discharge projections: 1986 to 2020

    SciTech Connect

    Heeb, C.M.; Walling, R.C.; Purcell, W.L.

    1987-03-01

    The creation of five reactor-specific spent fuel data bases that contain information on the projected amounts of spent fuel to be discharged from US commercial nuclear reactors through the year 2020 is described. The data bases contain detailed spent-fuel information from existing, planned, and projected pressurized water reactors (PWR) and boiling water reactors (BWR). The projections are based on individual reactor information supplied by the US reactor owners. The basic information is adjusted to conform to Energy Information Agency (EIA) forecasts for nuclear installed capacity, generation, and spent fuel discharged. The EIA cases considered are: (1) No new orders with extended burnup, (2) No new orders with constant burnup, (3) Upper reference (which assumes extended burnup), (4) Upper reference with constant burnup, and (5) Lower reference (which assumes extended burnup). Detailed, by-reactor tables are provided for annual discharged amounts of spent fuel, for storage requirements assuming maximum-at-reactor storage, and for storage requirements assuming maximum-at-reactor plus intra-utility transshipment of spent fuel. 6 refs., 8 figs., 8 tabs.

  14. Fuel Summary Report: Shippingport Light Water Breeder Reactor - Rev. 2

    SciTech Connect

    Olson, Gail Lynn; Mc Cardell, Richard Keith; Illum, Douglas Brent

    2002-09-01

    The Shippingport Light Water Breeder Reactor (LWBR) was developed by Bettis Atomic Power Laboratory to demonstrate the potential of a water-cooled, thorium oxide fuel cycle breeder reactor. The LWBR core operated from 1977-82 without major incident. The fuel and fuel components suffered minimal damage during operation, and the reactor testing was deemed successful. Extensive destructive and nondestructive postirradiation examinations confirmed that the fuel was in good condition with minimal amounts of cladding deformities and fuel pellet cracks. Fuel was placed in wet storage upon arrival at the Expended Core Facility, then dried and sent to the Idaho Nuclear Technology and Engineering Center for underground dry storage. It is likely that the fuel remains in good condition at its current underground dry storage location at the Idaho Nuclear Technology and Engineering Center. Reports show no indication of damage to the core associated with shipping, loading, or storage.

  15. Spent fuel utilization in a compact traveling wave reactor

    SciTech Connect

    Hartanto, Donny; Kim, Yonghee

    2012-06-06

    In recent years, several innovative designs of nuclear reactors are proposed. One of them is Traveling Wave Reactor (TWR). The unique characteristic of a TWR is the capability of breeding its own fuel in the reactor. The reactor is fueled by mostly depleted, natural uranium or spent nuclear fuel and a small amount of enriched uranium to initiate the fission process. Later on in the core, the reactor gradually converts the non-fissile material into the fissile in a process like a traveling wave. In this work, a TWR with spent nuclear fuel blanket was studied. Several parameters such as reactivity coefficients, delayed neutron fraction, prompt neutron generation lifetime, and fission power, were analyzed. The discharge burnup composition was also analyzed. The calculation is performed by a continuous energy Monte Carlo code McCARD.

  16. Preparation of high temperature gas-cooled reactor fuel element

    DOEpatents

    Bradley, Ronnie A.; Sease, John D.

    1976-01-01

    This invention relates to a method for the preparation of high temperature gas-cooled reactor (HTGR) fuel elements wherein uncarbonized fuel rods are inserted in appropriate channels of an HTGR fuel element block and the entire block is inserted in an autoclave for in situ carbonization under high pressure. The method is particularly applicable to remote handling techniques.

  17. Reactor-specific spent fuel discharge projections: 1985 to 2020

    SciTech Connect

    Heeb, C.M.; Libby, R.A.; Walling, R.C.; Purcell, W.L.

    1986-09-01

    The creation of four spent-fuel data bases that contain information on the projected amounts of spent fuel to be discharged from US commercial nuclear reactors through the year 2020 is described. The data bases contain detailed spent-fuel information from existing, planned, and projected pressurized water reactors (PWR) and boiling water reactors (BWR). The projections are based on individual reactor information supplied by the US reactor owners. The basic information is adjusted to conform to Energy Information Agency (EIA) forecasts for nuclear installed capacity, generation, and spent fuel discharged. The EIA cases considered are: (1) No New Orders with Extended Burnup, (2) No New Orders with Constant Burnup, (3) Middle Case with Extended Burnup, and (4) Middle Case with Constant Burnup. Detailed, by-reactor tables are provided for annual discharged amounts of spent fuel, for storage requirements assuming maximum-at-reactor storage, and for storage requirements assuming maximum-at-reactor plus intra-utility transshipment of spent fuel.

  18. High Performance Fuel Desing for Next Generation Pressurized Water Reactors

    SciTech Connect

    Mujid S. Kazimi; Pavel Hejzlar

    2006-01-31

    The use of internally and externally cooled annular fule rods for high power density Pressurized Water Reactors is assessed. The assessment included steady state and transient thermal conditions, neutronic and fuel management requirements, mechanical vibration issues, fuel performance issues, fuel fabrication methods and econmic assessment. The investigation was donducted by a team from MIT, Westinghouse, Gamma Engineering, Framatome ANP, and AECL. The analyses led to the conclusion that raising the power density by 50% may be possible with this advanced fuel. Even at the 150% power level, the fuel temperature would be a few hundred degrees lower than the current fuel temperatre. Significant economic and safety advantages can be obtained by using this fuel in new reactors. Switching to this type of fuel for existing reactors would yield safety advantages, but the economic return is dependent on the duration of plant shutdown to accommodate higher power production. The main feasiblity issue for the high power performance appears to be the potential for uneven splitting of heat flux between the inner and outer fuel surfaces due to premature closure of the outer fuel-cladding gap. This could be overcome by using a very narrow gap for the inner fuel surface and/or the spraying of a crushable zirconium oxide film at the fuel pellet outer surface. An alternative fuel manufacturing approach using vobropacking was also investigated but appears to yield lower than desirable fuel density.

  19. Control of autothermal reforming reactor of diesel fuel

    NASA Astrophysics Data System (ADS)

    Dolanc, Gregor; Pregelj, Boštjan; Petrovčič, Janko; Pasel, Joachim; Kolb, Gunther

    2016-05-01

    In this paper a control system for autothermal reforming reactor for diesel fuel is presented. Autothermal reforming reactors and the pertaining purification reactors are used to convert diesel fuel into hydrogen-rich reformate gas, which is then converted into electricity by the fuel cell. The purpose of the presented control system is to control the hydrogen production rate and the temperature of the autothermal reforming reactor. The system is designed in such a way that the two control loops do not interact, which is required for stable operation of the fuel cell. The presented control system is a part of the complete control system of the diesel fuel cell auxiliary power unit (APU).

  20. REACTOR PHYSICS MODELING OF SPENT NUCLEAR RESEARCH REACTOR FUEL FOR SNM ATTRIBUTION AND NUCLEAR FORENSICS

    SciTech Connect

    Sternat, M.; Beals, D.; Webb, R.; Nichols, T.

    2010-06-09

    Nuclear research reactors are the least safeguarded type of reactor; in some cases this may be attributed to low risk and in most cases it is due to difficulty from dynamic operation. Research reactors vary greatly in size, fuel type, enrichment, power and burnup providing a significant challenge to any standardized safeguard system. If a whole fuel assembly was interdicted, based on geometry and other traditional forensics work, one could identify the material's origin fairly accurately. If the material has been dispersed or reprocessed, in-depth reactor physics models may be used to help with the identification. Should there be a need to attribute research reactor fuel material, the Savannah River National Laboratory would perform radiochemical analysis of samples of the material as well as other non-destructive measurements. In depth reactor physics modeling would then be performed to compare to these measured results in an attempt to associate the measured results with various reactor parameters. Several reactor physics codes are being used and considered for this purpose, including: MONTEBURNS/ORIGEN/MCNP5, CINDER/MCNPX and WIMS. In attempt to identify reactor characteristics, such as time since shutdown, burnup, or power, various isotopes are used. Complexities arise when the inherent assumptions embedded in different reactor physics codes handle the isotopes differently and may quantify them to different levels of accuracy. A technical approach to modeling spent research reactor fuel begins at the assembly level upon acquiring detailed information of the reactor to be modeled. A single assembly is run using periodic boundary conditions to simulate an infinite lattice which may be repeatedly burned to produce input fuel isotopic vectors of various burnups for a core level model. A core level model will then be constructed using the assembly level results as inputs for the specific fuel shuffling pattern in an attempt to establish an equilibrium cycle. The

  1. Neutronics and fuel behavior of AIROX-processed fuel recycled into light water reactors

    SciTech Connect

    Allison, C.M.; Jahshan, S.N.; Wade, N.L.

    1993-08-01

    An evaluation of the Atomics International Reduction Oxidation (AIROX) process has begun to determine if the process could be used to recycle spent fuel to minimize high-level waste from commercial power reactors. This paper includes an evaluation of core neutronics to establish enrichment levels and expected in-reactor performance: a review of existing fuel behavior research to determine its applicability to AIROX-recycled fuels; and an evaluation of potential licensing issues unique to these fuels.

  2. Hanford single-pass reactor fuel storage basin demolition.

    PubMed

    Armstrong, Jason A

    2003-02-01

    The Environmental Restoration Contractor at the Hanford Site is tasked with removing auxiliary reactor structures and leaving the remaining concrete structure surrounding each reactor core. This is referred to as Interim Safe Storage. Part of placing the F Reactor into Interim Safe Storage is the demolition of the fuel storage basin, which was deactivated in 1970 by placing debris material into the basin prior to back filling with soil. Besides the debris material (wooden floor decking, handrails, and monorail pieces), the fuel storage basin contents included the possibility of spent nuclear fuel, fuel buckets, fuel spacers, process tubes, and tongs. Demolition of the fuel storage basin offered many unique radiological control challenges and innovative approaches to demolition. This paper describes how the total effective dose equivalent and contamination were controlled, how the use of a remote operated excavator was employed to remove high-dose-rate material, and how wireless technology was used to monitor changing radiological conditions.

  3. Molten salt reactors for burning dismantled weapons fuel

    SciTech Connect

    Gat, U.; Engel, J.R. ); Dodds, H.L. . Dept. of Nuclear Engineering)

    1992-12-01

    In this paper, the molten salt reactor (MSR) option for burning fissile fuel form dismantled weapons is examined. It is concluded that MSRs are potentially suitable for beneficial utilization of the dismantled fuel. the MSRs have the flexibility to utilize any fissile fuel in continuous operation with no special modifications, as demonstrated in the Molten Salt Reactor Experiment, while maintaining their economy. The MSRs further require a minimum of special fuel preparation and can tolerate denaturing and dilution of the fuel. Fuel shipments can be arbitrarily small, which may reduce the risk of diversion. The MSRs have inherent safety features that make them acceptable and attractive. They can burn a fuel type completely and convert it to other fuels. The 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 o nuclear power.

  4. Evaluation of Metal-Fueled Surface Reactor Concepts

    SciTech Connect

    Poston, David I.; Marcille, Thomas F.; Kapernick, Richard J.; Hiatt, Matthew T.; Amiri, Benjamin W.

    2007-01-30

    Surface fission power systems for use on the Moon and Mars may provide the first use of near-term reactor technology in space. Most near-term surface reactor concepts specify reactor temperatures <1000 K to allow the use of established material and power conversion technology and minimize the impact of the in-situ environment. Metal alloy fuels (e.g. U-10Zr and U-10Mo) have not traditionally been considered for space reactors because of high-temperature requirements, but they might be an attractive option for these lower temperature surface power missions. In addition to temperature limitations, metal fuels are also known to swell significantly at rather low fuel burnups ({approx}1 a/o), but near-term surface missions can mitigate this concern as well, because power and lifetime requirements generally keep fuel burnups <1 a/o. If temperature and swelling issues are not a concern, then a surface reactor concept may be able to benefit from the high uranium density and relative ease of manufacture of metal fuels. This paper investigates two reactor concepts that utilize metal fuels. It is found that these concepts compare very well to concepts that utilize other fuels (UN, UO2, UZrH) on a mass basis, while also providing the potential to simplify material safeguards issues.

  5. The DOE Advanced Gas Reactor Fuel Development and Qualification Program

    SciTech Connect

    David Petti

    2010-09-01

    The high outlet temperatures and high thermal-energy conversion efficiency of modular High Temperature Gas-cooled Reactors (HTGRs) enable an efficient and cost effective integration of the reactor system with non-electricity generation applications, such as process heat and/or hydrogen production, for the many petrochemical and other industrial processes that require temperatures between 300°C and 900°C. The Department of Energy (DOE) has selected the HTGR concept for the Next Generation Nuclear Plant (NGNP) Project as a transformative application of nuclear energy that will demonstrate emissions-free nuclear-assisted electricity, process heat, and hydrogen production, thereby reducing greenhouse-gas emissions and enhancing energy security. The objective of the DOE Advanced Gas Reactor (AGR) Fuel Development and Qualification program is to qualify tristructural isotropic (TRISO)-coated particle fuel for use in HTGRs. The Advanced Gas Reactor Fuel Development and Qualification Program consists of five elements: fuel manufacture, fuel and materials irradiations, post-irradiation examination (PIE) and safety testing, fuel performance modeling, and fission-product transport and source term evaluation. An underlying theme for the fuel development work is the need to develop a more complete, fundamental understanding of the relationship between the fuel fabrication process and key fuel properties, the irradiation and accident safety performance of the fuel, and the release and transport of fission products in the NGNP primary coolant system. An overview of the program and recent progress is presented.

  6. 4. VIEW LOOKING NORTHWEST OF FUEL HANDLING BUILDING (CENTER), REACTOR ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    4. VIEW LOOKING NORTHWEST OF FUEL HANDLING BUILDING (CENTER), REACTOR SERVICE BUILDING (RIGHT), MACHINE SHOP (LEFT) - Shippingport Atomic Power Station, On Ohio River, 25 miles Northwest of Pittsburgh, Shippingport, Beaver County, PA

  7. Characterization of Nuclear Fuel using Multivariate Statistical Analysis

    SciTech Connect

    Robel, M; Robel, M; Robel, M; Kristo, M J; Kristo, M J

    2007-11-27

    Various combinations of reactor type and fuel composition have been characterized using principle components analysis (PCA) of the concentrations of 9 U and Pu isotopes in the 10 fuel as a function of burnup. The use of PCA allows the reduction of the 9-dimensional data (isotopic concentrations) into a 3-dimensional approximation, giving a visual representation of the changes in nuclear fuel composition with burnup. Real-world variation in the concentrations of {sup 234}U and {sup 236}U in the fresh (unirradiated) fuel was accounted for. The effects of reprocessing were also simulated. The results suggest that, 15 even after reprocessing, Pu isotopes can be used to determine both the type of reactor and the initial fuel composition with good discrimination. Finally, partial least squares discriminant analysis (PSLDA) was investigated as a substitute for PCA. Our results suggest that PLSDA is a better tool for this application where separation between known classes is most important.

  8. Fuel Development For Gas-Cooled Fast Reactors

    SciTech Connect

    M. K. Meyer

    2006-06-01

    The Generation IV Gas-cooled Fast Reactor (GFR) concept is proposed to combine the advantages of high-temperature gas-cooled reactors (such as efficient direct conversion with a gas turbine and the potential for application of high-temperature process heat), with the sustainability advantages that are possible with a fast-spectrum reactor. The latter include the ability to fission all transuranics and the potential for breeding. The GFR is part of a consistent set of gas-cooled reactors that includes a medium-term Pebble Bed Modular Reactor (PBMR)-like concept, or concepts based on the Gas Turbine Modular Helium Reactor (GT-MHR), and specialized concepts such as the Very High Temperature Reactor (VHTR), as well as actinide burning concepts [ ]. To achieve the necessary high power density and the ability to retain fission gas at high temperature, the primary fuel concept proposed for testing in the United States is a dispersion coated fuel particles in a ceramic matrix. Alternative fuel concepts considered in the U.S. and internationally include coated particle beds, ceramic clad fuel pins, and novel ceramic ‘honeycomb’ structures. Both mixed carbide and mixed nitride-based solid solutions are considered as fuel phases.

  9. Fuel development for gas-cooled fast reactors

    NASA Astrophysics Data System (ADS)

    Meyer, M. K.; Fielding, R.; Gan, J.

    2007-09-01

    The Generation IV Gas-cooled Fast Reactor (GFR) concept is proposed to combine the advantages of high-temperature gas-cooled reactors (such as efficient direct conversion with a gas turbine and the potential for application of high-temperature process heat), with the sustainability advantages that are possible with a fast-spectrum reactor. The latter include the ability to fission all transuranics and the potential for breeding. The GFR is part of a consistent set of gas-cooled reactors that includes a medium-term Pebble Bed Modular Reactor (PBMR)-like concept, or concepts based on the Gas Turbine Modular Helium Reactor (GT-MHR), and specialized concepts such as the Very High-Temperature Reactor (VHTR), as well as actinide burning concepts [A Technology Roadmap for Generation IV Nuclear Energy Systems, US DOE Nuclear Energy Research Advisory Committee and the Generation IV International Forum, December 2002]. To achieve the necessary high power density and the ability to retain fission gas at high temperature, the primary fuel concept proposed for testing in the United States is dispersion coated fuel particles in a ceramic matrix. Alternative fuel concepts considered in the US and internationally include coated particle beds, ceramic clad fuel pins, and novel ceramic 'honeycomb' structures. Both mixed carbide and mixed nitride-based solid solutions are considered as fuel phases.

  10. Spent nuclear fuel discharges from US reactors 1993

    SciTech Connect

    Not Available

    1995-02-01

    The Energy Information Administration (EIA) of the U.S. Department of Energy (DOE) administers the Nuclear Fuel Data Survey, Form RW-859. This form is used to collect data on fuel assemblies irradiated at commercial nuclear reactors operating in the United States, and the current inventories and storage capacities of those reactors. These data are important to the design and operation of the equipment and facilities that DOE will use for the future acceptance, transportation, and disposal of spent fuels. The data collected and presented identifies trends in burnup, enrichment, and spent nuclear fuel discharged form commercial light-water reactor as of December 31, 1993. The document covers not only spent nuclear fuel discharges; but also site capacities and inventories; canisters and nonfuel components; and assembly type characteristics.

  11. Yttrium and rare earth stabilized fast reactor metal fuel

    DOEpatents

    Guon, Jerold; Grantham, LeRoy F.; Specht, Eugene R.

    1992-01-01

    To increase the operating temperature of a reactor, the melting point and mechanical properties of the fuel must be increased. For an actinide-rich fuel, yttrium, lanthanum and/or rare earth elements can be added, as stabilizers, to uranium and plutonium and/or a mixture of other actinides to raise the melting point of the fuel and improve its mechanical properties. Since only about 1% of the actinide fuel may be yttrium, lanthanum, or a rare earth element, the neutron penalty is low, the reactor core size can be reduced, the fuel can be burned efficiently, reprocessing requirements are reduced, and the nuclear waste disposal volumes reduced. A further advantage occurs when yttrium, lanthanum, and/or other rare earth elements are exposed to radiation in a reactor, they produce only short half life radioisotopes, which reduce nuclear waste disposal problems through much shorter assured-isolation requirements.

  12. Characterization of spent fuel approved testing material---ATM-105

    SciTech Connect

    Guenther, R.J.; Blahnik, D.E.; Campbell, T.K.; Jenquin, U.P.; Mendel, J.E.; Thomas, L.E.; Thornhill, C.K.

    1991-12-01

    The characterization data obtained to data are described for Approved Testing Material 105 (ATM-105), which is spent fuel from Bundles CZ346 and CZ348 of the Cooper Nuclear Power Plant, a boiling-water reactor. This report is one in a series being prepared by the Materials Characterization Center at Pacific Northwest Laboratory (PNL) on spent fuel ATMs. The ATMs are receiving extensive examinations to provide a source of well-characterized spent fuel for testing in the US Department of Energy Office of Civilian Radioactive Waste Management (OCRWM) Program. ATM-105 consists of 88 full-length irradiated fuel rods with rod-average burnups of about 2400 GJ/kgM (28 MWd/kgM) and expected fission gas release of about 1%. Characterization data include (1) descriptions of as-fabricated fuel design, irradiation history, and subsequent storage and handling; (2) isotopic gamma scans; (3) fission gas analyses; (4) ceramography of the fuel and metallography of the cladding; (5) special fuel studies involving analytical transmission electron microscopy (AEM); (6) calculated nuclide inventories and radioactivities in the fuel and cladding; and (7) radiochemical analyses of the fuel and cladding. Additional analyses of the fuel are being conducted and will be included in planned revisions of this report.

  13. Pyrolytic Waste Plastic Oil and Its Diesel Blend: Fuel Characterization.

    PubMed

    Khan, M Z H; Sultana, M; Al-Mamun, M R; Hasan, M R

    2016-01-01

    The authors introduced waste plastic pyrolysis oil (WPPO) as an alternative fuel characterized in detail and compared with conventional diesel. High density polyethylene, HDPE, was pyrolyzed in a self-designed stainless steel laboratory reactor to produce useful fuel products. HDPE waste was completely pyrolyzed at 330-490°C for 2-3 hours to obtain solid residue, liquid fuel oil, and flammable gaseous hydrocarbon products. Comparison of the fuel properties to the petrodiesel fuel standards ASTM D 975 and EN 590 revealed that the synthetic product was within all specifications. Notably, the fuel properties included a kinematic viscosity (40°C) of 1.98 cSt, density of 0.75 gm/cc, sulphur content of 0.25 (wt%), and carbon residue of 0.5 (wt%), and high calorific value represented significant enhancements over those of conventional petroleum diesel fuel. PMID:27433168

  14. Pyrolytic Waste Plastic Oil and Its Diesel Blend: Fuel Characterization.

    PubMed

    Khan, M Z H; Sultana, M; Al-Mamun, M R; Hasan, M R

    2016-01-01

    The authors introduced waste plastic pyrolysis oil (WPPO) as an alternative fuel characterized in detail and compared with conventional diesel. High density polyethylene, HDPE, was pyrolyzed in a self-designed stainless steel laboratory reactor to produce useful fuel products. HDPE waste was completely pyrolyzed at 330-490°C for 2-3 hours to obtain solid residue, liquid fuel oil, and flammable gaseous hydrocarbon products. Comparison of the fuel properties to the petrodiesel fuel standards ASTM D 975 and EN 590 revealed that the synthetic product was within all specifications. Notably, the fuel properties included a kinematic viscosity (40°C) of 1.98 cSt, density of 0.75 gm/cc, sulphur content of 0.25 (wt%), and carbon residue of 0.5 (wt%), and high calorific value represented significant enhancements over those of conventional petroleum diesel fuel.

  15. Pyrolytic Waste Plastic Oil and Its Diesel Blend: Fuel Characterization

    PubMed Central

    Sultana, M.; Al-Mamun, M. R.; Hasan, M. R.

    2016-01-01

    The authors introduced waste plastic pyrolysis oil (WPPO) as an alternative fuel characterized in detail and compared with conventional diesel. High density polyethylene, HDPE, was pyrolyzed in a self-designed stainless steel laboratory reactor to produce useful fuel products. HDPE waste was completely pyrolyzed at 330–490°C for 2-3 hours to obtain solid residue, liquid fuel oil, and flammable gaseous hydrocarbon products. Comparison of the fuel properties to the petrodiesel fuel standards ASTM D 975 and EN 590 revealed that the synthetic product was within all specifications. Notably, the fuel properties included a kinematic viscosity (40°C) of 1.98 cSt, density of 0.75 gm/cc, sulphur content of 0.25 (wt%), and carbon residue of 0.5 (wt%), and high calorific value represented significant enhancements over those of conventional petroleum diesel fuel. PMID:27433168

  16. Nuclear reactor vessel fuel thermal insulating barrier

    SciTech Connect

    Keegan, C. Patrick; Scobel, James H.; Wright, Richard F.

    2013-03-19

    The reactor vessel of a nuclear reactor installation which is suspended from the cold leg nozzles in a reactor cavity is provided with a lower thermal insulating barrier spaced from the reactor vessel that has a hemispherical lower section that increases in volume from the center line of the reactor to the outer extent of the diameter of the thermal insulating barrier and smoothly transitions up the side walls of the vessel. The space between the thermal insulating harrier and the reactor vessel forms a chamber which can be flooded with cooling water through passive valving to directly cool the reactor vessel in the event of a severe accident. The passive inlet valve for the cooling water includes a buoyant door that is normally maintained sealed under its own weight and floats open when the cavity is Hooded. Passively opening steam vents are also provided.

  17. Fuel element concept for long life high power nuclear reactors

    NASA Technical Reports Server (NTRS)

    Mcdonald, G. E.; Rom, F. E.

    1969-01-01

    Nuclear reactor fuel elements have burnups that are an order of magnitude higher than can currently be achieved by conventional design practice. Elements have greater time integrated power producing capacity per unit volume. Element design concept capitalizes on known design principles and observed behavior of nuclear fuel.

  18. NEUTRON REACTOR FUEL ELEMENT UTILIZING ZIRCONIUM-BASE ALLOYS

    DOEpatents

    Saller, H.A.; Keeler, J.R.; Szumachowski, E.R.

    1957-11-12

    This patent relates to clad fuel elements for use in neutronic reactors and is drawn to such a fuel element which consists of a core of fissionable material, comprised of an alloy of zirconium and U/sup 235/ enriched uranium, encased in a jacket of a binary zirconium-tin alloy in which the tin content ranges between 1 and 15% by weight.

  19. Code System for Reactor Physics and Fuel Cycle Simulation.

    1999-04-21

    Version 00 VSOP94 (Very Superior Old Programs) is a system of codes linked together for the simulation of reactor life histories. It comprises neutron cross section libraries and processing routines, repeated neutron spectrum evaluation, 2-D diffusion calculation based on neutron flux synthesis with depletion and shut-down features, in-core and out-of-pile fuel management, fuel cycle cost analysis, and thermal hydraulics (at present restricted to Pebble Bed HTRs). Various techniques have been employed to accelerate the iterativemore » processes and to optimize the internal data transfer. The code system has been used extensively for comparison studies of reactors, their fuel cycles, and related detailed features. In addition to its use in research and development work for the High Temperature Reactor, the system has been applied successfully to Light Water and Heavy Water Reactors.« less

  20. Code System for Reactor Physics and Fuel Cycle Simulation.

    SciTech Connect

    TEUCHERT, E.

    1999-04-21

    Version 00 VSOP94 (Very Superior Old Programs) is a system of codes linked together for the simulation of reactor life histories. It comprises neutron cross section libraries and processing routines, repeated neutron spectrum evaluation, 2-D diffusion calculation based on neutron flux synthesis with depletion and shut-down features, in-core and out-of-pile fuel management, fuel cycle cost analysis, and thermal hydraulics (at present restricted to Pebble Bed HTRs). Various techniques have been employed to accelerate the iterative processes and to optimize the internal data transfer. The code system has been used extensively for comparison studies of reactors, their fuel cycles, and related detailed features. In addition to its use in research and development work for the High Temperature Reactor, the system has been applied successfully to Light Water and Heavy Water Reactors.

  1. REACTOR PHYSICS MODELING OF SPENT RESEARCH REACTOR FUEL FOR TECHNICAL NUCLEAR FORENSICS

    SciTech Connect

    Nichols, T.; Beals, D.; Sternat, M.

    2011-07-18

    Technical nuclear forensics (TNF) refers to the collection, analysis and evaluation of pre- and post-detonation radiological or nuclear materials, devices, and/or debris. TNF is an integral component, complementing traditional forensics and investigative work, to help enable the attribution of discovered radiological or nuclear material. Research is needed to improve the capabilities of TNF. One research area of interest is determining the isotopic signatures of research reactors. Research reactors are a potential source of both radiological and nuclear material. Research reactors are often the least safeguarded type of reactor; they vary greatly in size, fuel type, enrichment, power, and burn-up. Many research reactors are fueled with highly-enriched uranium (HEU), up to {approx}93% {sup 235}U, which could potentially be used as weapons material. All of them have significant amounts of radiological material with which a radioactive dispersal device (RDD) could be built. Therefore, the ability to attribute if material originated from or was produced in a specific research reactor is an important tool in providing for the security of the United States. Currently there are approximately 237 operating research reactors worldwide, another 12 are in temporary shutdown and 224 research reactors are reported as shut down. Little is currently known about the isotopic signatures of spent research reactor fuel. An effort is underway at Savannah River National Laboratory (SRNL) to analyze spent research reactor fuel to determine these signatures. Computer models, using reactor physics codes, are being compared to the measured analytes in the spent fuel. This allows for improving the reactor physics codes in modeling research reactors for the purpose of nuclear forensics. Currently the Oak Ridge Research reactor (ORR) is being modeled and fuel samples are being analyzed for comparison. Samples of an ORR spent fuel assembly were taken by SRNL for analytical and radiochemical

  2. Technology gap analysis on sodium-cooled reactor fuel handling system supporting advanced burner reactor development.

    SciTech Connect

    Chikazawa, Y.; Farmer, M.; Grandy, C.; Nuclear Engineering Division

    2009-03-01

    The goals of the Global Nuclear Energy Partnership (GNEP) are to expand the use of nuclear energy to meet increasing global energy demand in an environmentally sustainable manner, to address nuclear waste management issues without making separated plutonium, and to address nonproliferation concerns. The advanced burner reactor (ABR) is a fast reactor concept which supports the GNEP fuel cycle system. Since the integral fast reactor (IFR) and advanced liquid-metal reactor (ALMR) projects were terminated in 1994, there has been no major development on sodium-cooled fast reactors in the United States. Therefore, in support of the GNEP fast reactor program, the history of sodium-cooled reactor development was reviewed to support the initiation of this technology within the United States and to gain an understanding of the technology gaps that may still remain for sodium fast reactor technology. The fuel-handling system is a key element of any fast reactor design. The major functions of this system are to receive, test, store, and then load fresh fuel into the core; unload from the core; then clean, test, store, and ship spent fuel. Major requirements are that the system must be reliable and relatively easy to maintain. In addition, the system should be designed so that it does not adversely impact plant economics from the viewpoints of capital investment or plant operations. In this gap analysis, information on fuel-handling operating experiences in the following reactor plants was carefully reviewed: EBR-I, SRE, HNPF, Fermi, SEFOR, FFTF, CRBR, EBR-II, DFR, PFR, Rapsodie, Phenix, Superphenix, KNK, SNR-300, Joyo, and Monju. The results of this evaluation indicate that a standardized fuel-handling system for a commercial fast reactor is yet to be established. However, in the past sodium-cooled reactor plants, most major fuel-handling components-such as the rotatable plug, in-vessel fuel-handling machine, ex-vessel fuel transportation cask, ex-vessel sodium-cooled storage

  3. Study of fueling requirements for the Engineering Test Reactor

    SciTech Connect

    Ho, S.K.; Perkins, L.J.

    1987-10-16

    An assessment of the fueling requirement for the TIBER Engineering Test Reactor is studied. The neutral shielding pellet ablation model with the inclusion of the effects of the alpha particles is used for our study. The high electron temperature in a reactor-grade plasma makes pellet penetration very difficult. The launch length has to be very large (several tens of meters) in order to avoid pellet breakage due to the low inertial strength of DT ''ice.'' The minimum repetition rate corresponding to the largest allowable pellet, is found to be about 1 Hz. A brief survey is done on the various operational and conceptual pellet injection schemes for plasma fueling. The underlying conclusion is that an alternative fueling scheme of coaxial compact-toroid plasma gun is very likely needed for effective central fueling of reactor-grade plasmas. 16 refs.

  4. Characterization of spent fuel approved testing material--ATM-104

    SciTech Connect

    Guenther, R.J.; Blahnik, D.E.; Jenquin, U.P.; Mendel, J.E.; Thomas, L.E.; Thornhill, C.K.

    1991-12-01

    The characterization data obtained to date are described for Approved Testing Material 104 (ATM-104), which is spent fuel from Assembly DO47 of the Calvert Cliffs Nuclear Power Plant (Unit 1), a pressurized-water reactor. This report is one in a series being prepared by the Materials Characterization Center at Pacific Northwest Laboratory (PNL) on spent fuel ATMs. The ATMs are receiving extensive examinations to provide a source of well-characterized spent fuel for testing in the US Department of Energy Office of Civilian Radioactive Waste Management (OCRWM) Program. ATM-104 consists of 128 full-length irradiated fuel rods with rod-average burnups of about 42 MWd/kgM and expected fission gas release of about 1%. A variety of analyses were performed to investigate cladding characteristics, radionuclide inventory, and redistribution of fission products. Characterization data include (1) fabricated fuel design, irradiation history, and subsequent storage and handling history; (2) isotopic gamma scans; (3) fission gas analyses; (4) ceramography of the fuel and metallography of the cladding; (5) special fuel studies involving analytical transmission electron microscopy (AEM) and electron probe microanalyses (EPMA); (6) calculated nuclide inventories and radioactivities in the fuel and cladding; and (7) radiochemical analyses of the fuel and cladding.

  5. An Implicit Solution Framework for Reactor Fuel Performance Simulation

    SciTech Connect

    Glen Hansen; Chris Newman; Derek Gaston; Cody Permann

    2009-08-01

    The simulation of nuclear reactor fuel performance involves complex thermomechanical processes between fuel pellets, made of fissile material, and the protective cladding that surrounds the pellets. An important design goal for a fuel is to maximize the life of the cladding thereby allowing the fuel to remain in the reactor for a longer period of time to achieve higher degrees of burnup. This presentation presents an initial approach for modeling the thermomechanical response of reactor fuel, and details of the solution method employed within INL's fuel performance code, BISON. The code employs advanced methods for solving coupled partial differential equation systems that describe multidimensional fuel thermomechanics, heat generation, and oxygen transport within the fuel. This discussion explores the effectiveness of a JFNK-based solution of a problem involving three dimensional fully coupled, nonlinear transient heat conduction and that includes pellet displacement and oxygen diffusion effects. These equations are closed using empirical data that is a function of temperature, density, and oxygen hyperstoichiometry. The method appears quite effective for the fuel pellet / cladding configurations examined, with excellent nonlinear convergence properties exhibited on the combined system. In closing, fully coupled solutions of three dimensional thermomechanics coupled with oxygen diffusion appear quite attractive using the JFNK approach described here, at least for configurations similar to those examined in this report.

  6. FUEL ELEMENT FOR A NEUTRONIC REACTOR

    DOEpatents

    McGeary, R.K.; Winslow, F.R.

    1963-08-13

    A method of making fuel elements wherein several individual fuel pellets are positioned into a cladding tube and the tape stretched longitudinally until the cladding tube grips each pellet and, in addition, necks down between each pellet is described. (AEC)

  7. Nuclear reactor fuel element having improved heat transfer

    DOEpatents

    Garnier, J.E.; Begej, S.; Williford, R.E.; Christensen, J.A.

    1982-03-03

    A nuclear reactor fuel element having improved heat transfer between fuel material and cladding is described. The element consists of an outer cladding tube divided into an upper fuel section containing a central core of fissionable or mixed fissionable and fertile fuel material, slightly smaller in diameter than the inner surface of the cladding tube and a small lower accumulator section, the cladding tube being which is filled with a low molecular weight gas to transfer heat from fuel material to cladding during irradiation. A plurality of essentially vertical grooves in the fuel section extend downward and communicate with the accumulator section. The radial depth of the grooves is sufficient to provide a thermal gradient between the hot fuel surface and the relatively cooler cladding surface to allow thermal segregation to take place between the low molecular weight heat transfer gas and high molecular weight fission product gases produced by the fuel material during irradiation.

  8. Composite nuclear fuel fabrication methodology for gas fast reactors

    NASA Astrophysics Data System (ADS)

    Vasudevamurthy, Gokul

    An advanced fuel form for use in Gas Fast Reactors (GFR) was investigated. Criteria for the fuel includes operation at high temperature (˜1400°C) and high burnup (˜150 MWD/MTHM) with effective retention of fission products even during transient temperatures exceeding 1600°C. The GFR fuel is expected to contain up to 20% transuranics for a closed fuel cycle. Earlier evaluations of reference fuels for the GFR have included ceramic-ceramic (cercer) dispersion type composite fuels of mixed carbide or nitride microspheres coated with SiC in a SiC matrix. Studies have indicated that ZrC is a potential replacement for SiC on account of its higher melting point, increased fission product corrosion resistance and better chemical stability. The present work investigated natural uranium carbide microspheres in a ZrC matrix instead of SiC. Known issues of minor actinide volatility during traditional fabrication procedures necessitated the investigation of still high temperature but more rapid fabrication techniques to minimize these anticipated losses. In this regard, fabrication of ZrC matrix by combustion synthesis from zirconium and graphite powders was studied. Criteria were established to obtain sufficient matrix density with UC microsphere volume fractions up to 30%. Tests involving production of microspheres by spark erosion method (similar to electrodischarge machining) showed the inability of the method to produce UC microspheres in the desired range of 300 to 1200 mum. A rotating electrode device was developed using a minimum current of 80A and rotating at speeds up to 1500 rpm to fabricate microspheres between 355 and 1200 mum. Using the ZrC process knowledge, UC electrodes were fabricated and studied for use in the rotating electrode device to produce UC microspheres. Fabrication of the cercer composite form was studied using microsphere volume fractions of 10%, 20%, and 30%. The macrostructure of the composite and individual components at various stages were

  9. Liquid fuel molten salt reactors for thorium utilization

    DOE PAGESBeta

    Gehin, Jess C.; Powers, Jeffrey J.

    2016-04-08

    Molten salt reactors (MSRs) represent a class of reactors that use liquid salt, usually fluoride- or chloride-based, as either a coolant with a solid fuel (such as fluoride salt-cooled high temperature reactors) or as a combined coolant and fuel with fuel dissolved in a carrier salt. For liquid-fuelled MSRs, the salt can be processed online or in a batch mode to allow for removal of fission products as well as introduction of fissile fuel and fertile materials during reactor operation. The MSR is most commonly associated with the 233U/thorium fuel cycle, as the nuclear properties of 233U combined with themore » online removal of parasitic absorbers allow for the ability to design a thermal-spectrum breeder reactor; however, MSR concepts have been developed using all neutron energy spectra (thermal, intermediate, fast, and mixed-spectrum zoned concepts) and with a variety of fuels including uranium, thorium, plutonium, and minor actinides. Early MSR work was supported by a significant research and development (R&D) program that resulted in two experimental systems operating at ORNL in the 1960s, the Aircraft Reactor Experiment and the Molten Salt Reactor Experiment. Subsequent design studies in the 1970s focusing on thermal-spectrum thorium-fueled systems established reference concepts for two major design variants: (1) a molten salt breeder reactor (MSBR), with multiple configurations that could breed additional fissile material or maintain self-sustaining operation; and (2) a denatured molten salt reactor (DMSR) with enhanced proliferation-resistance. T MSRs has been selected as one of six most promising Generation IV systems and development activities have been seen in fast-spectrum MSRs, waste-burning MSRs, MSRs fueled with low-enriched uranium (LEU), as well as more traditional thorium fuel cycle-based MSRs. This study provides an historical background of MSR R&D efforts, surveys and summarizes many of the recent development, and provides analysis comparing

  10. Technical Issues and Characterization for Fuel and Sludge in Hanford K Basins

    SciTech Connect

    MAKENAS, B.J.

    2000-06-01

    Technical Issues for the interim dry storage of N Reactor Spent Nuclear Fuel (SNF) are discussed. Characterization data from fuel, to support resolution of these issues, are reviewed and new results for the oxidation of fuel in a moist atmosphere and the drying of whole fuel elements are presented. Characterization of associated K basin sludge is also discussed in light of a newly adopted disposal pathway.

  11. Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems

    SciTech Connect

    D. E. Shropshire

    2009-01-01

    The Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems, prepared to support the U.S. Advanced Fuel Cycle Initiative (AFCI) systems analysis, provides a technology-oriented baseline system cost comparison between the open fuel cycle and closed fuel cycle systems. The intent is to understand their overall cost trends, cost sensitivities, and trade-offs. This analysis also improves the AFCI Program’s understanding of the cost drivers that will determine nuclear power’s cost competitiveness vis-a-vis other baseload generation systems. The common reactor-related costs consist of capital, operating, and decontamination and decommissioning costs. Fuel cycle costs include front-end (pre-irradiation) and back-end (post-iradiation) costs, as well as costs specifically associated with fuel recycling. This analysis reveals that there are large cost uncertainties associated with all the fuel cycle strategies, and that overall systems (reactor plus fuel cycle) using a closed fuel cycle are about 10% more expensive in terms of electricity generation cost than open cycle systems. The study concludes that further U.S. and joint international-based design studies are needed to reduce the cost uncertainties with respect to fast reactor, fuel separation and fabrication, and waste disposition. The results of this work can help provide insight to the cost-related factors and conditions needed to keep nuclear energy (including closed fuel cycles) economically competitive in the U.S. and worldwide. These results may be updated over time based on new cost information, revised assumptions, and feedback received from additional reviews.

  12. EXTENDING SODIUM FAST REACTOR DRIVER FUEL USE TO HIGHER TEMPERATURES

    SciTech Connect

    Douglas L. Porter

    2011-02-01

    Calculations of potential sodium-cooled fast reactor fuel temperatures were performed to estimate the effects of increasing the outlet temperature of a given fast reactor design by increasing pin power, decreasing assembly flow, or increasing inlet temperature. Based upon experience in the U.S., both metal and mixed oxide (MOX) fuel types are discussed in terms of potential performance effects created by the increased operating temperatures. Assembly outlet temperatures of 600, 650 and 700 °C were used as goal temperatures. Fuel/cladding chemical interaction (FCCI) and fuel melting, as well as challenges to the mechanical integrity of the cladding material, were identified as the limiting phenomena. For example, starting with a recent 1000 MWth fast reactor design, raising the outlet temperature to 650 °C through pin power increase increased the MOX centerline temperature to more than 3300 °C and the metal fuel peak cladding temperature to more than 700 °C. These exceeded limitations to fuel performance; fuel melting was limiting for MOX and FCCI for metal fuel. Both could be alleviated by design ‘fixes’, such as using a barrier inside the cladding to minimize FCCI in the metal fuel, or using annular fuel in the case of MOX. Both would also require an advanced cladding material with improved stress rupture properties. While some of these are costly, the benefits of having a high-temperature reactor which can support hydrogen production, or other missions requiring high process heat may make the extra costs justified.

  13. 75 FR 62695 - Physical Protection of Irradiated Reactor Fuel in Transit

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-10-13

    ... COMMISSION 10 CFR Part 73 RIN 3150-AI64 Physical Protection of Irradiated Reactor Fuel in Transit AGENCY...) is proposing to amend its security regulations pertaining to the transport of irradiated reactor fuel (for purposes of this rulemaking, the terms ``irradiated reactor fuel'' and ``spent nuclear fuel''...

  14. Reactor Physics Behavior of Transuranic-Bearing TRISO-Particle Fuel in a Pressurized Water Reactor

    SciTech Connect

    Michael A. Pope; R. Sonat Sen; Abderrafi M. Ougouag; Gilles Youinou; Brian Boer

    2012-04-01

    Calculations have been performed to assess the neutronic behavior of pins of Fully-Ceramic Micro-encapsulated (FCM) fuel in otherwise-conventional Pressurized Water Reactor (PWR) fuel pins. The FCM fuel contains transuranic (TRU)-only oxide fuel in tri-isotropic (TRISO) particles with the TRU loading coming from the spent fuel of a conventional LWR after 5 years of cooling. Use of the TRISO particle fuel would provide an additional barrier to fission product release in the event of cladding failure. Depletion calculations were performed to evaluate reactivity-limited burnup of the TRU-only FCM fuel. These calculations showed that due to relatively little space available for fuel, the achievable burnup with these pins alone is quite small. Various reactivity parameters were also evaluated at each burnup step including moderator temperature coefficient (MTC), Doppler, and soluble boron worth. These were compared to reference UO{sub 2} and MOX unit cells. The TRU-only FCM fuel exhibits degraded MTC and Doppler coefficients relative to UO{sub 2} and MOX. Also, the reactivity effects of coolant voiding suggest that the behavior of this fuel would be similar to a MOX fuel of very high plutonium fraction, which are known to have positive void reactivity. In general, loading of TRU-only FCM fuel into an assembly without significant quantities of uranium presents challenges to the reactor design. However, if such FCM fuel pins are included in a heterogeneous assembly alongside LEU fuel pins, the overall reactivity behavior would be dominated by the uranium pins while attractive TRU destruction performance levels in the TRU-only FCM fuel pins is. From this work, it is concluded that use of heterogeneous assemblies such as these appears feasible from a preliminary reactor physics standpoint.

  15. Sealed canister liquid sampling for fuel characterization shipments

    SciTech Connect

    Trimble, D.J.

    1995-01-12

    N-Reactor spent fuel elements will be removed from the sealed fuel storage canisters in the KW Basin and shipped to the 300 Area hot cells for characterization studies. Some of the canisters contain broken fuel elements exposing metallic uranium to the canister water. The exposed uranium surfaces have reacted with the water releasing fission product nuclides to the canister water. The extent of this release is unknown, but large radionuclide inventories could cause significant releases to the basin water. To avoid this, a method is needed to evaluate the magnitude of the canister water radioactivity to help make decisions about which canisters should not be opened for fuel sampling. The objective of this document is to provide a conceptual description of the process for obtaining and evaluating gas and liquid samples from KW fuel canisters in support of fuel characterization shipments. Also described are the decisions regarding the acceptability of a sample, the actions to be taken to obtain an acceptable sample, and the logical leading to decisions about opening canisters for fuel characterization sampling. This document will help guide a safety analysis of the liquid sampling process and the development of detailed procedures for the process. Outside the scope of this document are the processes for opening canisters, selecting fuel samples from opened canisters, shipping fuel samples, and closing canisters.

  16. High Density Fuel Development for Research Reactors

    SciTech Connect

    Daniel Wachs; Dennis Keiser; Mitchell Meyer; Douglas Burkes; Curtis Clark; Glenn Moore; Jan-Fong Jue; Totju Totev; Gerard Hofman; Tom Wiencek; Yeon So Kim; Jim Snelgrove

    2007-09-01

    An international effort to develop, qualify, and license high and very high density fuels has been underway for several years within the framework of multi-national RERTR programs. The current development status is the result of significant contributions from many laboratories, specifically CNEA in Argentina, AECL in Canada, CEA in France, TUM in Germany, KAERI in Korea, VNIIM, RDIPE, IPPE, NCCP and RIARR in Russia, INL, ANL and Y-12 in USA. These programs are mainly engaged with UMo dispersion fuels with densities from 6 to 8 gU/cm3 (high density fuel) and UMo monolithic fuel with density as high as 16 gU/cm3 (very high density fuel). This paper, mainly focused on the French & US programs, gives the status of high density UMo fuel development and perspectives on their qualification.

  17. System for fuel rod removal from a reactor module

    DOEpatents

    Matchett, Richard L.; Roof, David R.; Kikta, Thomas J.; Wilczynski, Rosemarie; Nilsen, Roy J.; Bacvinskas, William S.; Fodor, George

    1990-01-01

    A robotic system for remote underwater withdrawal of the fuel rods from fuel modules of a light water breeder reactor includes a collet/grapple assembly for gripping and removing fuel rods in each module, which is positioned by use of a winch and a radial support means attached to a vertical support tube which is mounted over the fuel module. A programmable logic controller in conjunction with a microcomputer, provides control for the accurate positioning and pulling force of the rod grapple assembly. Closed circuit television cameras are provided which aid in operator interface with the robotic system.

  18. System for fuel rod removal from a reactor module

    DOEpatents

    Matchett, R.L.; Fodor, G.; Kikta, T.J.; Bacvinsicas, W.S.; Roof, D.R.; Nilsen, R.J.; Wilczynski, R.

    1988-07-28

    A robotic system for remote underwater withdrawal of the fuel rods from fuel modules of a light water breeder reactor includes a collet/grapple assembly for gripping and removing fuel rods in each module, which is positioned by use of a winch and a radial support means attached to a vertical support tube which is mounted over the fuel module. A programmable logic controller in conjunction with a microcomputer, provides control for the accurate positioning and pulling force of the rod grapple assembly. Closed circuit television cameras are provided which aid in operator interface with the robotic system. 7 figs.

  19. Critical Issues for Particle-Bed Reactor Fuels

    NASA Astrophysics Data System (ADS)

    Evans, Robert S.; Husser, Dewayne L.; Jensen, Russell R.; Kerr, John M.

    1994-07-01

    Particle-Bed Reactors (PBRs) potentially offer performance advantages for nuclear thermal propulsion, including very high power densities, thrust-to-weight ratios, and specific impulses. A key factor in achieving all of these is the development of a very-high-temperature fuel. The critical issues for all such PBR fuels are uranium loading, thermomechanical and thermochemical stability, compatibility with contacting materials, fission product retention, manufacturability, and operational tolerance for particle failures. Each issue is discussed with respect to its importance to PBR operation, its status among current fuels, and additional development needs. Mixed-carbide-based fuels are recommended for further development to support high-performance PBRs.

  20. Reactor Physics Assessment of the Inclusion of Unseparated Neptunium in MOX Reactor Fuel

    SciTech Connect

    Ellis, Ronald James

    2009-01-01

    Reducing the number of actinide separation streams in a spent fuel recovery process would reduce the cost and complexity of the process, and lower the quantity and numbers of solvents needed. It is more difficult and costly to separate Np and recombine it with Am-Cm prior to co-conversion than to simply co-strip it with the U-Pu-Np. Inclusion of the Np in mixed oxide (MOX) fuel for light water reactor (LWR) applications should not seriously affect the operating behavior of the reactor, nor should it pose insurmountable fuel design issues. In this work, the U, Pu, and Np from typical discharged and cooled PWR spent nuclear fuel are assumed to be used together in the preparation of MOX fuel for use in a pressurized water reactor (PWR). The reactor grade Pu isotopic vector is used in the model and the relative mass ratio of the Pu and Np content (Np/Pu mass is 0.061) from the cooled spent fuel is maintained but the overall Pu-Np MOX wt% is adjusted with respect to the U content (assumed to be at 0.25 wt% 235U enrichment) to offset reactivity and cycle length effects. The SCALE 5.1 scientific package (especially modules TRITON, NEWT, ORIGEN-S, ORIGEN-ARP) was used for the calculations presented in this paper. A typical Westinghouse 17x17 fuel assembly design was modeled at nominal PWR operating conditions. It was seen that U-Pu-Np MOX fuel with NpO2 and PuO2 representing 11.5wt% of the total MOX fuel would be similar to standard MOX fuel in which PuO2 is 9wt% of the fuel. The reactivity, isotopic composition, and neutron and ? sources, and the decay heat details for the discharged MOX fuel are presented and discussed in this paper.

  1. Advanced Neutron Source Reactor thermal analysis of fuel plate defects

    SciTech Connect

    Giles, G.E.

    1995-08-01

    The Advanced Neutron Source Reactor (ANSR) is a research reactor designed to provide the highest continuous neutron beam intensity of any reactor in the world. The present technology for determining safe operations were developed for the High Flux Isotope Reactor (HFIR). These techniques are conservative and provide confidence in the safe operation of HFIR. However, the more intense requirements of ANSR necessitate the development of more accurate, but still conservative, techniques. This report details the development of a Local Analysis Technique (LAT) that provides an appropriate approach. Application of the LAT to two ANSR core designs are presented. New theories of the thermal and nuclear behavior of the U{sub 3}Si{sub 2} fuel are utilized. The implications of lower fuel enrichment and of modifying the inspection procedures are also discussed. Development of the computer codes that enable the automate execution of the LAT is included.

  2. Dismantled weapons fuel burning in molten salt reactors

    SciTech Connect

    Gat, U.; Engel, J.R.

    1993-10-01

    The advantages of burning fissile material from dismantled weapons in molten salt reactors (MSRs) are described. The fluid fuel MSRs with some, or full, processing are nondedicated reactors that generate energy and completely burn the fissile material on a continuous basis. No fuel fabrication is needed, and the entire dismantling can be done in a secure facility. Shipments are made in small, safe, and secure quantities. Denaturing, spiking, or mixing can be done at the source for added safety. MSRs are very safe reactors that help close the fuel cycle and simplify waste treatment, thereby contributing to acceptability. Additionally, MSRs are expected to be economically competitive as electric power stations. The safety, security, simplicity, economy, and proliferation resistant properties support the deployment in countries that have the need.

  3. BISON and MARMOT Development for Modeling Fast Reactor Fuel Performance

    SciTech Connect

    Gamble, Kyle Allan Lawrence; Williamson, Richard L.; Schwen, Daniel; Zhang, Yongfeng; Novascone, Stephen Rhead; Medvedev, Pavel G.

    2015-09-01

    BISON and MARMOT are two codes under development at the Idaho National Laboratory for engineering scale and lower length scale fuel performance modeling. It is desired to add capabilities for fast reactor applications to these codes. The fast reactor fuel types under consideration are metal (U-Pu-Zr) and oxide (MOX). The cladding types of interest include 316SS, D9, and HT9. The purpose of this report is to outline the proposed plans for code development and provide an overview of the models added to the BISON and MARMOT codes for fast reactor fuel behavior. A brief overview of preliminary discussions on the formation of a bilateral agreement between the Idaho National Laboratory and the National Nuclear Laboratory in the United Kingdom is presented.

  4. Feasibility study on the thorium fueled boiling water breeder reactor

    SciTech Connect

    PetrusTakaki, N.

    2012-07-01

    The feasibility of (Th,U)O 2 fueled, boiling water breeder reactor based on conventional BWR technology has been studied. In order to determine the potential use of water cooled thorium reactor as a competitive breeder, this study evaluated criticality, breeding and void reactivity coefficient in response to changes made in MFR and fissile enrichments. The result of the study shows that while using light water as moderator, low moderator to fuel volume ratio (MFR=0.5), it was possible to breed fissile fuel in negative void reactivity condition. However the burnup value was lower than the value of the current LWR. On the other hand, heavy water cooled reactor shows relatively wider feasible breeding region, which lead into possibility of designing a core having better neutronic and economic performance than light water with negative void reactivity coefficient. (authors)

  5. Sensitivity Analysis of Reprocessing Cooling Times on Light Water Reactor and Sodium Fast Reactor Fuel Cycles

    SciTech Connect

    R. M. Ferrer; S. Bays; M. Pope

    2008-04-01

    The purpose of this study is to quantify the effects of variations of the Light Water Reactor (LWR) Spent Nuclear Fuel (SNF) and fast reactor reprocessing cooling time on a Sodium Fast Reactor (SFR) assuming a single-tier fuel cycle scenario. The results from this study show the effects of different cooling times on the SFR’s transuranic (TRU) conversion ratio (CR) and transuranic fuel enrichment. Also, the decay heat, gamma heat and neutron emission of the SFR’s fresh fuel charge were evaluated. A 1000 MWth commercial-scale SFR design was selected as the baseline in this study. Both metal and oxide CR=0.50 SFR designs are investigated.

  6. Wastes from selected activities in two light-water reactor fuel cycles

    SciTech Connect

    Palmer, C.R.; Hill, O.F.

    1980-07-01

    This report presents projected volumes and radioactivities of wastes from the production of electrical energy using light-water reactors (LWR). The projections are based upon data developed for a recent environmental impact statement in which the transuranic wastes (i.e., those wastes containing certain long-lived alpha emitters at concentrations of at least 370 becquerels, or 10 nCi, per gram of waste) from fuel cycle activities were characterized. In addition, since the WG.7 assumed that all fuel cycle wastes except mill tailings are placed in a mined geologic repository, the nontransuranic wastes from several activities are included in the projections reported. The LWR fuel cycles considered are the LWR, once-through fuel cycle (Strategy 1), in which spent fuel is packaged in metal canisters and then isolated in geologic formations; and the LWR U/Pu recycle fuel cycle (Strategy 2), wherein spent fuel is reprocessed for recovery and recycle of uranium and plutonium in LWRs. The wastes projected for the two LWR fuel cycles are summarized. The reactor operations and decommissioning were found to dominate the rate of waste generation in each cycle. These activities account for at least 85% of the fuel cycle waste volume (not including head-end wastes) when normalized to per unit electrical energy generated. At 10 years out of reactor, however, spent fuel elements in Strategy 1 represent 98% of the fuel cycle activity but only 4% of the volume. Similarly, the packaged high-level waste, fuel hulls and hardware in Strategy 2 concentrate greater than 95% of the activity in 2% of the waste volume.

  7. Fuel pins with both target and fuel pellets in an isotope-production reactor

    DOEpatents

    Cawley, W.E.; Omberg, R.P.

    1982-08-19

    A method is described for producing tritium in a fast breeder reactor cooled with liquid metal. Lithium target pellets are placed in close contact with fissile fuel pellets in order to increase the tritium production rate.

  8. Characterization plan for Fort St. Vrain and Peach Bottom graphite fuels

    SciTech Connect

    Maarschman, S.C.; Berting, F.M.; Clemmer, R.G.; Gilbert, E.R.; Guenther, R.J.; Morgan, W.C.; Sliva, P.

    1993-09-01

    Part of Fort St. Vrain (FSV) and most of the Peach Bottom (PB) reactor spent fuels are currently stored at INEL and may remain in storage for many years before disposal. Three disposal pathways have been proposed: intact disposal, fuels partially disassembled and the high-level waste fraction conditioned prior to disposal, and fuels completed disassembled and conditioned prior to disposal. Many options exist within each of these pathways. PNL evaluated the literature and other reference to develop a fuels characterization plan for these fuels. This plan provides guidance for the characteristics of the fuel which will be needed to pursue any of the storage or disposal pathways. It also provides a suggested fuels monitoring program for the current storage facilities. This report recommends a minimum of 7 fuel elements be characterized: PB Core 1 fuel: one Type II nonfailed element, one Type II failed element, and one Type III nonfailed element; PB Core 2 fuel: two Type II nonfailed fuel elements; and FSV fuel: at least two fuel blocks from regions of high temperature and fluence and long in-reactor performance (preferably at reactor end-of- life). Selection of PB fuel elements should focus on these between radial core position 8 and 14 and on compacts between compact numbers 10 and 20. Selection of FSV fuel elements should focus on these from Fuel Zones II and III, located in Core Layers 6, 7, and possibly 8.

  9. Fission-suppressed blankets for fissile fuel breeding fusion reactors

    NASA Astrophysics Data System (ADS)

    Lee, J. D.; Moir, R. W.

    1981-07-01

    Two blanket concepts for deuterium-tritium (DT) fusion reactors are presented which maximize fissile fuel production while at the same time suppress fission reactions. By suppressing fission reactions, the reactor will be less hazardous, and therefore easier to design, develop, and license. A fusion breeder operating a given nuclear power level can produce much more fissile fuel by suppressing fission reactions. The two blankets described use beryllium for neutron multiplication. One blanket uses two separate circulating molten salts: one salt for tritium breeding and the other salt for U-233 breeding. The other uses separate solid forms of lithium and thorium for breeding and helium for cooling.

  10. Spent nuclear fuel discharges from US reactors 1992

    SciTech Connect

    Not Available

    1994-05-05

    This report provides current statistical data on every fuel assembly irradiated in commercial nuclear reactors operating in the United States. It also provides data on the current inventories and storage capacities of those reactors to a wide audience, including Congress, Federal and State agencies, the nuclear and electric industries and the general public. It uses data from the mandatory, ``Nuclear Fuel Data`` survey, Form RW-859 for 1992 and historical data collected by the Energy Information Administration (EIA) on previous Form RW-859 surveys. The report was prepared by the EIA under a Memorandum of Understanding with the Office of Civilian Radioactive Waste Management.

  11. Reactor-specific spent fuel discharge projections, 1984 to 2020

    SciTech Connect

    Heeb, C.M.; Libby, R.A.; Holter, G.M.

    1985-04-01

    The original spent fuel utility data base (SFDB) has been adjusted to produce agreement with the EIA nuclear energy generation forecast. The procedure developed allows the detail of the utility data base to remain intact, while the overall nuclear generation is changed to match any uniform nuclear generation forecast. This procedure adjusts the weight of the reactor discharges as reported on the SFDB and makes a minimal (less than 10%) change in the original discharge exposures in order to preserve discharges of an integral number of fuel assemblies. The procedure used in developing the reactor-specific spent fuel discharge projections, as well as the resulting data bases themselves, are described in detail in this report. Discussions of the procedure cover the following topics: a description of the data base; data base adjustment procedures; addition of generic power reactors; and accuracy of the data base adjustments. Reactor-specific discharge and storage requirements are presented. Annual and cumulative discharge projections are provided. Annual and cumulative requirements for additional storage are shown for the maximum at-reactor (AR) storage assumption, and for the maximum AR with transshipment assumption. These compare directly to the storage requirements from the utility-supplied data, as reported in the Spent Fuel Storage Requirements Report. The results presented in this report include: the disaggregated spent fuel discharge projections; and disaggregated projections of requirements for additional spent fuel storage capacity prior to 1998. Descriptions of the methodology and the results are included in this report. Details supporting the discussions in the main body of the report, including descriptions of the capacity and fuel discharge projections, are included. 3 refs., 6 figs., 12 tabs.

  12. Fuel leak detection apparatus for gas cooled nuclear reactors

    DOEpatents

    Burnette, Richard D.

    1977-01-01

    Apparatus is disclosed for detecting nuclear fuel leaks within nuclear power system reactors, such as high temperature gas cooled reactors. The apparatus includes a probe assembly that is inserted into the high temperature reactor coolant gaseous stream. The probe has an aperture adapted to communicate gaseous fluid between its inside and outside surfaces and also contains an inner tube for sampling gaseous fluid present near the aperture. A high pressure supply of noncontaminated gas is provided to selectively balance the pressure of the stream being sampled to prevent gas from entering the probe through the aperture. The apparatus includes valves that are operable to cause various directional flows and pressures, which valves are located outside of the reactor walls to permit maintenance work and the like to be performed without shutting down the reactor.

  13. Temperature measuring analysis of the nuclear reactor fuel assembly

    NASA Astrophysics Data System (ADS)

    F., Urban; Ľ., Kučák; Bereznai, J.; Závodný, Z.; Muškát, P.

    2014-08-01

    Study was based on rapid changes of measured temperature values from the thermocouple in the VVER 440 nuclear reactor fuel assembly. Task was to determine origin of fluctuations of the temperature values by experiments on physical model of the fuel assembly. During an experiment, heated water was circulating in the system and cold water inlet through central tube to record sensitivity of the temperature sensor. Two positions of the sensor was used. First, just above the central tube in the physical model fuel assembly axis and second at the position of the thermocouple in the VVER 440 nuclear reactor fuel assembly. Dependency of the temperature values on time are presented in the diagram form in the paper.

  14. Retrievable fuel pin end member for a nuclear reactor

    DOEpatents

    Rosa, Jerry M.

    1982-01-01

    A bottom end member (17b) on a retrievable fuel pin (13b) secures the pin (13b) within a nuclear reactor (12) by engaging on a transverse attachment rail (18) with a spring clip type of action. Removal and reinstallation if facilitated as only axial movement of the fuel pin (13b) is required for either operation. A pair of resilient axially extending blades (31) are spaced apart to define a slot (24) having a seat region (34) which receives the rail (18) and having a land region (37), closer to the tips (39) of the blades (31) which is normally of less width than the rail (18). Thus an axially directed force sufficient to wedge the resilient blades (31) apart is required to emplace or release the fuel pin (13b) such force being greater than the axial forces on the fuel pins (13b) which occur during operation of the reactor (12).

  15. Deep Burn: Development of Transuranic Fuel for High-Temperature Helium-Cooled Reactors- Monthly Highlights October 2010

    SciTech Connect

    Snead, Lance Lewis; Besmann, Theodore M; Collins, Emory D; Bell, Gary L

    2010-11-01

    The DB Program monthly highlights report for September 2010, ORNL/TM-2010/252, was distributed to program participants by email on October 26. This report discusses: (1) Core and Fuel Analysis; (2) Spent Fuel Management; (3) Fuel Cycle Integration of the HTR (high temperature helium-cooled reactor); (4) TRU (transuranic elements) HTR Fuel Qualification; (5) HTR Spent Fuel Recycle - (a) TRU Kernel Development (ORNL), (b) Coating Development (ORNL), (c) Characterization Development and Support, (d) ZrC Properties and Handbook; and (6) HTR Fuel Recycle.

  16. Hybrid fusion reactor for production of nuclear fuel with minimum radioactive contamination of the fuel cycle

    NASA Astrophysics Data System (ADS)

    Velikhov, E. P.; Kovalchuk, M. V.; Azizov, E. A.; Ignatiev, V. V.; Subbotin, S. A.; Tsibulskiy, V. F.

    2015-12-01

    The paper presents the results of the system research on the coordinated development of nuclear and fusion power engineering in the current century. Considering the increasing problems of resource procurement, including limited natural uranium resources, it seems reasonable to use fusion reactors as high-power neutron sources for production of nuclear fuel in a blanket. It is shown that the share of fusion sources in this structural configuration of the energy system can be relatively small. A fundamentally important aspect of this solution to the problem of closure of the fuel cycle is that recycling of highly active spent fuel can be abandoned. Radioactivity released during the recycling of the spent fuel from the hybrid reactor blanket is at least two orders of magnitude lower than during the production of the same number of fissile isotopes after the recycling of the spent fuel from a fast reactor.

  17. Hybrid fusion reactor for production of nuclear fuel with minimum radioactive contamination of the fuel cycle

    SciTech Connect

    Velikhov, E. P.; Kovalchuk, M. V.; Azizov, E. A. Ignatiev, V. V.; Subbotin, S. A. Tsibulskiy, V. F.

    2015-12-15

    The paper presents the results of the system research on the coordinated development of nuclear and fusion power engineering in the current century. Considering the increasing problems of resource procurement, including limited natural uranium resources, it seems reasonable to use fusion reactors as high-power neutron sources for production of nuclear fuel in a blanket. It is shown that the share of fusion sources in this structural configuration of the energy system can be relatively small. A fundamentally important aspect of this solution to the problem of closure of the fuel cycle is that recycling of highly active spent fuel can be abandoned. Radioactivity released during the recycling of the spent fuel from the hybrid reactor blanket is at least two orders of magnitude lower than during the production of the same number of fissile isotopes after the recycling of the spent fuel from a fast reactor.

  18. Testing of Gas Reactor Fuel and Materials in the Advanced Test Reactor

    SciTech Connect

    S. Blaine Grover

    2006-10-01

    The recent growth in interest for high temperature gas reactors has resulted in an increased need for materials and fuel testing for this type of reactor. The Advanced Test Reactor (ATR), located at the US Department of Energy’s Idaho National Laboratory, has long been involved in testing gas reactor fuel and materials, and has facilities and capabilities to provide the right environment for gas reactor irradiation experiments. These capabilities include both passive sealed capsule experiments, and instrumented/actively controlled experiments. The instrumented/actively controlled experiments typically contain thermocouples and control the irradiation temperature, but on-line measurements and controls for pressure and gas environment have also been performed in past irradiations. The ATR has an existing automated gas temperature control system that can maintain temperature in an irradiation experiment within very tight bounds, and has developed an on-line fission product monitoring system that is especially well suited for testing gas reactor particle fuel. The ATR’s control system, which consists primarily of vertical cylinders used to rotate neutron poisons/reflectors toward or away from the reactor core, provides a constant vertical flux profile over the duration of each operating cycle. This constant chopped cosine shaped axial flux profile, with a relatively flat peak at the vertical centre of the core, is more desirable for experiments than a constantly moving axial flux peak resulting from a control system of axially positioned control components which are vertically withdrawn from the core.

  19. FUEL ELEMENTS FOR THERMAL-FISSION NUCLEAR REACTORS

    DOEpatents

    Flint, O.

    1961-01-10

    Fuel elements for thermal-fission nuclear reactors are described. The fuel element is comprised of a core of alumina, a film of a metal of the class consisting of copper, silver, and nickel on the outer face of the core, and a coating of an oxide of a metal isotope of the class consisting of Un/sup 235/, U/ sup 233/, and Pu/sup 239/ on the metal f ilm.

  20. A Study of Fast Reactor Fuel Transmutation in a Candidate Dispersion Fuel Design

    SciTech Connect

    Mark DeHart; Hongbin Zhang; Eric Shaber; Matthew Jesse

    2010-11-01

    Dispersion fuels represent a significant departure from typical ceramic fuels to address swelling and radiation damage in high burnup fuel. Such fuels use a manufacturing process in which fuel particles are encapsulated within a non-fuel matrix. Dispersion fuels have been studied since 1997 as part of an international effort to develop and test very high density fuel types for the Reduced Enrichment for Research and Test Reactors (RERTR) program.[1] The Idaho National Laboratory is performing research in the development of an innovative dispersion fuel concept that will meet the challenges of transuranic (TRU) transmutation by providing an integral fission gas plenum within the fuel itself, to eliminate the swelling that accompanies the irradiation of TRU. In this process, a metal TRU vector produced in a separations process is atomized into solid microspheres. The dispersion fuel process overcoats the microspheres with a mixture of resin and hollow carbon microspheres to create a TRUC. The foam may then be heated and mixed with a metal power (e.g., Zr, Ti, or Si) and resin to form a matrix metal carbide, that may be compacted and extruded into fuel elements. In this paper, we perform reactor physics calculations for a core loaded with the conceptual fuel design. We will assume a “typical” TRU vector and a reference matrix density. We will employ a fuel and core design based on the Advanced Burner Test Reactor (ABTR) design.[2] Using the CSAS6 and TRITON modules of the SCALE system [3] for preliminary scoping studies, we will demonstrate the feasibility of reactor operations. This paper will describe the results of these analyses.

  1. Characterization of thermal-hydraulic and ignition phenomena in prototypic, full-length boiling water reactor spent fuel pool assemblies after a complete loss-of-coolant accident.

    SciTech Connect

    Lindgren, Eric Richard; Durbin, Samuel G

    2007-04-01

    The objective of this project was to provide basic thermal-hydraulic data associated with a SFP complete loss-of-coolant accident. The accident conditions of interest for the SFP were simulated in a full-scale prototypic fashion (electrically-heated, prototypic assemblies in a prototypic SFP rack) so that the experimental results closely represent actual fuel assembly responses. A major impetus for this work was to facilitate code validation (primarily MELCOR) and reduce questions associated with interpretation of the experimental results. It was necessary to simulate a cluster of assemblies to represent a higher decay (younger) assembly surrounded by older, lower-power assemblies. Specifically, this program provided data and analysis confirming: (1) MELCOR modeling of inter-assembly radiant heat transfer, (2) flow resistance modeling and the natural convective flow induced in a fuel assembly as it heats up in air, (3) the potential for and nature of thermal transient (i.e., Zircaloy fire) propagation, and (4) mitigation strategies concerning fuel assembly management.

  2. Means for supporting fuel elements in a nuclear reactor

    DOEpatents

    Andrews, Harry N.; Keller, Herbert W.

    1980-01-01

    A grid structure for a nuclear reactor fuel assembly comprising a plurality of connecting members forming at least one longitudinally extending opening peripheral and inner fuel element openings through each of which openings at least one nuclear fuel element extends, said connecting members forming wall means surrounding said each peripheral and inner fuel element opening, a pair of rigid projections longitudinally spaced from one another extending from a portion of said wall means into said each peripheral and inner opening for rigidly engaging said each fuel element, respectively, yet permit individual longitudinal slippage thereof, and resilient means formed integrally on and from said wall means and positioned in said each peripheral and inner opening in opposed relationship with said projections and located to engage said fuel element to bias the latter into engagement with said rigid projections, respectively

  3. Proliferation resistant fuel for pebble bed modular reactors

    SciTech Connect

    Ronen, Y.; Aboudy, M.; Regev, D.; Gilad, E.

    2012-07-01

    We show that it is possible to denature the Plutonium produced in Pebble Bed Modular Reactors (PBMR) by doping the nuclear fuel with either 3050 ppm of {sup 237}Np or 2100 ppm of Am vector. A correct choice of these isotopes concentration yields denatured Plutonium with isotopic ratio {sup 238}Pu/Pu {>=} 6%, for the entire fuel burnup cycle. The penalty for introducing these isotopes into the nuclear fuel is a subsequent shortening of the fuel burnup cycle, with respect to a non-doped reference fuel, by 41.2 Full Power Days (FPDs) and 19.9 FPDs, respectively, which correspond to 4070 MWd/ton and 1965 MWd/ton reduction in fuel discharge burnup. (authors)

  4. Inertial fusion energy power reactor fuel recovery system

    SciTech Connect

    Gentile, C. A.; Kozub, T.; Langish, S. W.; Ciebiera, L. P.; Nobile, A.; Wermer, J.; Sessions, K.

    2008-07-15

    A conceptual design is proposed to support the recovery of un-expended fuel, ash, and associated post-detonation products resident in plasma exhaust from a {approx}2 GWIFE direct drive power reactor. The design includes systems for the safe and efficient collection, processing, and purification of plasma exhaust fuel components. The system has been conceptually designed and sized such that tritium bred within blankets, lining the reactor target chamber, can also be collected, processed, and introduced into the fuel cycle. The system will nominally be sized to process {approx}2 kg of tritium per day and is designed to link directly to the target chamber vacuum pumping system. An effort to model the fuel recovery system (FRS) using the Aspen Plus engineering code has commenced. The system design supports processing effluent gases from the reactor directly from the exhaust of the vacuum pumping system or in batch mode, via a buffer vessel in the Receiving and Analysis System. Emphasis is on nuclear safety, reliability, and redundancy as to maximize availability. The primary goal of the fuel recovery system design is to economically recycle components of direct drive IFE fuel. The FRS design is presented as a facility sub-system in the context of supporting the larger goal of producing safe and economical IFE power. (authors)

  5. Silicon carbide composite for light water reactor fuel assembly applications

    NASA Astrophysics Data System (ADS)

    Yueh, Ken; Terrani, Kurt A.

    2014-05-01

    The feasibility of using SiCf-SiCm composites in light water reactor (LWR) fuel designs was evaluated. The evaluation was motivated by the desire to improve fuel performance under normal and accident conditions. The Fukushima accident once again highlighted the need for improved fuel materials that can maintain fuel integrity to higher temperatures for longer periods of time. The review identified many benefits as well as issues in using the material. Issues perceived as presenting the biggest challenges to the concept were identified to be flux gradient induced differential volumetric swelling, fragmentation and thermal shock resistance. The oxidation of silicon and its release into the coolant as silica has been identified as an issue because existing plant systems have limited ability for its removal. Detailed evaluation using available literature data and testing as part of this evaluation effort have eliminated most of the major concerns. The evaluation identified Boiling Water Reactor (BWR) channel, BWR fuel water tube, and Pressurized Water Reactor (PWR) guide tube as feasible applications for SiC composite. A program has been initiated to resolve some of the remaining issues and to generate physical property data to support the design of commercial fuel components.

  6. LIGHT WATER REACTOR ACCIDENT TOLERANT FUELS IRRADIATION TESTING

    SciTech Connect

    Carmack, William Jonathan; Barrett, Kristine Eloise; Chichester, Heather Jean MacLean

    2015-09-01

    The purpose of Accident Tolerant Fuels (ATF) experiments is to test novel fuel and cladding concepts designed to replace the current zirconium alloy uranium dioxide (UO2) fuel system. The objective of this Research and Development (R&D) is to develop novel ATF concepts that will be able to withstand loss of active cooling in the reactor core for a considerably longer time period than the current fuel system while maintaining or improving the fuel performance during normal operations, operational transients, design basis, and beyond design basis events. It was necessary to design, analyze, and fabricate drop-in capsules to meet the requirements for testing under prototypic LWR temperatures in Idaho National Laboratory's Advanced Test Reactor (ATR). Three industry led teams and one DOE team from Oak Ridge National Laboratory provided fuel rodlet samples for their new concepts for ATR insertion in 2015. As-built projected temperature calculations were performed on the ATF capsules using the BISON fuel performance code. BISON is an application of INL’s Multi-physics Object Oriented Simulation Environment (MOOSE), which is a massively parallel finite element based framework used to solve systems of fully coupled nonlinear partial differential equations. Both 2D and 3D models were set up to examine cladding and fuel performance.

  7. Exploratory Design of a Reactor/Fuel Cycle Using Spent Nuclear Fuel Without Conventional Reprocessing - 13579

    SciTech Connect

    Bertch, Timothy C.; Schleicher, Robert W.; Rawls, John D.

    2013-07-01

    General Atomics has started design of a waste to energy nuclear reactor (EM2) that can use light water reactor (LWR) spent nuclear fuel (SNF). This effort addresses two problems: using an advanced small reactor with long core life to reduce nuclear energy overnight cost and providing a disposal path for LWR SNF. LWR SNF is re-fabricated into new EM2 fuel using a dry voloxidation process modeled on AIROX/ OREOX processes which remove some of the fission products but no heavy metals. By not removing all of the fission products the fuel remains self-protecting. By not separating heavy metals, the process remains proliferation resistant. Implementation of Energy Multiplier Module (EM2) fuel cycle will provide low cost nuclear energy while providing a long term LWR SNF disposition path which is important for LWR waste confidence. With LWR waste confidence recent impacts on reactor licensing, an alternate disposition path is highly relevant. Centered on a reactor operating at 250 MWe, the compact electricity generating system design maximizes site flexibility with truck transport of all system components and available dry cooling features that removes the need to be located near a body of water. A high temperature system using helium coolant, electricity is efficiently produced using an asynchronous high-speed gas turbine while the LWR SNF is converted to fission products. Reactor design features such as vented fuel and silicon carbide cladding support reactor operation for decades between refueling, with improved fuel utilization. Beyond the reactor, the fuel cycle is designed so that subsequent generations of EM2 reactor fuel will use the previous EM2 discharge, providing its own waste confidence plus eliminating the need for enrichment after the first generation. Additional LWR SNF is added at each re-fabrication to replace the removed fission products. The fuel cycle uses a dry voloxidation process for both the initial LWR SNF re-fabrication and later for EM2

  8. Nuclear breeder reactor fuel element with silicon carbide getter

    DOEpatents

    Christiansen, David W.; Karnesky, Richard A.

    1987-01-01

    An improved cesium getter 28 is provided in a breeder reactor fuel element or pin in the form of an extended surface area, low density element formed in one embodiment as a helically wound foil 30 located with silicon carbide, and located at the upper end of the fertile material upper blanket 20.

  9. Method of controlling crystallite size in nuclear-reactor fuels

    DOEpatents

    Lloyd, Milton H.; Collins, Jack L.; Shell, Sam E.

    1985-01-01

    Improved spherules for making enhanced forms of nuclear-reactor fuels are prepared by internal gelation procedures within a sol-gel operation and are accomplished by first boiling the concentrated HMTA-urea feed solution before engaging in the spherule-forming operation thereby effectively controlling crystallite size in the product spherules.

  10. Method of controlling crystallite size in nuclear-reactor fuels

    DOEpatents

    Lloyd, M.H.; Collins, J.L.; Shell, S.E.

    Improved spherules for making enhanced forms of nuclear-reactor fuels are prepared by internal gelation procedures within a sol-gel operation and are accomplished by first boiling the concentrated HMTA-urea feed solution before engaging in the spherule-forming operation thereby effectively controlling crystallite size in the product spherules.

  11. Yttrium and rare earth stabilized fast reactor metal fuel

    SciTech Connect

    Guon, J.; Grantham, L.F.; Specht, E.R.

    1992-05-12

    This patent describes an improved metal alloy reactor fuel consisting essentially of uranium, plutonium, and at least one element from the group consisting of yttrium, lanthanum, cerium, praseodymium, neodymium, promethium, samarium, europium, gadolinium, terbium, dysprosium, holmium, erbium, thulium, ytterbium and lutetium.

  12. Treatment of Off-Gases from Melting Reactor Fuel

    SciTech Connect

    Peacock, H.B.

    2002-05-02

    The ''Melt-Dilute'' process has been studied as a method for preparing spent highly enriched reactor fuels for repository disposition. This study was undertaken to determine the behavior of cesium under projected process conditions, and to investigate techniques for its recovery and disposition.

  13. METHOD OF FORMING A FUEL ELEMENT FOR A NUCLEAR REACTOR

    DOEpatents

    Layer, E.H. Jr.; Peet, C.S.

    1962-01-23

    A method is given for preparing a fuel element for a nuclear reactor. The method includes the steps of sandblasting a body of uranium dioxide to roughen the surface thereof, depositing a thin layer of carbon thereon by thermal decomposition of methane, and cladding the uranium dioxide body with zirconium by gas pressure bonding. (AEC)

  14. METHOD OF PREPARING A FUEL ELEMENT FOR A NUCLEAR REACTOR

    DOEpatents

    Hauth, J.J.; Anicetti, R.J.

    1962-12-01

    A method is described for preparing a fuel element for a nuclear reactor. According to the patent uranium dioxide is compacted in a metal tabe by directlng intense sound waves at the tabe prior to tamp packing or vibration compaction of the powder. (AEC)

  15. Method of locating a leaking fuel element in a fast breeder power reactor

    DOEpatents

    Honekamp, John R.; Fryer, Richard M.

    1978-01-01

    Leaking fuel elements in a fast reactor are identified by measuring the ratio of .sup.134 Xe to .sup.133 Xe in the reactor cover gas following detection of a fuel element leak, this ratio being indicative of the power and burnup of the failed fuel element. This procedure can be used to identify leaking fuel elements in a power breeder reactor while continuing operation of the reactor since the ratio measured is that of the gases stored in the plenum of the failed fuel element. Thus, use of a cleanup system for the cover gas makes it possible to identify sequentially a multiplicity of leaking fuel elements without shutting the reactor down.

  16. Advanced Gas Reactor (AGR)-5/6/7 Fuel Irradiation Experiments in the Advanced Test Reactor

    SciTech Connect

    A. Joseph Palmer; David A. Petti; S. Blaine Grover

    2014-04-01

    The United States Department of Energy’s Very High Temperature Reactor (VHTR) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating up to seven separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which each consist of at least five separate capsules, are being irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gases also have on-line fission product monitoring the effluent from each capsule to track performance of the fuel during irradiation. The first two experiments (designated AGR-1 and AGR-2), have been completed. The third and fourth experiments have been combined into a single experiment designated AGR-3/4, which started its irradiation in December 2011 and is currently scheduled to be completed in April 2014. The design of the fuel qualification experiment, designated AGR-5/6/7, is well underway and incorporates lessons learned from the three previous experiments. Various design issues will be discussed with particular details related to selection of thermometry.

  17. Simulation on reactor TRIGA Puspati core kinetics fueled with thorium (Th) based fuel element

    NASA Astrophysics Data System (ADS)

    Mohammed, Abdul Aziz; Pauzi, Anas Muhamad; Rahman, Shaik Mohmmed Haikhal Abdul; Zin, Muhamad Rawi Muhammad; Jamro, Rafhayudi; Idris, Faridah Mohamad

    2016-01-01

    In confronting global energy requirement and the search for better technologies, there is a real case for widening the range of potential variations in the design of nuclear power plants. Smaller and simpler reactors are attractive, provided they can meet safety and security standards and non-proliferation issues. On fuel cycle aspect, thorium fuel cycles produce much less plutonium and other radioactive transuranic elements than uranium fuel cycles. Although not fissile itself, Th-232 will absorb slow neutrons to produce uranium-233 (233U), which is fissile. By introducing Thorium, the numbers of highly enriched uranium fuel element can be reduced while maintaining the core neutronic performance. This paper describes the core kinetic of a small research reactor core like TRIGA fueled with a Th filled fuel element matrix using a general purpose Monte Carlo N-Particle (MCNP) code.

  18. The DOE advanced gas reactor fuel development and qualification program

    NASA Astrophysics Data System (ADS)

    Petti, David; Maki, John; Hunn, John; Pappano, Pete; Barnes, Charles; Saurwein, John; Nagley, Scott; Kendall, Jim; Hobbins, Richard

    2010-09-01

    The high outlet temperatures and high thermal-energy conversion efficiency of modular high-temperature gas-cooled reactors (HTGRs) enable an efficient and cost-effective integration of the reactor system with non-electricity-generation applications, such as process heat and/or hydrogen production, for the many petrochemical and other industrial processes that require temperatures between 300°C and 900°C. The U.S. Department of Energy (DOE) has selected the HTGR concept for the Next Generation Nuclear Plant (NGNP) Project as a transformative application of nuclear energy that will demonstrate emissions-free nuclear-assisted electricity, process heat, and hydrogen production, thereby reducing greenhouse-gas emissions and enhancing energy security. The objective of the DOE Advanced Gas Reactor (AGR) Fuel Development and Qualification program is to qualify tristructural isotropic (TRISO)-coated particle fuel for use in HTGRs. An overview of the program and recent progress is presented.

  19. Electrolysis cell for reprocessing plutonium reactor fuel

    DOEpatents

    Miller, W.E.; Steindler, M.J.; Burris, L.

    1985-01-04

    An electrolytic cell for refining a mixture of metals including spent fuel containing U and Pu contaminated with other metals is claimed. The cell includes a metallic pot containing a metallic pool as one anode at a lower level, a fused salt as the electrolyte at an intermediate level and a cathode and an anode basket in spaced-apart positions in the electrolyte with the cathode and anode being retractable to positions above the electrolyte during which spent fuel may be added to the anode basket. The anode basket is extendable into the lower pool to dissolve at least some metallic contaminants; the anode basket contains the spent fuel acting as a second anode when in the electrolyte.

  20. Electrolysis cell for reprocessing plutonium reactor fuel

    DOEpatents

    Miller, William E.; Steindler, Martin J.; Burris, Leslie

    1986-01-01

    An electrolytic cell for refining a mixture of metals including spent fuel containing U and Pu contaminated with other metals, the cell including a metallic pot containing a metallic pool as one anode at a lower level, a fused salt as the electrolyte at an intermediate level and a cathode and an anode basket in spaced-apart positions in the electrolyte with the cathode and anode being retractable to positions above the electrolyte during which spent fuel may be added to the anode basket and the anode basket being extendable into the lower pool to dissolve at least some metallic contaminants, the anode basket containing the spent fuel acting as a second anode when in the electrolyte.

  1. Chemicals from heavy fuels in millisecond catalytic reactors

    NASA Astrophysics Data System (ADS)

    Krummenacher, Jakob Jose

    The role that millisecond catalytic reactors play on the development of the hydrogen economy is crucial because they have higher throughputs and smaller volumes than their traditional counter parts. Issues with the storage and production of hydrogen give an advantage to these reactors because their compact and small size makes them highly portable. As with the development of any new technology challenges that threaten the operation of the millisecond catalytic reactor with liquid fuels were encountered. First, the autoignition of linear alkanes is higher than their boiling points. This makes it impossible to simply vaporize the fuel and mix it with air. Second, analyzing such a diverse range of products using gas chromatography is not possible with conventional equipment. The first results of the especially engineered reactor show the successful partial oxidation of linear alkanes. Decane and hexadecane over rhodium catalysts produce synthesis gas with selectivities exceeding 80%. Experiments confirmed the highly tunable nature of the reactor. A fuel concentrated feed produces ethylene and alpha-olefins with high selectivities. A proposed mechanism and experimental results obtained suggest that these olefins are formed by homogeneous endothermic cracking. A single fuel, decane, was used to study further the millisecond catalytic reactor versatility. Several operating conditions such as catalyst porosity, effect of wash-coat, addition of steam, and addition of hydrogen were studied. Rhodium, platinum, and platinum-rhodium mixtures were coated on supports of different porosity to determine the differences in metal reactivities. The results show that rhodium is the best catalyst for making hydrogen or synthesis gas and that adding steam improves its performance. The highest yields of ethylene and a-olefins were obtained using a platinum catalyst coated with a small amount of rhodium on the front face. A rough economic analysis gives a sound idea of the potential that

  2. Design of the Advanced Gas Reactor Fuel Experiments for Irradiation in the Advanced Test Reactor

    SciTech Connect

    S. Blaine Grover

    2005-10-01

    The United States Department of Energy’s Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating eight particle fuel tests in the Advanced Test Reactor (ATR) located at the newly formed Idaho National Laboratory (INL) to support development of the next generation Very High Temperature Reactor (VHTR) in the United States. The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the new United States Department of Energy’s lead laboratory for nuclear energy development. These AGR fuel experiments will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The experiments will be irradiated in an inert sweep gas atmosphere with on-line temperature monitoring and control combined with on-line fission product monitoring of the sweep gas. The final design phase has just been completed on the first experiment (AGR-1) in this series and the support systems and fission product monitoring system that will monitor and control the experiment during irradiation. This paper discusses the development of the experimental hardware and support system designs and the status of the experiment.

  3. 10 CFR 71.97 - Advance notification of shipment of irradiated reactor fuel and nuclear waste.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... notification of shipment of irradiated reactor fuel and nuclear waste. (a)(1) As specified in paragraphs (b... shipment of irradiated reactor fuel or nuclear waste must contain the following information: (1) The name... nuclear waste shipment; (2) A description of the irradiated reactor fuel or nuclear waste contained in...

  4. 10 CFR 71.97 - Advance notification of shipment of irradiated reactor fuel and nuclear waste.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... notification of shipment of irradiated reactor fuel and nuclear waste. (a) As specified in paragraphs (b), (c... of the shipper, carrier, and receiver of the irradiated reactor fuel or nuclear waste shipment; (2) A description of the irradiated reactor fuel or nuclear waste contained in the shipment, as specified in...

  5. 10 CFR 71.97 - Advance notification of shipment of irradiated reactor fuel and nuclear waste.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... notification of shipment of irradiated reactor fuel and nuclear waste. (a) As specified in paragraphs (b), (c... of the shipper, carrier, and receiver of the irradiated reactor fuel or nuclear waste shipment; (2) A description of the irradiated reactor fuel or nuclear waste contained in the shipment, as specified in...

  6. 10 CFR 71.97 - Advance notification of shipment of irradiated reactor fuel and nuclear waste.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... notification of shipment of irradiated reactor fuel and nuclear waste. (a)(1) As specified in paragraphs (b... shipment of irradiated reactor fuel or nuclear waste must contain the following information: (1) The name... nuclear waste shipment; (2) A description of the irradiated reactor fuel or nuclear waste contained in...

  7. Testing of Gas Reactor Materials and Fuel in the Advanced Test Reactor

    SciTech Connect

    S. Blaine Grover

    2004-10-01

    The Advanced Test Reactor (ATR) has long been involved in testing gas reactor materials, and has developed facilities well suited for providing the right conditions and environment for gas reactor tests. This paper discusses the different types of irradiation hardware that have been utilized in past ATR irradiation tests of gas reactor materials. The new Gas Test Loop facility currently being developed for the ATR is discussed and the different approaches being considered in the design of the facility. The different options for an irradiation experiment such as active versus passive temperature control, neutron spectrum tailoring, and different types of lead experiment sweep gas monitors are also discussed. The paper is then concluded with examples of different past and present gas reactor material and fuel irradiations.

  8. TESTING OF GAS REACTOR MATERIALS AND FUEL IN THE ADVANCED TEST REACTOR

    SciTech Connect

    Grover, S.B.

    2004-10-06

    The Advanced Test Reactor (ATR) has long been involved in testing gas reactor materials, and has developed facilities well suited for providing the right conditions and environment for gas reactor tests. This paper discusses the different types of irradiation hardware that have been utilized in past ATR irradiation tests of gas reactor materials. The new Gas Test Loop facility currently being developed for the ATR is discussed and the different approaches being considered in the design of the facility. The different options for an irradiation experiment such as active versus passive temperature control, neutron spectrum tailoring, and different types of lead experiment sweep gas monitors are also discussed. The paper is then concluded with examples of different past and present gas reactor material and fuel irradiations.

  9. Overview of past and current activities on fuels for fast reactors at the Institute for Transuranium Elements

    NASA Astrophysics Data System (ADS)

    Fernandez, A.; McGinley, J.; Somers, J.; Walter, M.

    2009-07-01

    Nuclear energy has the potential to provide a secure and sustainable electricity supply at a competitive price and to make a significant contribution to the reduction of greenhouse gas emissions. The renewal of interest in fast neutron spectra reactors to meet more ambitious sustainable development criteria (i.e., resource maximisation and waste minimisation), opens a favourable framework for R&D activities in this area. The Institute for Transuranium Elements has extensive experience in the fabrication, characterization and irradiation testing (Phénix, Dounreay, Rapsodie) of fast reactor fuels, in oxide, nitride and carbide forms. An overview of these past and current activities on fast reactor fuels is presented.

  10. ALLOY FOR FUEL OF NEUTRONIC REACTORS

    DOEpatents

    Bloomster, C.H.; Katayama, Y.B.

    1963-04-23

    This patent deals with an aluminum alloy suitable as nuclear fuel and consisting mainly of from 1 to 10 wt% of plutonium, from 2 to 3.5 wt% of nickel, the balance being aluminum. The alloy may also contain from 0.9 to 1.1 wt% of silicon and up to 0.7% of iron. (AEC)

  11. Thermal barrier and support for nuclear reactor fuel core

    DOEpatents

    Betts, Jr., William S.; Pickering, J. Larry; Black, William E.

    1987-01-01

    A thermal barrier/core support for the fuel core of a nuclear reactor having a metallic cylinder secured to the reactor vessel liner and surrounded by fibrous insulation material. A top cap is secured to the upper end of the metallic cylinder that locates and orients a cover block and post seat. Under normal operating conditions, the metallic cylinder supports the entire load exerted by its associated fuel core post. Disposed within the metallic cylinder is a column of ceramic material, the height of which is less than that of the metallic cylinder, and thus is not normally load bearing. In the event of a temperature excursion beyond the design limits of the metallic cylinder and resulting in deformation of the cylinder, the ceramic column will abut the top cap to support the fuel core post.

  12. A cermet fuel reactor for nuclear thermal propulsion

    NASA Technical Reports Server (NTRS)

    Kruger, Gordon

    1991-01-01

    Work on the cermet fuel reactor done in the 1960's by General Electric (GE) and the Argonne National Laboratory (ANL) that had as its goal the development of systems that could be used for nuclear rocket propulsion as well as closed cycle propulsion system designs for ship propulsion, space nuclear propulsion, and other propulsion systems is reviewed. It is concluded that the work done in the 1960's has demonstrated that we can have excellent thermal and mechanical performance with cermet fuel. Thousands of hours of testing were performed on the cermet fuel at both GE and AGL, including very rapid transients and some radiation performance history. We conclude that there are no feasibility issues with cermet fuel. What is needed is reactivation of existing technology and qualification testing of a specific fuel form. We believe this can be done with a minimum development risk.

  13. Deep-Burn Modular Helium Reactor Fuel Development Plan

    SciTech Connect

    McEachern, D

    2002-12-02

    This document contains the workscope, schedule and cost for the technology development tasks needed to satisfy the fuel and fission product transport Design Data Needs (DDNs) for the Gas Turbine-Modular Helium Reactor (GT-MHR), operating in its role of transmuting transuranic (TRU) nuclides in spent fuel discharged from commercial light-water reactors (LWRs). In its application for transmutation, the GT-MHR is referred to as the Deep-Burn MHR (DB-MHR). This Fuel Development Plan (FDP) describes part of the overall program being undertaken by the U.S. Department of Energy (DOE), utilities, and industry to evaluate the use of the GT-MHR to transmute transuranic nuclides from spent nuclear fuel. The Fuel Development Plan (FDP) includes the work on fuel necessary to support the design and licensing of the DB-MHR. The FDP is organized into ten sections. Section 1 provides a summary of the most important features of the plan, including cost and schedule information. Section 2 describes the DB-MHR concept, the features of its fuel and the plan to develop coated particle fuel for transmutation. Section 3 describes the knowledge base for fabrication of coated particles, the experience with irradiation performance of coated particle fuels, the database for fission product transport in HTGR cores, and describes test data and calculations for the performance of coated particle fuel while in a repository. Section 4 presents the fuel performance requirements in terms of as-manufactured quality and performance of the fuel coatings under irradiation and accident conditions. These requirements are provisional because the design of the DB-MHR is in an early stage. However, the requirements are presented in this preliminary form to guide the initial work on the fuel development. Section 4 also presents limits on the irradiation conditions to which the coated particle fuel can be subjected for the core design. These limits are based on past irradiation experience. Section 5 describes

  14. Sodium fast reactor fuels and materials : research needs.

    SciTech Connect

    Denman, Matthew R.; Porter, Douglas; Wright, Art; Lambert, John; Hayes, Steven; Natesan, Ken; Ott, Larry J.; Garner, Frank; Walters, Leon; Yacout, Abdellatif

    2011-09-01

    An expert panel was assembled to identify gaps in fuels and materials research prior to licensing sodium cooled fast reactor (SFR) design. The expert panel considered both metal and oxide fuels, various cladding and duct materials, structural materials, fuel performance codes, fabrication capability and records, and transient behavior of fuel types. A methodology was developed to rate the relative importance of phenomena and properties both as to importance to a regulatory body and the maturity of the technology base. The technology base for fuels and cladding was divided into three regimes: information of high maturity under conservative operating conditions, information of low maturity under more aggressive operating conditions, and future design expectations where meager data exist.

  15. Moving bed reactor for solar thermochemical fuel production

    SciTech Connect

    Ermanoski, Ivan

    2013-04-16

    Reactors and methods for solar thermochemical reactions are disclosed. Embodiments of reactors include at least two distinct reactor chambers between which there is at least a pressure differential. In embodiments, reactive particles are exchanged between chambers during a reaction cycle to thermally reduce the particles at first conditions and oxidize the particles at second conditions to produce chemical work from heat. In embodiments, chambers of a reactor are coupled to a heat exchanger to pre-heat the reactive particles prior to direct exposure to thermal energy with heat transferred from reduced reactive particles as the particles are oppositely conveyed between the thermal reduction chamber and the fuel production chamber. In an embodiment, particle conveyance is in part provided by an elevator which may further function as a heat exchanger.

  16. Design of a new portable fork detector for research reactor spent fuel

    SciTech Connect

    Hsue, S.T.; Menlove, H.O.; Rinard, P.M.

    1995-02-01

    There are many situations in nonproliferation and international safeguards when one needs to verify spent research-reactor fuel. Special inspections, a reactor coming under safeguards for the first time, and failed surveillance are prime examples. Several years ago, Los Alamos developed the FORK detector for the IAEA and EURATOM. This detector, together with the GRAND electronics package, is used routinely by inspectors to verify light-water-reactor spent fuels. Both the FORK detector and the GRAND electronics technologies have been transferred and are now commercially available. Recent incidents in the world indicate that research-reactor fuel is potentially a greater concern for proliferation than light-water-reactor fuels. A device similar to the FORK/GRAND should be developed to verify research-reactor spent fuels because the signals from light-water-reactor spent fuel are quite different than those from research-reactor fuels.

  17. Toxicity of irradiated advanced heavy water reactor fuels.

    PubMed

    Priest, N D; Richardson, R B; Edwards, G W R

    2013-02-01

    The good neutron economy and online refueling capability of the CANDU® heavy water moderated reactor (HWR) enable it to use many different fuels such as low enriched uranium (LEU), plutonium, or thorium, in addition to its traditional natural uranium (NU) fuel. The toxicity and radiological protection methods for these proposed fuels, unlike those for NU, are not well established. This study uses software to compare the fuel composition and toxicity of irradiated NU fuel against those of two irradiated advanced HWR fuel bundles as a function of post-irradiation time. The first bundle investigated is a CANFLEX® low void reactor fuel (LVRF), of which only the dysprosium-poisoned central element, and not the outer 42 LEU elements, is specifically analyzed. The second bundle investigated is a heterogeneous high-burnup (LEU,Th)O(2) fuelled bundle, whose two components (LEU in the outer 35 elements and thorium in the central eight elements) are analyzed separately. The LVRF central element was estimated to have a much lower toxicity than that of NU at all times after shutdown. Both the high burnup LEU and the thorium fuel had similar toxicity to NU at shutdown, but due to the creation of such inhalation hazards as (238)Pu, (240)Pu, (242)Am, (242)Cm, and (244)Cm (in high burnup LEU), and (232)U and (228)Th (in irradiated thorium), the toxicity of these fuels was almost double that of irradiated NU after 2,700 d of cooling. New urine bioassay methods for higher actinoids and the analysis of thorium in fecal samples are recommended to assess the internal dose from these two fuels. PMID:23274823

  18. Toxicity of irradiated advanced heavy water reactor fuels.

    PubMed

    Priest, N D; Richardson, R B; Edwards, G W R

    2013-02-01

    The good neutron economy and online refueling capability of the CANDU® heavy water moderated reactor (HWR) enable it to use many different fuels such as low enriched uranium (LEU), plutonium, or thorium, in addition to its traditional natural uranium (NU) fuel. The toxicity and radiological protection methods for these proposed fuels, unlike those for NU, are not well established. This study uses software to compare the fuel composition and toxicity of irradiated NU fuel against those of two irradiated advanced HWR fuel bundles as a function of post-irradiation time. The first bundle investigated is a CANFLEX® low void reactor fuel (LVRF), of which only the dysprosium-poisoned central element, and not the outer 42 LEU elements, is specifically analyzed. The second bundle investigated is a heterogeneous high-burnup (LEU,Th)O(2) fuelled bundle, whose two components (LEU in the outer 35 elements and thorium in the central eight elements) are analyzed separately. The LVRF central element was estimated to have a much lower toxicity than that of NU at all times after shutdown. Both the high burnup LEU and the thorium fuel had similar toxicity to NU at shutdown, but due to the creation of such inhalation hazards as (238)Pu, (240)Pu, (242)Am, (242)Cm, and (244)Cm (in high burnup LEU), and (232)U and (228)Th (in irradiated thorium), the toxicity of these fuels was almost double that of irradiated NU after 2,700 d of cooling. New urine bioassay methods for higher actinoids and the analysis of thorium in fecal samples are recommended to assess the internal dose from these two fuels.

  19. Transportation impact analysis for shipment of irradiated N-reactor fuel and associated materials

    SciTech Connect

    Daling, P.M.; Harris, M.S.

    1994-12-01

    An analysis of the radiological and nonradiological impacts of highway transportation of N-Reactor irradiated fuel (N-fuel) and associated materials is described in this report. N-fuel is proposed to be transported from its present locations in the 105-KE and 105-KW Basins, and possibly the PUREX Facility, to the 327 Building for characterization and testing. Each of these facilities is located on the Hanford Site, which is near Richland, Washington. The projected annual shipping quantity is 500 kgU/yr for 5 years for a total of 2500 kgU. It was assumed the irradiated fuel would be returned to the K- Basins following characterization, so the total amount of fuel shipped was assumed to be 5000 kgU. The shipping campaign may also include the transport and characterization of liquids, gases, and sludges from the storage basins, including fuel assembly and/or canister parts that may also be present in the basins. The impacts of transporting these other materials are bounded by the impacts of transporting 5000 kgU of N-fuel. This report was prepared to support an environmental assessment of the N-fuel characterization program. The RADTRAN 4 and GENII computer codes were used to evaluate the radiological impacts of the proposed shipping campaign. RADTRAN 4 was used to calculate the routine exposures and accident risks to workers and the general public from the N-fuel shipments. The GENII computer code was used to calculate the consequences of the maximum credible accident. The results indicate that the transportation of N-fuel in support of the characterization program should not cause excess radiological-induced latent cancer fatalities or traffic-related nonradiological accident fatalities. The consequences of the maximum credible accident are projected to be small and result in no excess latent cancer fatalities.

  20. Electrolytic recovery of reactor metal fuel

    DOEpatents

    Miller, W.E.; Tomczuk, Z.

    1994-09-20

    A new electrolytic process and apparatus are provided using sodium, cerium or a similar metal in alloy or within a sodium beta or beta[double prime]-alumina sodium ion conductor to electrolytically displace each of the spent fuel metals except for cesium and strontium on a selective basis from the electrolyte to an inert metal cathode. Each of the metals can be deposited separately. An electrolytic transfer of spent fuel into the electrolyte includes a sodium or cerium salt in the electrolyte with sodium or cerium alloy being deposited on the cathode during the transfer of the metals from the spent fuel. The cathode with the deposit of sodium or cerium alloy is then shunted to an anode and the reverse transfer is carried out on a selective basis with each metal being deposited separately at the cathode. The result is that the sodium or cerium needed for the process is regenerated in the first step and no additional source of these reactants is required. 2 figs.

  1. Electrolytic recovery of reactor metal fuel

    DOEpatents

    Miller, William E.; Tomczuk, Zygmunt

    1994-01-01

    A new electrolytic process and apparatus are provided using sodium, cerium or a similar metal in alloy or within a sodium beta or beta"-alumina sodium ion conductor to electrolytically displace each of the spent fuel metals except for cesium and strontium on a selective basis from the electrolyte to an inert metal cathode. Each of the metals can be deposited separately. An electrolytic transfer of spent fuel into the electrolyte includes a sodium or cerium salt in the electrolyte with sodium or cerium alloy being deposited on the cathode during the transfer of the metals from the spent fuel. The cathode with the deposit of sodium or cerium alloy is then chanted to an anode and the reverse transfer is carried out on a selective basis with each metal being deposited separately at the cathode. The result is that the sodium or cerium needed for the process is regenerated in the first step and no additional source of these reactants is required.

  2. Electrolytic recovery of reactor metal fuel

    DOEpatents

    Miller, W.E.; Tomczuk, Z.

    1993-02-03

    This invention is comprised of a new electrolytic process and apparatus using sodium, cerium or a similar metal in an alloy or within a sodium beta or beta-alumina sodium ion conductor to electrolytically displace each of the spent fuel metals except for Cesium and strontium on a selective basis from the electrolyte to an inert metal cathode. Each of the metals can be deposited separately. An electrolytic transfer of spent fuel into the electrolyte includes a sodium or cerium salt in the electrolyte with sodium or cerium alloy being deposited on the cathode during the transfer of the metals from the spent fuel. The cathode with the deposit of sodium or cerium alloy is then changed to an anode and the reverse transfer is carried out on a selective basis with each metal being deposited separately at the cathode. The result is that the sodium or cerium needed for the process is regenerated in the first step and no additional source of these reactants is required.

  3. Performance optimization considerations for thermionic fuel elements in a heat pipe cooled thermionic reactor

    NASA Astrophysics Data System (ADS)

    Bellis, Elizabeth A.

    1992-01-01

    A heat pipe-cooled, in-core thermionic (HPTI) reactor design has been proposed in support of the Air Force Thermionic Space Nuclear Power Program. As part of this design, the performance of the power conversion system has been characterized. This paper focuses on the performance optimization studies carried out of a thermionic fuel element (TFE) which will be used in a reactor design capable of producing 40 kWe over a 10 year operating life. The technical approach to the optimization studies closely couples converter lifetime constraints with converter performance to produce the best possible design choice.

  4. Laser shockwave technique for characterization of nuclear fuel plate interfaces

    SciTech Connect

    Perton, M.; Levesque, D.; Monchalin, J.-P.; Lord, M.; Smith, J. A.; Rabin, B. H.

    2013-01-25

    The US National Nuclear Security Agency is tasked with minimizing the worldwide use of high-enriched uranium. One aspect of that effort is the conversion of research reactors to monolithic fuel plates of low-enriched uranium. The manufacturing process includes hot isostatic press bonding of an aluminum cladding to the fuel foil. The Laser Shockwave Technique (LST) is here evaluated for characterizing the interface strength of fuel plates using depleted Uranium/Mo foils. LST is a non-contact method that uses lasers for the generation and detection of large amplitude acoustic waves and is therefore well adapted to the quality assurance of this process. Preliminary results show a clear signature of well-bonded and debonded interfaces and the method is able to classify/rank the bond strength of fuel plates prepared under different HIP conditions.

  5. Laser Shockwave Technique For Characterization Of Nuclear Fuel Plate Interfaces

    SciTech Connect

    James A. Smith; Barry H. Rabin; Mathieu Perton; Daniel Lévesque; Jean-Pierre Monchalin; Martin Lord

    2012-07-01

    The US National Nuclear Security Agency is tasked with minimizing the worldwide use of high-enriched uranium. One aspect of that effort is the conversion of research reactors to monolithic fuel plates of low-enriched uranium. The manufacturing process includes hot isostatic press bonding of an aluminum cladding to the fuel foil. The Laser Shockwave Technique (LST) is here evaluated for characterizing the interface strength of fuel plates using depleted Uranium/Mo foils. LST is a non-contact method that uses lasers for the generation and detection of large amplitude acoustic waves and is therefore well adapted to the quality assurance of this process. Preliminary results show a clear signature of well-bonded and debonded interfaces and the method is able to classify/rank the bond strength of fuel plates prepared under different HIP conditions.

  6. Laser shockwave technique for characterization of nuclear fuel plate interfaces

    NASA Astrophysics Data System (ADS)

    Perton, M.; Lévesque, D.; Monchalin, J.-P.; Lord, M.; Smith, J. A.; Rabin, B. H.

    2013-01-01

    The US National Nuclear Security Agency is tasked with minimizing the worldwide use of high-enriched uranium. One aspect of that effort is the conversion of research reactors to monolithic fuel plates of low-enriched uranium. The manufacturing process includes hot isostatic press bonding of an aluminum cladding to the fuel foil. The Laser Shockwave Technique (LST) is here evaluated for characterizing the interface strength of fuel plates using depleted Uranium/Mo foils. LST is a non-contact method that uses lasers for the generation and detection of large amplitude acoustic waves and is therefore well adapted to the quality assurance of this process. Preliminary results show a clear signature of well-bonded and debonded interfaces and the method is able to classify/rank the bond strength of fuel plates prepared under different HIP conditions.

  7. Synergistic Smart Fuel For In-pile Nuclear Reactor Measurements

    SciTech Connect

    James A. Smith; Dale K. Kotter; Randall A. Ali; Steven L . Garrett

    2013-10-01

    In March 2011, an earthquake of magnitude 9.0 on the Richter scale struck Japan with its epicenter on the northeast coast, near the Tohoku region. In addition to the immense physical destruction and casualties across the country, several nuclear power plants (NPP) were affected. It was the Fukushima Daiichi NPP that experienced the most severe and irreversible damage. The earthquake brought the reactors at Fukushima to an automatic shutdown and because the power transmission lines were damaged, emergency diesel generators (EDGs) were activated to ensure that there was continued cooling of the reactors and spent fuel pools. The situation was being successfully managed until the tsunami hit about forty-five minutes later with a maximum wave height of approximately 15 m. The influx of water submerged the EDGs, the electrical switchgear, and dc batteries, resulting in the total loss of power to the reactors.2 At this point, the situation became critical. There was a loss of the sensors and instrumentation within the reactor that could have provided valuable information to guide the operators to make informed decisions and avoid the unfortunate events that followed. In the light of these events, we have developed and tested a potential self-powered thermoacoustic system, which will have the ability to serve as a temperature sensor and can transmit data independently of electronic networks. Such a device is synergistic with the harsh environment of the nuclear reactor as it utilizes the heat from the nuclear fuel to provide the input power.

  8. Delayed neutron signal characterization in a fast reactor

    SciTech Connect

    Gross, K.C.; Strain, R.V.

    1980-01-01

    Experimental and analytical techniques have been developed for delayed neutron (DN) signal analysis and characterization that can provide diagnostic information to augment data from cover-gas analyses in the detection and identification of breached elements in an LMFBR. Eleven flow reduction tests have been run in EBR-II to provide base data support for predicting DN signal characteristics during exposed fuel operation. Results from the tests demonstrate the feasibility and practicability of response-analysis techniques for determining the transit time, T/sub tr/, for DN emitters traveling from the core to the detector, and the isotopic holdup time, T/sub h/, of DN precursors in the fuel element. T/sub tr/ has been found to vary with the relative grid location of the DN source, and T/sub h/ is affected by the form of fuel exposed to the coolant as well as the condition of the breach site. These parameters are incorporated into a mathematical formulism that enables one to compute for any exposed-fuel test an equivalent recoil area. This concept provides a basis for comparison of different run-beyond-cladding-breach tests in fast reactors.

  9. Fresh Fuel Characterization of U-Mo Alloys

    SciTech Connect

    D. E. Burkes; D. M. Wachs; D. D. Keiser; J. Gan; F. J. Rice; R. Prabhakaran; B. D. Miller; M. A. Okuniewski

    2008-10-01

    The need to provide more accurate property information on U-Mo fuel alloys to reactor operators, modelers, researchers, fabricators, and regulators increases as success of the RERTR program continues. This presentation will provide an overview of fresh fuel U-Mo characterization activities on monolithic fuel occurring at the Idaho National Laboratory. The overview will particularly be focused on properties available through current and previous research, and also on the type of information still needed. The presentation will deal with mechanical, physical, and microstructural properties in terms of both integrated and separate effects. Appropriate discussion in terms of fabrication characteristics, impurities, thermodynamic response, and the effects on the topic areas will be provided, along with a brief background on the characterization techniques being used or being developed to obtain the information. Furthermore, efforts to measure similar characteristics as a function of irradiation conditions and determine end-of-life observations with beginning-of-life behavior will be discussed.

  10. Proposed power upgrade of the Hot Fuel Examination Facility's neutron radiography reactor. [NRAD reactor

    SciTech Connect

    Pruett, D.P.; Richards, W.J.; Heidel, C.C.

    1984-01-01

    The Hot Fuel Examination Facility, HFEF, is one of several facilities located at the Argonne Site. HFEF comprises a large hot cell where both non-destructive and destructive examination of highly-irradiated reactor fuels are conducted in support of the LMFBR program. One of the non-destructive examination techniques utilized at HFEF is neutron radiography. Neutron radiography is provided by the NRAD reactor facility, which is located beneath the HFEF hot cell. The NRAD reactor is a TRIGA reactor and is operated at a steady state power level of 250 kW solely for neutron radiography and the development of radiography techniques. When the NRAD facility was designed and constructed, an operating power level of 250 kW was considered to be adequate for obtaining radiographs of the type of specimens envisaged at that time. A typical radiograph required approximately a twenty-minute exposure time. Specimens were typically single fuel rods placed in an aluminum tray. Since that time, however, several things have occurred that have tended to increase radiography exposure times to as much as 90 minutes each. In order to decrease exposure times, the reactor power level is to be increased from 250 kw to 1 MW. This increase in power will necessitate several engineering and design changes. These changes are described.

  11. Support grid for fuel elements in a nuclear reactor

    DOEpatents

    Finch, Lester M.

    1977-01-01

    A support grid is provided for holding nuclear fuel rods in a rectangular array. Intersecting sheet metal strips are interconnected using opposing slots in the strips to form a rectangular cellular grid structure for engaging the sides of a multiplicity of fuel rods. Spring and dimple supports for engaging fuel and guide rods extending through each cell in the support grid are formed in the metal strips with the springs thus formed being characterized by nonlinear spring rates.

  12. A Parametric Study of the DUPIC Fuel Cycle to Reflect Pressurized Water Reactor Fuel Management Strategy

    SciTech Connect

    Rozon, Daniel; Shen Wei

    2001-05-15

    For both pressurized water reactor (PWR) and Canada deuterium uranium (CANDU) tandem analysis, the Direct Use of spent PWR fuel In CANDU reactor (DUPIC) fuel cycle in a CANDU 6 reactor is studied using the DRAGON/DONJON chain of codes with the ENDF/B-V and ENDF/B-VI libraries. The reference feed material is a 17 x 17 French standard 900-MW(electric) PWR fuel. The PWR spent-fuel composition is obtained from two-dimensional DRAGON assembly transport and depletion calculations. After a number of years of cooling, this defines the initial fuel nuclide field in the CANDU unit cell calculations in DRAGON, where it is further depleted with the same neutron group structure. The resulting macroscopic cross sections are condensed and tabulated to be used in a full-core model of a CANDU 6 reactor to find an optimized channel fueling rate distribution on a time-average basis. Assuming equilibrium refueling conditions and a particular refueling sequence, instantaneous full-core diffusion calculations are finally performed with the DONJON code, from which both the channel power peaking factors and local parameter effects are estimated. A generic study of the DUPIC fuel cycle is carried out using the linear reactivity model for initial enrichments ranging from 3.2 to 4.5 wt% in a PWR. Because of the uneven power histories of the spent PWR assemblies, the spent PWR fuel composition is expected to differ from one assembly to the next. Uneven mixing of the powder during DUPIC fuel fabrication may lead to uncertainties in the composition of the fuel bundle and larger peaking factors in CANDU. A mixing method for reducing composition uncertainties is discussed.

  13. Automated Characterization of Spent Fuel through the Multi-Isotope Process (MIP) Monitor

    SciTech Connect

    Coble, Jamie B.; Orton, Christopher R.; Schwantes, Jon M.

    2012-07-31

    This research developed an algorithm for characterizing spent nuclear fuel (SNF) samples based on simulated gamma spectra. The gamma spectra for a variety of light water reactor fuels typical of those found in the United States were simulated. Fuel nuclide concentrations were simulated in ORIGEN-ARP for 1296 fuel samples with a variety of reactor designs, initial enrichments, burn ups, and cooling times. The results of the ORIGEN-ARP simulation were then input to SYNTH to simulate the gamma spectrum for each sample. These spectra were evaluated with partial least squares (PLS)-based multivariate analysis methods to characterize the fuel according to reactor type (pressurized or boiling water reactor), enrichment, burn up, and cooling time. Characterizing some of the features in series by using previously estimated features in the prediction greatly improves the performance. By first classifying the spent fuel reactor type and then using type-specific models, the prediction error for enrichment, burn up, and cooling time improved by a factor of two to four. For some features, the prediction was further improved by including additional information, such as including the predicted burn up in the estimation of cooling time. The optimal prediction flow was determined based on the simulated data. A PLS discriminate analysis model was developed which perfectly classified SNF reactor type. Burn up was predicted within 0.1% root mean squared percent error (RMSPE) and both cooling time and initial enrichment within approximately 2% RMSPE.

  14. Conceptual design of fuel transfer cask for Reactor TRIGA PUSPATI (RTP)

    NASA Astrophysics Data System (ADS)

    Muhamad, Shalina Sheik; Hamzah, Mohd Arif Arif B.

    2014-02-01

    Spent fuel transfer cask is used to transfer a spent fuel from the reactor tank to the spent fuel storage or for spent fuel inspection. Typically, the cask made from steel cylinders that are either welded or bolted closed. The cylinder is enclosed with additional steel, concrete, or other material to provide radiation shielding and containment of the spent fuel. This paper will discuss the Conceptual Design of fuel transfer cask for Reactor TRIGA Puspati (RTP).

  15. Tools for LWR spent fuel characterization: Assembly classes and fuel designs

    SciTech Connect

    Moore, R.S. Automated Sciences Group, Inc., Oak Ridge, TN ); Notz, K.J. )

    1991-01-01

    The Characteristics Data Base (CDB) is sponsored by the DOE's Office of Civilian Radioactive Waste Management (OCRWM). The CDB provides a single, comprehensive source of data pertaining to radioactive wastes that will or may require geologic disposal, including detailed data describing the physical, quantitative, and radiological characteristics of light-water reactor (LWR) spent fuel. In developing the CDB, tools for the classification of fuel assembly types have been developed. The assembly class scheme is particularly useful for size- and handling-based describes these tools and presents results of their applications in the areas of fuel assembly type identification, characterization of projected discharges, cask accommodation analyses, and defective fuel analyses. Suggestions for additional applications are also made. 7 refs., 1 fig., 2 tabs.

  16. Advanced fuels for plutonium management in pressurized water reactors

    NASA Astrophysics Data System (ADS)

    Vasile, A.; Dufour, Ph; Golfier, H.; Grouiller, J. P.; Guillet, J. L.; Poinot, Ch; Youinou, G.; Zaetta, A.

    2003-06-01

    Several fuel concepts are under investigation at CEA with the aim of manage plutonium inventories in pressurized water reactors. This options range from the use of mature technologies like MOX adapted in the case of MOX-EUS (enriched uranium support) and COmbustible Recyclage A ILot (CORAIL) assemblies to more innovative technologies using IMF like DUPLEX and advanced plutonium assembly (APA). The plutonium burning performances reported to the electrical production go from 7 to 60 kg (TW h) -1. More detailed analysis covering economic, sustainability, reliability and safety aspects and their integration in the whole fuel cycle would allow identifying the best candidate.

  17. Some Tooling for Manufacturing Research Reactor Fuel Plates

    SciTech Connect

    Knight, R.W.

    1999-10-03

    This paper will discuss some of the tooling necessary to manufacture aluminum-based research reactor fuel plates. Most of this tooling is intended for use in a high-production facility. Some of the tools shown have manufactured more than 150,000 pieces. The only maintenance has been sharpening. With careful design, tools can be made to accommodate the manufacture of several different fuel elements, thus, reducing tooling costs and maintaining tools that the operators are trained to use. An important feature is to design the tools using materials with good lasting quality. Good tools can increase return on investment.

  18. Fuel subassembly leak test chamber for a nuclear reactor

    DOEpatents

    Divona, Charles J.

    1978-04-04

    A container with a valve at one end is inserted into a nuclear reactor coolant pool. Once in the pool, the valve is opened by a mechanical linkage. An individual fuel subassembly is lifted into the container by a gripper; the valve is then closed providing an isolated chamber for the subassembly. A vacuum is drawn on the chamber to encourage gaseous fission product leakage through any defects in the cladding of the fuel rods comprising the subassembly; this leakage may be detected by instrumentation, and the need for replacement of the assembly ascertained.

  19. FUEL SUBASSEMBLY CONSTRUCTION FOR RADIAL FLOW IN A NUCLEAR REACTOR

    DOEpatents

    Treshow, M.

    1962-12-25

    An assembly of fuel elements for a boiling water reactor arranged for radial flow of the coolant is described. The ingress for the coolant is through a central header tube, perforated with parallel circumferertial rows of openings each having a lip to direct the coolant flow downward. Around the central tube there are a number of equally spaced concentric trays, closely fitiing the central header tube. Cylindrical fuel elements are placed in a regular pattern around the central tube, piercing the trays. A larger tube encloses the arrangement, with space provided for upward flow of coolart beyond the edge of the trays. (AEC)

  20. Secure automated fabrication: remote fabrication of breeder-reactor fuel

    SciTech Connect

    Gerber, E.W.; Rice, L.H.; Horgos, R.M.; Nagamoto, T.T.; Graham, R.A.

    1981-05-01

    The Secure Automated Fabrication (SAF) Program was initiated at the Hanford Engineering Development Laboratory (HEDL) to develop and demonstrate an advanced manufacturing line (SAF line) for plutonium oxide breeder reactor fuel pins. The SAF line is to be installed in the Fuels and Materials Examination Facility (FMEF) at Hanford and will utilize technology that focuses on improved safety features for plant operating personnel, the public, and the environment. Equipment and process improvements incorporated by the SAF line will yield significant gains in nuclear materials safeguards, product quality and productivity.

  1. Nuclear reactor fuel element with vanadium getter on cladding

    DOEpatents

    Johnson, Carl E.; Carroll, Kenneth G.

    1977-01-01

    A nuclear reactor fuel element is described which has an outer cladding, a central core of fissionable or mixed fissionable and fertile fuel material and a layer of vanadium as an oxygen getter on the inner surface of the cladding. The vanadium reacts with oxygen released by the fissionable material during irradiation of the core to prevent the oxygen from reacting with and corroding the cladding. Also described is a method for coating the inner surface of small diameter tubes of cladding with a layer of vanadium.

  2. FABRICATION OF TUBE TYPE FUEL ELEMENT FOR NUCLEAR REACTORS

    DOEpatents

    Loeb, E.; Nicklas, J.H.

    1959-02-01

    A method of fabricating a nuclear reactor fuel element is given. It consists essentially of fixing two tubes in concentric relationship with respect to one another to provide an annulus therebetween, filling the annulus with a fissionablematerial-containing powder, compacting the powder material within the annulus and closing the ends thereof. The powder material is further compacted by swaging the inner surface of the inner tube to increase its diameter while maintaining the original size of the outer tube. This process results in reduced fabrication costs of powdered fissionable material type fuel elements and a substantial reduction in the peak core temperatures while materially enhancing the heat removal characteristics.

  3. Catalytic reactor for low-Btu fuels

    SciTech Connect

    Smith, Lance; Etemad, Shahrokh; Karim, Hasan; Pfefferle, William C.

    2009-04-21

    An improved catalytic reactor includes a housing having a plate positioned therein defining a first zone and a second zone, and a plurality of conduits fabricated from a heat conducting material and adapted for conducting a fluid therethrough. The conduits are positioned within the housing such that the conduit exterior surfaces and the housing interior surface within the second zone define a first flow path while the conduit interior surfaces define a second flow path through the second zone and not in fluid communication with the first flow path. The conduit exits define a second flow path exit, the conduit exits and the first flow path exit being proximately located and interspersed. The conduits define at least one expanded section that contacts adjacent conduits thereby spacing the conduits within the second zone and forming first flow path exit flow orifices having an aggregate exit area greater than a defined percent of the housing exit plane area. Lastly, at least a portion of the first flow path defines a catalytically active surface.

  4. The design and performance of the research reactor fuel counter

    SciTech Connect

    Abhold, M.E.; Hsue, S.T.; Menlove, H.O.; Walton, G.; Holt, S.

    1996-09-01

    This paper describes the design features, hardware specifications, and performance characteristics of the Research Reactor Fuel Counter (RRFC) System. The system is an active mode neutron coincidence counter intended to assay material test reactor fuel assemblies under water. The RRFC contains 12 {sup 3}He tubes, each with its own preamplifier, and a single ion chamber. The neutron counting electronics are based on the Los Alamos Portable Shift Register (PSR) and the gamma readout is a manual-range pico-ammeter of Los Alamos design. The RRFC is connected to the surface by a 20-m-long cable bundle. The PSR is controlled by a portable IBM computer running a modified version of the Los Alamos neutron coincidence counting code also called RRFC. There is a manual that describes the RRFC software.

  5. Characterization of used nuclear fuel with multivariate analysis for process monitoring

    NASA Astrophysics Data System (ADS)

    Dayman, Kenneth J.; Coble, Jamie B.; Orton, Christopher R.; Schwantes, Jon M.

    2014-01-01

    This paper presents initial development of a reactor-type classifier that is used to select a reactor-specific partial least squares model to predict used nuclear fuel burnup. Nuclide activities for prototypic used fuel samples were generated in ORIGEN-ARP and used to investigate techniques to characterize used nuclear fuel in terms of reactor type (pressurized or boiling water reactor) and burnup. A variety of reactor type classification algorithms, including k-nearest neighbors, linear and quadratic discriminant analyses, and support vector machines, were evaluated to differentiate used fuel from pressurized and boiling water reactors. Then, reactor type-specific partial least squares models were developed to predict the burnup of the fuel. Using these reactor type-specific models instead of a model trained for all light water reactors improved the accuracy of burnup predictions. The developed classification and prediction models were combined and applied to a large dataset that included eight fuel assembly designs, two of which were not used in training the models, and spanned the range of the initial 235U enrichment, cooling time, and burnup values expected of future commercial used fuel for reprocessing. Error rates were consistent across the range of considered enrichment, cooling time, and burnup values. Average absolute relative errors in burnup predictions for validation data both within and outside the training space were 0.0574% and 0.0597%, respectively. The errors seen in this work are artificially low, because the models were trained, optimized, and tested on simulated, noise-free data. However, these results indicate that the developed models may generalize well to new data and that the proposed approach constitutes a viable first step in developing a fuel characterization algorithm based on gamma spectra.

  6. Thermal analysis for fuel handling system for sodium cooled reactor considering minor actinide-bearing metal fuel.

    SciTech Connect

    Chikazawa, Y.; Grandy, C.; Nuclear Engineering Division

    2009-03-01

    The Advanced Burner Reactor (ABR) is one of the components of the Global Nuclear Energy Partnership (GNEP) used to close the fuel cycle. ABR is a sodium-cooled fast reactor that is used to consume transuranic elements resulting from the reprocessing of light water reactor spent nuclear fuel. ABR-1000 [1000 MW(thermal)] is a fast reactor concept created at Argonne National Laboratory to be used as a reference concept for various future trade-offs. ABR-1000 meets the GNEP goals although it uses what is considered base sodium fast reactor technology for its systems and components. One of the considerations of any fast reactor plant concept is the ability to perform fuel-handling operations with new and spent fast reactor fuel. The transmutation fuel proposed as the ABR fuel has a very little experience base, and thus, this paper investigates a fuel-handling concept and potential issues of handling fast reactor fuel containing minor actinides. In this study, two thermal analyses supporting a conceptual design study on the ABR-1000 fuel-handling system were carried out. One analysis investigated passive dry spent fuel storage, and the other analysis investigated a fresh fuel shipping cask. Passive dry storage can be made suitable for the ABR-1000 spent fuel storage with sodium-bonded metal fuel. The thermal analysis shows that spent fast reactor fuel with a decay heat of 2 kW or less can be stored passively in a helium atmosphere. The 2-kW value seems to be a reasonable and practical level, and a combination of reasonably-sized in-sodium storage followed by passive dry storage could be a candidate for spent fuel storage for the next-generation sodium-cooled reactor with sodium-bonded metal fuel. Requirements for the shipping casks for minor actinide-bearing fuel with a high decay heat level are also discussed in this paper. The shipping cask for fresh sodium-cooled-reactor fuel should be a dry type to reduce the reaction between residual moisture on fresh fuel and the

  7. Detachable connection for a nuclear reactor fuel assembly

    DOEpatents

    Christiansen, D.W.; Karnesky, R.A.

    1983-08-29

    A locking connection for releasably attaching a handling socket to the duct tube of a fuel assembly for a nuclear reactor. The connection comprises a load pad housing mechanically attached to the duct tube and a handling socket threadably secured within the housing. A retaining ring is interposed between the housing and the handling socket and is formed with a projection and depression engagable within a cavity and groove of the housing and handling socket, respectively, to form a detachable interlocked connection assembly.

  8. Detachable connection for a nuclear reactor fuel assembly

    DOEpatents

    Christiansen, David W.; Karnesky, Richard A.

    1986-01-01

    A locking connection for releasably attaching a handling socket to the duct tube of a fuel assembly for a nuclear reactor. The connection comprises a load pad housing mechanically attached to the duct tube and a handling socket threadably secured within the housing. A retaining ring is interposed between the housing and the handling socket and is formed with a projection and depression engageable within a cavity and groove of the housing and handling socket, respectively, to form a detachable interlocked connection assembly.

  9. Development potential for thermal reactors and their fuel cycles

    SciTech Connect

    Dodds, H.L.; Gat, U.

    1997-08-01

    The advantages of molten salt reactors (MSRs) for power production are very briefly described in this paper. The MSRs considered are those with on-line fuel processing, external cooling, and fluoride salt separation. Characteristics noted include lack of meltdown potential, small radioactive source terms, and complete burnup of fissile material. The burnup capability of MSRs would allow them to be used to dispose of plutonium while producing energy. 8 refs.

  10. FUEL ELEMENTS FOR NUCLEAR REACTORS AND PROCESS OF MAKING

    DOEpatents

    Roake, W.E.

    1958-08-19

    A process is described for producing uranium metal granules for use in reactor fuel elements. The granules are made by suspending powdered uramiunn metal or uranium hydride in a viscous, non-reactive liquid, such as paraffin oil, aad pouring the resulting suspension in droplet, on to a bed of powdered absorbent. In this manner the liquid vehicle is taken up by the sorbent and spherical pellets of uranium metal are obtained. The

  11. Reactor Fuel Isotopics and Code Validation for Nuclear Applications

    SciTech Connect

    Francis, Matthew W.; Weber, Charles F.; Pigni, Marco T.; Gauld, Ian C.

    2015-02-01

    Experimentally measured isotopic concentrations of well characterized spent nuclear fuel (SNF) samples have been collected and analyzed by previous researchers. These sets of experimental data have been used extensively to validate the accuracy of depletion code predictions for given sets of burnups, initial enrichments, and varying power histories for different reactor types. The purpose of this report is to present the diversity of data in a concise manner and summarize the current accuracy of depletion modeling. All calculations performed for this report were done using the Oak Ridge Isotope GENeration (ORIGEN) code, an internationally used irradiation and decay code solver within the SCALE comprehensive modeling and simulation code. The diversity of data given in this report includes key actinides, stable fission products, and radioactive fission products. In general, when using the current ENDF/B-VII.0 nuclear data libraries in SCALE, the major actinides are predicted to within 5% of the measured values. Large improvements were seen for several of the curium isotopes when using improved cross section data found in evaluated nuclear data file ENDF/B-VII.0 as compared to ENDF/B-V-based results. The impact of the flux spectrum on the plutonium isotope concentrations as a function of burnup was also shown. The general accuracy noted for the actinide samples for reactor types with burnups greater than 5,000 MWd/MTU was not observed for the low-burnup Hanford B samples. More work is needed in understanding these large discrepancies. The stable neodymium and samarium isotopes were predicted to within a few percent of the measured values. Large improvements were seen in prediction for a few of the samarium isotopes when using the ENDF/B-VII.0 libraries compared to results obtained with ENDF/B-V libraries. Very accurate predictions were obtained for 133Cs and 153Eu. However, the predicted values for the stable ruthenium and rhodium isotopes varied

  12. Fuel Breeding and Core Behavior Analyses on In Core Fuel Management of Water Cooled Thorium Reactors

    NASA Astrophysics Data System (ADS)

    Permana, Sidik; Sekimoto, Hiroshi; Waris, Abdul; Subhki, Muhamad Nurul; Ismail

    2010-12-01

    Thorium fuel cycle with recycled U-233 has been widely recognized having some contributions to improve the water-cooled breeder reactor program which has been shown by a feasible area of breeding and negative void reactivity which confirms that fissile of 233U contributes to better fuel breeding and effective for obtaining negative void reactivity coefficient as the main fissile material. The present study has the objective to estimate the effect of whole core configuration as well as burnup effects to the reactor core profile by adopting two dimensional model of fuel core management. About more than 40 months of cycle period has been employed for one cycle fuel irradiation of three batches fuel system for large water cooled thorium reactors. All position of fuel arrangement contributes to the total core conversion ratio which gives conversion ratio less than unity of at the BOC and it contributes to higher than unity (1.01) at the EOC after some irradiation process. Inner part and central part give the important part of breeding contribution with increasing burnup process, while criticality is reduced with increasing the irradiation time. Feasibility of breeding capability of water-cooled thorium reactors for whole core fuel arrangement has confirmed from the obtained conversion ratio which shows higher than unity. Whole core analysis on evaluating reactivity change which is caused by the change of voided condition has been employed for conservative assumption that 100% coolant and moderator are voided. It obtained always a negative void reactivity coefficient during reactor operation which shows relatively more negative void coefficient at BOC (fresh fuel composition), and it becomes less negative void coefficient with increasing the operation time. Negative value of void reactivity coefficient shows the reactor has good safety properties in relation to the reactivity profile which is the main parameter in term of criticality safety analysis. Therefore, this

  13. Fuel Breeding and Core Behavior Analyses on In Core Fuel Management of Water Cooled Thorium Reactors

    SciTech Connect

    Permana, Sidik; Sekimoto, Hiroshi; Waris, Abdul; Subhki, Muhamad Nurul; Ismail,

    2010-12-23

    Thorium fuel cycle with recycled U-233 has been widely recognized having some contributions to improve the water-cooled breeder reactor program which has been shown by a feasible area of breeding and negative void reactivity which confirms that fissile of 233U contributes to better fuel breeding and effective for obtaining negative void reactivity coefficient as the main fissile material. The present study has the objective to estimate the effect of whole core configuration as well as burnup effects to the reactor core profile by adopting two dimensional model of fuel core management. About more than 40 months of cycle period has been employed for one cycle fuel irradiation of three batches fuel system for large water cooled thorium reactors. All position of fuel arrangement contributes to the total core conversion ratio which gives conversion ratio less than unity of at the BOC and it contributes to higher than unity (1.01) at the EOC after some irradiation process. Inner part and central part give the important part of breeding contribution with increasing burnup process, while criticality is reduced with increasing the irradiation time. Feasibility of breeding capability of water-cooled thorium reactors for whole core fuel arrangement has confirmed from the obtained conversion ratio which shows higher than unity. Whole core analysis on evaluating reactivity change which is caused by the change of voided condition has been employed for conservative assumption that 100% coolant and moderator are voided. It obtained always a negative void reactivity coefficient during reactor operation which shows relatively more negative void coefficient at BOC (fresh fuel composition), and it becomes less negative void coefficient with increasing the operation time. Negative value of void reactivity coefficient shows the reactor has good safety properties in relation to the reactivity profile which is the main parameter in term of criticality safety analysis. Therefore, this

  14. Fuel Summary Report: Shippingport Light Water Breeder Reactor

    SciTech Connect

    Illum, D.B.; Olson, G.L.; McCardell, R.K.

    1999-01-01

    The Shippingport Light Water Breeder Reactor (LWBR) was a small water cooled, U-233/Th-232 cycle breeder reactor developed by the Pittsburgh Naval Reactors to improve utilization of the nation's nuclear fuel resources in light water reactors. The LWBR was operated at Shippingport Atomic Power Station (APS), which was a Department of Energy (DOE) (formerly Atomic Energy Commission)-owned reactor plant. Shippingport APS was the first large-scale, central-station nuclear power plant in the United States and the first plant of such size in the world operated solely to produce electric power. The Shippingport LWBR was operated successfully from 1977 to 1982 at the APS. During the five years of operation, the LWBR generated more than 29,000 effective full power hours (EFPH) of energy. After final shutdown, the 39 core modules of the LWBR were shipped to the Expended Core Facility (ECF) at Naval Reactors Facility at the Idaho National Engineering and Environmental Laboratory (INEEL). At ECF, 12 of the 39 modules were dismantled and about 1000 of more than 17,000 rods were removed from the modules of proof-of-breeding and fuel performance testing. Some of the removed rods were kept at ECF, some were sent to Argonne National Laboratory-West (ANL-W) in Idaho and some to ANL-East in Chicago for a variety of physical, chemical and radiological examinations. All rods and rod sections remaining after the experiments were shipped back to ECF, where modules and loose rods were repackaged in liners for dry storage. In a series of shipments, the liners were transported from ECF to Idaho Nuclear Technology Engineering Center (INTEC), formerly the Idaho Chemical Processing Plant (ICPP). The 47 liners containing the fully-rodded and partially-derodded core modules, the loose rods, and the rod scraps, are now stored in underground dry wells at CPP-749.

  15. Development of fast breeder reactor fuel reprocessing technology at the Power Reactor and Nuclear Fuel Development Corporation

    SciTech Connect

    Kawata, T.; Takeda, H.; Togashi, A.; Hayashi, S. . Tokai Works); Stradley, J.G. )

    1991-01-01

    For the past two decades, a broad range of research development (R D) programs to establish fast breeder reactor (FBR) system and its associated fuel cycle technology have been pursued by the Power Reactor and Nuclear Fuel Development Corporation (PNC). Developmental activities for FBR fuel reprocessing technology have been primarily conducted at PNC Tokai Works where many important R D facilities for nuclear fuel cycle are located. These include cold and uranium tests for process equipment development in the Engineering Demonstration Facilities (EDF)-I and II, and laboratory-scale hot tests in the Chemical Processing Facility (CPF) where fuel dissolution and solvent extraction characteristics are being investigated with irradiated FBR fuel pins whose burn-up ranges up to 100,000 MWd/t. An extensive effort has also been made at EDF-III to develop advanced remote technology which enables to increase plant availability and to decrease radiation exposures to the workers in future reprocessing plants. The PNC and the United States Department of Energy (USDOE) entered into the joint collaboration in which the US shares the R Ds to support FBR fuel reprocessing program at the PNC. Several important R Ds on advanced process equipment such as a rotary dissolver and a centrifugal contactor system are in progress in a joint effort with the Oak Ridge National Laboratory (ORNL) Consolidated Fuel Reprocessing Program (CFRP). In order to facilitate hot testing on advanced processes and equipment, the design of a new engineering-scale hot test facility is now in progress aiming at the start of hot operation in late 90's. 31 refs., 2 tabs.

  16. Reprocessing of research reactor fuel the Dounreay option

    SciTech Connect

    Cartwright, P.

    1997-08-01

    Reprocessing is a proven process for the treatment of spent U/Al Research Reactor fuel. At Dounreay 12679 elements have been reprocessed during the past 30 years. For reactors converting to LEU fuel the uranium recovered in reprocessing can be blended down to less than 20% U{sub 235}, enrichment and be fabricated into new elements. For reactors already converted to LEU it is technically possible to reprocess spent silicide fuel to reduce the U{sub 235} burden and present to a repository only stable conditioned waste. The main waste stream from reprocessing which contains the Fission products is collected in underground storage tanks where it is kept for a period of at least five years before being converted to a stable solid form for return to the country of origin for subsequent storage/disposal. Discharges to the environment from reprocessing are low and are limited to the radioactive gases contained in the spent fuel and a low level liquid waste steam. Both of these discharges are independently monitored, and controlled within strict discharge limits set by the UK Government`s Scottish Office. Transportation of spent fuel to Dounreay has been undertaken using many routes from mainland Europe and has utilised over the past few years both chartered and scheduled vessel services. Several different transport containers have been handled and are currently licensed in the UK. This paper provides a short history of MTR reprocessing at Dounreay, and provides information to show reprocessing can satisfy the needs of MTR operators, showing that reprocessing is a valuable asset in non-proliferation terms, offers a complete solution and is environmentally acceptable.

  17. Determining reactor flux from xenon-136 and cesium-135 in spent fuel

    NASA Astrophysics Data System (ADS)

    Hayes, A. C.; Jungman, Gerard

    2012-10-01

    The ability to infer reactor flux from spent fuel or seized fissile material would enhance the tools of nuclear forensics and nuclear nonproliferation significantly. We show that reactor flux can be inferred from the ratios of xenon-136 to xenon-134 and cesium-135 to cesium-137. If the average flux of a reactor is known, the flux inferred from measurements of spent fuel could help determine whether that spent fuel was loaded as a blanket or close to the mid-plane of the reactor. The cesium ratio also provides information on reactor shutdowns during the irradiation of fuel, which could prove valuable for identifying the reactor in question through comparisons with satellite reactor heat monitoring data. We derive analytic expressions for these correlations and compare them to experimental data and to detailed reactor burn simulations. The enrichment of the original uranium fuel affects the correlations by up to 3%, but only at high flux.

  18. Decommissioning of nuclear reactor fuel channels using laser technology

    NASA Astrophysics Data System (ADS)

    Panchenko, Vladislav Y.; Zabelin, Alexandre M.; Slepokon, Yu. I.; Ryahin, V. M.; Kuznetsov, P. P.; Panasyuk, V. F.; Korotchenko, A. V.; Kislov, V. S.; Loktev, S. V.

    2000-07-01

    Decommissioning of nuclear reactors using laser remote dismounting and welding was experimentally proved at a nuclear reactor of Kursk Nuclear Power Plant. The main reason of laser beam application in this case is the marked decrease of radioactive exposure of the service personnel. The use of a high-power laser beam provided for laser cutting and welding processes realization at a distance up to 35 m between the laser and the workstation placed behind a radiation shield. By application of laser cutting gas and dust contamination is ten-fold decreased. Some results of decommissioning application of a stationary laser workstation based upon a 5 kW fast-transverse-flow discharge CW CO2 laser TL-5M installed at a nuclear reactor site are presented. A special high-beam- quality model of the laser was developed to satisfy the needs of decommissioning. Laser cutting process was applied to decommissioning of fuel channels (FC) of RBMK-1000 reactor, after their extractor from the reactor active zone during the procedure of channels replacement.

  19. Pressurized water reactor fuel crud and corrosion modeling

    NASA Astrophysics Data System (ADS)

    Deshon, Jeff; Hussey, Dennis; Kendrick, Brian; McGurk, John; Secker, Jeff; Short, Michael

    2011-08-01

    Pressurized water reactors circulate high-temperature water that slowly corrodes Inconel and stainless steel system surfaces, and the nickel/iron based corrosion products deposit in regions of the fuel where sub-cooled nucleate boiling occurs. The deposited corrosion products, called `crud', can have an adverse impact on fuel performance. Boron can concentrate within the crud in the boiling regions of the fuel leading to a phenomenon known as axial offset anomaly (AOA). In rare cases, fuel clad integrity can be compromised because of crud-induced localized corrosion (CILC) of the zirconium-based alloy. Westinghouse and the Electric Power Research Institute have committed to understanding the crud transport process and develop a risk assessment software tool called boron-induced offset anomaly (BOA) to avoid AOA and CILC. This paper reviews the history of the BOA model development and new efforts to develop a micro-scale model called MAMBA for use in the Consortium for Advanced Light Water Reactor Simulation (CASL) program.

  20. Fuel qualification plan for the Advanced Neutron Source Reactor

    SciTech Connect

    Copeland, G.L.

    1995-07-01

    This report describes the development and qualification plan for the fuel for the Advanced Neutron Source. The reference fuel is U{sub 3}Si{sub 2}, dispersed in aluminum and clad in 6061 aluminum. This report was prepared in May 1994, at which time the reference design was for a two-element core containing highly enriched uranium (93% {sup 235}U) . The reactor was in the process of being redesigned to accommodate lowered uranium enrichment and became a three-element core containing a higher volume fraction of uranium enriched to 50% {sup 235}U. Consequently, this report was not issued at that time and would have been revised to reflect the possibly different requirements of the lower-enrichment, higher-volume fraction fuel. Because the reactor is now being canceled, this unrevised report is being issued for archival purposes. The report describes the fabrication and inspection development plan, the irradiation tests and performance modeling to qualify performance, the transient testing that is part of the safety program, and the interactions and interfaces of the fuel development with other tasks.

  1. Breeding nuclear fuels with accelerators: replacement for breeder reactors

    SciTech Connect

    Grand, P.; Takahashi, H.

    1984-01-01

    One application of high energy particle accelerators has been, and still is, the production of nuclear fuel for the nuclear energy industry; tantalizing because it would create a whole new industry. This approach to producing fissile from fertile material was first considered in the early 1950's in the context of the nuclear weapons program. A considerable development effort was expended before discovery of uranium ore in New Mexico put an end to the project. Later, US commitment to the Liquid Metal Fast Breeder Reactors (LMFBR) killed any further interest in pursuing accelerator breeder technology. Interest in the application of accelerators to breed nuclear fuels, and possibly burn nuclear wastes, revived in the late 1970's, when the LMFBR came under attack during the Carter administration. This period gave the opportunity to revisit the concept in view of the present state of the technology. This evaluation and the extensive calculational modeling of target designs that have been carried out are promising. In fact, a nuclear fuel cycle of Light Water Reactors and Accelerator Breeders is competitive to that of the LMFBR. At this time, however, the relative abundance of uranium reserves vs electricity demand and projected growth rate render this study purely academic. It will be for the next generation of accelerator builders to demonstate the competitiveness of this technology versus that of other nuclear fuel cycles, such as LMFBR's or Fusion Hybrid systems. 22 references, 1 figure, 5 tables.

  2. Expert system for identification of simultaneous and sequential reactor fuel failures with gas tagging

    DOEpatents

    Gross, Kenny C.

    1994-01-01

    Failure of a fuel element in a nuclear reactor core is determined by a gas tagging failure detection system and method. Failures are catalogued and characterized after the event so that samples of the reactor's cover gas are taken at regular intervals and analyzed by mass spectroscopy. Employing a first set of systematic heuristic rules which are applied in a transformed node space allows the number of node combinations which must be processed within a barycentric algorithm to be substantially reduced. A second set of heuristic rules treats the tag nodes of the most recent one or two leakers as "background" gases, further reducing the number of trial node combinations. Lastly, a "fuzzy" set theory formalism minimizes experimental uncertainties in the identification of the most likely volumes of tag gases. This approach allows for the identification of virtually any number of sequential leaks and up to five simultaneous gas leaks from fuel elements.

  3. Expert system for identification of simultaneous and sequential reactor fuel failures with gas tagging

    DOEpatents

    Gross, K.C.

    1994-07-26

    Failure of a fuel element in a nuclear reactor core is determined by a gas tagging failure detection system and method. Failures are catalogued and characterized after the event so that samples of the reactor's cover gas are taken at regular intervals and analyzed by mass spectroscopy. Employing a first set of systematic heuristic rules which are applied in a transformed node space allows the number of node combinations which must be processed within a barycentric algorithm to be substantially reduced. A second set of heuristic rules treats the tag nodes of the most recent one or two leakers as background'' gases, further reducing the number of trial node combinations. Lastly, a fuzzy'' set theory formalism minimizes experimental uncertainties in the identification of the most likely volumes of tag gases. This approach allows for the identification of virtually any number of sequential leaks and up to five simultaneous gas leaks from fuel elements. 14 figs.

  4. TEM CHARACTERIZATION OF IRRADIATED U3SI2/AL DISPERSION FUEL

    SciTech Connect

    J. Gan; B. Miller; D. Keiser; A. Robinson; P. Medvedev; D. Wachs

    2010-10-01

    The silicide dispersion fuel of U3Si2/Al has been recognized as a reasonably good performance fuel for nuclear research and test reactors except that it requires the use of high enrichment uranium. An irradiated U3Si2/Al dispersion fuel (~75% enrichment) from the high flux side of a RERTR-8 (U0R040) plate was characterized using transmission electron microscopy (TEM). The fuel plate was irradiated in the advanced test reactor (ATR) for 105 days. The average irradiation temperature and fission density of the fuel particles for the TEM sample are estimated to be approximately ~110 degrees C and 5.4 x 10-21 f/cm3. The characterization was performed using a 200KV TEM with a LaB6 filament. Detailed microstructural information along with composition analysis is obtained. The results and their implication on the performance of this silicide fuel are discussed.

  5. Distribution of characteristics of LWR [light water reactor] spent fuel

    SciTech Connect

    Reich, W.J.; Notz, K.J.; Moore, R.S.

    1991-01-01

    The purpose of this report is to develop a collective description of the entire spent fuel inventory in terms of various fuel properties relevant to Approved Testing Materials (ATMs) using information available from the Characteristics Data Base (CBD), which is sponsored by the US Department of Energy`s (DOE`s) Office of Civilian Radioactive Waste Management. A number of light-water reactor (LWR) characteristics were analyzed including assembly class representation, fuel burnup, enrichment, fuel fabrication data, defective fuel quantities, and, at PNL`s specific request, linear heat generation rate (LHGR) and the utilization of burnable poisons. A quantitative relationships was developed between burnup and enrichment for BWRs and PWRs. The relationship shows that the existing BWR ATM is near the center of the burnup-enrichment distribution, while the four PWR ATMs bracket the center of the burnup range but are on the low side of the enrichment range. Fuel fabrication data are based on vendor specifications for new fuel. Defective fuel distributions were analyzed in terms of assembly class and vendor design. LHGR values were calculated from utility data on burnup and effective full-power days; these calculations incorporate some unavoidable assumptions which may compromise the value of the results. Only a limited amount of data are available on burnable poisons at this time. Based on this distribution study, suggestions for additional ATMs are made. These are based on the class and design concepts and include BWR/2,3 barrier fuel, and the WE 17 {times} 17 class with integral burnable poison. Both should be at relatively high burnups. 16 refs., 5 figs., 15 tabs.

  6. Multi-cycle boiling water reactor fuel cycle optimization

    SciTech Connect

    Ottinger, K.; Maldonado, G.I.

    2013-07-01

    In this work a new computer code, BWROPT (Boiling Water Reactor Optimization), is presented. BWROPT uses the Parallel Simulated Annealing (PSA) algorithm to solve the out-of-core optimization problem coupled with an in-core optimization that determines the optimum fuel loading pattern. However it uses a Haling power profile for the depletion instead of optimizing the operating strategy. The result of this optimization is the optimum new fuel inventory and the core loading pattern for the first cycle considered in the optimization. Several changes were made to the optimization algorithm with respect to other nuclear fuel cycle optimization codes that use PSA. Instead of using constant sampling probabilities for the solution perturbation types throughout the optimization as is usually done in PSA optimizations the sampling probabilities are varied to get a better solution and/or decrease runtime. The new fuel types available for use can be sorted into an array based on any number of parameters so that each parameter can be incremented or decremented, which allows for more precise fuel type selection compared to random sampling. Also, the results are sorted by the new fuel inventory of the first cycle for ease of comparing alternative solutions. (authors)

  7. A fuel for sub-critical fast reactor

    SciTech Connect

    Moiseenko, V. E.; Chernitskiy, S. V.; Agren, O.; Noack, K.

    2012-06-19

    Along with the problem of the nuclear waste transmutation, the problem of minimization of waste production is of current interest. It is not possible to eliminate production of waste at a nuclear power plant, but, as is shown in this report, it is in principle possible to arrange a fuel composition with no net production of transuranic elements. The idea is to find the transuranic elements composition to which the depleted uranium is continuously supplied during frequent reprocessing, and amount of each other transuranic fuel component remains unchanged in time. For each transuranic component, the balance is achieved by equating burnup and production rates. The production is due to neutron capture by the neighboring lighter isotope and subsequent beta-decay. The burnup includes fission, neutron capture and decays. For the calculations a simplified burnup model which accounts for 9 isotopes of uranium, neptunium, plutonium and americium is used. The calculated fuel composition consists mainly of uranium with minority of plutonium isotopes. Such a fuel, after usage in a sub-critical fast reactor, should be reprocessed. The fission product content increases during burnup, representing a net production of waste, while the transuranic elements and {sup 238}U should be recycled into a new fuel. For such a fuel cycle, the net consumption is only for 238U, and the net waste production is just fission products.

  8. A fuel for sub-critical fast reactor

    NASA Astrophysics Data System (ADS)

    Moiseenko, V. E.; Chernitskiy, S. V.; Ågren, O.; Noack, K.

    2012-06-01

    Along with the problem of the nuclear waste transmutation, the problem of minimization of waste production is of current interest. It is not possible to eliminate production of waste at a nuclear power plant, but, as is shown in this report, it is in principle possible to arrange a fuel composition with no net production of transuranic elements. The idea is to find the transuranic elements composition to which the depleted uranium is continuously supplied during frequent reprocessing, and amount of each other transuranic fuel component remains unchanged in time. For each transuranic component, the balance is achieved by equating burnup and production rates. The production is due to neutron capture by the neighboring lighter isotope and subsequent beta-decay. The burnup includes fission, neutron capture and decays. For the calculations a simplified burnup model which accounts for 9 isotopes of uranium, neptunium, plutonium and americium is used. The calculated fuel composition consists mainly of uranium with minority of plutonium isotopes. Such a fuel, after usage in a sub-critical fast reactor, should be reprocessed. The fission product content increases during burnup, representing a net production of waste, while the transuranic elements and 238U should be recycled into a new fuel. For such a fuel cycle, the net consumption is only for 238U, and the net waste production is just fission products.

  9. Development of a Monolithic Research Reactor Fuel Type at Argonne National Laboratory

    SciTech Connect

    Clark, C.R.; Briggs, R.J.

    2004-10-06

    The Reduced Enrichment for Research and Test Reactors (RERTR) program has been tasked with the conversion of research reactors from highly enriched to low-enriched uranium (LEU). To convert several high power reactors, monolithic fuel, a new fuel type, is being developed. This fuel type replaces the standard fuel dispersion with a fuel alloy foil, which allows for fuel densities far in excess of that found in dispersion fuel. The single-piece fuel foil also contains a significantly lower interface area between the fuel and the aluminum in the plate than the standard fuel type, limiting the amount of detrimental fuel-aluminum interaction that can occur. Implementation of monolithic fuel is dependant on the development of a suitable fabrication method as traditional roll-bonding techniques are inadequate.

  10. Computational fluid dynamic simulations of chemical looping fuel reactors utilizing gaseous fuels

    SciTech Connect

    Mahalatkar, K.; Kuhlman, J.; Huckaby, E.D.; O'Brien, T.

    2011-01-01

    A computational fluid dynamic(CFD) model for the fuel reactor of chemical looping combustion technology has been developed,withspecialfocusonaccuratelyrepresentingtheheterogeneous chemicalreactions.Acontinuumtwo-fluidmodelwasusedtodescribeboththegasandsolidphases. Detailedsub-modelstoaccountforfluid–particleandparticle–particleinteractionforceswerealso incorporated.Twoexperimentalcaseswereanalyzedinthisstudy(Son andKim,2006; Mattisonetal., 2001). SimulationswerecarriedouttotestthecapabilityoftheCFDmodeltocapturechangesinoutletgas concentrationswithchangesinnumberofparameterssuchassuperficialvelocity,metaloxide concentration,reactortemperature,etc.Fortheexperimentsof Mattissonetal.(2001), detailedtime varyingoutletconcentrationvalueswerecompared,anditwasfoundthatCFDsimulationsprovideda reasonablematchwiththisdata.

  11. Deep Burn: Development of Transuranic Fuel for High-Temperature Helium-Cooled Reactors- Monthly Highlights September 2010

    SciTech Connect

    Snead, Lance Lewis; Besmann, Theodore M; Collins, Emory D; Bell, Gary L

    2010-10-01

    The DB Program monthly highlights report for August 2010, ORNL/TM-2010/184, was distributed to program participants by email on September 17. This report discusses: (1) Core and Fuel Analysis - (a) Core Design Optimization in the HTR (high temperature helium-cooled reactor) Prismatic Design (Logos), (b) Core Design Optimization in the HTR Pebble Bed Design (INL), (c) Microfuel analysis for the DB HTR (INL, GA, Logos); (2) Spent Fuel Management - (a) TRISO (tri-structural isotropic) repository behavior (UNLV), (b) Repository performance of TRISO fuel (UCB); (3) Fuel Cycle Integration of the HTR (high temperature helium-cooled reactor) - Synergy with other reactor fuel cycles (GA, Logos); (4) TRU (transuranic elements) HTR Fuel Qualification - (a) Thermochemical Modeling, (b) Actinide and Fission Product Transport, (c) Radiation Damage and Properties; (5) HTR Spent Fuel Recycle - (a) TRU Kernel Development (ORNL), (b) Coating Development (ORNL), (c) Characterization Development and Support, (d) ZrC Properties and Handbook; and (6) HTR Fuel Recycle - (a) Graphite Recycle (ORNL), (b) Aqueous Reprocessing, (c) Pyrochemical Reprocessing METROX (metal recovery from oxide fuel) Process Development (ANL).

  12. Nuclear breeder reactor fuel element with axial tandem stacking and getter

    DOEpatents

    Gibby, Ronald L.; Lawrence, Leo A.; Woodley, Robert E.; Wilson, Charles N.; Weber, Edward T.; Johnson, Carl E.

    1981-01-01

    A breeder reactor fuel element having a tandem arrangement of fissile and fertile fuel with a getter for fission product cesium disposed between the fissile and fertile sections. The getter is effective at reactor operating temperatures to isolate the cesium generated by the fissile material from reacting with the fertile fuel section.

  13. Development of a Robust Tri-Carbide Fueled Reactor for Multimegawatt Space Power and Propulsion Applications

    SciTech Connect

    Samim Anghaie; Travis W. Knight; Johann Plancher; Reza Gouw

    2004-08-11

    An innovative reactor core design based on advanced, mixed carbide fuels was analyzed for nuclear space power applications. Solid solution, mixed carbide fuels such as (U,Zr,Nb)c and (U,Zr, Ta)C offer great promise as an advanced high temperature fuel for space power reactors.

  14. Consolidated fuel reprocessing program: Criticality experiments with fast test reactor fuel pins in an organic moderator

    SciTech Connect

    Bierman, S.R.

    1986-12-01

    The results obtained in a series of criticality experiments performed as part of a joint program on criticality data development between the United States Department of Energy and the Power Reactor and Nuclear Fuel Development Corporation of Japan are presented in this report along with a complete description of the experiments. The experiments involved lattices of Fast Test Reactor (FTR) fuel pins in an organic moderator mixture similar to that used in the solvent extraction stage of fuel reprocessing. The experiments are designed to provide data for direct comparison with previously performed experimental measurements with water moderated lattices of FTR fuel pins. The same lattice arrangements and FTR fuel pin types are used in these organic moderated experimental assemblies as were used in the water moderated experiments. The organic moderator is a mixture of 38 wt % tributylphosphate in a normal paraffin hydrocarbon mixture of C{sub 11}H{sub 24} to C{sub 15}H{sub 32} molecules. Critical sizes of 1054.8, 599.2, 301.8, 199.5 and 165.3 fuel pins were obtained respectively for organic moderated lattices having 0.761 cm, 0.968 cm, 1.242 cm, 1.537 cm and 1.935 cm square lattice pitches as compared to 1046.9, 571.9, 293.9, 199.7 and 165.1 fuel pins for the same lattices water moderated.

  15. Physics Characterization of a Heterogeneous Sodium Fast Reactor Transmutation System

    SciTech Connect

    Samuel E. Bays

    2007-09-01

    The threshold-fission (fertile) nature of Am-241 is used to destroy this minor actinide by capitalizing upon neutron capture instead of fission within a sodium fast reactor. This neutron-capture and its subsequent decay chain leads to the breeding of even mass number plutonium isotopes. A slightly moderated target design is proposed for breeding plutonium in an axial blanket located above the active “fast reactor” driver fuel region. A parametric study on the core height and fuel pin diameter-to-pitch ratio is used to explore the reactor and fuel cycle aspects of this design. This study resulted in both a non-flattened and a pancake core geometry. Both of these designs demonstrated a high capacity for removing americium from the fuel cycle. A reactivity coefficient analysis revealed that this heterogeneous design will have comparable safety aspects to a homogeneous reactor of the same size.

  16. Depletion analysis of mixed-oxide fuel pins in light water reactors and the Advanced Test Reactor

    SciTech Connect

    Chang, G.S.; Ryskamp, J.M.

    2000-03-01

    An experiment containing weapons-grade mixed-oxide (WG-MOX) fuel has been designed and is being irradiated in the Advanced Test Reactor (ATR) at the Idaho National Engineering and Environmental Laboratory (INEEL). The ability to accurately predict fuel pin performance is an essential requirement for the MOX fuel test assembly design. Detailed radial fission power and temperature profile effects and fission gas release in the fuel pin are a function of the fuel pin's temperature, fission power, and fission product ad actinide concentration profiles. In addition, the burnup-dependent profile analyses in irradiated fuel pins is important for fuel performance analysis to support the potential licensing of the MOX fuel made from WG-plutonium and depleted uranium for use in US reactors. The MCNP Coupling With ORIGEN2 burnup calculation code (MCWO) can analyze the detailed burnup profiles of WG-MOX and reactor-grade mixed-oxide (RG-MOX) fuel pins. The validated code MCWO can provide the best-estimate neutronic characteristics of fuel burnup performance analysis. Applying this capability with a new minicell method allows calculation of detailed nuclide concentration and power distributions within the MOX pins as a function of burnup. This methodology was applied to MOX fuel in a commercial pressurized water reactor and in an experiment currently being irradiated in the ATR. The prediction of nuclide concentration profiles and power distributions in irradiated MOX plus via this new methodology can provide insights into MOX fuel performance.

  17. The IAEA international conference on fast reactors and related fuel cycles: highlights and main outcomes

    SciTech Connect

    Monti, S.; Toti, A.

    2013-07-01

    The 'International Conference on Fast Reactors and Related Fuel Cycles', which is regularly held every four years, represents the main international event dealing with fast reactors technology and related fuel cycles options. Main topics of the conference were new fast reactor concepts, design and simulation capabilities, safety of fast reactors, fast reactor fuels and innovative fuel cycles, analysis of past experience, fast reactor knowledge management. Particular emphasis was put on safety aspects, considering the current need of developing and harmonizing safety standards for fast reactors at the international level, taking also into account the lessons learned from the accident occurred at the Fukushima- Daiichi nuclear power plant in March 2011. Main advances in the several key areas of technological development were presented through 208 oral presentations during 41 technical sessions which shows the importance taken by fast reactors in the future of nuclear energy.

  18. Freeze-casting as a Novel Manufacturing Process for Fast Reactor Fuels. Final Report

    SciTech Connect

    Wegst, Ulrike G.K.; Allen, Todd; Sridharan, Kumar

    2014-04-07

    Advanced burner reactors are designed to reduce the amount of long-lived radioactive isotopes that need to be disposed of as waste. The input feedstock for creating advanced fuel forms comes from either recycle of used light water reactor fuel or recycle of fuel from a fast burner reactor. Fuel for burner reactors requires novel fuel types based on new materials and designs that can achieve higher performance requirements (higher burn up, higher power, and greater margins to fuel melting) then yet achieved. One promising strategy to improved fuel performance is the manufacture of metal or ceramic scaffolds which are designed to allow for a well-defined placement of the fuel into the host, and this in a manner that permits greater control than that possible in the production of typical CERMET fuels.

  19. Chemical Looping Combustion System-Fuel Reactor Modeling

    SciTech Connect

    Gamwo, I.K.; Jung, J.; Anderson, R.R.; Soong, Y.

    2007-04-01

    Chemical looping combustion (CLC) is a process in which an oxygen carrier is used for fuel combustion instead of air or pure oxygen as shown in the figure below. The combustion is split into air and fuel reactors where the oxidation of the oxygen carrier and the reduction of the oxidized metal occur respectively. The CLC system provides a sequestration-ready CO2 stream with no additional energy required for separation. This major advantage places combustion looping at the leading edge of a possible shift in strict control of CO2 emissions from power plants. Research in this novel technology has been focused in three distinct areas: techno-economic evaluations, integration of the system into power plant concepts, and experimental development of oxygen carrier metals such as Fe, Ni, Mn, Cu, and Ca. Our recent thorough literature review shows that multiphase fluid dynamics modeling for CLC is not available in the open literature. Here, we have modified the MFIX code to model fluid dynamic in the fuel reactor. A computer generated movie of our simulation shows bubble behavior consistent with experimental observations.

  20. Ultrasonic decontamination of prototype fast breeder reactor fuel pins.

    PubMed

    Kumar, Aniruddha; Bhatt, R B; Behere, P G; Afzal, Mohd

    2014-04-01

    Fuel pin decontamination is the process of removing particulates of radioactive material from its exterior surface. It is an important process step in nuclear fuel fabrication. It assumes more significance with plutonium bearing fuel known to be highly radio-toxic owing to its relatively longer biological half life and shorter radiological half life. Release of even minute quantity of plutonium oxide powder in the atmosphere during its handling can cause alarming air borne activity and may pose a severe health hazard to personnel working in the vicinity. Decontamination of fuel pins post pellet loading operation is thus mandatory before they are removed from the glove box for further processing and assembly. This paper describes the setting up of ultrasonic decontamination process, installed inside a custom built fume-hood in the production line, comprising of a cleaning tank with transducers, heaters, pin handling device and water filtration system and its application in cleaning of fuel pins for prototype fast breeder reactor. The cleaning process yielded a typical decontamination efficiency of more than 99%.

  1. Desludging of N Reactor fuel canisters: Analysis, Test, and data requirements

    SciTech Connect

    Johnson, A.B. Jr.

    1996-01-01

    The N Reactor fuel is currently stored in canisters in the K East (KE) and K West (KW) Basins. In KE, the canisters have open tops; in KW, the cans have sealed lids, but are vented to release gases. Corrosion products have formed on exposed uranium metal fuel, on carbon steel basin component surfaces, and on aluminum alloy canister surfaces. Much of the corrosion product is retained on the corroding surfaces; however, large inventories of particulates have been released. Some of the corrosion product particulates form sludge on the basin floors; some particulates are retained within the canisters. The floor sludge inventories are much greater in the KE Basin than in the KW Basin because KE Basin operated longer and its water chemistry was less controlled. Another important factor is the absence of lids on the KE canisters, allowing uranium corrosion products to escape and water-borne species, principally iron oxides, to settle in the canisters. The inventories of corrosion products, including those released as particulates inside the canisters, are only beginning to be characterized for the closed canisters in KW Basin. The dominant species in the KE floor sludge are oxides of aluminum, iron, and uranium. A large fraction of the aluminum and uranium floor sludge particulates may have been released during a major fuel segregation campaign in the 1980s, when fuel was emptied from 4990 canisters. Handling and jarring of the fuel and aluminum canisters seems likely to have released particulates from the heavily corroded surfaces. Four candidate methods are discussed for dealing with canister sludge emerged in the N Reactor fuel path forward: place fuel in multi-canister overpacks (MCOs) without desludging; drill holes in canisters and drain; drill holes in canisters and flush with water; and remove sludge and repackage the fuel.

  2. A combined gas cooled nuclear reactor and fuel cell cycle

    NASA Astrophysics Data System (ADS)

    Palmer, David J.

    Rising oil costs, global warming, national security concerns, economic concerns and escalating energy demands are forcing the engineering communities to explore methods to address these concerns. It is the intention of this thesis to offer a proposal for a novel design of a combined cycle, an advanced nuclear helium reactor/solid oxide fuel cell (SOFC) plant that will help to mitigate some of the above concerns. Moreover, the adoption of this proposal may help to reinvigorate the Nuclear Power industry while providing a practical method to foster the development of a hydrogen economy. Specifically, this thesis concentrates on the importance of the U.S. Nuclear Navy adopting this novel design for its nuclear electric vessels of the future with discussion on efficiency and thermodynamic performance characteristics related to the combined cycle. Thus, the goals and objectives are to develop an innovative combined cycle that provides a solution to the stated concerns and show that it provides superior performance. In order to show performance, it is necessary to develop a rigorous thermodynamic model and computer program to analyze the SOFC in relation with the overall cycle. A large increase in efficiency over the conventional pressurized water reactor cycle is realized. Both sides of the cycle achieve higher efficiencies at partial loads which is extremely important as most naval vessels operate at partial loads as well as the fact that traditional gas turbines operating alone have poor performance at reduced speeds. Furthermore, each side of the cycle provides important benefits to the other side. The high temperature exhaust from the overall exothermic reaction of the fuel cell provides heat for the reheater allowing for an overall increase in power on the nuclear side of the cycle. Likewise, the high temperature helium exiting the nuclear reactor provides a controllable method to stabilize the fuel cell at an optimal temperature band even during transients helping

  3. Plutonium Discharge Rates and Spent Nuclear Fuel Inventory Estimates for Nuclear Reactors Worldwide

    SciTech Connect

    Brian K. Castle; Shauna A. Hoiland; Richard A. Rankin; James W. Sterbentz

    2012-09-01

    This report presents a preliminary survey and analysis of the five primary types of commercial nuclear power reactors currently in use around the world. Plutonium mass discharge rates from the reactors’ spent fuel at reload are estimated based on a simple methodology that is able to use limited reactor burnup and operational characteristics collected from a variety of public domain sources. Selected commercial reactor operating and nuclear core characteristics are also given for each reactor type. In addition to the worldwide commercial reactors survey, a materials test reactor survey was conducted to identify reactors of this type with a significant core power rating. Over 100 material or research reactors with a core power rating >1 MW fall into this category. Fuel characteristics and spent fuel inventories for these material test reactors are also provided herein.

  4. Control rod system useable for fuel handling in a gas-cooled nuclear reactor

    DOEpatents

    Spurrier, Francis R.

    1976-11-30

    A control rod and its associated drive are used to elevate a complete stack of fuel blocks to a position above the core of a gas-cooled nuclear reactor. A fuel-handling machine grasps the control rod and the drive is unlatched from the rod. The stack and rod are transferred out of the reactor, or to a new location in the reactor, by the fuel-handling machine.

  5. Results from ORNL Characterization of HRB-21 Reference Fuel

    SciTech Connect

    Hunn, John D

    2004-08-01

    This document is a compilation of the characterization data produced by ORNL for coated UCO fuel particles (350 {micro}m kernel diameter) fabricated by GA for the HRB-21 irradiation test capsule. The archived fuel particles form the batch used in HRB-21 were obtained by the Advanced Gas Reactor Fuel Development and Qualification (AGR) program for use as a reference material. The GA characterization data for this batch of fuel particles is presented in document GT-HTGR-88357, Rev. C, 'Capsule HRB-21 Preirradiation Report'. The archived fuel particles obtained by the AGR Program are from batch 8876-70, which was the parent batch for batch 8876-70-O, which was actually irradiated in HRB-21. The difference between batches 8876-70 and 8876-70-L is that the particles in batch 8876-70 do not have the seal coat and protective pyrocarbon coating (PPyC) that were deposited over the OPyC layer in batch 8876-70-O. The ORNL designation for the material characterized is AGR-10. This document summarizes characterization of the HRB-21 fuel for size, shape, coating thickness, and density. Fracture behavior and microstructural analysis of the layers and interfaces is compared to previous analyses of the German proof test particles (EU 2358-2365) published in ORNL/CF-04/06. Further detailed comparative study of the microstructure of these two reference materials would be valuable to continue to define the property differences between particles which exhibited good irradiation performance (the German particles) and poor irradiation performance (the HRB-21 particles).

  6. Characterization and Testing of Monolithic RERTR Fuel Plates

    SciTech Connect

    D. D. Keiser; J. F. Jue; D. E. Burkes

    2007-03-01

    Monolithic fuel plates are being developed for application in research reactors throughout the world. These fuel plates are comprised of a U-Mo alloy foil encased in aluminum alloy cladding. Three different fabrication techniques have been looked at for producing monolithic fuel plates: hot isostatic pressing (HIP), transient liquid phase bonding (TLPB), and friction stir welding (FSW). Of these three techniques, HIP and FSW are currently being emphasized. As part of the development of these fabrication techniques, fuel plates are characterized and tested to determine properties like hardness and the bond strength at the interface between the fuel and cladding. Testing of HIPed samples indicates that the foil/cladding interaction behavior depends on the Mo content in the U-Mo foil, the measured hardness values are quite different for the fuel, cladding, and interaction zone phase and Ti, Zr and Nb are the most effective diffusion barriers. For FSW samples, there is a dependence of the bond strength at the foil/cladding interface on the type of tool that is employed for performing the actual FSW process.

  7. Dual phase magnesia-zirconia ceramics for light water reactor inert matrix fuel.

    SciTech Connect

    Medvedev, P.; Meyer, M. K.; Nuclear Technology

    2005-01-01

    To address the low thermal conductivity of the ZrO{sub 2}-based inert matrix fuel and the instability in water of the MgO-based inert matrix fuel, the dual-phase MgO-ZrO{sub 2} ceramics are proposed as a matrix for light water reactor fuel for actinide transmutation and Pu burning. It is envisioned that in a dual-phase system MgO will act as efficient heat conductor while ZrO{sub 2} will provide protection from the coolant attack. This paper describes results of fabrication, characterization and hydration testing of MgO-ZrO{sub 2} ceramics containing 30-70 wt% of MgO.

  8. Dry Storage of Research Reactor Spent Nuclear Fuel - 13321

    SciTech Connect

    Adams, T.M.; Dunsmuir, M.D.; Leduc, D.R.; Severynse, T.F.; Sindelar, R.L.; Moore, E.N.

    2013-07-01

    Spent fuel from domestic and foreign research reactors is received and stored at the Savannah River Site's L Area Material Storage (L Basin) Facility. This DOE-owned fuel consists primarily of highly enriched uranium in metal, oxide or silicide form with aluminum cladding. Upon receipt, the fuel is unloaded and transferred to basin storage awaiting final disposition. Disposition alternatives include processing via the site's H Canyon facility for uranium recovery, or packaging and shipment of the spent fuel to a waste repository. A program has been developed to provide a phased approach for dry storage of the L Basin fuel. The initial phase of the dry storage program will demonstrate loading, drying, and storage of fuel in twelve instrumented canisters to assess fuel performance. After closure, the loaded canisters are transferred to pad-mounted concrete overpacks, similar to those used for dry storage of commercial fuel. Unlike commercial spent fuel, however, the DOE fuel has high enrichment, very low to high burnup, and low decay heat. The aluminum cladding presents unique challenges due to the presence of an oxide layer that forms on the cladding surface, and corrosion degradation resulting from prolonged wet storage. The removal of free and bound water is essential to the prevention of fuel corrosion and radiolytic generation of hydrogen. The demonstration will validate models predicting pressure, temperature, gas generation, and corrosion performance, provide an engineering scale demonstration of fuel handling, drying, leak testing, and canister backfill operations, and establish 'road-ready' storage of fuel that is suitable for offsite repository shipment or retrievable for onsite processing. Implementation of the Phase I demonstration can be completed within three years. Phases II and III, leading to the de-inventory of L Basin, would require an additional 750 canisters and 6-12 years to complete. Transfer of the fuel from basin storage to dry storage

  9. Statistical Methods Handbook for Advanced Gas Reactor Fuel Materials

    SciTech Connect

    J. J. Einerson

    2005-05-01

    Fuel materials such as kernels, coated particles, and compacts are being manufactured for experiments simulating service in the next generation of high temperature gas reactors. These must meet predefined acceptance specifications. Many tests are performed for quality assurance, and many of these correspond to criteria that must be met with specified confidence, based on random samples. This report describes the statistical methods to be used. The properties of the tests are discussed, including the risk of false acceptance, the risk of false rejection, and the assumption of normality. Methods for calculating sample sizes are also described.

  10. An improved characterization method for international accountancy measurements of fresh and irradiated mixed oxide (MOX) fuel: helping achieve continual monitoring and safeguards through the fuel cycle

    SciTech Connect

    Evans, Louise G; Croft, Stephen; Swinhoe, Martyn T; Tobin, S. J.; Menlove, H. O.; Schear, M. A.; Worrall, Andrew

    2011-01-13

    Nuclear fuel accountancy measurements are conducted at several points through the nuclear fuel cycle to ensure continuity of knowledge (CofK) of special nuclear material (SNM). Non-destructive assay (NDA) measurements are performed on fresh fuel (prior to irradiation in a reactor) and spent nuclear fuel (SNF) post-irradiation. We have developed a fuel assembly characterization system, based on the novel concept of 'neutron fingerprinting' with multiplicity signatures to ensure detailed CofK of nuclear fuel through the entire fuel cycle. The neutron fingerprint in this case is determined by the measurement of the various correlated neutron signatures, specific to fuel isotopic composition, and therefore offers greater sensitivity to variations in fissile content among fuel assemblies than other techniques such as gross neutron counting. This neutron fingerprint could be measured at the point of fuel dispatch (e.g. from a fuel fabrication plant prior to irradiation, or from a reactor site post-irradiation), monitored during transportation of the fuel assembly, and measured at a subsequent receiving site (e.g. at the reactor site prior to irradiation, or reprocessing facility post-irradiation); this would confirm that no unexpected changes to the fuel composition or amount have taken place during transportation and/ or reactor operations. Changes may indicate an attempt to divert material for example. Here, we present the current state of the practice of fuel measurements for both fresh mixed oxide (MOX) fuel and SNF (both MOX and uranium dioxide). This is presented in the framework of international safeguards perspectives from the US and UK. We also postulate as to how the neutron fingerprinting concept could lead to improved fuel characterization (both fresh MOX and SNF) resulting in: (a) assured CofK of fuel across the nuclear fuel cycle, (b) improved detection of SNM diversion, and (c) greater confidence in safeguards of SNF transportation.

  11. An improved characterization method for international accountancy measurements of fresh and irradiated mixed oxide (MOX) fuel: helping achieve continual monitoring and safeguards through the fuel cycle

    SciTech Connect

    Evans, Louise G; Croft, Stephen; Swinhoe, Martyn T; Tobin, S. J.; Boyer, B. D.; Menlove, H. O.; Schear, M. A.; Worrall, Andrew

    2010-11-24

    Nuclear fuel accountancy measurements are conducted at several points through the nuclear fuel cycle to ensure continuity of knowledge (CofK) of special nuclear material (SNM). Non-destructive assay (NDA) measurements are performed on fresh fuel (prior to irradiation in a reactor) and spent nuclear fuel (SNF) post-irradiation. We have developed a fuel assembly characterization system, based on the novel concept of 'neutron fingerprinting' with multiplicity signatures to ensure detailed CofK of nuclear fuel through the entire fuel cycle. The neutron fingerprint in this case is determined by the measurement of the various correlated neutron signatures, specific to fuel isotopic composition, and therefore offers greater sensitivity to variations in fissile content among fuel assemblies than other techniques such as gross neutron counting. This neutron fingerprint could be measured at the point of fuel dispatch (e.g. from a fuel fabrication plant prior to irradiation, or from a reactor site post-irradiation), monitored during transportation of the fuel assembly, and measured at a subsequent receiving site (e.g. at the reactor site prior to irradiation, or reprocessing facility post-irradiation); this would confirm that no unexpected changes to the fuel composition or amount have taken place during transportation and/or reactor operations. Changes may indicate an attempt to divert material for example. Here, we present the current state of the practice of fuel measurements for both fresh mixed oxide (MOX) fuel and SNF (both MOX and uranium dioxide). This is presented in the framework of international safeguards perspectives from the US and UK. We also postulate as to how the neutron fingerprinting concept could lead to improved fuel characterization (both fresh MOX and SNF) resulting in: (a) assured CofK of fuel across the nuclear fuel cycle, (b) improved detection of SNM diversion, and (c) greater confidence in safeguards of SNF transportation.

  12. Feasibility assessment of burnup credit in the criticality analysis of shipping casks with boiling water reactor spent fuel

    SciTech Connect

    Broadhead, B.L.

    1991-08-01

    Considerable interest in the allowance of reactivity credit for the exposure history of power reactor fuel currently exists. This burnup credit'' issue has the potential to greatly reduce risk and cost when applied to the design and certification of spent fuel casks used for transportation and storage. Recently, analyses have demonstrated the technical feasibility and estimated the risk and economic incentives for allowing burnup credit in pressurized water reactor (PWR) spent fuel shipping cask applications. This report summarizes the extension of the previous PWR technical feasibility assessment to boiling water reactor (BWR) fuel. This feasibility analysis aims to apply simple methods that adequately characterize the time-dependent isotopic compositions of typical BWR fuel. An initial analysis objective was to identify a simple and reliable method for characterizing BWR spent fuel. Two different aspects of fuel characterization were considered:l first, the generation of burn- up dependent material interaction probabilities; second, the prediction of material inventories over time (depletion). After characterizing the spent fuel at various stages of exposure and decay, three dimensional (3-D) models for an infinite array of assemblies and, in several cases, infinite arrays of assemblies in a typical shipping cask basket were analyzed. Results for assemblies without a basket provide reactivity control requirements as a function of burnup and decay, while results including the basket allow assessment of typical basket configurations to provide sufficient reactivity control for spent BWR fuel. Resulting basket worths and reactivity trends over time are then evaluated to determine whether burnup credit is needed and feasible in BWR applications.

  13. 75 FR 67636 - Physical Protection of Shipments of Irradiated Reactor Fuel

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-11-03

    ... rule was published on October 13, 2010 (75 FR 62695), and the public comment period runs through... ``irradiated reactor fuel'' and ``spent nuclear fuel'' (SNF) are used interchangeably). The NRC has prepared...

  14. CHARACTERIZATION OF HYDROGEN CONTENT IN ZIRCALOY-4 NUCLEAR FUEL CLADDING

    SciTech Connect

    Pfeif, E. A.; Mishra, B.; Olson, D. L.; Lasseigne, A. N.; Krzywosz, K.; Mader, E. V.

    2010-02-22

    Assessment of hydrogen uptake of underwater nuclear fuel clad and component materials will enable improved monitoring of fuel health. Zirconium alloys are used in nuclear reactors as fuel cladding, fuel channels, guide tubes and spacer grids, and are available for inspection in spent fuel pools. With increasing reactor exposure zirconium alloys experience hydrogen ingress due to neutron interactions and water-side corrosion that is not easily quantified without destructive hot cell examination. Contact and non-contact nondestructive techniques, using Seebeck coefficient measurements and low frequency impedance spectroscopy, to assess the hydrogen content and hydride formation within zircaloy 4 material that are submerged to simulate spent fuel pools are presented.

  15. An integrated approach for the verification of fresh mixed oxide fuel (MOX) assemblies at light water reactor MOX recycle reactors

    SciTech Connect

    Menlove, Howard O; Lee, Sang - Yoon

    2009-01-01

    This paper presents an integrated approach for the verification of mixed oxide (MOX) fuel assemblies prior to their being loaded into the reactor. There is a coupling of the verification approach that starts at the fuel fabrication plant and stops with the transfer of the assemblies into the thermal reactor. The key measurement points are at the output of the fuel fabrication plant, the receipt at the reactor site, and the storage in the water pool as fresh fuel. The IAEA currently has the capability to measure the MOX fuel assemblies at the output of the fuel fabrication plants using a passive neutron coincidence counting systems of the passive neutron collar (PNCL) type. Also. at the MOX reactor pool, the underwater coincidence counter (UWCC) has been developed to measure the MOX assemblies in the water. The UWCC measurement requires that the fuel assembly be lifted about two meters up in the storage rack to avoid interference from the fuel that is stored in the rack. This paper presents a new method to verify the MOX fuel assemblies that are in the storage rack without the necessity of moving the fuel. The detector system is called the Underwater MOX Verification System (UMVS). The integration and relationship of the three measurements systems is described.

  16. Breeding nuclear fuels with accelerators — replacement for breeder reactors

    NASA Astrophysics Data System (ADS)

    Grand, P.; Takahashi, H.

    1985-05-01

    High-current, high-energy linear accelerators are potential competitors to breeder reactors and fusion devices for the production of fissile fuel. Accelerator breeder studies, conducted at Chalk River (Canada) and Brookhaven National Laboratory during the last decade, indicate that the technology is available and system's costs competitive with that of the LMFBR and Fusion-Hybrid systems. This paper addresses the technical issues to be faced in an accelerator specifically designed for commercial operation of this kind, the neutronics and engineering feasibility of target systems, and related fuel-cycle cost/benefit analysis. A nearly optimized accelerator breeder concept consisting of a 1.5 GeV, 300 mA linear accelerator, and of a H 2O-cooled, U-metal target (or thorium) has been designed and costed. Accelerator efficiency, beam-to-plug, is estimated at 50% and target power generation efficiency, beam-to-thermal output, at about 600% (U-metal case). Fuel costs produced by the facility are practically entirely driven by capital investment costs. This is also true for any competing system. On the basis, the accelerator breeder is economically competitive with the LMFBR while offering real advantages in terms of safety, fuel cycle flexibility, and resource independence.

  17. Axially staggered seed-blanket reactor fuel module construction

    DOEpatents

    Cowell, Gary K.; DiGuiseppe, Carl P.

    1985-01-01

    A heterogeneous nuclear reactor of the seed-blanket type is provided wher the fissile (seed) and fertile (blanket) nuclear fuels are segregated axially within each fuel element such that fissile and fertile regions occur in an alternating pattern along the length of the fuel element. Further, different axial stacking patterns are used for the fuel elements of at least two module types such that when modules of different types are positioned adjacent to one another, the fertile regions of the modules are offset or staggered. Thus, when a module of one type is surrounded by modules of the second type the fertile regions thereof will be surrounded on all sides by fissile material. This provides enhanced neutron communication both radially and axially, thereby resulting in greater power oscillation stability than other axial arrangements. The arrangements of the fissile and fertile regions in an alternating axial manner minimizes the radial power peaking factors and provides a more optional thermal-hydraulic design than is afforded by radial arrangements.

  18. Individual source positioning mechanism for a nuclear reactor fuel assembly

    SciTech Connect

    Wilson, J.F.; Gjertsen, R.K.; Cerni, S.

    1987-07-07

    A nuclear reactor is described including a fuel assembly, at lest one elongated neutron source rod and an upper core plate. The fuel assembly has top and bottom nozzles with a guide thimbles extending between and interconnecting the nozzles. The upper core plate is positioned adjacent to and above the top nozzle of the fuel assembly and having flow openings to allow passage of coolant from the fuel assembly. At least some of the openings is aligned over respective ones of the guide thimbles with seating means defined about the openings on a lower side of the core plate, a separate mechanism for positioning each individual neutron source rod in a respective guide thimble aligned with one of the openings defined through the upper core plate, comprising: (a) locating means registering against the core plate seating means; and (b) resilient holddown means extending partially into the guide thimble and coupling the source rod with the locating means in a manner which restrains the source rod in a lateral direction and positions the rod in a stationary axial relationship within the guide thimble.

  19. Reactor Physics and Criticality Benchmark Evaluations for Advanced Nuclear Fuel - Final Technical Report

    SciTech Connect

    William Anderson; James Tulenko; Bradley Rearden; Gary Harms

    2008-09-11

    The nuclear industry interest in advanced fuel and reactor design often drives towards fuel with uranium enrichments greater than 5 wt% 235U. Unfortunately, little data exists, in the form of reactor physics and criticality benchmarks, for uranium enrichments ranging between 5 and 10 wt% 235U. The primary purpose of this project is to provide benchmarks for fuel similar to what may be required for advanced light water reactors (LWRs). These experiments will ultimately provide additional information for application to the criticality-safety bases for commercial fuel facilities handling greater than 5 wt% 235U fuel.

  20. Advanced Fuels Campaign Light Water Reactor Accident Tolerant Fuel Performance Metrics

    SciTech Connect

    Brad Merrill; Melissa Teague; Robert Youngblood; Larry Ott; Kevin Robb; Michael Todosow; Chris Stanek; Mitchell Farmer; Michael Billone; Robert Montgomery; Nicholas Brown; Shannon Bragg-Sitton

    2014-02-01

    The safe, reliable and economic operation of the nation’s nuclear power reactor fleet has always been a top priority for the United States’ nuclear industry. As a result, continual improvement of technology, including advanced materials and nuclear fuels, remains central to industry’s success. Decades of research combined with continual operation have produced steady advancements in technology and yielded an extensive base of data, experience, and knowledge on light water reactor (LWR) fuel performance under both normal and accident conditions. In 2011, following the Great East Japan Earthquake, resulting tsunami, and subsequent damage to the Fukushima Daiichi nuclear power plant complex, enhancing the accident tolerance of LWRs became a topic of serious discussion. As a result of direction from the U.S. Congress, the U.S. Department of Energy Office of Nuclear Energy (DOE-NE) initiated an Accident Tolerant Fuel (ATF) Development program. The complex multiphysics behavior of LWR nuclear fuel makes defining specific material or design improvements difficult; as such, establishing qualitative attributes is critical to guide the design and development of fuels and cladding with enhanced accident tolerance. This report summarizes a common set of technical evaluation metrics to aid in the optimization and down selection of candidate designs. As used herein, “metrics” describe a set of technical bases by which multiple concepts can be fairly evaluated against a common baseline and against one another. Furthermore, this report describes a proposed technical evaluation methodology that can be applied to assess the ability of each concept to meet performance and safety goals relative to the current UO2 – zirconium alloy system and relative to one another. The resultant ranked evaluation can then inform concept down-selection, such that the most promising accident tolerant fuel design option(s) can continue to be developed for lead test rod or lead test assembly

  1. Mass estimates of very small reactor cores fueled by Uranium-235, U-233 and Cm-245

    NASA Astrophysics Data System (ADS)

    Wright, Steven A.; Lipinski, Ronald J.

    2001-02-01

    This paper explores the possibility of manufacturing very small reactors from U-235, U-233 and Cm-245. Pin type reactor systems fueled with uranium or curium metal zirconium hydride (UZrH or CmZrH) are compared with similar designs using U-235. Criticality measurements of homogeneous water uranium systems, suggest that reactor subsystem masses have a broad minimum for hydrogen-to-uranium atom ratios that vary from 25-250. This paper compares the masses of metal-hydride fueled reactor systems that use U-235, U-233, and Cm-245 fuel with hydrogen-to-metal atom ratios from 20-300 when cooled by gas (HeXe), liquid metal (Na), and water. The results indicate that water cooled reactors in general have the smallest reactor subsystem mass. For gas and liquid-metal cooled reactors U-233 subsystems have total masses that are about 1/2 those of similarly designed U-235 fuel reactors. Reactor subsystems consisting of 11.2% enriched Cm-245 (balance Cm-244) that can be obtained from fuel reprocessing have system masses comparable to that of U-233. The smallest reactor subsystem masses were on the order of 60-80 kg for U-233 fueled water cooled reactors. .

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

    SciTech Connect

    Williams, Kent Alan; Shropshire, David E.

    2009-01-01

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

  3. Corrosion of the AlFeNi alloy used for the fuel cladding in the Jules Horowitz research reactor

    NASA Astrophysics Data System (ADS)

    Wintergerst, M.; Dacheux, N.; Datcharry, F.; Herms, E.; Kapusta, B.

    2009-09-01

    The AlFeNi aluminium alloy (1 wt% Fe, 1 wt% Ni, 1 wt% Mg) is expected to be used as nuclear fuel cladding for the Jules Horowitz experimental reactor. To guarantee a safe behaviour of the fuel, a good understanding of the fuel clad corrosion mechanisms is required. In this field, the experimental characterization of the selected alloy was performed. Then experimental studies of the aluminium alloy corrosion product obtained in autoclaves have shown an oxide film composed of two layers. This duplex structure results from a mixed growth mechanism: an anionic growth to develop the inner oxide and a cationic diffusion parallel to a dissolution-precipitation process to form the outer zone. Dynamic experiments at 70 °C have demonstrated that a solid diffusion step controls the release kinetic. Then post-irradiation exams performed on irradiated fuel plates were used to investigate the effects of the irradiation on the corrosion behaviour in the reactor core.

  4. Dealing with Historical Discrepancies: The Recovery of National Research Experiment (NRX) Reactor Fuel Rods at Chalk River Laboratories (CRL) - 13324

    SciTech Connect

    Vickerd, Meggan

    2013-07-01

    Following the 1952 National Research Experiment (NRX) Reactor accident, fuel rods which had short irradiation histories were 'temporarily' buried in wooden boxes at the 'disposal grounds' during the cleanup effort. The Nuclear Legacy Liabilities Program (NLLP), funded by Natural Resources Canada (NRCan), strategically retrieves legacy waste and restores lands affected by Atomic Energy of Canada Limited (AECL) early operations. Thus under this program the recovery of still buried NRX reactor fuel rods and their relocation to modern fuel storage was identified as a priority. A suspect inventory of NRX fuels was compiled from historical records and various research activities. Site characterization in 2005 verified the physical location of the fuel rods and determined the wooden boxes they were buried in had degraded such that the fuel rods were in direct contact with the soil. The fuel rods were recovered and transferred to a modern fuel storage facility in 2007. Recovered identification tags and measured radiation fields were used to identify the inventory of these fuels. During the retrieval activity, a discrepancy was discovered between the anticipated number of fuel rods and the number found during the retrieval. A total of 32 fuel rods and cans of cut end pieces were recovered from the specified site, which was greater than the anticipated 19 fuel rods and cans. This discovery delayed the completion of the project, increased the associated costs, and required more than anticipated storage space in the modern fuel storage facility. A number of lessons learned were identified following completion of this project, the most significant of which was the potential for discrepancies within the historical records. Historical discrepancies are more likely to be resolved by comprehensive historical record searches and site characterizations. It was also recommended that a complete review of the wastes generated, and the total affected lands as a result of this historic

  5. Characterization and supply of coal based fuels

    SciTech Connect

    Not Available

    1992-06-01

    Studies and data applicable for fuel markets and coal resource assessments were reviewed and evaluated to provide both guidelines and specifications for premium quality coal-based fuels. The fuels supplied under this contract were provided for testing of advanced combustors being developed under Pittsburgh Energy Technology Center (PETC) sponsorship for use in the residential, commercial and light industrial (RCLI) market sectors. The requirements of the combustor development contractors were surveyed and periodically updated to satisfy the evolving needs based on design and test experience. Available coals were screened and candidate coals were selected for further detailed characterization and preparation for delivery. A team of participants was assembled to provide fuels in both coal-water fuel (CWF) and dry ultrafine coal (DUC) forms. Information about major US coal fields was correlated with market needs analysis. Coal fields with major reserves of low sulfur coal that could be potentially amenable to premium coal-based fuels specifications were identified. The fuels requirements were focused in terms of market, equipment and resource constraints. With this basis, the coals selected for developmental testing satisfy the most stringent fuel requirements and utilize available current deep-cleaning capabilities.

  6. Characterization of Used Nuclear Fuel with Multivariate Analysis for Process Monitoring

    SciTech Connect

    Dayman, Kenneth J.; Coble, Jamie B.; Orton, Christopher R.; Schwantes, Jon M.

    2014-01-01

    The Multi-Isotope Process (MIP) Monitor combines gamma spectroscopy and multivariate analysis to detect anomalies in various process streams in a nuclear fuel reprocessing system. Measured spectra are compared to models of nominal behavior at each measurement location to detect unexpected changes in system behavior. In order to improve the accuracy and specificity of process monitoring, fuel characterization may be used to more accurately train subsequent models in a full analysis scheme. This paper presents initial development of a reactor-type classifier that is used to select a reactor-specific partial least squares model to predict fuel burnup. Nuclide activities for prototypic used fuel samples were generated in ORIGEN-ARP and used to investigate techniques to characterize used nuclear fuel in terms of reactor type (pressurized or boiling water reactor) and burnup. A variety of reactor type classification algorithms, including k-nearest neighbors, linear and quadratic discriminant analyses, and support vector machines, were evaluated to differentiate used fuel from pressurized and boiling water reactors. Then, reactor type-specific partial least squares models were developed to predict the burnup of the fuel. Using these reactor type-specific models instead of a model trained for all light water reactors improved the accuracy of burnup predictions. The developed classification and prediction models were combined and applied to a large dataset that included eight fuel assembly designs, two of which were not used in training the models, and spanned the range of the initial 235U enrichment, cooling time, and burnup values expected of future commercial used fuel for reprocessing. Error rates were consistent across the range of considered enrichment, cooling time, and burnup values. Average absolute relative errors in burnup predictions for validation data both within and outside the training space were 0.0574% and 0.0597%, respectively. The errors seen in this

  7. On the possibility of using uranium-beryllium oxide fuel in a VVER reactor

    NASA Astrophysics Data System (ADS)

    Kovalishin, A. A.; Prosyolkov, V. N.; Sidorenko, V. D.; Stogov, Yu. V.

    2014-12-01

    The possibility of using UO2-BeO fuel in a VVER reactor is considered with allowance for the thermophysical properties of this fuel. Neutron characteristics of VVER fuel assemblies with UO2-BeO fuel pellets are estimated.

  8. On the possibility of using uranium-beryllium oxide fuel in a VVER reactor

    SciTech Connect

    Kovalishin, A. A.; Prosyolkov, V. N.; Sidorenko, V. D.; Stogov, Yu. V.

    2014-12-15

    The possibility of using UO{sub 2}-BeO fuel in a VVER reactor is considered with allowance for the thermophysical properties of this fuel. Neutron characteristics of VVER fuel assemblies with UO{sub 2}-BeO fuel pellets are estimated.

  9. Nuclear fuel cycle analysis of the SABR fusion-fission hybrid transmutation reactor

    NASA Astrophysics Data System (ADS)

    Sommer, Chris; Stacey, Weston; Petrovic, Bojan

    2009-11-01

    Various fuel cycles have been designed and analyzed for the Subcritical Advanced Burner Reactor (SABR). SABR is a sodium cooled fast reactor fueled with transuranics (TRU) from spent fuel of light water reactors and driven by a tokamak fusion neutron source based on ITER physics and technology. SABR employs a four batch fuel cycle using an out-to-in shuffling pattern, with the fuel being reprocessed at the end of each cycle. The reprocessing method assumes recovery rates of 99.9% of the actinides and 0.1% of the fission products remain in the recycled fuel. The reprocessing fuel cycles were analyzed to find an optimal cycle length in terms of burn up, power distribution, and materials limitations. Fuel cycles are analyzed using CEA's ERANOS2.0 code, with fuel residence times limited by radiation damage at 100, 150 and 200 dpa.

  10. MODELING ASSUMPTIONS FOR THE ADVANCED TEST REACTOR FRESH FUEL SHIPPING CONTAINER

    SciTech Connect

    Rick J. Migliore

    2009-09-01

    The Advanced Test Reactor Fresh Fuel Shipping Container (ATR FFSC) is currently licensed per 10 CFR 71 to transport a fresh fuel element for either the Advanced Test Reactor, the University of Missouri Research Reactor (MURR), or the Massachusetts Institute of Technology Research Reactor (MITR-II). During the licensing process, the Nuclear Regulatory Commission (NRC) raised a number of issues relating to the criticality analysis, namely (1) lack of a tolerance study on the fuel and packaging, (2) moderation conditions during normal conditions of transport (NCT), (3) treatment of minor hydrogenous packaging materials, and (4) treatment of potential fuel damage under hypothetical accident conditions (HAC). These concerns were adequately addressed by modifying the criticality analysis. A tolerance study was added for both the packaging and fuel elements, full-moderation was included in the NCT models, minor hydrogenous packaging materials were included, and fuel element damage was considered for the MURR and MITR-II fuel types.

  11. CORAL: a stepping stone for establishing the Indian fast reactor fuel reprocessing technology

    SciTech Connect

    Venkataraman, M.; Natarajan, R.; Raj, Baldev

    2007-07-01

    The reprocessing of spent fuel from Fast Breeder Test Reactor (FBTR) has been successfully demonstrated in the pilot plant, CORAL (COmpact Reprocessing facility for Advanced fuels in Lead shielded cell). Since commissioning in 2003, spent mixed carbide fuel from FBTR of different burnups and varying cooling period, have been reprocessed in this facility. Reprocessing of the spent fuel with a maximum burnup of 100 GWd/t has been successfully carried out so far. The feed backs from these campaigns with progressively increasing specific activities, have been useful in establishing a viable process flowsheet for reprocessing the Prototype Fast Breeder Reactor (PFBR) spent fuel. Also, the design of various equipments and processes for the future plants, which are either under design for construction, namely, the Demonstration Fast Reactor Fuel Reprocessing Plant (DFRP) and the Fast reactor fuel Reprocessing Plant (FRP) could be finalized. (authors)

  12. Design and operation of gamma scan and fission gas sampling systems for characterization of irradiated commercial nuclear fuel

    SciTech Connect

    Knox, C.A.; Thornhill, R.E.; Mellinger, G.B.

    1989-09-01

    One of the primary objectives of the Materials Characterization Center (MCC) is to acquire and characterize spent fuels used in waste form testing related to nuclear waste disposal. The initial steps in the characterization of a fuel rod consist of gamma scanning the rod and sampling the gas contained in the fuel rod (referred to as fission gas sampling). The gamma scan and fission gas sampling systems used by the MCC are adaptable to a wide range of fuel types and have been successfully used to characterize both boiling water reactor (BWR) and pressurized water reactor (PWR) fuel rods. This report describes the design and operation of systems used to gamma scan and fission gas sample full-length PWR and BWR fuel rods. 1 ref., 10 figs., 1 tab.

  13. Supplemental Reactor Physics Calculations and Analysis of ELF Mk 1A Fuel

    SciTech Connect

    Michael A. Pope

    2014-10-01

    These calculations supplement previous the reactor physics work evaluating the Enhanced Low Enriched Uranium (LEU) Fuel (ELF) Mk 1A element. This includes various additional comparisons between the current Highly Enriched Uranium (HEU) and LEU along with further characterization of the performance of the ELF fuel. The excess reactivity to be held down at BOC for ELF Mk 1A fuel is estimated to be approximately $2.75 greater than with HEU for a typical cycle. This is a combined effect of the absence of burnable poison in the ELF fuel and the reduced neck shim worth in LEU fuel compared to HEU. Burnable poison rods were conceptualized for use in the small B positions containing Gd2O3 absorber. These were shown to provide $2.37 of negative reactivity at BOC and to burn out in less than half of a cycle. The worth of OSCCs is approximately the same between HEU and ELF Mk 1A (LEU) fuels in the representative loading evaluated. This was evaluated by rotating all banks simultaneously. The safety rod worth is relatively unchanged between HEU and ELF Mk 1A (LEU) fuels in the representative loading evaluated. However, this should be reevaluated with different loadings. Neutron flux, both total and fast (>1 MeV), is either the same or reduced upon changing from HEU to ELF Mk 1A (LEU) fuels in the representative loading evaluated. This is consistent with the well-established trend of lower neutron fluxes for a given power in LEU than HEU.The IPT loop void reactivity is approximately the same or less positive with ELF Mk 1A (LEU) fuel than HEU in the representative loading evaluated.

  14. Preliminary Advanced Test Reactor LEU Fuel Conversion Feasibility Study

    SciTech Connect

    G. S. Chang; R. G. Ambrosek

    2005-11-01

    The Advanced Test Reactor (ATR) is a high power density, high neutron flux research reactor operating in the United States. The ATR has large irradiation test volumes located in high flux areas. Powered with highly enriched uranium (HEU), the ATR has a maximum thermal power rating of 250 MWth with a maximum unperturbed thermal neutron flux rating of 1.0 x 1015 n/cm2–s. As a result, the ATR is a representative candidate for assessing the necessary modifications and evaluating the subsequent operating effects associated with low-enriched uranium (LEU) fuel conversion. A detailed plate-by-plate MCNP ATR 1/8th core model was developed for the fuel cycle burnup comparison analysis. Using the current HEU 235U enrichment of 93.0 % as a baseline, an analysis can be performed to determine the LEU uranium density and 235U enrichment required in the fuel meat to yield an equivalent Keff between the HEU core and a LEU core versus effective full power days (EFPD). The MCNP ATR 1/8th core model will be used to optimize the 235U loading in the LEU core, such that the differences in Keff between the HEU and LEU core can be minimized for operation at 150 EFPD with a total core power of 115 MW. The Monte-Carlo with ORIGEN-2 (MCWO) method was used to calculate Keff versus EFPDs. The MCWO-calculated results for the LEU case demonstrated adequate excess reactivity such that the LEU core conversion designer should be able to optimize the 235U content of each fuel plate, so that the Keff and relative radial fission heat flux profile are similar to the reference ATR HEU case. However, to demonstrate that the LEU core fuel cycle performance can meet the Upgraded Final Safety Analysis Report (UFSAR) safety requirements, a further study will be required in order to investigate the detailed radial, axial, and azimuthal heat flux profile variations versus EFPDs.

  15. Operation of N Reactor and Fuels Fabrication Facilities, Hanford Reservation, Richland, Benton County, Washington: Environmental assessment

    SciTech Connect

    Not Available

    1980-08-01

    Environmental data, calculations and analyses show no significant adverse radiological or nonradiological impacts from current or projected future operations resulting from N Reactor, Fuels Fabrication and Spent Fuel Storage Facilities. Nonoccupational radiation exposures resulting from 1978 N Reactor operations are summarized and compared to allowable exposure limits.

  16. Technology Implementation Plan. Fully Ceramic Microencapsulated Fuel for Commercial Light Water Reactor Application

    SciTech Connect

    Snead, Lance Lewis; Terrani, Kurt A.; Powers, Jeffrey J.; Worrall, Andrew; Robb, Kevin R.; Snead, Mary A.

    2015-04-01

    This report is an overview of the implementation plan for ORNL's fully ceramic microencapsulated (FCM) light water reactor fuel. The fully ceramic microencapsulated fuel consists of tristructural isotropic (TRISO) particles embedded inside a fully dense SiC matrix and is intended for utilization in commercial light water reactor application.

  17. Pyrometallurgical processing of Integral Fast Reactor metal fuels

    SciTech Connect

    Battles, J.E.; Miller, W.E.; Gay, E.C.

    1991-01-01

    The pyrometallurgical process for recycling spent metal fuels from the Integral Fast Reactor is now in an advanced state of development. This process involves electrorefining spent fuel with a cadmium anode, solid and liquid cathodes, and a molten salt electrolyte (LiCl-KCl) at 500{degrees}C. The initial process feasibility and flowsheet verification studies have been conducted in a laboratory-scale electrorefiner. Based on these studies, a dual cathode approach has been adopted, where uranium is recovered on a solid cathode mandrel and uranium-plutonium is recovered in a liquid cadmium cathode. Consolidation and purification (salt and cadmium removal) of uranium and uranium-plutonium products from the electrorefiner have been successful. The process is being developed with the aid of an engineering-scale electrorefiner, which has been successfully operated for more than three years. In this electrorefiner, uranium has been electrotransported from the cadmium anode to a solid cathode in 10 kg quantities. Also, anodic dissolution of 10 kg batches of chopped, simulated fuel (U--10% Zr) has been demonstrated. Development of the liquid cadmium cathode for recovering uranium-plutonium is under way.

  18. Closing nuclear fuel cycle with fast reactors: problems and prospects

    SciTech Connect

    Shadrin, A.; Dvoeglazov, K.; Ivanov, V.

    2013-07-01

    The closed nuclear fuel cycle (CNFC) with fast reactors (FR) is the most promising way of nuclear energetics development because it prevents spent nuclear fuel (SNF) accumulation and minimizes radwaste volume due to minor actinides (MA) transmutation. CNFC with FR requires the elaboration of safety, environmentally acceptable and economically effective methods of treatment of SNF with high burn-up and low cooling time. The up-to-date industrially implemented SNF reprocessing technologies based on hydrometallurgical methods are not suitable for the reprocessing of SNF with high burn-up and low cooling time. The alternative dry methods (such as electrorefining in molten salts or fluoride technologies) applicable for such SNF reprocessing have not found implementation at industrial scale. So the cost of SNF reprocessing by means of dry technologies can hardly be estimated. Another problem of dry technologies is the recovery of fissionable materials pure enough for dense fuel fabrication. A combination of technical solutions performed with hydrometallurgical and dry technologies (pyro-technology) is proposed and it appears to be a promising way for the elaboration of economically, ecologically and socially accepted technology of FR SNF management. This paper deals with discussion of main principle of dry and aqueous operations combination that probably would provide safety and economic efficiency of the FR SNF reprocessing. (authors)

  19. Core conversion of the Portuguese research reactor to LEU fuel

    SciTech Connect

    Marques, J.G.; Ramos, A.R.; Kocher, A.

    2008-07-15

    Core conversion of the Portuguese Research Reactor (RPI) to LEU fuel is being performed within IAEA's Technical Cooperation project POR/4/016, with financial support from the US and Portugal. CERCA was selected as manufacturer of the LEU assemblies by the IAEA after an international call for bids. CERCA provided a comprehensive package to the RPI which included the mechanical verification of the design of the assemblies, their manufacture and arrangements for a joint inspection of the finished assemblies. The LEU fuel assemblies were manufactured within 8 months upon final approval of the design. The safety analyses for the core conversion to LEU fuel were made with the assistance of the RERTR program and were submitted for review by the IAEA and by Portuguese authorities in January 2007. Revised documents were submitted in June 2007 addressing the issues raised during review. Regulatory approval was received in early August and core conversion was done in early September. All measured safety parameters are within the defined acceptance limits. Operation at full power is expected by the end of October. (author)

  20. Impact of conversion to mixed-oxide fuels on reactor structural components

    SciTech Connect

    Yahr, G.T.

    1997-04-01

    The use of mixed-oxide (MOX) fuel to replace conventional uranium fuel in commercial light-water power reactors will result in an increase in the neutron flux. The impact of the higher flux on the structural integrity of reactor structural components must be evaluated. This report briefly reviews the effects of radiation on the mechanical properties of metals. Aging degradation studies and reactor operating experience provide a basis for determining the areas where conversion to MOX fuels has the potential to impact the structural integrity of reactor components.

  1. Reactor physics and safety aspects of various design options of a Russian light water reactor with rock-like fuels

    NASA Astrophysics Data System (ADS)

    Bondarenko, A. V.; Komissarov, O. V.; Kozmenkov, Ya. K.; Matveev, Yu. V.; Orekhov, Yu. I.; Pivovarov, V. A.; Sharapov, V. N.

    2003-06-01

    This paper presents results of analytical studies on weapons grade plutonium incineration in VVER (640) medium size light water reactors using a special composition of rock-like fuel (ROX-fuel) to assure spent fuel long-term storage without its reprocessing. The main goal is to achieve high degree of plutonium incineration in once-through cycle. In this paper we considered two fuel compositions. In both compositions weapons grade plutonium is used as fissile material. Spinel (MgAl 2O 4) is used as the 'preserving' material assuring safe storage of the spent fuel. Besides an inert matrix, the option of rock-like fuel with thorium dioxide was studied. One of principal problems in the realization of the proposed approach is the substantial change of properties of the light water reactor core when passing to the use of the ROX-fuel, in particular: (i) due to the absence of 238U the Doppler effect playing a crucial role in reactor's self-regulation and limiting the consequences of reactivity accidents, decreases significantly, (ii) no fuel breeding on one hand, and the quest to attain the maximum plutonium burnup on the other hand, would result in a drastical change of the fuel assembly power during the lifetime and, as a consequence, the rise in irregularity of the power density of fuel assemblies, (iii) both the control rods worth and dissolved boron worth decrease in view of neutron spectrum hardening brought on by the larger absorption cross-section of plutonium as compared to uranium, (iv) βeff is markedly reduced. All these distinctive features are potentially detrimental to the reactor nuclear safety. The principal objective of this work is that to identify a variant of the fuel composition and the reactor layout, which would permit neutralize the negative effect of the above-mentioned distinctive features.

  2. Fuel and cladding nano-technologies based solutions for long life heat-pipe based reactors

    SciTech Connect

    Popa-Simil, L.

    2012-07-01

    A novel nuclear reactor concept, unifying the fuel pipe with fuel tube functionality has been developed. The structure is a quasi-spherical modular reactor, designed for a very long life. The reactor module unifies the fuel tube with the heat pipe and a graphite beryllium reflector. It also uses a micro-hetero-structure that allows the fission products to be removed in the heat pipe flow and deposited in a getter area in the cold zone of the heat pipe, but outside the neutron flux. The reactor operates as a breed and burn reactor - it contains the fuel pipe with a variable enrichment, starting from the hot-end of the pipe, meant to assure the initial criticality, and reactor start-up followed by area with depleted uranium or thorium that get enriched during the consumption of the first part of the enriched uranium. (authors)

  3. Conversion and standardization of US university reactor fuels using LEU, status 1989

    SciTech Connect

    Brown, K.R.; Matos, J.E.; Argonne National Lab., IL )

    1989-01-01

    In 1986, the US Department of Energy initiated a program to change the fuel used in most of the US university research reactors using HEU (93%) to LEU({lt}20{percent}) in order to minimize the risk of theft or diversion of this weapons-useable material. An important consideration in the LEU conversion planning process has been the desire to standardize the fuels that are used and to enhance the performance and utilization of the reactors. This paper describes the current status of this conversion process and the plans and schedules to complete an orderly transition from HEU to LEU fuel in most of these reactors. To date, three university reactors have been converted to LEU fuel, completed safety documentation for three reactors is being evaluated by the USNRC, and work on the safety documentation for six reactors is in progress. 13 refs., 9 tabs.

  4. Light water reactor fuel reprocessing: dissolution studies of voloxidized and nonvoloxidized fuel

    SciTech Connect

    Johnson, D.R.; Stone, J.A.

    1980-04-01

    Small-scale tests with irradiated Zircaloy-clad fuels from Robinson, Oconee, Saxton, and Point Beach reactors with burnups from about 200 to 28,000 MWD/MTHM have been made to determine the dissolution behavior of both voloxidized (U{sub 3}O{sub 8}) and nonvoloxidized (UO{sub 2}) fuel. No significant technical problems were encountered in batch-dissolving of either form. Dissolution rates were well-controlled in all tests. Significant characteristics of U{sub 3}O{sub 8} dissolution that differed from UO{sub 2} dissolution included: (1) reduced tritium and ruthenium ({sup 106}Ru) concentrations in product solutions, (2) increased insoluble noble metal fission product residue (about 2.2X greater), and (3) increased insoluble plutonium in the fission product residue. The insoluble plutonium is easily leached from the residue by 10M HNO{sub 3}. The weight of the fission product residue collected from both U{sub 3}O{sub 8} and UO{sub 2} fuels increased aproximately linearly with fuel burnup. A major fraction (>83%) of the {sup 85}Kr was evolved from U{sub 3}O{sub 8} fuel during dissolution rather than voloxidation. The {sup 85}Kr evolution rate was an appropriate monitor of fuel dissolution rate. Virtually all of the {sup 129}I was evolved by air sparging of the dissolver solution during dissolution. 30 tables, 18 figures.

  5. Assessment of Startup Fuel Options for the GNEP Advanced Burner Reactor (ABR)

    SciTech Connect

    Jon Carmack; Kemal O. Pasamehmetoglu; David Alberstein

    2008-02-01

    The Global Nuclear Energy Program (GNEP) includes a program element for the development and construction of an advanced sodium cooled fast reactor to demonstrate the burning (transmutation) of significant quantities of minor actinides obtained from a separations process and fabricated into a transuranic bearing fuel assembly. To demonstrate and qualify transuranic (TRU) fuel in a fast reactor, an Advanced Burner Reactor (ABR) prototype is needed. The ABR would necessarily be started up using conventional metal alloy or oxide (U or U, Pu) fuel. Startup fuel is needed for the ABR for the first 2 to 4 core loads of fuel in the ABR. Following start up, a series of advanced TRU bearing fuel assemblies will be irradiated in qualification lead test assemblies in the ABR. There are multiple options for this startup fuel. This report provides a description of the possible startup fuel options as well as possible fabrication alternatives available to the program in the current domestic and international facilities and infrastructure.

  6. Development of an Integrated Performance Model for TRISO-Coated Gas Reactor Particle Fuel

    SciTech Connect

    Petti, David Andrew; Miller, Gregory Kent; Martin, David George; Maki, John Thomas

    2005-05-01

    The success of gas reactors depends upon the safety and quality of the coated particle fuel. The understanding and evaluation of this fuel requires development of an integrated mechanistic fuel performance model that fully describes the mechanical and physico-chemical behavior of the fuel particle under irradiation. Such a model, called PARFUME (PARticle Fuel ModEl), is being developed at the Idaho National Engineering and Environmental Laboratory. PARFUME is based on multi-dimensional finite element modeling of TRISO-coated gas reactor fuel. The goal is to represent all potential failure mechanisms and to incorporate the statistical nature of the fuel. The model is currently focused on carbide, oxide nd oxycarbide uranium fuel kernels, while the coating layers are the classical IPyC/SiC/OPyC. This paper reviews the current status of the mechanical aspects of the model and presents results of calculations for irradiations from the New Production Modular High Temperature Gas Reactor program.

  7. Hybrid fusion-fission reactor with a thorium blanket: Its potential in the fuel cycle of nuclear reactors

    NASA Astrophysics Data System (ADS)

    Shmelev, A. N.; Kulikov, G. G.; Kurnaev, V. A.; Salahutdinov, G. H.; Kulikov, E. G.; Apse, V. A.

    2015-12-01

    Discussions are currently going on as to whether it is suitable to employ thorium in the nuclear fuel cycle. This work demonstrates that the 231Pa-232U-233U-Th composition to be produced in the thorium blanket of a hybrid thermonuclear reactor (HTR) as a fuel for light-water reactors opens up the possibility of achieving high, up to 30% of heavy metals (HM), or even ultrahigh fuel burnup. This is because the above fuel composition is able to stabilize its neutron-multiplying properties in the process of high fuel burnup. In addition, it allows the nuclear fuel cycle (NFC) to be better protected against unauthorized proliferation of fissile materials owing to an unprecedentedly large fraction of 232U (several percent!) in the uranium bred from the Th blanket, which will substantially hamper the use of fissile materials in a closed NFC for purposes other than power production.

  8. Hybrid fusion–fission reactor with a thorium blanket: Its potential in the fuel cycle of nuclear reactors

    SciTech Connect

    Shmelev, A. N. Kulikov, G. G. Kurnaev, V. A. Salahutdinov, G. H. Kulikov, E. G. Apse, V. A.

    2015-12-15

    Discussions are currently going on as to whether it is suitable to employ thorium in the nuclear fuel cycle. This work demonstrates that the {sup 231}Pa–{sup 232}U–{sup 233}U–Th composition to be produced in the thorium blanket of a hybrid thermonuclear reactor (HTR) as a fuel for light-water reactors opens up the possibility of achieving high, up to 30% of heavy metals (HM), or even ultrahigh fuel burnup. This is because the above fuel composition is able to stabilize its neutron-multiplying properties in the process of high fuel burnup. In addition, it allows the nuclear fuel cycle (NFC) to be better protected against unauthorized proliferation of fissile materials owing to an unprecedentedly large fraction of {sup 232}U (several percent!) in the uranium bred from the Th blanket, which will substantially hamper the use of fissile materials in a closed NFC for purposes other than power production.

  9. Failed fuel monitoring and surveillance techniques for liquid metal cooled fast reactors

    SciTech Connect

    Lambert, J.D.B.; Mikaili, R.; Gross, K.C.; Strain, R.V.; Aoyama, T.; Ukai, S.; Nomura, S.; Nakae, N.

    1995-05-01

    The Experimental Breeder Reactor II (EBR-II) has been used as a facility for irradiation of LMR fuels and components for thirty years. During this time many tests of experimental fuel were continued to cladding breach in order to study modes of element failure; the methods used to identify such failures are described in a parallel paper. This paper summarizes experience of monitoring the delayed-neutron (DN) and fission-gas (FG) release behavior of a smaller number of elements that continued operation in the run-beyond-cladding-breach (RBCB) mode. The scope of RBCB testing, the methods developed to characterize failures on-line, and examples of DN/FG behavior are described.

  10. Fuel and Additive Characterization for HCCI Combustion

    SciTech Connect

    Aceves, S M; Flowers, D; Martinez-Frias, J; Espinosa-Loza, F; Pitz, W J; Dibble, R

    2003-02-12

    This paper shows a numerical evaluation of fuels and additives for HCCl combustion. First, a long list of candidate HCCl fuels is selected. For all the fuels in the list, operating conditions (compression ratio, equivalence ratio and intake temperature) are determined that result in optimum performance under typical operation for a heavy-duty engine. Fuels are also characterized by presenting Log(p)-Log(T) maps for multiple fuels under HCCl conditions. Log(p)-Log(T) maps illustrate important processes during HCCl engine operation, including compression, low temperature heat release and ignition. Log(p)-Log(T) diagrams can be used for visualizing these processes and can be used as a tool for detailed analysis of HCCl combustion. The paper also includes a ranking of many potential additives. Experiments and analyses have indicated that small amounts (a few parts per million) of secondary fuels (additives) may considerably affect HCCl combustion and may play a significant role in controlling HCCl combustion. Additives are ranked according to their capability to advance HCCl ignition. The best additives are listed and an explanation of their effect on HCCl combustion is included.

  11. Simulated Verification of Fuel Element Inventory in a Small Reactor Core Using the Nuclear Materials Identification System (NMIS)

    SciTech Connect

    Grogan, Brandon R; Mihalczo, John T

    2009-01-01

    The International Panel on Climate Change projects that by 2050 the world energy demand may double. Although the primary focus for new nuclear power plants in industrialized nations is on large plants in the 1000-1600 MWe range, there is an increasing demand for small and medium reactors (SMRs). About half of the innovative SMR concepts are for small (<300 MWe) reactors with a 5-30 year life without on-site refueling. This type of reactor is also known as a battery-type reactor. These reactors are particularly attractive to countries with small power grids and for non-electrical purposes such as heating, hydrogen production, and seawater desalination. Traditionally, this type of reactor has been used in a nautical propulsion role. This type of reactor is designed as a permanently sealed unit to prevent the diversion of the uranium in the core by the user. However, after initial fabrication it will be necessary to verify that the newly fabricated reactor core contains the quantity of uranium that initially entered the fuel fabrication plant. In most instances, traditional inspection techniques can be used to perform this verification, but in certain situations the core design will be considered sensitive. Non-intrusive verification techniques must be utilized in these situations. The Nuclear Materials Identification System (NMIS) with imaging uses active interrogation and a fast time correlation processor to characterize fissile material. The MCNP-PoliMi computer code was used to simulate NMIS measurements of a small, sealed reactor core. Because most battery-type reactor designs are still in the early design phase, a more traditional design based on a Russian icebreaker core was used in the simulations. These simulations show how the radiography capabilities of the NMIS could be used to detect the diversion of fissile material by detecting void areas in the assembled core where fuel elements have been removed.

  12. On the Criticality Safety of Transuranic Sodium Fast Reactor Fuel Transport Casks

    SciTech Connect

    Samuel Bays; Ayodeji Alajo

    2010-05-01

    This work addresses the neutronic performance and criticality safety issues of transport casks for fuel pertaining to low conversion ratio sodium cooled fast reactors, conventionally known as Advanced Burner Reactors. The criticality of a one, three, seven and 19-assembly cask capacity is presented. Both dry “helium” and flooded “water” filled casks are considered. No credit for fuel burnup or fission products was assumed. As many as possible of the conservatisms used in licensing light water reactor universal transport casks were incorporated into this SFR cask criticality design and analysis. It was found that at 7-assemblies or more, adding moderator to the SFR cask increases criticality margin. Also, removal of MAs from the fuel increases criticality margin of dry casks and takes a slight amount of margin away for wet casks. Assuming credit for borated fuel tube liners, this design analysis suggests that as many as 19 assemblies can be loaded in a cask if limited purely by criticality safety. If no credit for boron is assumed, the cask could possibly hold seven assemblies if low conversion ratio fast reactor grade fuel and not breeder reactor grade fuel is assumed. The analysis showed that there is a need for new cask designs for fast reactors spent fuel transportation. There is a potential of modifying existing transportation cask design as the starting point for fast reactor spent fuel transportation.

  13. Physical characteristics of GE (General Electric) BWR (boiling-water reactor) fuel assemblies

    SciTech Connect

    Moore, R.S.; Notz, K.J.

    1989-06-01

    The physical characteristics of fuel assemblies manufactured by the General Electric Company for boiling-water reactors are classified and described. The classification into assembly types is based on the GE reactor product line, the Characteristics Data Base (CDB) assembly class, and the GE fuel design. Thirty production assembly types are identified. Detailed physical data are presented for each assembly type in an appendix. Descriptions of special (nonstandard) fuels are also reported. 52 refs., 1 fig., 6 tabs.

  14. On the flexibility of high temperature reactor cores for high-and low-enriched fuel

    SciTech Connect

    Bzandes, S.; Lonhert, G.

    1982-07-01

    The operational flexibility of a high temperature reactor (HTR) is not restricted to either a low- or a high-enriched fuel cycle. Both fuel cycles are possible for the same core design. The fuel cycle cost is, however, penalized for low-enriched fuel; in addition, higher uranium consumption is required. Hence, an HTR is most economical to operate in the high-enriched thorium-uranium fuel cycle.

  15. Performance of low smeared density sodium-cooled fast reactor metal fuel

    NASA Astrophysics Data System (ADS)

    Porter, D. L.; Chichester, H. J. M.; Medvedev, P. G.; Hayes, S. L.; Teague, M. C.

    2015-10-01

    An experiment was performed in the Experimental Breeder Rector-II (EBR-II) in the 1990s to show that metallic fast reactor fuel could be used in reactors with a single, once-through core. To prove the long duration, high burnup, high neutron exposure capability an experiment where the fuel pin was designed with a very large fission gas plenum and very low fuel smeared density (SD). The experiment, X496, operated to only 8.3 at.% burnup because the EBR-II reactor was scheduled for shut-down at that time. Many of the examinations of the fuel pins only funded recently with the resurgence of reactor designs using very high-burnup fuel. The results showed that, despite the low smeared density of 59% the fuel swelled radially to contact the cladding, fission gas release appeared to be slightly higher than demonstrated in conventional 75%SD fuel tests and axial growth was about the same as 75% SD fuel. There were axial positions in some of the fuel pins which showed evidence of fuel restructuring and an absence of fission products with low melting points and gaseous precursors (Cs and Rb). A model to investigate whether these areas may have overheated due to a loss of bond sodium indicates that it is a possible explanation for the fuel restructuring and something to be considered for fuel performance modeling of low SD fuel.

  16. Performance of low smeared density sodium-cooled fast reactor metal fuel

    SciTech Connect

    Porter, D. L.; H. J. M. Chichester; Medvedev, P. G.; Hayes, S. L.; Teague, M. C.

    2015-06-17

    An experiment was performed in the Experimental Breeder Rector-II (EBR-II) in the 1990s to show that metallic fast reactor fuel could be used in reactors with a single, once-through core. To prove the long duration, high burnup, high neutron exposure capability an experiment where the fuel pin was designed with a very large fission gas plenum and very low fuel smeared density (SD). The experiment, X496, operated to only 8.3 at. % burnup because the EBR-II reactor was scheduled for shut-down at that time. Many of the examinations of the fuel pins only funded recently with the resurgence of reactor designs using very high-burnup fuel. The results showed that, despite the low smeared density of 59% the fuel swelled radially to contact the cladding, fission gas release appeared to be slightly higher than demonstrated in conventional 75%SD fuel tests and axial growth was about the same as 75% SD fuel. There were axial positions in some of the fuel pins which showed evidence of fuel restructuring and an absence of fission products with low metaling points and gaseous precursors (Cs and Rb). Lastly, a model to investigate whether these areas may have overheated due to a loss of bond sodium indicates that it is a possible explanation for the fuel restructuring and something to be considered for fuel performance modeling of low SD fuel.

  17. Comparison of REMIX vs. MOX fuel characteristics in multiple recycling in VVER reactor

    SciTech Connect

    Dekusar, V.M.; Kalashnikov, A.G.; Kapranova, E.N.; Korobitsyn, V.E.; Puzakov, A.Y.

    2013-07-01

    Multiple recycling of regenerated uranium-plutonium fuel in thermal reactors of VVER-1000 type with high enriched uranium feeding (REMIX-fuel) gives a possibility to terminate the accumulation of spent nuclear fuels (SNF) and Pu and decrease the accumulation of irradiated uranium by an order of magnitude. Results of comparison of VVER-1000 nuclear fuel cycle characteristics vs different fuel types such as UOX, MOX and REMIX-fuel have been presented. REMIX fuel (Regenerated Mixture of U-, Pu oxides) is the mixture of plutonium and uranium extracted from SNF and refined from other actinides and fission products with the addition of enriched uranium to provide the power potential necessary. The savings in terms of uranium quantities and separation works in the nuclear energy system (NES) with reactors using REMIX-fuel compared to the NES with uranium-fuelled reactors are shown to be of about 30% and 8%, respectively. For the NES with thermal reactors partially loaded with MOX-fuel, the uranium and separation works saving of about 14% would be obtained. Production of neptunium and americium in reactors with REMIX-fuel in steady state increases by a factor 3, and production of curium - by 10 compared to the reactors with UOX-fuel. This increase of minor actinide buildup is owed to the multiple recycling of plutonium. It should be noted that in this case all fuel assemblies contain high-background plutonium, and their manufacturing involves an expensive technology. Besides, management of REMIX-fuel will require special protection measures even during the fresh fuel manufacturing phase. The above-said gives ground to state that the use of REMIX fuel would be questionable in economic aspect.

  18. N-Reactor (U-metal) Fuel Characteristics for Disposal Criticality Analysis

    SciTech Connect

    Taylor, Larry Lorin

    2000-05-01

    DOE-owned spent nuclear fuels encompass many fuel types. In an effort to facilitate criticality analysis for these various fuel types, they were categorized into nine characteristic fuel groups with emphasis on fuel matrix composition. Out of each fuel group, a representative fuel type was chosen for analysis as a bounding case within that fuel group. Generally, burnup data, fissile enrichments, and total fuel and fissile mass govern the selection of the representative or candidate fuel within that group. Additionally, the criticality analysis will also require data to support design of the canister internals, thermal, and radiation shielding. The purpose of this report is to consolidate and provide in a concise format, material and information/data needed to perform supporting analyses to qualify N-Reactor fuels for acceptance into the designated repository. The N Reactor fuels incorporate zirconium cladding and uranium metal with unique fabrication details in terms of physical size, and method of construction. The fuel construction and post-irradiation handling have created attendant issues relative to cladding failure in the underwater storage environment. These fuels were comprised of low-enriched metal (0.947 to 1.25 wt% 235U) that were originally intended to generate weapons-grade plutonium for national defense. Modifications in subsequent fuel design and changes in the mode of reactor operation in later years were focused more toward power production.

  19. Fuel Burnup and Fuel Pool Shielding Analysis for Bushehr Nuclear Reactor VVER-1000

    NASA Astrophysics Data System (ADS)

    Hadad, Kamal; Ayobian, Navid

    Bushehr Nuclear power plant (BNPP) is currently under construction. The VVER-1000 reactor will be loaded with 126 tons of about 4% enriched fuel having 3-years life cycle. The spent fuel (SF) will be transferred into the spent fuel pool (SPF), where it stays for 8 years before being transferred to Russia. The SPF plays a crucial role during 8 years when the SP resides in there. This paper investigates the shielding of this structure as it is designed to shield the SF radiation. In this study, the SF isotope inventory, for different cycles and with different burnups, was calculated using WIMS/4D transport code. Using MCNP4C nuclear code, the intensity of γ rays was obtained in different layers of SFP shields. These layers include the water above fuel assemblies (FA) in pool, concrete wall of the pool and water laid above transferring fuels. Results show that γ rays leakage from the shield in the mentioned layers are in agreement with the plant's PSAR data. Finally we analyzed an accident were the water height above the FA in the pool drops to 47 cm. In this case it was observed that exposure dose above pool, 10 and 30 days from the accident, are still high and in the levels of 1000 and 758 R/hr.

  20. Transuranic Waste Burning Potential of Thorium Fuel in a Fast Reactor - 12423

    SciTech Connect

    Wenner, Michael; Franceschini, Fausto; Ferroni, Paolo; Sartori, Alberto; Ricotti, Marco

    2012-07-01

    Westinghouse Electric Company (referred to as 'Westinghouse' in the rest of this paper) is proposing a 'back-to-front' approach to overcome the stalemate on nuclear waste management in the US. In this approach, requirements to further the societal acceptance of nuclear waste are such that the ultimate health hazard resulting from the waste package is 'as low as reasonably achievable'. Societal acceptability of nuclear waste can be enhanced by reducing the long-term radiotoxicity of the waste, which is currently driven primarily by the protracted radiotoxicity of the transuranic (TRU) isotopes. Therefore, a transition to a more benign radioactive waste can be accomplished by a fuel cycle capable of consuming the stockpile of TRU 'legacy' waste contained in the LWR Used Nuclear Fuel (UNF) while generating waste which is significantly less radio-toxic than that produced by the current open U-based fuel cycle (once through and variations thereof). Investigation of a fast reactor (FR) operating on a thorium-based fuel cycle, as opposed to the traditional uranium-based is performed. Due to a combination between its neutronic properties and its low position in the actinide chain, thorium not only burns the legacy TRU waste, but it does so with a minimal production of 'new' TRUs. The effectiveness of a thorium-based fast reactor to burn legacy TRU and its flexibility to incorporate various fuels and recycle schemes according to the evolving needs of the transmutation scenario have been investigated. Specifically, the potential for a high TRU burning rate, high U-233 generation rate if so desired and low concurrent production of TRU have been used as metrics for the examined cycles. Core physics simulations of a fast reactor core running on thorium-based fuels and burning an external TRU feed supply have been carried out over multiple cycles of irradiation, separation and reprocessing. The TRU burning capability as well as the core isotopic content have been characterized

  1. Analysis and Development of A Robust Fuel for Gas-Cooled Fast Reactors

    SciTech Connect

    Knight, Travis W

    2010-01-31

    The focus of this effort was on the development of an advanced fuel for gas-cooled fast reactor (GFR) applications. This composite design is based on carbide fuel kernels dispersed in a ZrC matrix. The choice of ZrC is based on its high temperature properties and good thermal conductivity and improved retention of fission products to temperatures beyond that of traditional SiC based coated particle fuels. A key component of this study was the development and understanding of advanced fabrication techniques for GFR fuels that have potential to reduce minor actinide (MA) losses during fabrication owing to their higher vapor pressures and greater volatility. The major accomplishments of this work were the study of combustion synthesis methods for fabrication of the ZrC matrix, fabrication of high density UC electrodes for use in the rotating electrode process, production of UC particles by rotating electrode method, integration of UC kernels in the ZrC matrix, and the full characterization of each component. Major accomplishments in the near-term have been the greater characterization of the UC kernels produced by the rotating electrode method and their condition following the integration in the composite (ZrC matrix) following the short time but high temperature combustion synthesis process. This work has generated four journal publications, one conference proceeding paper, and one additional journal paper submitted for publication (under review). The greater significance of the work can be understood in that it achieved an objective of the DOE Generation IV (GenIV) roadmap for GFR Fuel—namely the demonstration of a composite carbide fuel with 30% volume fuel. This near-term accomplishment is even more significant given the expected or possible time frame for implementation of the GFR in the years 2030 -2050 or beyond.

  2. REACTOR

    DOEpatents

    Roman, W.G.

    1961-06-27

    A pressurized water reactor in which automatic control is achieved by varying the average density of the liquid moderator-cooiant is patented. Density is controlled by the temperature and power level of the reactor ftself. This control can be effected by the use of either plate, pellet, or tubular fuel elements. The fuel elements are disposed between upper and lower coolant plenum chambers and are designed to permit unrestricted coolant flow. The control chamber has an inlet opening communicating with the lower coolant plenum chamber and a restricted vapor vent communicating with the upper coolant plenum chamber. Thus, a variation in temperature of the fuel elements will cause a variation in the average moderator density in the chamber which directly affects the power level of the reactor.

  3. Modeling of the performance of weapons MOX fuel in light water reactors

    SciTech Connect

    Alvis, J.; Bellanger, P.; Medvedev, P.G.; Peddicord, K.L.; Gellene, G.I.

    1999-05-01

    Both the Russian Federation and the US are pursing mixed uranium-plutonium oxide (MOX) fuel in light water reactors (LWRs) for the disposition of excess plutonium from disassembled nuclear warheads. Fuel performance models are used which describe the behavior of MOX fuel during irradiation under typical power reactor conditions. The objective of this project is to perform the analysis of the thermal, mechanical, and chemical behavior of weapons MOX fuel pins under LWR conditions. If fuel performance analysis indicates potential questions, it then becomes imperative to assess the fuel pin design and the proposed operating strategies to reduce the probability of clad failure and the associated release of radioactive fission products into the primary coolant system. Applying the updated code to anticipated fuel and reactor designs, which would be used for weapons MOX fuel in the US, and analyzing the performance of the WWER-100 fuel for Russian weapons plutonium disposition are addressed in this report. The COMETHE code was found to do an excellent job in predicting fuel central temperatures. Also, despite minor predicted differences in thermo-mechanical behavior of MOX and UO{sub 2} fuels, the preliminary estimate indicated that, during normal reactor operations, these deviations remained within limits foreseen by fuel pin design.

  4. Characterization of Suspect Fuel Rod Pieces from the 105 K West Basin

    SciTech Connect

    Delegard, Calvin H.; Schmidt, Andrew J.; Pool, Karl N.; Thornton, Brenda M.

    2006-07-25

    This report provides physical and radiochemical characterization results from examinations and laboratory analyses performed on {approx}0.55-inch diameter rod pieces found in the 105 K West (KW) Basin that were suspected to be from nuclear reactor fuel. The characterization results will be used to establish the technical basis for adding this material to the contents of one of the final Multi-Canister Overpacks (MCOs) that will be loaded out of the KW Basin in late FY2006 or at a later time depending on project priorities. Fifteen fuel rod pieces were found during the clean out of the KW Basin. Based on lack of specific credentials, documentation, or obvious serial numbers, none of the items could be positively identified nor could their sources or compositions be described. Item weights and dimensions measured in the KW Basin indicated densities consistent with the suspect fuel rods containing uranium dioxide (UO2), uranium metal, or being empty. Extensive review of the Hanford Site technical literature led to the postulation that these pieces likely were irradiated test fuel prepared to support of the development of the Hanford ''New Production Reactor'', later called N Reactor. To obtain definitive data on the composition of the suspect fuel, 4 representative fuel rod pieces, with densities corresponding to oxide fuel were selected from the 15 items, and shipped from the KW Basin to the Pacific Northwest National Laboratory's (PNNL) Radiological Processing Laboratory (RPL; also known at the 325 Building) for examinations and characterization. The three fuel rod that were characterized appear to contain slightly irradiated UO2 fuel, originally of natural enrichment, with zirconium cladding. The uranium-235 isotopic concentrations decreased by the irradiation and become slightly lower than the natural enrichment of 0.72% to range from 0.67 to 0.71 atom%. The plutonium concentrations, ranged from about 200 to 470 grams per metric ton of uranium and ranged in

  5. Very High Temperature Reactor (VHTR) Deep Burn Core and Fuel Analysis -- Complete Design Selection for the Pebble Bed Reactor

    SciTech Connect

    B. Boer; A. M. Ougouag

    2010-09-01

    The Deep-Burn (DB) concept focuses on the destruction of transuranic nuclides from used light water reactor fuel. These transuranic nuclides are incorporated into TRISO coated fuel particles and used in gas-cooled reactors with the aim of a fractional fuel burnup of 60 to 70% in fissions per initial metal atom (FIMA). This high performance is expected through the use of multiple recirculation passes of the fuel in pebble form without any physical or chemical changes between passes. In particular, the concept does not call for reprocessing of the fuel between passes. In principle, the DB pebble bed concept employs the same reactor designs as the presently envisioned low-enriched uranium core designs, such as the 400 MWth Pebble Bed Modular Reactor (PBMR-400). Although it has been shown in the previous Fiscal Year (2009) that a PuO2 fueled pebble bed reactor concept is viable, achieving a high fuel burnup, while remaining within safety-imposed prescribed operational limits for fuel temperature, power peaking and temperature reactivity feedback coefficients for the entire temperature range, is challenging. The presence of the isotopes 239-Pu, 240-Pu and 241-Pu that have resonances in the thermal energy range significantly modifies the neutron thermal energy spectrum as compared to a ”standard,” UO2-fueled core. Therefore, the DB pebble bed core exhibits a relatively hard neutron energy spectrum. However, regions within the pebble bed that are near the graphite reflectors experience a locally softer spectrum. This can lead to power and temperature peaking in these regions. Furthermore, a shift of the thermal energy spectrum with increasing temperature can lead to increased absorption in the resonances of the fissile Pu isotopes. This can lead to a positive temperature reactivity coefficient for the graphite moderator under certain operating conditions. The effort of this task in FY 2010 has focused on the optimization of the core to maximize the pebble discharge

  6. Advanced reactors and associated fuel cycle facilities: safety and environmental impacts.

    PubMed

    Hill, R N; Nutt, W M; Laidler, J J

    2011-01-01

    The safety and environmental impacts of new technology and fuel cycle approaches being considered in current U.S. nuclear research programs are contrasted to conventional technology options in this paper. Two advanced reactor technologies, the sodium-cooled fast reactor (SFR) and the very high temperature gas-cooled reactor (VHTR), are being developed. In general, the new reactor technologies exploit inherent features for enhanced safety performance. A key distinction of advanced fuel cycles is spent fuel recycle facilities and new waste forms. In this paper, the performance of existing fuel cycle facilities and applicable regulatory limits are reviewed. Technology options to improve recycle efficiency, restrict emissions, and/or improve safety are identified. For a closed fuel cycle, potential benefits in waste management are significant, and key waste form technology alternatives are described.

  7. Advanced reactors and associated fuel cycle facilities: safety and environmental impacts.

    PubMed

    Hill, R N; Nutt, W M; Laidler, J J

    2011-01-01

    The safety and environmental impacts of new technology and fuel cycle approaches being considered in current U.S. nuclear research programs are contrasted to conventional technology options in this paper. Two advanced reactor technologies, the sodium-cooled fast reactor (SFR) and the very high temperature gas-cooled reactor (VHTR), are being developed. In general, the new reactor technologies exploit inherent features for enhanced safety performance. A key distinction of advanced fuel cycles is spent fuel recycle facilities and new waste forms. In this paper, the performance of existing fuel cycle facilities and applicable regulatory limits are reviewed. Technology options to improve recycle efficiency, restrict emissions, and/or improve safety are identified. For a closed fuel cycle, potential benefits in waste management are significant, and key waste form technology alternatives are described. PMID:21399407

  8. OECD NEA Benchmark Database of Spent Nuclear Fuel Isotopic Compositions for World Reactor Designs

    SciTech Connect

    Gauld, Ian C; Sly, Nicholas C; Michel-Sendis, Franco

    2014-01-01

    Experimental data on the isotopic concentrations in irradiated nuclear fuel represent one of the primary methods for validating computational methods and nuclear data used for reactor and spent fuel depletion simulations that support nuclear fuel cycle safety and safeguards programs. Measurement data have previously not been available to users in a centralized or searchable format, and the majority of accessible information has been, for the most part, limited to light-water-reactor designs. This paper describes a recent initiative to compile spent fuel benchmark data for additional reactor designs used throughout the world that can be used to validate computer model simulations that support nuclear energy and nuclear safeguards missions. Experimental benchmark data have been expanded to include VVER-440, VVER-1000, RBMK, graphite moderated MAGNOX, gas cooled AGR, and several heavy-water moderated CANDU reactor designs. Additional experimental data for pressurized light water and boiling water reactor fuels has also been compiled for modern assembly designs and more extensive isotopic measurements. These data are being compiled and uploaded to a recently revised structured and searchable database, SFCOMPO, to provide the nuclear analysis community with a centrally-accessible resource of spent fuel compositions that can be used to benchmark computer codes, models, and nuclear data. The current version of SFCOMPO contains data for eight reactor designs, 20 fuel assembly designs, more than 550 spent fuel samples, and measured isotopic data for about 80 nuclides.

  9. Irradiation performance of AGR-1 high temperature reactor fuel

    SciTech Connect

    Demkowicz, Paul A.; Hunn, John D.; Ploger, Scott A.; Morris, Robert N.; Baldwin, Charles A.; Harp, Jason M.; Winston, Philip L.; Gerczak, Tyler J.; van Rooyen, Isabella J.; Montgomery, Fred C.; Silva, Chinthaka M.

    2015-10-23

    The AGR-1 experiment contained 72 low-enriched uranium oxide/uranium carbide TRISO coated particle fuel compacts in six capsules irradiated to burnups of 11.2 to 19.6% FIMA, with zero TRISO coating failures detected during the irradiation. The irradiation performance of the fuel including the extent of fission product release and the evolution of kernel and coating microstructures was evaluated based on detailed examination of the irradiation capsules, the fuel compacts, and individual particles. Fractional release of 110mAg from the fuel compacts was often significant, with capsule-average values ranging from 0.01 to 0.38. Analysis of silver release from individual compacts indicated that it was primarily dependent on fuel temperature history. Europium and strontium were released in small amounts through intact coatings, but were found to be significantly retained in the outer pyrocarbon and compact matrix. The capsule-average fractional release from the compacts was 1 × 10–4 to 5 × 10–4 for 154Eu and 8 × 10–7 to 3 × 10–5 for 90Sr. The average 134Cs fractional release from compacts was <3 × 10–6 when all particles maintained intact SiC. An estimated four particles out of 2.98 × 105 in the experiment experienced partial cesium release due to SiC failure during the irradiation, driving 134Cs fractional release in two capsules to approximately 10–5. Identification and characterization of these particles has provided unprecedented insight into the nature and causes of SiC coating failure in high-quality TRISO fuel. In general, changes in coating morphology were found to be dominated by the behavior of the buffer and inner pyrolytic carbon (IPyC), and infrequently observed SiC layer damage was usually related to cracks in the IPyC. Palladium attack of the SiC layer was relatively minor, except for the particles that

  10. Irradiation performance of AGR-1 high temperature reactor fuel

    DOE PAGESBeta

    Demkowicz, Paul A.; Hunn, John D.; Ploger, Scott A.; Morris, Robert N.; Baldwin, Charles A.; Harp, Jason M.; Winston, Philip L.; Gerczak, Tyler J.; van Rooyen, Isabella J.; Montgomery, Fred C.; et al

    2015-10-23

    The AGR-1 experiment contained 72 low-enriched uranium oxide/uranium carbide TRISO coated particle fuel compacts in six capsules irradiated to burnups of 11.2 to 19.6% FIMA, with zero TRISO coating failures detected during the irradiation. The irradiation performance of the fuel including the extent of fission product release and the evolution of kernel and coating microstructures was evaluated based on detailed examination of the irradiation capsules, the fuel compacts, and individual particles. Fractional release of 110mAg from the fuel compacts was often significant, with capsule-average values ranging from 0.01 to 0.38. Analysis of silver release from individual compacts indicated that itmore » was primarily dependent on fuel temperature history. Europium and strontium were released in small amounts through intact coatings, but were found to be significantly retained in the outer pyrocarbon and compact matrix. The capsule-average fractional release from the compacts was 1 × 10–4 to 5 × 10–4 for 154Eu and 8 × 10–7 to 3 × 10–5 for 90Sr. The average 134Cs fractional release from compacts was <3 × 10–6 when all particles maintained intact SiC. An estimated four particles out of 2.98 × 105 in the experiment experienced partial cesium release due to SiC failure during the irradiation, driving 134Cs fractional release in two capsules to approximately 10–5. Identification and characterization of these particles has provided unprecedented insight into the nature and causes of SiC coating failure in high-quality TRISO fuel. In general, changes in coating morphology were found to be dominated by the behavior of the buffer and inner pyrolytic carbon (IPyC), and infrequently observed SiC layer damage was usually related to cracks in the IPyC. Palladium attack of the SiC layer was relatively minor, except for the particles that released cesium during irradiation, where SiC corrosion was found adjacent to IPyC cracks. In conclusion, palladium, silver, and

  11. Review of consequences of uranium hydride formation in N-Reactor fuel elements stored in the K-Basins

    SciTech Connect

    Weber, J.W.

    1994-09-28

    The 105-K Basins on the Hanford site are used to store uranium fuel elements and assemblies irradiated in and discharged from N Reactor. The storage cylinders in KW Basin are known to have some broken N reactor fuel elements in which the exposed uranium is slowly reacting chemically with water in the cylinder. The products of these reactions are uranium oxide, hydrogen, and potentially some uranium hydride. The purpose of this report is to document the results f the latest review of potential, but highly unlikely accidents postulated to occur as closed cylinders containing N reactor fuel assemblies are opened under water in the KW basin and as a fuel assembly is raised from the basin in a shipping cask for transportation to the 327 Building for examination as part of the SNF Characterization Program. The postulated accidents reviews in this report are considered to bound all potential releases of radioactivity and hydrogen. These postulated accidents are: (1) opening and refill of a cylinder containing significant amounts of hydrogen and uranium hydride; and (2) draining of the single element can be used to keep the fuel element submerged in water after the cask containing the can and element is lifted from the KW Basin. Analysis shows the release of radioactivity to the site boundary is significantly less than that allowed by the K Basin Safety Evaluation. Analysis further shows there would be no damage to the K Basin structure nor would there be injury to personnel for credible events.

  12. Conversion of hydrocarbon fuel in thermal protection reactors of hypersonic aircraft

    NASA Astrophysics Data System (ADS)

    Kuranov, A. L.; Mikhaylov, A. M.; Korabelnikov, A. V.

    2016-07-01

    Thermal protection of heat-stressed surfaces of a high-speed vehicle flying in dense layers of atmosphere is one of the topical issues. Not of a less importance is also the problem of hydrocarbon fuel combustion in a supersonic air flow. In the concept under development, it is supposed that in the most high-stressed parts of airframe and engine, catalytic thermochemical reactors will be installed, wherein highly endothermic processes of steam conversion of hydrocarbon fuel take place. Simultaneously with heat absorption, hydrogen generation will occur in the reactors. This paper presents the results of a study of conversion of hydrocarbon fuel in a slit reactor.

  13. Design and Performance of the UO2 Fuel for the Topaz-II Reactor

    NASA Astrophysics Data System (ADS)

    Hoth, Carl W.; Degaltsev, Yuri; Gontar, Alexander; Rakitskaya, Elena

    1994-07-01

    The Russian Topaz-II space reactor system is being modified for use for the United States - Nuclear Electric Propulsion Space Test Project (NEPSTP). The nuclear reactor fuel consists of annular UO2 fuel pellets with 17 mm diameter and 9 mm height. The fuel is fabricated to high purity and high density (96% theoretical density). The reactor core contains 37 single-cell thermionic fuel elements (TFEs), each with approximately 40 fuel pellets. The fuel pellets are contained within an emitter tube of single crystal Mo-3%Nb with a tungsten coating on the outer surface to enhance thermionic emission. The fuel has evolved through an ongoing development program to provide low swelling and very high mechanical strength. Extensive irradiation testing, chemical compatibility testing, and vibration testing were conducted in Russia to verify the performance and lifetime capabilities of the fuel and TFE. This paper summarizes the fuel design, fuel development, and performance testing of the UO2 fuel for the TOPAZ-II reactor.

  14. Advanced Fuels Campaign Light Water Reactor Accident Tolerant Fuel Performance Metrics Executive Summary

    SciTech Connect

    Shannon Bragg-Sitton

    2014-02-01

    Research and development (R&D) activities on advanced, higher performance Light Water Reactor (LWR) fuels have been ongoing for the last few years. Following the unfortunate March 2011 events at the Fukushima Nuclear Power Plant in Japan, the R&D shifted toward enhancing the accident tolerance of LWRs. Qualitative attributes for fuels with enhanced accident tolerance, such as improved reaction kinetics with steam resulting in slower hydrogen generation rate, provide guidance for the design and development of fuels and cladding with enhanced accident tolerance. A common set of technical metrics should be established to aid in the optimization and down selection of candidate designs on a more quantitative basis. “Metrics” describe a set of technical bases by which multiple concepts can be fairly evaluated against a common baseline and against one another. This report describes a proposed technical evaluation methodology that can be applied to evaluate the ability of each concept to meet performance and safety goals relative to the current UO2 – zirconium alloy system and relative to one another. The resultant ranked evaluation can then inform concept down-selection, such that the most promising accident tolerant fuel design option(s) can continue to be developed toward qualification.

  15. Parametric analyses of single-zone thorium-fueled molten salt reactor fuel cycle options

    SciTech Connect

    Powers, J.J.; Worrall, A.; Gehin, J.C.; Harrison, T.J.; Sunny, E.E.

    2013-07-01

    Analyses of fuel cycle options based on thorium-fueled Molten Salt Reactors (MSRs) have been performed in support of fuel cycle screening and evaluation activities for the United States Department of Energy. The MSR options considered are based on thermal spectrum MSRs with 3 different separations levels: full recycling, limited recycling, and 'once-through' operation without active separations. A single-fluid, single-zone 2250 MWth (1000 MWe) MSR concept consisting of a fuel-bearing molten salt with graphite moderator and reflectors was used as the basis for this study. Radiation transport and isotopic depletion calculations were performed using SCALE 6.1 with ENDF/B-VII nuclear data. New methodology developed at Oak Ridge National Laboratory (ORNL) enables MSR analysis using SCALE, modeling material feed and removal by taking user-specified parameters and performing multiple SCALE/TRITON simulations to determine the resulting equilibrium operating conditions. Parametric analyses examined the sensitivity of the performance of a thorium MSR to variations in the separations efficiency for protactinium and fission products. Results indicate that self-sustained operation is possible with full or limited recycling but once-through operation would require an external neutron source. (authors)

  16. Summary engineering description of underwater fuel storage facility for foreign research reactor spent nuclear fuel

    SciTech Connect

    Dahlke, H.J.; Johnson, D.A.; Rawlins, J.K.; Searle, D.K.; Wachs, G.W.

    1994-10-01

    This document is a summary description for an Underwater Fuel Storage Facility (UFSF) for foreign research reactor (FRR) spent nuclear fuel (SNF). A FRR SNF environmental Impact Statement (EIS) is being prepared and will include both wet and dry storage facilities as storage alternatives. For the UFSF presented in this document, a specific site is not chosen. This facility can be sited at any one of the five locations under consideration in the EIS. These locations are the Idaho National Engineering Laboratory, Savannah River Site, Hanford, Oak Ridge National Laboratory, and Nevada Test Site. Generic facility environmental impacts and emissions are provided in this report. A baseline fuel element is defined in Section 2.2, and the results of a fission product analysis are presented. Requirements for a storage facility have been researched and are summarized in Section 3. Section 4 describes three facility options: (1) the Centralized-UFSF, which would store the entire fuel element quantity in a single facility at a single location, (2) the Regionalized Large-UFSF, which would store 75% of the fuel element quantity in some region of the country, and (3) the Regionalized Small-UFSF, which would store 25% of the fuel element quantity, with the possibility of a number of these facilities in various regions throughout the country. The operational philosophy is presented in Section 5, and Section 6 contains a description of the equipment. Section 7 defines the utilities required for the facility. Cost estimates are discussed in Section 8, and detailed cost estimates are included. Impacts to worker safety, public safety, and the environment are discussed in Section 9. Accidental releases are presented in Section 10. Standard Environmental Impact Forms are included in Section 11.

  17. Performance of low smeared density sodium-cooled fast reactor metal fuel

    DOE PAGESBeta

    Porter, D. L.; H. J. M. Chichester; Medvedev, P. G.; Hayes, S. L.; Teague, M. C.

    2015-06-17

    An experiment was performed in the Experimental Breeder Rector-II (EBR-II) in the 1990s to show that metallic fast reactor fuel could be used in reactors with a single, once-through core. To prove the long duration, high burnup, high neutron exposure capability an experiment where the fuel pin was designed with a very large fission gas plenum and very low fuel smeared density (SD). The experiment, X496, operated to only 8.3 at. % burnup because the EBR-II reactor was scheduled for shut-down at that time. Many of the examinations of the fuel pins only funded recently with the resurgence of reactormore » designs using very high-burnup fuel. The results showed that, despite the low smeared density of 59% the fuel swelled radially to contact the cladding, fission gas release appeared to be slightly higher than demonstrated in conventional 75%SD fuel tests and axial growth was about the same as 75% SD fuel. There were axial positions in some of the fuel pins which showed evidence of fuel restructuring and an absence of fission products with low metaling points and gaseous precursors (Cs and Rb). Lastly, a model to investigate whether these areas may have overheated due to a loss of bond sodium indicates that it is a possible explanation for the fuel restructuring and something to be considered for fuel performance modeling of low SD fuel.« less

  18. FLOWSHEET EVALUATION FOR THE DISSOLVING AND NEUTRALIZATION OF SODIUM REACTOR EXPERIMENT USED NUCLEAR FUEL

    SciTech Connect

    Daniel, W. E.; Hansen, E. K.; Shehee, T. C.

    2012-10-30

    This report includes the literature review, hydrogen off-gas calculations, and hydrogen generation tests to determine that H-Canyon can safely dissolve the Sodium Reactor Experiment (SRE; thorium fuel), Ford Nuclear Reactor (FNR; aluminum alloy fuel), and Denmark Reactor (DR-3; silicide fuel, aluminum alloy fuel, and aluminum oxide fuel) assemblies in the L-Bundles with respect to the hydrogen levels in the projected peak off-gas rates. This is provided that the number of L-Bundles charged to the dissolver is controlled. Examination of SRE dissolution for potential issues has aided in predicting the optimal batching scenario. The calculations detailed in this report demonstrate that the FNR, SRE, and DR-3 used nuclear fuel (UNF) are bounded by MURR UNF and may be charged using the controls outlined for MURR dissolution in a prior report.

  19. The evaluation of the use of metal alloy fuels in pressurized water reactors. Final report

    SciTech Connect

    Lancaster, D.

    1992-10-26

    The use of metal alloy fuels in a PWR was investigated. It was found that it would be feasible and competitive to design PWRs with metal alloy fuels but that there seemed to be no significant benefits. The new technology would carry with it added economic uncertainty and since no large benefits were found it was determined that metal alloy fuels are not recommended. Initially, a benefit was found for metal alloy fuels but when the oxide core was equally optimized the benefit faded. On review of the optimization of the current generation of ``advanced reactors,`` it became clear that reactor design optimization has been under emphasized. Current ``advanced reactors`` are severely constrained. The AP-600 required the use of a fuel design from the 1970`s. In order to find the best metal alloy fuel design, core optimization became a central effort. This work is ongoing.

  20. Characterization of the Annular Core Research Reactor (ACRR) Neutron Radiography System Imaging Plane

    NASA Astrophysics Data System (ADS)

    Kaiser, Krista; Chantel Nowlen, K.; DePriest, K. Russell

    2016-02-01

    The Annular Core Research Reactor (ACRR) at Sandia National Laboratories (SNL) is an epithermal pool-type research reactor licensed up to a thermal power of 2.4 MW. The ACRR facility has a neutron radiography facility that is used for imaging a wide range of items including reactor fuel and neutron generators. The ACRR neutron radiography system has four apertures (65:1, 125:1, 250:1, and 500:1) available to experimenters. The neutron flux and spectrum as well as the gamma dose rate were characterized at the imaging plane for the ACRR's neutron radiography system for the 65:1, 125:1 and 250:1 apertures.

  1. Irradiation of the First Advanced Gas Reactor Fuel Development and Qualification Experiment in the Advanced Test Reactor

    SciTech Connect

    S. Blaine Grover; David A. Petti

    2008-10-01

    The United States Department of Energy’s Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating eight separate tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the United States Department of Energy’s lead laboratory for nuclear energy development. These AGR fuel experiments will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The experiments, which will each consist of six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control for each capsule. The swept gas will also have on-line fission product monitoring to track performance of the fuel in each individual capsule during irradiation. The design of the first experiment (designated AGR-1) was completed in 2005, and the fabrication and assembly of the test train as well as the support systems and fission product monitoring system that monitor and control the experiment during irradiation were completed in September 2006. The experiment was inserted in the ATR in December 2006, and is serving as a shakedown test of the multi-capsule experiment design that will be used in the subsequent irradiations as well as a test of the early variants of the fuel produced under this program. The experiment test train as well as the monitoring, control, and data collection systems are discussed and the status of the experiment is provided.

  2. Irradiation performance of AGR-1 high temperature reactor fuel

    SciTech Connect

    Paul A. Demkowicz; John D. Hunn; Robert N. Morris; Charles A. Baldwin; Philip L. Winston; Jason M. Harp; Scott A. Ploger; Tyler Gerczak; Isabella J. van Rooyen; Fred C. Montgomery; Chinthaka M. Silva

    2014-10-01

    The AGR-1 experiment contained 72 low-enriched uranium oxide/uranium carbide TRISO-coated particle fuel compacts in six capsules irradiated to burnups of 11.2 to 19.5% FIMA, with zero TRISO coating failures detected during the irradiation. The irradiation performance of the fuel–including the extent of fission product release and the evolution of kernel and coating microstructures–was evaluated based on detailed examination of the irradiation capsules, the fuel compacts, and individual particles. Fractional release of 110mAg from the fuel compacts was often significant, with capsule-average values ranging from 0.01 to 0.38. Analysis of silver release from individual compacts indicated that it was primarily dependent on fuel temperature history. Europium and strontium were released in small amounts through intact coatings, but were found to be significantly retained in the outer pyrocrabon and compact matrix. The capsule-average fractional release from the compacts was 1×10 4 to 5×10 4 for 154Eu and 8×10 7 to 3×10 5 for 90Sr. The average 134Cs release from compacts was <3×10 6 when all particles maintained intact SiC. An estimated four particles out of 2.98×105 experienced partial cesium release due to SiC failure during the irradiation, driving 134Cs release in two capsules to approximately 10 5. Identification and characterization of these particles has provided unprecedented insight into the nature and causes of SiC coating failure in high-quality TRISO fuel. In general, changes in coating morphology were found to be dominated by the behavior of the buffer and inner pyrolytic carbon (IPyC), and infrequently observed SiC layer damage was usually related to cracks in the IPyC. Palladium attack of the SiC layer was relatively minor, except for the particles that released cesium during irradiation, where SiC corrosion was found adjacent to IPyC cracks. Palladium, silver, and uranium were found in the SiC layer of irradiated particles, and characterization

  3. The Advanced High-Temperature Reactor (AHTR) for Producing Hydrogen to Manufacture Liquid Fuels

    SciTech Connect

    Forsberg, C.W.; Peterson, P.F.; Ott, L.

    2004-10-06

    Conventional world oil production is expected to peak within a decade. Shortfalls in production of liquid fuels (gasoline, diesel, and jet fuel) from conventional oil sources are expected to be offset by increased production of fuels from heavy oils and tar sands that are primarily located in the Western Hemisphere (Canada, Venezuela, the United States, and Mexico). Simultaneously, there is a renewed interest in liquid fuels from biomass, such as alcohol; but, biomass production requires fertilizer. Massive quantities of hydrogen (H2) are required (1) to convert heavy oils and tar sands to liquid fuels and (2) to produce fertilizer for production of biomass that can be converted to liquid fuels. If these liquid fuels are to be used while simultaneously minimizing greenhouse emissions, nonfossil methods for the production of H2 are required. Nuclear energy can be used to produce H2. The most efficient methods to produce H2 from nuclear energy involve thermochemical cycles in which high-temperature heat (700 to 850 C) and water are converted to H2 and oxygen. The peak nuclear reactor fuel and coolant temperatures must be significantly higher than the chemical process temperatures to transport heat from the reactor core to an intermediate heat transfer loop and from the intermediate heat transfer loop to the chemical plant. The reactor temperatures required for H2 production are at the limits of practical engineering materials. A new high-temperature reactor concept is being developed for H2 and electricity production: the Advanced High-Temperature Reactor (AHTR). The fuel is a graphite-matrix, coated-particle fuel, the same type that is used in modular high-temperature gas-cooled reactors (MHTGRs). The coolant is a clean molten fluoride salt with a boiling point near 1400 C. The use of a liquid coolant, rather than helium, reduces peak reactor fuel and coolant temperatures 100 to 200 C relative to those of a MHTGR. Liquids are better heat transfer fluids than gases

  4. Thermal Aspects of Using Alternative Nuclear Fuels in Supercritical Water-Cooled Reactors

    NASA Astrophysics Data System (ADS)

    Grande, Lisa Christine

    A SuperCritical Water-cooled Nuclear Reactor (SCWR) is a Generation IV concept currently being developed worldwide. Unique to this reactor type is the use of light-water coolant above its critical point. The current research presents a thermal-hydraulic analysis of a single fuel channel within a Pressure Tube (PT)-type SCWR with a single-reheat cycle. Since this reactor is in its early design phase many fuel-channel components are being investigated in various combinations. Analysis inputs are: steam cycle, Axial Heat Flux Profile (AHFP), fuel-bundle geometry, and thermophysical properties of reactor coolant, fuel sheath and fuel. Uniform and non-uniform AHFPs for average channel power were applied to a variety of alternative fuels (mixed oxide, thorium dioxide, uranium dicarbide, uranium nitride and uranium carbide) enclosed in an Inconel-600 43-element bundle. The results depict bulk-fluid, outer-sheath and fuel-centreline temperature profiles together with the Heat Transfer Coefficient (HTC) profiles along the heated length of fuel channel. The objective is to identify the best options in terms of fuel, sheath material and AHFPS in which the outer-sheath and fuel-centreline temperatures will be below the accepted temperature limits of 850°C and 1850°C respectively. The 43-element Inconel-600 fuel bundle is suitable for SCWR use as the sheath-temperature design limit of 850°C was maintained for all analyzed cases at average channel power. Thoria, UC2, UN and UC fuels for all AHFPs are acceptable since the maximum fuel-centreline temperature does not exceed the industry accepted limit of 1850°C. Conversely, the fuel-centreline temperature limit was exceeded for MOX at all AHFPs, and UO2 for both cosine and downstream-skewed cosine AHFPs. Therefore, fuel-bundle modifications are required for UO2 and MOX to be feasible nuclear fuels for SCWRs.

  5. MANAGEMENT OF RESEARCH AND TEST REACTOR ALUMINUM SPENT NUCLEAR FUEL - A TECHNOLOGY ASSESSMENT

    SciTech Connect

    Vinson, D.

    2010-07-11

    The Department of Energy's Environmental Management (DOE-EM) Program is responsible for the receipt and storage of aluminum research reactor spent nuclear fuel or used fuel until ultimate disposition. Aluminum research reactor used fuel is currently being stored or is anticipated to be returned to the U.S. and stored at DOE-EM storage facilities at the Savannah River Site and the Idaho Nuclear Technology and Engineering Center. This paper assesses the technologies and the options for safe transportation/receipt and interim storage of aluminum research reactor spent fuel and reviews the comprehensive strategy for its management. The U.S. Department of Energy uses the Appendix A, Spent Nuclear Fuel Acceptance Criteria, to identify the physical, chemical, and isotopic characteristics of spent nuclear fuel to be returned to the United States under the Foreign Research Reactor Spent Nuclear Fuel Acceptance Program. The fuel is further evaluated for acceptance through assessments of the fuel at the foreign sites that include corrosion damage and handleability. Transport involves use of commercial shipping casks with defined leakage rates that can provide containment of the fuel, some of which are breached. Options for safe storage include wet storage and dry storage. Both options must fully address potential degradation of the aluminum during the storage period. This paper focuses on the various options for safe transport and storage with respect to technology maturity and application.

  6. Radiotoxicity Characterization of Multi-Recycled Thorium Fuel - 12394

    SciTech Connect

    Franceschini, F.; Wenner, M.; Fiorina, C.; Huang, M.; Petrovic, B.; Krepel, J.

    2012-07-01

    As described in companion papers, Westinghouse is proposing the implementation of a thorium based fuel cycle to burn the transuranic (TRU) contained in the used nuclear fuel. The potential of thorium as a TRU burner is described in another paper presented at this conference. This paper analyzes the long-term impact of thorium on the front-end and backend of the fuel cycle. This is accomplished by an assessment of the isotopic make-up of Th in a closed cycle and its impact on representative metrics, such as radiotoxicity, decay heat and gamma heat. The behavior in both thermal and fast neutron energy ranges has been investigated. Irradiation in a Th fuel PWR has been assumed as representative of the thermal range, while a Th fuel fast reactor (FR) has been employed to characterize the behavior in the high-energy range. A comparison with a U-fuel closed-cycle FR has been undertaken in an attempt of a more comprehensive evaluation of each cycle's long-term potential. As the Th fuel undergoes multiple cycles of irradiation, the isotopic composition of the recycled fuel changes. Minor Th isotopes are produced; U-232 and Pa-231 build up; the U vector gradually shifts towards increasing amounts of U-234, U-235 etc., eventually leading to the production of non negligible amounts of TRU isotopes, especially Pu-238. The impact of the recycled fuel isotopic makeup on the in-core behavior is mild, and for some aspects beneficial, i.e. the reactivity swing during irradiation is reduced as the fertile characteristics of the fuel increase. On the other hand, the front and the back-end of the fuel cycle are negatively affected due to the presence of Th-228 and U-232 and the build-up of higher actinides (Pu-238 etc.). The presence of U-232 can also be seen as advantageous as it represents an obstacle to potential proliferators. Notwithstanding the increase in the short-term radiotoxicity and decay heat in the multi-recycled fuel, the Th closed cycle has some potentially substantial

  7. Water storage of liquid-metal fast-breeder-reactor fuel

    SciTech Connect

    Meacham, S.A.

    1982-01-01

    The purpose of this paper is to present a general overview of a concept proposed for receiving and storing liquid metal fast breeder reactor (LMFBR) spent fuel. This work was done as part of the Consolidated Fuel Reprocessing Program (CFRP) at the Oak Ridge National Laboratory (ORNL). The CFRP has as its major objective the development of technology for reprocessing advanced nuclear reactor fuels. The program plans that research and development will be carried through to a sufficient scale, using irradiated spent fuel under plant operating conditions, to establish a basis for confident projection of reprocessing capability to support a breeder industry.

  8. PREPARING THE HIGH FLUX ISOTOPE REACTOR FOR CONVERSION TO LOW ENRICHED URANIUM FUEL ? EXTENDING CYCLE BURNUP

    SciTech Connect

    Primm, Trent; Chandler, David

    2009-01-01

    Reactor performance studies have been completed for conceptual plate designs and show that maintaining reactor performance while converting HFIR from high enriched to low enriched uranium (20 wt % 235U) fuel requires extending the end-of-life burnup value for HFIR fuel from the current nominal value of 2200 MWD to 2600 MWD. The current fuel fabrication procedure is discussed and changes that would be required to this procedure are identified. Design and safety related analyses that are required for the certification of a new fuel are identified. Qualification tests and comments regarding the regulatory approval process are provided along with a conceptual schedule.

  9. Micro-reactors for characterization of nanostructure-based sensors

    NASA Astrophysics Data System (ADS)

    Savu, R.; Silveira, J. V.; Flacker, A.; Vaz, A. R.; Joanni, E.; Pinto, A. C.; Gobbi, A. L.; Santos, T. E. A.; Rotondaro, A. L. P.; Moshkalev, S. A.

    2012-05-01

    Fabrication and testing of micro-reactors for the characterization of nanosensors is presented in this work. The reactors have a small volume (100 μl) and are equipped with gas input/output channels. They were machined from a single piece of kovar in order to avoid leaks in the system due to additional welding. The contact pins were electrically insulated from the body of the reactor using a borosilicate sealing glass and the reactor was hermetically sealed using a lid and an elastomeric o-ring. One of the advantages of the reactor lies in its simple assembly and ease of use with any vacuum/gas system, allowing the connection of more than one device. Moreover, the lid can be modified in order to fit a window for in situ optical characterization. In order to prove its versatility, carbon nanotube-based sensors were tested using this micro-reactor. The devices were fabricated by depositing carbon nanotubes over 1 μm thick gold electrodes patterned onto Si/SiO2 substrates. The sensors were tested using oxygen and nitrogen atmospheres, in the pressure range between 10-5 and 10-1 mbar. The small chamber volume allowed the measurement of fast sensor characteristic times, with the sensors showing good sensitivity towards gas and pressure as well as high reproducibility.

  10. Comparison of the radiological hazard of thorium and uranium spent fuels from VVER-1000 reactor

    NASA Astrophysics Data System (ADS)

    Frybort, Jan

    2014-11-01

    Thorium fuel is considered as a viable alternative to the uranium fuel used in the current generation of nuclear power plants. Switch from uranium to thorium means a complete change of composition of the spent nuclear fuel produced as a result of the fuel depletion during operation of a reactor. If the Th-U fuel cycle is implemented, production of minor actinides in the spent fuel is negligible. This is favourable for the spent fuel disposal. On the other hand, thorium fuel utilisation is connected with production of 232U, which decays via several alpha decays into a strong gamma emitter 208Tl. Presence of this nuclide might complicate manipulations with the irradiated thorium fuel. Monte-Carlo computation code MCNPX can be used to simulate thorium fuel depletion in a VVER-1000 reactor. The calculated actinide composition will be analysed and dose rate from produced gamma radiation will be calculated. The results will be compared to the reference uranium fuel. Dependence of the dose rate on time of decay after the end of irradiation in the reactor will be analysed. This study will compare the radiological hazard of the spent thorium and uranium fuel handling.

  11. Completion of the first NGNP Advanced Gas Reactor Fuel Irradiation Experiment, AGR-1, in the Advanced Test Reactor

    SciTech Connect

    Blaine Grover; John Maki; David Petti

    2010-10-01

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating up to seven separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the United States Department of Energy’s lead laboratory for nuclear energy development. The ATR is one of the world’s premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States, and will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of at least six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and completed a very successful irradiation in early November 2009. The design of AGR-1 test train and support systems used to monitor and control the experiment during

  12. Assemblies with both target and fuel pins in an isotope-production reactor

    DOEpatents

    Cawley, W.E.; Omberg, R.P.

    1982-08-19

    A method is described for producing tritium in a fast breeder reactor cooled with liquid metal. Lithium target material is placed in pins adjacent to fuel pins in order to increase the tritium production rate.

  13. Environmental data energy technology characterizations: synthetic fuels

    SciTech Connect

    Not Available

    1980-04-01

    Environmental Data Energy Technology Characterizations are publications which are intended to provide policy analysts and technical analysts with basic environmental data associated with key energy technologies. This publication provides documentation on synthetic fuels (coal-derived and oil shale). The transformation of the energy in coal and oil shale into a more useful form is described in this publication in terms of major activity areas in the synthetic fuel cycles, that is, in terms of activities which produce either an energy product or a fuel leading to the production of an energy product in a different form. The activities discussed in this document are coal liquefaction, coal gasification, in-situ gasification, and oil shales. These activities represent both well-documented and advanced activity areas. The former activities are characterized in terms of actual operating data with allowance for future modification where appropriate. Emissions are assumed to conform to environmental standards. The advanced activity areas examined are those like coal liquefaction and in-situ retorting of oil shale. For these areas, data from pilot or demonstration plants were used where available; otherwise, engineering studies provided the data. The organization of the chapters in this volume is designed to support the tabular presentation in the summary volume. Each chapter begins with a brief description of the activity under consideration. The standard characteristics, size, availability, mode of functioning and place in the fuel cycle are presented. Next, major legislative and/or technological factors influencing the commercial operation of the activity are offered. Discussions of resources consumed, residuals produced, and economics follow. To aid in comparing and linking the different activity areas, data for each area are normalized to 10/sup 12/ Btu of energy output from the activity.

  14. Radio-toxicity of spent fuel of the advanced heavy water reactor.

    PubMed

    Anand, S; Singh, K D S; Sharma, V K

    2010-01-01

    The Advanced Heavy Water Reactor (AHWR) is a new power reactor concept being developed at Bhabha Atomic Research Centre, Mumbai. The reactor retains many desirable features of the existing Pressurised Heavy Water Reactor (PHWR), while incorporating new, advanced safety features. The reactor aims to utilise the vast thorium resources available in India. The reactor core will use plutonium as the make-up fuel, while breeding (233)U in situ. On account of this unique combination of fuel materials, the operational characteristics of the fuel as determined by its radioactivity, decay heat and radio-toxicity are being viewed with great interest. Radio-toxicity of the spent fuel is a measure of potential radiological hazard to the members of the public and also important from the ecological point of view. The radio-toxicity of the AHWR fuel is extremely high to start with, being approximately 10(4) times that of the fresh natural U fuel used in a PHWR, and continues to remain relatively high during operation and subsequent cooling. A unique feature of this fuel is the peak observed in its radio-toxicity at approximately 10(5) y of decay cooling. The delayed increase in fuel toxicity has been traced primarily to a build-up of (229)Th, (230)Th and (226)Ra. This phenomenon has been observed earlier for thorium-based fuels and is confirmed for the AHWR fuel. This paper presents radio-toxicity data for AHWR spent fuel up to a period of 10(6) y and the results are compared with the radio-toxicity of PHWR. PMID:19776247

  15. Radio-toxicity of spent fuel of the advanced heavy water reactor.

    PubMed

    Anand, S; Singh, K D S; Sharma, V K

    2010-01-01

    The Advanced Heavy Water Reactor (AHWR) is a new power reactor concept being developed at Bhabha Atomic Research Centre, Mumbai. The reactor retains many desirable features of the existing Pressurised Heavy Water Reactor (PHWR), while incorporating new, advanced safety features. The reactor aims to utilise the vast thorium resources available in India. The reactor core will use plutonium as the make-up fuel, while breeding (233)U in situ. On account of this unique combination of fuel materials, the operational characteristics of the fuel as determined by its radioactivity, decay heat and radio-toxicity are being viewed with great interest. Radio-toxicity of the spent fuel is a measure of potential radiological hazard to the members of the public and also important from the ecological point of view. The radio-toxicity of the AHWR fuel is extremely high to start with, being approximately 10(4) times that of the fresh natural U fuel used in a PHWR, and continues to remain relatively high during operation and subsequent cooling. A unique feature of this fuel is the peak observed in its radio-toxicity at approximately 10(5) y of decay cooling. The delayed increase in fuel toxicity has been traced primarily to a build-up of (229)Th, (230)Th and (226)Ra. This phenomenon has been observed earlier for thorium-based fuels and is confirmed for the AHWR fuel. This paper presents radio-toxicity data for AHWR spent fuel up to a period of 10(6) y and the results are compared with the radio-toxicity of PHWR.

  16. A Blueprint for GNEP Advanced Burner Reactor Startup Fuel Fabrication Facility

    SciTech Connect

    S. Khericha

    2010-12-01

    The purpose of this article is to identify the requirements and issues associated with design of GNEP Advanced Burner Reactor Fuel Facility. The report was prepared in support of providing data for preparation of a NEPA Environmental Impact Statement in support the U. S. Department of Energy (DOE) Global Nuclear Energy Partnership (GNEP). One of the GNEP objectives was to reduce the inventory of long lived actinide from the light water reactor (LWR) spent fuel. The LWR spent fuel contains Plutonium (Pu) -239 and other transuranics (TRU) such as Americium-241. One of the options is to transmute or burn these actinides in fast neutron spectra as well as generate the electricity. A sodium-cooled Advanced Recycling Reactor (ARR) concept was proposed to achieve this goal. However, fuel with relatively high TRU content has not been used in the fast reactor. To demonstrate the utilization of TRU fuel in a fast reactor, an Advanced Burner Reactor (ABR) prototype of ARR was proposed, which would necessarily be started up using weapons grade (WG) Pu fuel. The WG Pu is distinguished by relatively highest proportions of Pu-239 and lesser amount of other actinides. The WG Pu was assumed to be used as the startup fuel along with TRU fuel in lead test assemblies. Because such fuel is not currently being produced in the US, a new facility (or new capability in an existing facility) was being considered for fabrication of WG Pu fuel for the ABR. It was estimated that the facility will provide the startup fuel for 10-15 years and would take 3 to 5 years to construct.

  17. Annular core liquid-salt cooled reactor with multiple fuel and blanket zones

    DOEpatents

    Peterson, Per F.

    2013-05-14

    A liquid fluoride salt cooled, high temperature reactor having a reactor vessel with a pebble-bed reactor core. The reactor core comprises a pebble injection inlet located at a bottom end of the reactor core and a pebble defueling outlet located at a top end of the reactor core, an inner reflector, outer reflector, and an annular pebble-bed region disposed in between the inner reflector and outer reflector. The annular pebble-bed region comprises an annular channel configured for receiving pebble fuel at the pebble injection inlet, the pebble fuel comprising a combination of seed and blanket pebbles having a density lower than the coolant such that the pebbles have positive buoyancy and migrate upward in said annular pebble-bed region toward the defueling outlet. The annular pebble-bed region comprises alternating radial layers of seed pebbles and blanket pebbles.

  18. Multiple recycle of REMIX fuel based on reprocessed uranium and plutonium mixture in thermal reactors

    SciTech Connect

    Fedorov, Y.S.; Bibichev, B.A.; Zilberman, B.Y.; Baryshnikov, M.V.; Kryukov, O.V.; Khaperskaya, A.V.

    2013-07-01

    REMIX fuel consumption in WWER-1000 is considered. REMIX fuel is fabricated from non-separated mixture of uranium and plutonium obtained during NPP spent fuel reprocessing with further makeup by enriched natural uranium. It makes possible to recycle several times the total amount of uranium and plutonium obtained from spent fuel with 100% loading of the WWER-1000 core. The stored SNF could be also involved in REMIX fuel cycle by enrichment of regenerated uranium. The same approach could be applied to closing the fuel cycle of CANDU reactors. (authors)

  19. Environmental Assessment: Relocation and storage of TRIGA{reg_sign} reactor fuel, Hanford Site, Richland, Washington

    SciTech Connect

    1995-08-01

    In order to allow the shutdown of the Hanford 308 Building in the 300 Area, it is proposed to relocate fuel assemblies (101 irradiated, three unirradiated) from the Mark I TRIGA Reactor storage pool. The irradiated fuel assemblies would be stored in casks in the Interim Storage Area in the Hanford 400 Area; the three unirradiated ones would be transferred to another TRIGA reactor. The relocation is not expected to change the offsite exposure from all Hanford Site 300 and 400 Area operations.

  20. Thorium: Uranium fuel cycle in safe reactors, the time is now

    SciTech Connect

    Gat, Uri

    1995-12-31

    The thorium-uranium fuel cycle has several advantages that make it attractive. Some of these beneficial properties are of particular interest now as they help alleviate current concerns. The Th-U cycle has neutronic advantages when utilized in thermal or epithermal reactors. Some of these reactors enjoy extraordinary safety qualities. The combination of these traits suggest that now is an appropriate time to deploy and begin exploiting the Th-U fuel cycle.

  1. Combustion and fuel characterization of coal-water fuels

    SciTech Connect

    Chow, O.K.; Patel, R.L.; Levasseur, A.A.

    1987-07-01

    Pittsburgh Energy Technology Center (PETC) of the Department of Energy initiated a comprehensive effort in 1982 to develop the necessary performance and cost data and to assess the commercial viability of coal-water fuels (CWFs) as applied to representative utility and industrial units. The effort comprised six tasks beginning with coal resource evaluation and culminating in the assessment of the technical and economic consequences of switching representative commercial units from oil to state-of-the-art CWF firing. Extensive bench, pilot and commercial-scale tests were performed to develop necessary CWF combustion and fireside performance data for the subsequent boiler performance analyses and retrofit cost estimates. Discussions on transport, rheology, combustion properties, and ash characterization are included. 11 refs., 9 figs., 7 tabs.

  2. Thorium plutonium (TREX) fuel for weapons-grade plutonium disposition in pressurized water reactors

    SciTech Connect

    Comfort, C.; Ferguson, C.; Klima, S.; Lilly, D.E.; Rahnema, F.

    1996-12-31

    The goal of this study was to create a pressurized water reactor (PWR) reactor assembly (17 x 17) that would burn weapons-grade plutonium (WGP). Current designs of mixed-oxide (MOX) fuels combine WGP with uranium as the fuel. MOX fuel assemblies will destroy plutonium, but only 40 to 50% of the plutonium present in the fuel. This percentage is limited by the presence of {sup 238}U in the core, which becomes {sup 239}Pu by absorption and decay. The production of plutonium counteracts the disposition of WGP in current MOX fuel designs. This problem can be overcome by replacing the uranium in a MOX design with thorium. This loss of uranium (primarily {sup 238}U) halts the production of {sup 239}Pu in the thorium plutonium (TREX) fuel. The absence of {sup 239}Pu production allows the TREX design to burn up to 85 wt% of the {sup 239}Pu, originally loaded in the fuel.

  3. A long term radiological risk model for plutonium-fueled and fission reactor space nuclear system

    SciTech Connect

    Bartram, B.W.; Dougherty, D.K.

    1987-01-01

    This report describes the optimization of the RISK III mathematical model, which provides risk assessment for the use of a plutonium-fueled, fission reactor in space systems. The report discusses possible scenarios leading to radiation releases on the ground; distinctions are made for an intact reactor and a dispersed reactor. Also included are projected dose equivalents for various accident situations. 54 refs., 31 figs., 11 tabs. (TEM)

  4. Microstructural Characterization of the U-9.1Mo Fuel/AA6061 Cladding Interface in Friction-Bonded Monolithic Fuel Plates Irradiated in the RERTR-6 Experiment

    SciTech Connect

    Keiser, Dennis D.; Jue, Jan-Fong; Miller, Brandon; Gan, Jian; Robinson, Adam; Medvedev, Pavel; Madden, James; Wachs, Dan; Clark, Curtis; Meyer, Mitch

    2015-09-03

    Low-enrichment (U-235 < 20%) U-Mo monolithic fuel is being developed for use in research and test reactors. The earliest design for this fuel that was investigated via reactor testing was comprised of a nominally U-10Mo fuel foil encased in AA6061 (Al-6061) cladding. For a fuel design to be deemed adequate for final use in a reactor, it must maintain dimensional stability and retain fission products throughout irradiation, which means that there must be good integrity at the fuel foil/cladding interface. To investigate the nature of the fuel/cladding interface for this fuel type after irradiation, fuel plates that were tested in INL's Advanced Test Reactor (ATR) were subsequently characterized using optical metallography, scanning electron microscopy, and transmission electron microscopy. Results of this characterization showed that the fuel/cladding interaction layers present at the U-Mo fuel/AA6061 cladding interface after fabrication became amorphous during irradiation. Up to two main interaction layers, based on composition, could be found at the fuel/cladding interface, depending on location. After irradiation, an Al-rich layer contained very few fission gas bubbles, but did exhibit Xe enrichment near the AA6061 cladding interface. Another layer, which contained more Si, had more observable fission gas bubbles. Adjacent to the AA6061 cladding were Mg-rich precipitates, which was in close proximity to the region where Xe is observed to be enriched. In samples produced using a focused ion beam at the interaction zone/AA6061 cladding interface were possible indications of porosity/debonding, which suggested that the interface in this location is relatively weak.

  5. Microstructural Characterization of the U-9.1Mo Fuel/AA6061 Cladding Interface in Friction-Bonded Monolithic Fuel Plates Irradiated in the RERTR-6 Experiment

    DOE PAGESBeta

    Keiser, Dennis D.; Jue, Jan-Fong; Miller, Brandon; Gan, Jian; Robinson, Adam; Medvedev, Pavel; Madden, James; Wachs, Dan; Clark, Curtis; Meyer, Mitch

    2015-09-03

    Low-enrichment (U-235 < 20%) U-Mo monolithic fuel is being developed for use in research and test reactors. The earliest design for this fuel that was investigated via reactor testing was comprised of a nominally U-10Mo fuel foil encased in AA6061 (Al-6061) cladding. For a fuel design to be deemed adequate for final use in a reactor, it must maintain dimensional stability and retain fission products throughout irradiation, which means that there must be good integrity at the fuel foil/cladding interface. To investigate the nature of the fuel/cladding interface for this fuel type after irradiation, fuel plates that were tested inmore » INL's Advanced Test Reactor (ATR) were subsequently characterized using optical metallography, scanning electron microscopy, and transmission electron microscopy. Results of this characterization showed that the fuel/cladding interaction layers present at the U-Mo fuel/AA6061 cladding interface after fabrication became amorphous during irradiation. Up to two main interaction layers, based on composition, could be found at the fuel/cladding interface, depending on location. After irradiation, an Al-rich layer contained very few fission gas bubbles, but did exhibit Xe enrichment near the AA6061 cladding interface. Another layer, which contained more Si, had more observable fission gas bubbles. Adjacent to the AA6061 cladding were Mg-rich precipitates, which was in close proximity to the region where Xe is observed to be enriched. In samples produced using a focused ion beam at the interaction zone/AA6061 cladding interface were possible indications of porosity/debonding, which suggested that the interface in this location is relatively weak.« less

  6. A document review to characterize Atomic International SNAP fuels shipped to INEL 1966--1973

    SciTech Connect

    Wahnschaffe, S.D.; Lords, R.E.; Kneff, D.W.; Nagel, W.E.; Pearlman, H.; Schaubert, V.J.

    1995-09-01

    This report provides the results of a document search and review study to obtain information on the spent fuels for the following six Nuclear Auxiliary Power (SNAP) reactor cores now stored at the Idaho National Engineering Laboratory (INEL): SNAP-2 Experimental Reactor, SNAP-2 Development Reactor, SNAP-10A Ground Test Reactor, SNAP-8 Experimental Reactor, SNAP-8 Development Reactor, and Shield Test Reactor. The report also covers documentation on SNAP fuel materials from four in-pile materials tests: NAA-82-1, NAA-115-2, NAA-117-1, and NAA-121. Pieces of these fuel materials are also stored at INEL as part of the SNAP fuel shipments.

  7. Fuel-Cycle and Nuclear Material Disposition Issues Associated with High-Temperature Gas Reactors

    SciTech Connect

    Shropshire, D.E.; Herring, J.S.

    2004-10-03

    The objective of this paper is to facilitate a better understanding of the fuel-cycle and nuclear material disposition issues associated with high-temperature gas reactors (HTGRs). This paper reviews the nuclear fuel cycles supporting early and present day gas reactors, and identifies challenges for the advanced fuel cycles and waste management systems supporting the next generation of HTGRs, including the Very High Temperature Reactor, which is under development in the Generation IV Program. The earliest gas-cooled reactors were the carbon dioxide (CO2)-cooled reactors. Historical experience is available from over 1,000 reactor-years of operation from 52 electricity-generating, CO2-cooled reactor plants that were placed in operation worldwide. Following the CO2 reactor development, seven HTGR plants were built and operated. The HTGR came about from the combination of helium coolant and graphite moderator. Helium was used instead of air or CO2 as the coolant. The helium gas has a significant technical base due to the experience gained in the United States from the 40-MWe Peach Bottom and 330-MWe Fort St. Vrain reactors designed by General Atomics. Germany also built and operated the 15-MWe Arbeitsgemeinschaft Versuchsreaktor (AVR) and the 300-MWe Thorium High-Temperature Reactor (THTR) power plants. The AVR, THTR, Peach Bottom and Fort St. Vrain all used fuel containing thorium in various forms (i.e., carbides, oxides, thorium particles) and mixtures with highly enriched uranium. The operational experience gained from these early gas reactors can be applied to the next generation of nuclear power systems. HTGR systems are being developed in South Africa, China, Japan, the United States, and Russia. Elements of the HTGR system evaluated included fuel demands on uranium ore mining and milling, conversion, enrichment services, and fuel fabrication; fuel management in-core; spent fuel characteristics affecting fuel recycling and refabrication, fuel handling, interim

  8. 10 CFR 73.35 - Requirements for physical protection of irradiated reactor fuel (100 grams or less) in transit.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... 10 Energy 2 2014-01-01 2014-01-01 false Requirements for physical protection of irradiated reactor... Transit § 73.35 Requirements for physical protection of irradiated reactor fuel (100 grams or less) in... quantity of irradiated reactor fuel weighing 100 grams (0.22 pounds) or less in net weight of...

  9. The behaviour of transuranic mixed oxide fuel in a Candu-900 reactor

    SciTech Connect

    Morreale, A. C.; Ball, M. R.; Novog, D. R.; Luxat, J. C.

    2012-07-01

    The production of transuranic actinide fuels for use in current thermal reactors provides a useful intermediary step in closing the nuclear fuel cycle. Extraction of actinides reduces the longevity, radiation and heat loads of spent material. The burning of transuranic fuels in current reactors for a limited amount of cycles reduces the infrastructure demand for fast reactors and provides an effective synergy that can result in a reduction of as much as 95% of spent fuel waste while reducing the fast reactor infrastructure needed by a factor of almost 13.5 [1]. This paper examines the features of actinide mixed oxide fuel, TRUMOX, in a CANDU{sup R}* nuclear reactor. The actinide concentrations used were based on extraction from 30 year cooled spent fuel and mixed with natural uranium in 3.1 wt% actinide MOX fuel. Full lattice cell modeling was performed using the WIMS-AECL code, super-cell calculations were analyzed in DRAGON and full core analysis was executed in the RFSP 2-group diffusion code. A time-average full core model was produced and analyzed for reactor coefficients, reactivity device worth and online fuelling impacts. The standard CANDU operational limits were maintained throughout operations. The TRUMOX fuel design achieved a burnup of 27.36 MWd/kg HE. A full TRUMOX fuelled CANDU was shown to operate within acceptable limits and provided a viable intermediary step for burning actinides. The recycling, reprocessing and reuse of spent fuels produces a much more sustainable and efficient nuclear fuel cycle. (authors)

  10. The DOE Advanced Gas Reactor (AGR) Fuel Development and Qualification Program

    SciTech Connect

    David Petti; Hans Gougar; Gary Bell

    2005-05-01

    The Department of Energy has established the Advanced Gas Reactor Fuel Development and Qualification Program to address the following overall goals: Provide a baseline fuel qualification data set in support of the licensing and operation of the Next Generation Nuclear Plant (NGNP). Gas-reactor fuel performance demonstration and qualification comprise the longest duration research and development (R&D) task for the NGNP feasibility. The baseline fuel form is to be demonstrated and qualified for a peak fuel centerline temperature of 1250°C. Support near-term deployment of an NGNP by reducing market entry risks posed by technical uncertainties associated with fuel production and qualification. Utilize international collaboration mechanisms to extend the value of DOE resources. The Advanced Gas Reactor Fuel Development and Qualification Program consists of five elements: fuel manufacture, fuel and materials irradiations, postirradiation examination (PIE) and safety testing, fuel performance modeling, and fission product transport and source term evaluation. An underlying theme for the fuel development work is the need to develop a more complete fundamental understanding of the relationship between the fuel fabrication process, key fuel properties, the irradiation performance of the fuel, and the release and transport of fission products in the NGNP primary coolant system. Fuel performance modeling and analysis of the fission product behavior in the primary circuit are important aspects of this work. The performance models are considered essential for several reasons, including guidance for the plant designer in establishing the core design and operating limits, and demonstration to the licensing authority that the applicant has a thorough understanding of the in-service behavior of the fuel system. The fission product behavior task will also provide primary source term data needed for licensing. An overview of the program and recent progress will be presented.

  11. Ambient Laboratory Coater for Advanced Gas Reactor Fuel Development

    SciTech Connect

    Duane D. Bruns; Robert M. Counce; Irma D. Lima Rojas

    2010-06-09

    this research is targeted at developing improved experimentally-based scaling relationships for the hydrodynamics of shallow, gas-spouted beds of dense particles. The work is motivated by the need to more effctively scale up shallow spouted beds used in processes such as in the coating of nuclear fuel particles where precise control of solids and gas circulation is critically important. Experimental results reported here are for a 50 mm diameter spouted bed containing two different types of bed solids (alumina and zirconia) at different static bed depths and fluidized by air and helium. Measurements of multiple local average pressures, inlet gas pressure fluctuations, and spout height were used to characterize the bed hydrodynamics for each operating condition. Follow-on studies are planned that include additional variations in bed size, particle properties, and fluidizing gas. The ultimate objective is to identify the most important non-dimensional hydrodynamic scaling groups and possible spouted-bed design correlations based on these groups.

  12. Use of freeze-casting in advanced burner reactor fuel design

    SciTech Connect

    Lang, A. L.; Yablinsky, C. A.; Allen, T. R.; Burger, J.; Hunger, P. M.; Wegst, U. G. K.

    2012-07-01

    This paper will detail the modeling of a fast reactor with fuel pins created using a freeze-casting process. Freeze-casting is a method of creating an inert scaffold within a fuel pin. The scaffold is created using a directional solidification process and results in open porosity for emplacement of fuel, with pores ranging in size from 300 microns to 500 microns in diameter. These pores allow multiple fuel types and enrichments to be loaded into one fuel pin. Also, each pore could be filled with varying amounts of fuel to allow for the specific volume of fission gases created by that fuel type. Currently fast reactors, including advanced burner reactors (ABR's), are not economically feasible due to the high cost of operating the reactors and of reprocessing the fuel. However, if the fuel could be very precisely placed, such as within a freeze-cast scaffold, this could increase fuel performance and result in a valid design with a much lower cost per megawatt. In addition to competitive costs, freeze-cast fuel would also allow for selective breeding or burning of actinides within specific locations in fast reactors. For example, fast flux peak locations could be utilized on a minute scale to target specific actinides for transmutation. Freeze-cast fuel is extremely flexible and has great potential in a variety of applications. This paper performs initial modeling of freeze-cast fuel, with the generic fast reactor parameters for this model based on EBR-II. The core has an assumed power of 62.5 MWt. The neutronics code used was Monte Carlo N-Particle (MCNP5) transport code. Uniform pore sizes were used in increments of 100 microns. Two different freeze-cast scaffold materials were used: ceramic (MgO-ZrO{sub 2}) and steel (SS316L). Separate models were needed for each material because the freeze-cast ceramic and metal scaffolds have different structural characteristics and overall porosities. Basic criticality results were compiled for the various models. Preliminary

  13. Drop-in capsule testing of plutonium-based fuels in the Advanced Test Reactor

    SciTech Connect

    Chang, G.S.; Ryskamp, J.M.; Terry, W.K.; Ambrosek, R.G.; Palmer, A.J.; Roesener, R.A.

    1996-09-01

    The most attractive way to dispose of weapons-grade plutonium (WGPu) is to use it as fuel in existing light water reactors (LWRs) in the form of mixed oxide (MOX) fuel - i.e., plutonia (PuO[sub 2]) mixed with urania (UO[sub 2]). Before U.S. reactors could be used for this purpose, their operating licenses would have to be amended. Numerous technical issues must be resolved before LWR operating licenses can be amended to allow the use of MOX fuel. The proposed weapons-grade MOX fuel is unusual, even relative to ongoing foreign experience with reactor-grade MOX power reactor fuel. Some demonstration of the in- reactor thermal, mechanical, and fission gas release behavior of the prototype fuel will most likely be required in a limited number of test reactor irradiations. The application to license operation with MOX fuel must be amply supported by experimental data. The Advanced Test Reactor (ATR) at the Idaho National Engineering Laboratory (INEL) is capable of playing a key role in the irradiation, development, and licensing of these new fuel types. The ATR is a 250- MW (thermal) LWR designed to study the effects of intense radiation on reactor fuels and materials. For 25 years, the primary role of the ATR has been to serve in experimental investigations for the development of advanced nuclear fuels. Both large- and small-volume test positions in the ATR could be used for MOX fuel irradiation. The ATR would be a nearly ideal test bed for developing data needed to support applications to license LWRs for operation with MOX fuel made from weapons-grade plutonium. Furthermore, these data can be obtained more quickly by using ATR instead of testing in a commercial LWR. Our previous work in this area has demonstrated that it is technically feasible to perform MOX fuel testing in the ATR. This report documents our analyses of sealed drop-in capsules containing plutonium-based test specimens placed in various ATR positions.

  14. PRELIMINARY DATA CALL REPORT ADVANCED BURNER REACTOR START UP FUEL FABRICATION FACILITY

    SciTech Connect

    S. T. Khericha

    2007-04-01

    The purpose of this report is to provide data for preparation of a NEPA Environmental Impact Statement in support the U. S. Department of Energy (DOE) Global Nuclear Energy Partnership (GNEP). One of the GNEP objectives is to reduce the inventory of long lived actinide from the light water reactor (LWR) spent fuel. The LWR spent fuel contains Plutonium (Pu) -239 and other transuranics (TRU) such as Americium-241. One of the options is to transmute or burn these actinides in fast neutron spectra as well as generate the electricity. A sodium-cooled Advanced Recycling Reactor (ARR) concept has been proposed to achieve this goal. However, fuel with relatively high TRU content has not been used in the fast reactor. To demonstrate the utilization of TRU fuel in a fast reactor, an Advanced Burner Reactor (ABR) prototype of ARR is proposed, which would necessarily be started up using weapons grade (WG) Pu fuel. The WG Pu is distinguished by relatively highest proportions of Pu-239 and lesser amount of other actinides. The WG Pu will be used as the startup fuel along with TRU fuel in lead test assemblies. Because such fuel is not currently being produced in the US, a new facility (or new capability in an existing facility) is being considered for fabrication of WG Pu fuel for the ABR. This report is provided in response to ‘Data Call’ for the construction of startup fuel fabrication facility. It is anticipated that the facility will provide the startup fuel for 10-15 years and will take to 3 to 5 years to construct.

  15. Design of small gas cooled fast reactor with two region of natural Uranium fuel fraction

    NASA Astrophysics Data System (ADS)

    Ariani, Menik; Su'ud, Zaki; Waris, Abdul; Khairurrijal, Monado, Fiber; Sekimoto, Hiroshi; Nakayama, Sinsuke

    2012-06-01

    A design study of small Gas Cooled Fast Reactor with two region fuel has been performed. In this study, design GCFR with Helium coolant which can be continuously operated by supplying mixed Natural Uranium without fuel enrichment plant or fuel reprocessing plant. The active reactor cores are divided into two region fuel i.e. 60% fuel fraction of Natural Uranium as inner core and 65% fuel fraction of Natural Uranium as outer core. Each fuel core regions are subdivided into ten parts (region-1 until region-10) with the same volume in the axial direction. The fresh Natural Uranium initially put in region-1, after one cycle of 10 years of burn-up it is shifted to region-2 and the each region-1 filled by fresh Natural Uranium. This concept is basically applied to all regions in both cores area, i.e. shifted the core of ith region into i+1 region after the end of 10 years burn-up cycle. For the next cycles, we will add only Natural Uranium on each region-1. The burn-up calculation is performed using collision probability method PIJ (cell burn-up calculation) in SRAC code which then given eight energy group macroscopic cross section data to be used in two dimensional R-Z geometry multi groups diffusion calculation in CITATION code. This reactor can results power thermal 600 MWth with average power density i.e. 80 watt/cc. After reactor start-up the operation, furthermore reactor only needs Natural Uranium supply for continue operation along 100 years. This calculation result then compared with one region fuel design i.e. 60% and 65% fuel fraction. This core design with two region fuel fraction can be an option for fuel optimization.

  16. Fabrication procedures for manufacturing High Flux Isotope Reactor fuel elements - 2

    SciTech Connect

    Knight, R.W.; Morin, R.A.

    1999-12-01

    The original fabrication procedures written in 1968 delineated the manufacturing procedures at that time. Since 1968, there have been a number of procedural changes. This rewrite of the fabrication procedures incorporates these changes. The entire fuel core of this reactor is made up of two fuel elements. Each element consists of one annular array of fuel plates. These annuli are identified as the inner and outer fuel elements, since one fits inside the other. The inner element consists of 171 identical fuel plates, and the outer element contains 369 identical fuel plates differing slightly from those in the inner element. Both sets of fuel plates contain U{sub 3}O{sub 8} powder as the fuel, dispersed in an aluminum powder matrix and clad with aluminum. Procedures for manufacturing and inspection of the fuel elements are described and illustrated.

  17. 10 CFR 71.97 - Advance notification of shipment of irradiated reactor fuel and nuclear waste.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... nuclear waste contained in the shipment, as specified in the regulations of DOT in 49 CFR 172.202 and 172... fuel and nuclear waste. 71.97 Section 71.97 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PACKAGING... notification of shipment of irradiated reactor fuel and nuclear waste. (a) As specified in paragraphs (b),...

  18. Advanced Fuel/Cladding Testing Capabilities in the ORNL High Flux Isotope Reactor

    SciTech Connect

    Ott, Larry J; Ellis, Ronald James; McDuffee, Joel Lee; Spellman, Donald J; Bevard, Bruce Balkcom

    2009-01-01

    The ability to test advanced fuels and cladding materials under reactor operating conditions in the United States is limited. The Oak Ridge National Laboratory (ORNL) High Flux Isotope Reactor (HFIR) and the newly expanded post-irradiation examination (PIE) capability at the ORNL Irradiated Fuels Examination Laboratory provide unique support for this type of advanced fuel/cladding development effort. The wide breadth of ORNL's fuels and materials research divisions provides all the necessary fuel development capabilities in one location. At ORNL, facilities are available from test fuel fabrication, to irradiation in HFIR under either thermal or fast reactor conditions, to a complete suite of PIEs, and to final product disposal. There are very few locations in the world where this full range of capabilities exists. New testing capabilities at HFIR have been developed that allow testing of advanced nuclear fuels and cladding materials under prototypic operating conditions (i.e., for both fast-spectrum conditions and light-water-reactor conditions). This paper will describe the HFIR testing capabilities, the new advanced fuel/cladding testing facilities, and the initial cooperative irradiation experiment that begins this year.

  19. HETEROGENEOUS NUCLEAR REACTOR EMPLOYING SMALL UNCLAD BODIES OF FISSIONABLE MATERIAL AS FUEL

    DOEpatents

    Hyman, H.H.; Katz, J.J.

    1961-05-01

    A nuclear reactor in which fuel pellets are continuously dissolved in a moderator liquid is described. The fuel pellets are fed into the top of elongated baskets which are submerged in moderator liquid, and a portion of the moderator liquid is continuously withdrawn and processed to recove r reaction products.

  20. Analyses for conversion of the Georgia Tech Research Reactor from HEU to LEU fuel

    SciTech Connect

    Matos, J.E.; Mo, S.C.; Woodruff, W.L.

    1992-09-01

    This document presents information concerning: analyses for conversion of the Georgia Tech Research Reactor from HEU to LEU; changes to technical specifications mandated by the conversion of the GTRR to low enrichment fuel; changes in the Safety Analysis Report mandated by the conversion of the GTRR to low enrichment fuel; and copies of all changed pages of the SAR and the technical specifications.

  1. Non-Proliferative, Thorium-Based, Core and Fuel Cycle for Pressurized Water Reactors

    SciTech Connect

    Todosow M.; Todosow M.; Raitses, G. Galperin, A.

    2009-07-12

    Two of the major barriers to the expansion of worldwide adoption of nuclear power are related to proliferation potential of the nuclear fuel cycle and issues associated with the final disposal of spent fuel. The Radkowsky Thorium Fuel (RTF) concept proposed by Professor A. Radkowsky offers a partial solution to these problems. The main idea of the concept is the utilization of the seed-blanket unit (SBU) fuel assembly geometry which is a direct replacement for a 'conventional' assembly in either a Russian pressurized water reactor (VVER-1000) or a Western pressurized water reactor (PWR). The seed-blanket fuel assembly consists of a fissile (U) zone, known as seed, and a fertile (Th) zone known as blanket. The separation of fissile and fertile allows separate fuel management schemes for the thorium part of the fuel (a subcritical 'blanket') and the 'driving' part of the core (a supercritical 'seed'). The design objective for the blanket is an efficient generation and in-situ fissioning of the U233 isotope, while the design objective for the seed is to supply neutrons to the blanket in a most economic way, i.e. with minimal investment of natural uranium. The introduction of thorium as a fertile component in the nuclear fuel cycle significantly reduces the quantity of plutonium production and modifies its isotopic composition, reducing the overall proliferation potential of the fuel cycle. Thorium based spent fuel also contains fewer higher actinides, hence reducing the long-term radioactivity of the spent fuel. The analyses show that the RTF core can satisfy the requirements of fuel cycle length, and the safety margins of conventional pressurized water reactors. The coefficients of reactivity are comparable to currently operating VVER's/PWR's. The major feature of the RTF cycle is related to the total amount of spent fuel discharged for each cycle from the reactor core. The fuel management scheme adopted for RTF core designs allows a significant decrease in the

  2. Multiphase CFD-based models for chemical looping combustion process: Fuel reactor modeling

    SciTech Connect

    Jung, Jonghwun; Gamwo, I.K.

    2008-04-21

    Chemical looping combustion (CLC) is a flameless two-step fuel combustion that produces a pure CO2 stream, ready for compression and sequestration. The process is composed of two interconnected fluidized bed reactors. The air reactor which is a conventional circulating fluidized bed and the fuel reactor which is a bubbling fluidized bed. The basic principle is to avoid the direct contact of air and fuel during the combustion by introducing a highly-reactive metal particle, referred to as oxygen carrier, to transport oxygen from the air to the fuel. In the process, the products from combustion are kept separated from the rest of the flue gases namely nitrogen and excess oxygen. This process eliminates the energy intensive step to separate the CO2 from nitrogen-rich flue gas that reduce the thermal efficiency. Fundamental knowledge of multiphase reactive fluid dynamic behavior of the gas–solid flow is essential for the optimization and operation of a chemical looping combustor. Our recent thorough literature review shows that multiphase CFD-based models have not been adapted to chemical looping combustion processes in the open literature. In this study, we have developed the reaction kinetics model of the fuel reactor and implemented the kinetic model into a multiphase hydrodynamic model, MFIX, developed earlier at the National Energy Technology Laboratory. Simulated fuel reactor flows revealed high weight fraction of unburned methane fuel in the flue gas along with CO2 and H2O. This behavior implies high fuel loss at the exit of the reactor and indicates the necessity to increase the residence time, say by decreasing the fuel flow rate, or to recirculate the unburned methane after condensing and removing CO2.

  3. 10 CFR 830 Major Modification Determination for Advanced Test Reactor LEU Fuel Conversion

    SciTech Connect

    Boyd D. Christensen; Michael A. Lehto; Noel R. Duckwitz

    2012-05-01

    The Advanced Test Reactor (ATR), located in the ATR Complex of the Idaho National Laboratory (INL), was constructed in the 1960s for the purpose of irradiating reactor fuels and materials. Other irradiation services, such as radioisotope production, are also performed at ATR. The ATR is fueled with high-enriched uranium (HEU) matrix (UAlx) in an aluminum sandwich plate cladding. The National Nuclear Security Administration Global Threat Reduction Initiative (GTRI) strategic mission includes efforts to reduce and protect vulnerable nuclear and radiological material at civilian sites around the world. Converting research reactors from using HEU to low-enriched uranium (LEU) was originally started in 1978 as the Reduced Enrichment for Research and Test Reactors (RERTR) Program under the U.S. Department of Energy (DOE) Office of Science. Within this strategic mission, GTRI has three goals that provide a comprehensive approach to achieving this mission: The first goal, the driver for the modification that is the subject of this determination, is to convert research reactors from using HEU to LEU. Thus the mission of the ATR LEU Fuel Conversion Project is to convert the ATR and Advanced Test Reactor Critical facility (ATRC) (two of the six U.S. High-Performance Research Reactors [HPRR]) to LEU fuel by 2017. The major modification criteria evaluation of the project pre-conceptual design identified several issues that lead to the conclusion that the project is a major modification.

  4. Integrated Decision-Making Tool to Develop Spent Fuel Strategies for Research Reactors

    SciTech Connect

    Beatty, Randy L; Harrison, Thomas J

    2016-01-01

    IAEA Member States operating or having previously operated a Research Reactor are responsible for the safe and sustainable management and disposal of associated radioactive waste, including research reactor spent nuclear fuel (RRSNF). This includes the safe disposal of RRSNF or the corresponding equivalent waste returned after spent fuel reprocessing. One key challenge to developing general recommendations lies in the diversity of spent fuel types, locations and national/regional circumstances rather than mass or volume alone. This is especially true given that RRSNF inventories are relatively small, and research reactors are rarely operated at a high power level or duration typical of commercial power plants. Presently, many countries lack an effective long-term policy for managing RRSNF. This paper presents results of the International Atomic Energy Agency (IAEA) Coordinated Research Project (CRP) #T33001 on Options and Technologies for Managing the Back End of the Research Reactor Nuclear Fuel Cycle which includes an Integrated Decision Making Tool called BRIDE (Back-end Research reactor Integrated Decision Evaluation). This is a multi-attribute decision-making tool that combines the Total Estimated Cost of each life-cycle scenario with Non-economic factors such as public acceptance, technical maturity etc and ranks optional back-end scenarios specific to member states situations in order to develop a specific member state strategic plan with a preferred or recommended option for managing spent fuel from Research Reactors.

  5. Fresh and Spent Nuclear Fuel Repatriation from the IRT-2000 Research Reactor Facility, Sofia, Bulgaria

    SciTech Connect

    K. J. Allen; T. G. Apostolov; I. S. Dimitrov

    2009-03-01

    The IRT 2000 research reactor, operated by the Bulgarian Institute for Nuclear Research and Nuclear Energy (INRNE), safely shipped all of their Russian-origin nuclear fuel from the Republic of Bulgaria to the Russian Federation beginning in 2003 and completing in 2008. These fresh and spent fuel shipments removed all highly enriched uranium (HEU) from Bulgaria. The fresh fuel was shipped by air in December 2003 using trucks and a commercial cargo aircraft. One combined spent fuel shipment of HEU and low enriched uranium (LEU) was completed in July 2008 using high capacity VPVR/M casks transported by truck, barge, and rail. The HEU shipments were assisted by the Russian Research Reactor Fuel Return Program (RRRFR) and the LEU spent fuel shipment was funded by Bulgaria. This report describes the work, approvals, organizations, equipment, and agreements required to complete these shipments and concludes with several major lessons learned.

  6. Establishing Specifications for Low Enriched Uranium Fuel Operations Conducted Outside the High Flux Isotope Reactor Site

    SciTech Connect

    Pinkston, Daniel; Primm, Trent; Renfro, David G; Sease, John D

    2010-10-01

    The National Nuclear Security Administration (NNSA) has funded staff at Oak Ridge National Laboratory (ORNL) to study the conversion of the High Flux Isotope Reactor (HFIR) from the current, high enriched uranium fuel to low enriched uranium fuel. The LEU fuel form is a metal alloy that has never been used in HFIR or any HFIR-like reactor. This report provides documentation of a process for the creation of a fuel specification that will meet all applicable regulations and guidelines to which UT-Battelle, LLC (UTB) the operating contractor for ORNL - must adhere. This process will allow UTB to purchase LEU fuel for HFIR and be assured of the quality of the fuel being procured.

  7. Fabrication of advanced oxide fuels containing minor actinide for use in fast reactors

    SciTech Connect

    Miwa, Shuhei; Osaka, Masahiko; Tanaka, Kosuke; Ishi, Yohei; Yoshimochi, Hiroshi; Tanaka, Kenya

    2007-07-01

    R and D of advanced fuel containing minor actinide for use in fast reactors is described related to the composite fuel with MgO matrix. Fabrication tests of MgO composite fuels containing Am were done by a practical process that could be adapted to the presently used commercial manufacturing technology. Am-containing MgO composite fuels having good characteristics, i.e., having no defects, a high density, a homogeneous dispersion of host phase, were obtained. As related technology, burn-up characteristics of a fast reactor core loaded with the present MgO composite fuel were also analyzed, mainly in terms of core criticality. Furthermore, phase relations of MA oxide which was assumed to be contained in MgO matrix fuel were experimentally investigated. (authors)

  8. Meeting Summary Advanced Light Water Reactor Fuels Industry Meeting Washington DC October 27 - 28, 2011

    SciTech Connect

    Not Listed

    2011-11-01

    The Advanced LWR Fuel Working Group first met in November of 2010 with the objective of looking 20 years ahead to the role that advanced fuels could play in improving light water reactor technology, such as waste reduction and economics. When the group met again in March 2011, the Fukushima incident was still unfolding. After the March meeting, the focus of the program changed to determining what we could do in the near term to improve fuel accident tolerance. Any discussion of fuels with enhanced accident tolerance will likely need to consider an advanced light water reactor with enhanced accident tolerance, along with the fuel. The Advanced LWR Fuel Working Group met in Washington D.C. on October 72-18, 2011 to continue discussions on this important topic.

  9. Microstructure of 50 year old SCK CEN BR1 research reactor fuel

    SciTech Connect

    Leenaers, A.; Berghe, S. van den

    2008-07-15

    The BR1 research reactor at SCK CEN, Mol (Belgium) has a graphite core matrix loaded with fuel rods consisting of a natural uranium slug in an aluminum cladding. Fabrication reports show the application of a so-called AlSi bonding layer and an U(Al,Si){sub 3} anti-diffusion layer on the natural uranium fuel slug to limit the interaction between the uranium fuel and aluminum cladding. The BR1 reactor is in operation since 1956 and still contains its original fuel rods. After more than 50 years irradiation at low temperature, the integrity of some of the fuel rods is investigated. The microstructure of the fuel, bonding and anti-diffusion layer and cladding is analysed using optical microscopy (OM), scanning electron microscopy (SEM) and electron microprobe analysis (EPMA). (author)

  10. End-of-life nondestructive examination of Light Water Breeder Reactor fuel rods (LWBR Development Program)

    SciTech Connect

    Gorscak, D.A.; Campbell, W.R.; Clayton, J.C.

    1987-10-01

    In-bundle and out-of-bundle (single rod) nondestructive examinations of Light Water Breeder Reactor fuel rods were performed. In-bundle examinations included visual examination and measurement of rod bow, rod-to-rod gaps, and rod removal forces. Out-of-bundle examinations included rod visuals and measurement of fuel rod length, diameter and ovality, cladding oxide and crud thickness, support grid induced cladding wear mark depth and volume, and fuel rod free hanging bow. The out-of-bundle examination also included ultrasonic inspection for cladding defects, neutron radiography for pellet integrity and plenum gap measurements, and gamma scans for instack axial gap screening and binary fuel stack length measurements. The measurements confirmed design predictions of fuel rod performance and provided evidence of excellent fuel rod performance for operation of Light Water Breeder Reactor to 29,047 effective full power hours (EFPH).

  11. Fuel provision for nonbreeding deuterium-tritium fusion reactors

    SciTech Connect

    Jassby, D.L.; Katsurai, M.

    1980-01-01

    Nonbreeding D-T reactors have decisive advantages in minimum size, unit cost, variety of applications, and ease of heat removal over reactors using any other fusion cycle, and significant advantages in environmental and safety characteristics over breeding D-T reactors. Considerations of relative energy production demonstrate that the most favorable source of tritium for a widely deployed system of nonbreeding D-T reactors is the very large (approx. 10 GW thermal) semi-catalyzed-deuterium (SCD), or sub-SCD reactor, where none of the escaping /sup 3/He (> 95%) or tritium (< 25%) is reinjected for burn-up. Feasibility of the ignited SCD tokamak reactor requires spatially averaged betas of 15 to 20% with a magnetic field at the TF coils of 12 to 13 Tesla.

  12. Ultra-thin 242mAm fuel elements in nuclear reactors. II

    NASA Astrophysics Data System (ADS)

    Ronen, Y.; Raitses, G.

    2004-04-01

    There is growing interest in using 242mAm as a nuclear fuel for space reactors and nuclear batteries. In this paper, we discuss different 242mAm enrichments, as well as fuel weight requirements, to produce a critical reactor. It was found that relatively low enrichments of 242mAm, about 10 w/o, are enough to guarantee criticality. Such low enrichments might eliminate the need for a 242mAm enrichment process. It was also found that the best results for low 242mAm requirements are obtained with a moderator to fuel volume ratio of 10,000.

  13. Fuel performance models for high-temperature gas-cooled reactor core design

    SciTech Connect

    Stansfield, O.M.; Simon, W.A.; Baxter, A.M.

    1983-09-01

    Mechanistic fuel performance models are used in high-temperature gas-cooled reactor core design and licensing to predict failure and fission product release. Fuel particles manufactured with defective or missing SiC, IPyC, or fuel dispersion in the buffer fail at a level of less than 5 x 10/sup -4/ fraction. These failed particles primarily release metallic fission products because the OPyC remains intact on 90% of the particles and retains gaseous isotopes. The predicted failure of particles using performance models appears to be conservative relative to operating reactor experience.

  14. Compact, high-power nuclear reactor systems based on small diameter particulate fuel

    NASA Astrophysics Data System (ADS)

    Powell, J. R.; Botts, T. E.

    Two compact, high-power nuclear reactor concepts are discussed. Both are gas-cooled cavity-type reactors which utilize particulate fuel of the type now used in HTGR reactors. Unshielded reactor volumes are on the order of one cubic meter. The Fixed Bed Reactor operating temperature is limited to 2500 K and the output power to 250 MW(e). In the Rotating Bed Reactor fuel is held within a rotating porous metal drum as a rotating fluidized bed. Rotating Bed Reactor outlet temperatures up to 3000 K and output power levels up to 1000 MW(e) are achievable. Both reactors can be brought up from stand by to full power in times on the order of a few seconds, due primarily to the short thermal time constant for the fuel particles. Turbine and MHD Brayton are the power conversion cycles of choice. Open cycle operation is generally favored for applications operating at less than 1000 sec of equivalent integrated full power. At power levels above 1 MW(e), the liquid droplet radiator is the favored means of heat rejection. Power system specific power levels of 10 kW(e)/kg (not including shield) appears to be quite feasible.

  15. Fuel, Structural Material and Coolant for an Advanced Fast Micro-Reactor

    NASA Astrophysics Data System (ADS)

    Do Nascimento, J. A.; Duimarães, L. N. F.; Ono, S.

    The use of nuclear reactors in space, seabed or other Earth hostile environment in the future is a vision that some Brazilian nuclear researchers share. Currently, the USA, a leader in space exploration, has as long-term objectives the establishment of a permanent Moon base and to launch a manned mission to Mars. A nuclear micro-reactor is the power source chosen to provide energy for life support, electricity for systems, in these missions. A strategy to develop an advanced micro-reactor technologies may consider the current fast reactor technologies as back-up and the development of advanced fuel, structural and coolant materials. The next generation reactors (GEN-IV) for terrestrial applications will operate with high output temperature to allow advanced conversion cycle, such as Brayton, and hydrogen production, among others. The development of an advanced fast micro-reactor may create a synergy between the GEN-IV and space reactor technologies. Considering a set of basic requirements and materials properties this paper discusses the choice of advanced fuel, structural and coolant materials for a fast micro-reactor. The chosen candidate materials are: nitride, oxide as back-up, for fuel, lead, tin and gallium for coolant, ferritic MA-ODS and Mo alloys for core structures. The next step will be the neutronic and burnup evaluation of core concepts with this set of materials.

  16. Development of remote disassembly technology for liquid-metal reactor (LMR) fuel

    SciTech Connect

    Bradley, E.C.; Evans, J.H.; Metz, C.F. III; Weil, B.S.

    1990-01-01

    A major objective of the Consolidated Fuel Reprocessing Program (CFRP) is to develop equipment and demonstrate technology to reprocess fast breeder reactor fuel. Experimental work on fuel disassembly cutting methods began in the 1970s. High-power laser cutting was selected as the preferred cutting method for fuel disassembly. Remotely operated development equipment was designed, fabricated, installed, and tested at Oak Ridge National Laboratory (ORNL). Development testing included remote automatic operation, remote maintenance testing, and laser cutting process development. This paper summarizes the development work performed at ORNL on remote fuel disassembly. 2 refs., 1 fig.

  17. Fuel assembly for the production of tritium in light water reactors

    DOEpatents

    Cawley, William E.; Trapp, Turner J.

    1985-01-01

    A nuclear fuel assembly is described for producing tritium in a light water moderated reactor. The assembly consists of two intermeshing arrays of subassemblies. The first subassemblies comprise concentric annular elements of an outer containment tube, an annular target element, an annular fuel element, and an inner neutron spectrums shifting rod. The second subassemblies comprise an outer containment tube and an inner rod of either fuel, target, or neutron spectrum shifting neutral.

  18. Feasibility and Safety Assessment for Advanced Reactor Concepts Using Vented Fuel

    SciTech Connect

    Klein, Andrew; Matthews, Topher; Lenhof, Renae; Deason, Wesley; Harter, Jackson

    2015-01-16

    Recent interest in fast reactor technology has led to renewed analysis of past reactor concepts such as Gas Fast Reactors and Sodium Fast Reactors. In an effort to make these reactors more economic, the fuel is required to stay in the reactor for extended periods of time; the longer the fuel stays within the core, the more fertile material is converted into usable fissile material. However, as burnup of the fuel-rod increases, so does the internal pressure buildup due to gaseous fission products. In order to reach the 30 year lifetime requirements of some reactor designs, the fuel pins must have a vented-type design to allow the buildup of fission products to escape. The present work aims to progress the understanding of the feasibility and safety issues related to gas reactors that incorporate vented fuel. The work was separated into three different work-scopes: 1. Quantitatively determine fission gas release from uranium carbide in a representative helium cooled fast reactor; 2. Model the fission gas behavior, transport, and collection in a Fission Product Vent System; and, 3. Perform a safety analysis of the Fission Product Vent System. Each task relied on results from the previous task, culminating in a limited scope Probabilistic Risk Assessment (PRA) of the Fission Product Vent System. Within each task, many key parameters lack the fidelity needed for comprehensive or accurate analysis. In the process of completing each task, the data or methods that were lacking were identified and compiled in a Gap Analysis included at the end of the report.

  19. Transmutation, Burn-Up and Fuel Fabrication Trade-Offs in Reduced-Moderation Water Reactor Thorium Fuel Cycles - 13502

    SciTech Connect

    Lindley, Benjamin A.; Parks, Geoffrey T.; Franceschini, Fausto

    2013-07-01

    Multiple recycle of long-lived actinides has the potential to greatly reduce the required storage time for spent nuclear fuel or high level nuclear waste. This is generally thought to require fast reactors as most transuranic (TRU) isotopes have low fission probabilities in thermal reactors. Reduced-moderation LWRs are a potential alternative to fast reactors with reduced time to deployment as they are based on commercially mature LWR technology. Thorium (Th) fuel is neutronically advantageous for TRU multiple recycle in LWRs due to a large improvement in the void coefficient. If Th fuel is used in reduced-moderation LWRs, it appears neutronically feasible to achieve full actinide recycle while burning an external supply of TRU, with related potential improvements in waste management and fuel utilization. In this paper, the fuel cycle of TRU-bearing Th fuel is analysed for reduced-moderation PWRs and BWRs (RMPWRs and RBWRs). RMPWRs have the advantage of relatively rapid implementation and intrinsically low conversion ratios. However, it is challenging to simultaneously satisfy operational and fuel cycle constraints. An RBWR may potentially take longer to implement than an RMPWR due to more extensive changes from current BWR technology. However, the harder neutron spectrum can lead to favourable fuel cycle performance. A two-stage fuel cycle, where the first pass is Th-Pu MOX, is a technically reasonable implementation of either concept. The first stage of the fuel cycle can therefore be implemented at relatively low cost as a Pu disposal option, with a further policy option of full recycle in the medium term. (authors)

  20. Impact of thorium based molten salt reactor on the closure of the nuclear fuel cycle

    NASA Astrophysics Data System (ADS)

    Jaradat, Safwan Qasim Mohammad

    Molten salt reactor (MSR) is one of six reactors selected by the Generation IV International Forum (GIF). The liquid fluoride thorium reactor (LFTR) is a MSR concept based on thorium fuel cycle. LFTR uses liquid fluoride salts as a nuclear fuel. It uses 232Th and 233U as the fertile and fissile materials, respectively. Fluoride salt of these nuclides is dissolved in a mixed carrier salt of lithium and beryllium (FLiBe). The objective of this research was to complete feasibility studies of a small commercial thermal LFTR. The focus was on neutronic calculations in order to prescribe core design parameter such as core size, fuel block pitch (p), fuel channel radius, fuel path, reflector thickness, fuel salt composition, and power. In order to achieve this objective, the applicability of Monte Carlo N-Particle Transport Code (MCNP) to MSR modeling was verified. Then, a prescription for conceptual small thermal reactor LFTR and relevant calculations were performed using MCNP to determine the main neutronic parameters of the core reactor. The MCNP code was used to study the reactor physics characteristics for the FUJI-U3 reactor. The results were then compared with the results obtained from the original FUJI-U3 using the reactor physics code SRAC95 and the burnup analysis code ORIPHY2. The results were comparable with each other. Based on the results, MCNP was found to be a reliable code to model a small thermal LFTR and study all the related reactor physics characteristics. The results of this study were promising and successful in demonstrating a prefatory small commercial LFTR design. The outcome of using a small core reactor with a diameter/height of 280/260 cm that would operate for more than five years at a power level of 150 MWth was studied. The fuel system 7LiF - BeF2 - ThF4 - UF4 with a (233U/ 232Th) = 2.01 % was the candidate fuel for this reactor core.

  1. Neutronic analysis stochastic distribution of fuel particles in Very High Temperature Gas-Cooled Reactors

    NASA Astrophysics Data System (ADS)

    Ji, Wei

    The Very High Temperature Gas-Cooled Reactor (VHTR) is a promising candidate for Generation IV designs due to its inherent safety, efficiency, and its proliferation-resistant and waste minimizing fuel cycle. A number of these advantages stem from its unique fuel design, consisting of a stochastic mixture of tiny (0.78mm diameter) microspheres with multiple coatings. However, the microsphere fuel regions represent point absorbers for resonance energy neutrons, resulting in the "double heterogeneity" for particle fuel. Special care must be taken to analyze this fuel in order to predict the spatial and spectral dependence of the neutron population in a steady-state reactor configuration. The challenges are considerable and resist brute force computation: there are over 1010 microspheres in a typical reactor configuration, with no hope of identifying individual microspheres in this stochastic mixture. Moreover, when individual microspheres "deplete" (e.g., burn the fissile isotope U-235 or transmute the fertile isotope U-238 (eventually) to Pu-239), the stochastic time-dependent nature of the depletion compounds the difficulty posed by the stochastic spatial mixture of the fuel, resulting in a prohibitive computational challenge. The goal of this research is to develop a methodology to analyze particle fuel randomly distributed in the reactor, accounting for the kernel absorptions as well as the stochastic depletion of the fuel mixture. This Ph.D. dissertation will address these challenges by developing a methodology for analyzing particle fuel that will be accurate enough to properly model stochastic particle fuel in both static and time-dependent configurations and yet be efficient enough to be used for routine analyses. This effort includes creation of a new physical model, development of a simulation algorithm, and application to real reactor configurations.

  2. Modeling the Pyrochemical Reduction of Spent UO2 Fuel in a Pilot-Scale Reactor

    SciTech Connect

    Steven D. Herrmann; Michael F. Simpson

    2006-08-01

    A kinetic model has been derived for the reduction of oxide spent nuclear fuel in a radial flow reactor. In this reaction, lithium dissolved in molten LiCl reacts with UO2 and fission product oxides to form a porous, metallic product. As the reaction proceeds, the depth of the porous layer around the exterior of each fuel particle increases. The observed rate of reaction has been found to be only dependent upon the rate of diffusion of lithium across this layer, consistent with a classic shrinking core kinetic model. This shrinking core model has been extended to predict the behavior of a hypothetical, pilot-scale reactor for oxide reduction. The design of the pilot-scale reactor includes forced flow through baskets that contain the fuel particles. The results of the modeling indicate that this is an essential feature in order to minimize the time needed to achieve full conversion of the fuel.

  3. Characterization of candidate DOE sites for fabricating MOX fuel for lead assemblies

    SciTech Connect

    Holdaway, R.F.; Miller, J.W.; Sease, J.D.; Moses, R.J.; O`Connor, D.G.; Carrell, R.D.; Jaeger, C.D.; Thompson, M.L.; Strasser, A.A.

    1998-03-01

    The Office of Fissile Materials Disposition (MD) of the Department of Energy (DOE) is directing the program to disposition US surplus weapons-usable plutonium. For the reactor option for disposition of this surplus plutonium, MD is seeking to contract with a consortium, which would include a mixed-oxide (MOX) fuel fabricator and a commercial US reactor operator, to fabricate and burn MOX fuel in existing commercial nuclear reactors. This option would entail establishing a MOX fuel fabrication facility under the direction of the consortium on an existing DOE site. Because of the lead time required to establish a MOX fuel fabrication facility and the need to qualify the MOX fuel for use in a commercial reactor, MD is considering the early fabrication of lead assemblies (LAs) in existing DOE facilities under the technical direction of the consortium. The LA facility would be expected to produce a minimum of 1 metric ton heavy metal per year and must be operational by June 2003. DOE operations offices were asked to identify candidate sites and facilities to be evaluated for suitability to fabricate MOX fuel LAs. Savannah River Site, Argonne National Laboratory-West, Hanford, Lawrence Livermore National Laboratory, and Los Alamos National Laboratory were identified as final candidates to host the LA project. A Site Evaluation Team (SET) worked with each site to develop viable plans for the LA project. SET then characterized the suitability of each of the five plans for fabricating MOX LAs using 28 attributes and documented the characterization to aid DOE and the consortium in selecting the site for the LA project. SET concluded that each option has relative advantages and disadvantages in comparison with other options; however, each could meet the requirements of the LA project as outlined by MD and SET.

  4. Investigation of stainless steel clad fuel rod failures and fuel performance in the Connecticut Yankee Reactor. Final report

    SciTech Connect

    Pasupathi, V.; Klingensmith, R. W.

    1981-11-01

    Significant levels of fuel rod failures were observed in the batch 8 fuel assemblies of the Connecticut Yankee reactor. Failure of 304 stainless steel cladding in a PWR environment was not expected. Therefore a detailed poolside and hot cell examination program was conducted to determine the cause of failure and identify differences between batch 8 fuel and previous batches which had operated without failures. Hot cell work conducted consisted of detailed nondestructive and destructive examination of fuel rods from batches 7 and 8. The results indicate that the batch 8 failure mechanism was stress corrosion cracking initiating on the clad outer surface. The sources of cladding stresses are believed to be (a) fuel pellet chips wedged in the cladding gap, (b) swelling of highly nondensifying batch 8 fuel and (c) potentially harmful effects of a power change event that occurred near the end of the second cycle of irradiation for batch 8.

  5. Nuclear proliferation and civilian nuclear power. Report of the Nonproliferation Alternative Systems Assessment Program. Volume IX. Reactor and fuel cycle description

    SciTech Connect

    Not Available

    1980-06-01

    The Nonproliferation Alterntive Systems Assessment Program (NASAP) has characterized and assessed various reactor/fuel-cycle systems. Volume IX provides, in summary form, the technical descriptions of the reactor/fuel-cycle systems studied. This includes the status of the system technology, as well as a discussion of the safety, environmental, and licensing needs from a technical perspective. This information was then used in developing the research, development, and demonstration (RD and D) program, including its cost and time frame, to advance the existing technology to the level needed for commercial use. Wherever possible, the cost data are given as ranges to reflect the uncertainties in the estimates.

  6. Low-Enriched Fuel Design Concept for the Prismatic Very High Temperature Reactor Core

    SciTech Connect

    Sterbentz, James W

    2007-05-01

    A new non-TRISO fuel and clad design concept is proposed for the prismatic, heliumcooled Very High Temperature Reactor core. The new concept could substantially reduce the current 10-20 wt% TRISO uranium enrichments down to 4-6 wt% for both initial and reload cores. The proposed fuel form would be a high-temperature, high-density uranium ceramic, for example UO2, configured into very small diameter cylindrical rods. The small diameter fuel rods significantly increase core reactivity through improved neutron moderation and fuel lumping. Although a high-temperature clad system for the concept remains to be developed, recent success in tube fabrication and preliminary irradiation testing of silicon carbide (SiC) cladding for light water reactor applications offers good potential for this application, and for future development of other carbide clad designs. A high-temperature ceramic fuel, together with a high-temperature clad material, could also lead to higher thermal safety margins during both normal and transient reactor conditions relative to TRISO fuel. The calculated neutronic results show that the lowenrichment, small diameter fuel rods and low thermal neutron absorbing clad retain the strong negative Doppler fuel temperature coefficient of reactivity that ensures inherent safe operation of the VHTR, and depletion studies demonstrate that an 18-month power cycle can be achieved with the lower enrichment fuel.

  7. Reactor Physics Characterization of Transmutation Targeting Options in a Sodium Fast Reactor

    SciTech Connect

    Samuel E. Bays

    2007-04-01

    In sodium fast reactor designs, the fuel related inherent negative reactivity feedback is accomplished mainly through parasitic capture in U-238. However for an efficient minor actinide burning system, it is desirable to reduce or eliminate U-238 entirely to suppress further transuranic actinide generation. Consequently, reactivity feedback is accomplished by enhancing axial neutron streaming during a loss of coolant void situation. This is done by flattening “pancake” the active core geometry. Flattening the reactor also increases axial leakage which removes neutrons that could otherwise be used to destroy minor actinides. Therefore, it is important to tailor the neutron spectrum in the core for optimized feedback and minor actinide destruction simultaneously by using minor actinide and fission product targets.

  8. Metallic fast reactor fuel fabrication for the global nuclear energy partnership

    NASA Astrophysics Data System (ADS)

    Burkes, Douglas E.; Fielding, Randall S.; Porter, Douglas L.

    2009-07-01

    Fast reactors are once again being considered for nuclear power generation, in addition to transmutation of long-lived fission products resident in spent nuclear fuels. This re-consideration follows with intense developmental programs for both fuel and reactor design. One of the two leading candidates for next generation fast reactor fuel is metal alloys, resulting primarily from the successes achieved in the 1960s to early 1990s with both the experimental breeding reactor-II and the fast flux test facility. The goal of the current program is to develop and qualify a nuclear fuel system that performs all of the functions of a conventional, fast-spectrum nuclear fuel while destroying recycled actinides, thereby closing the nuclear fuel cycle. In order to meet this goal, the program must develop efficient and safe fuel fabrication processes designed for remote operation. This paper provides an overview of advanced casting processes investigated in the past, and the development of a gaseous diffusion calculation that demonstrates how straightforward process parameter modification can mitigate the loss of volatile minor actinides in the metal alloy melt.

  9. High Temperature Reactor (HTR) Deep Burn Core and Fuel Analysis: Design Selection for the Prismatic Block Reactor

    SciTech Connect

    Francesco Venneri; Chang-Keun Jo; Jae-Man Noh; Yonghee Kim; Claudio Filippone; Jonghwa Chang; Chris Hamilton; Young-Min Kim; Ji-Su Jun; Moon-Sung Cho; Hong-Sik Lim; MIchael A. Pope; Abderrafi M. Ougouag; Vincent Descotes; Brian Boer

    2010-09-01

    The Deep Burn (DB) Project is a U.S. Department of Energy sponsored feasibility study of Transuranic Management using high burnup fuel in the high temperature helium cooled reactor (HTR). The DB Project consists of seven tasks: project management, core and fuel analysis, spent fuel management, fuel cycle integration, TRU fuel modeling, TRU fuel qualification, and HTR fuel recycle. In the Phase II of the Project, we conducted nuclear analysis of TRU destruction/utilization in the HTR prismatic block design (Task 2.1), deep burn fuel/TRISO microanalysis (Task 2.3), and synergy with fast reactors (Task 4.2). The Task 2.1 covers the core physics design, thermo-hydraulic CFD analysis, and the thermofluid and safety analysis (low pressure conduction cooling, LPCC) of the HTR prismatic block design. The Task 2.3 covers the analysis of the structural behavior of TRISO fuel containing TRU at very high burnup level, i.e. exceeding 50% of FIMA. The Task 4.2 includes the self-cleaning HTR based on recycle of HTR-generated TRU in the same HTR. Chapter IV contains the design and analysis results of the 600MWth DB-HTR core physics with the cycle length, the average discharged burnup, heavy metal and plutonium consumptions, radial and axial power distributions, temperature reactivity coefficients. Also, it contains the analysis results of the 450MWth DB-HTR core physics and the analysis of the decay heat of a TRU loaded DB-HTR core. The evaluation of the hot spot fuel temperature of the fuel block in the DB-HTR (Deep-Burn High Temperature Reactor) core under full operating power conditions are described in Chapter V. The investigated designs are the 600MWth and 460MWth DB-HTRs. In Chapter VI, the thermo-fluid and safety of the 600MWth DB-HTRs has been analyzed to investigate a thermal-fluid design performance at the steady state and a passive safety performance during an LPCC event. Chapter VII describes the analysis results of the TRISO fuel microanalysis of the 600MWth and 450

  10. Characterization of exposure-dependent eigenvalue drift using Monte Carlo based nuclear fuel management

    NASA Astrophysics Data System (ADS)

    Xoubi, Ned

    2005-12-01

    The ability to accurately predict the multiplication factor (keff) of a nuclear reactor core as a function of exposure continues to be an elusive task for core designers despite decades of advances in computational methods. The difference between a predicted eigenvalue (target) and the actual eigenvalue at critical reactor conditions is herein referred to as the "eigenvalue drift." This dissertation studies exposure-dependent eigenvalue drift using MCNP-based fuel management analysis of the ORNL High Flux Isotope Reactor core. Spatial-dependent burnup is evaluated using the MONTEBURNS and ALEPH codes to link MCNP to ORIGEN to help analyze the behavior of keff as a function of fuel exposure. Understanding the exposure-dependent eigenvalue drift of a nuclear reactor is of particular relevance when trying to predict the impact of major design changes upon fuel cycle behavior and length. In this research, the design of an advanced HFIR core with a fuel loading of 12 kg of 235U is contrasted against the current loading of 9.4 kg. The goal of applying exposure dependent eigenvalue characterization is to produce a more accurate prediction of the fuel cycle length than prior analysis techniques, and to improve our understanding of the reactivity behavior of the core throughout the cycle. This investigation predicted a fuel cycle length of 40 days, representing a 50% increase in the cycle length in response to a 25% increase in fuel loading. The average burnup increased by about 48 MWd/kg U and it was confirmed that the excess reactivity can be controlled with the present design and arrangement of control elements throughout the core's life. Another major design change studied was the effect of installing an internal beryllium reflector upon cycle length. Exposure dependent eigenvalue predictions indicate that the actual benefit could be twice as large as that originally assessed via beginning-of-life (BOL) analyses.

  11. FEASIBILITY OF RECYCLING PLUTONIUM AND MINOR ACTINIDES IN LIGHT WATER REACTORS USING HYDRIDE FUEL

    SciTech Connect

    Greenspan, Ehud; Todreas, Neil; Taiwo, Temitope

    2009-03-10

    The objective of this DOE NERI program sponsored project was to assess the feasibility of improving the plutonium (Pu) and minor actinide (MA) recycling capabilities of pressurized water reactors (PWRs) by using hydride instead of oxide fuels. There are four general parts to this assessment: 1) Identifying promising hydride fuel assembly designs for recycling Pu and MAs in PWRs 2) Performing a comprehensive systems analysis that compares the fuel cycle characteristics of Pu and MA recycling in PWRs using the promising hydride fuel assembly designs identified in Part 1 versus using oxide fuel assembly designs 3) Conducting a safety analysis to assess the likelihood of licensing hydride fuel assembly designs 4) Assessing the compatibility of hydride fuel with cladding materials and water under typical PWR operating conditions Hydride fuel was found to offer promising transmutation characteristics and is recommended for further examination as a possible preferred option for recycling plutonium in PWRs.

  12. A dynamic fuel cycle analysis for a heterogeneous thorium-DUPIC recycle in CANDU reactors

    SciTech Connect

    Jeong, C. J.; Park, C. J.; Choi, H.

    2006-07-01

    A heterogeneous thorium fuel recycle scenario in a Canada deuterium uranium (CANDU) reactor has been analyzed by the dynamic analysis method. The thorium recycling is performed through a dry process which has a strong proliferation resistance. In the fuel cycle model, the existing nuclear power plant construction plan was considered up to 2016, while the nuclear demand growth rate from the year 2016 was assumed to be 0%. In this analysis, the spent fuel inventory as well as the amount of plutonium, minor actinides, and fission products of a multiple thorium recycling fuel cycle were estimated and compared to those of the once-through fuel cycle. The analysis results have shown that the heterogeneous thorium fuel cycle can be constructed through the dry process technology. It is also shown that the heterogeneous thorium fuel cycle can reduce the spent fuel inventory and save on the natural uranium resources when compared with the once-through cycle. (authors)

  13. Fuel Thermo-physical Characterization Project: Evaluation of Models to Calculate Thermal Diffusivity of Layered Composites

    SciTech Connect

    Burkes, Douglas; Casella, Amanda J.; Gardner, Levi D.; Casella, Andrew M.; Huber, Tanja K.; Breitkreutz, Harald

    2015-02-11

    The Office of Material Management and Minimization Fuel Thermo-physical Characterization Project at Pacific Northwest National Laboratory (PNNL) is tasked with using PNNL facilities and processes to receive irradiated low enriched uranium-molybdenum fuel plate samples and perform analyses in support of the Office of Material Management and Minimization Reactor Conversion Program. This work is in support of the Fuel Development Pillar that is managed by Idaho National Laboratory. A key portion of the scope associated with this project was to measure the thermal properties of fuel segments harvested from plates that were irradiated in the Advanced Test Reactor. Thermal diffusivity of samples prepared from the fuel segments was measured using laser flash analysis. Two models, one developed by PNNL and the other developed by the Technische Universität München (TUM), were evaluated to extract the thermal diffusivity of the uranium-molybdenum alloy from measurements made on the irradiated, layered composites. The experimental data of the “TC” irradiated fuel segment was evaluated using both models considering a three-layer and five-layer system. Both models are in acceptable agreement with one another and indicate that the zirconium diffusion barrier has a minimal impact on the overall thermal diffusivity of the monolithic U-Mo fuel.

  14. Development of an internally cooled annular fuel bundle for pressurized heavy water reactors

    SciTech Connect

    Hamilton, H.; Armstrong, J.; Kittmer, A.; Zhuchkova, A.; Xu, R.; Hyland, B.; King, M.; Nava-Dominguez, A.; Livingstone, S.; Bergeron, A.

    2013-07-01

    A number of preliminary studies have been conducted at Atomic Energy of Canada Limited to explore the potential of using internally cooled annular fuel (ICAF) in CANDU reactors including finite element thermo-mechanical modelling, reactor physics, thermal hydraulics, fabrication and mechanical design. The most compelling argument for this design compared to the conventional solid-rod design is the significant reduction in maximum fuel temperature for equivalent LERs (linear element ratings). This feature presents the potential for power up-rating or higher burnup and a decreased defect probability due to in-core power increases. The thermal-mechanical evaluation confirmed the significant reduction in maximum fuel temperatures for ICAF fuel compared to solid-rod fuel for equivalent LER. The maximum fuel temperature increase as a function of LER increase is also significantly less for ICAF fuel. As a result, the sheath stress induced by an equivalent power increase is approximately six times less for ICAF fuel than solid-rod fuel. This suggests that the power-increase thresholds to failure (due to stress-corrosion cracking) for ICAF fuel should be well above those for solid-rod fuel, providing improvement in operation flexibility and safety.

  15. PREPARING THE HIGH FLUX ISOTOPE REACTOR FOR CONVERSION TO LOW ENRICHED URANIUM FUEL ? RETURN TO 100 MW

    SciTech Connect

    Smith, Kevin Arthur; Primm, Trent

    2009-01-01

    The feasibility of low-enriched uranium (LEU) fuel as a replacement for the current, high enriched uranium (HEU) fuel for the High Flux Isotope Reactor (HFIR) has been under study since 2006. Reactor performance studies have been completed for conceptual plate designs and show that maintaining reactor performance while converting to LEU fuel requires returning the reactor power to 100 MW from 85 MW. The analyses required to up-rate the reactor power and the methods to perform these analyses are discussed. Comments regarding the regulatory approval process are provided along with a conceptual schedule.

  16. Fuel cell operation characterization using simulation

    NASA Astrophysics Data System (ADS)

    Schott, Pascal; Baurens, Pierre

    The first step of fuel cell stack characterization consists in experimental testing of its global electrochemical performances under various working conditions. The second step is devoted to link the collected data to local species concentrations, membrane hydration and resulting current density in the stack. In the project PICOS, we have to characterize a 1.5 kW 12-cells' stack of 330 cm 2. The stack performances are measured on a specific test bench and analyzed using a fuel cell-stack modeling tool. The test bench allows to control the oxygen (or air) and hydrogen flow rates and humidity at the inlet of the stack, the pressure at the outlet, and the delivered current. Each cell potential and the total stack potential are monitored, the produced water is condensed at the outlet in order to measure its amount. The model used is a mono-dimensional description of a single cell. It is based on a semi-empirical law for the local electrochemical response computation. The local pressures, temperatures, gas compositions and membrane hydration are computed dynamically. A semi-empirical electrochemical model of the active layer is fitted to experimental data, the gas diffusion layer porosity and electro-osmosis coefficient of the membrane have been adjusted to obtain the best fit. The simulations calculate the cell potential and the current density profile in the cell for various feeding conditions. The simulations allow us to explain the effect of gas hydration on the stack potential and to predict possible drying out of the membrane.

  17. Development of an on-line fuel failure monitoring system for CANDU reactors

    NASA Astrophysics Data System (ADS)

    Livingstone, Stephen Jason

    Although relatively rare in CANDU plants, fuel defects have always been an important operating concern for CANDU fuel operation and behaviour, and play a critical role in health, safety, and the economics of an operating reactor. A fuel defect occurs when a fuel element has a breach in its sheath resulting in fission product (FP) release and/or uranium fuel loss into the reactor primary heat transport system (PHTS). The unintended release of FP and fuel material into the PHTS creates elevated radiation fields and hazards for the reactor operator. The intent of this thesis is to develop an online real-time system that can analyse PHTS activity and infer information relating to otherwise unknown defect(s) in the reactor core. A MATLABRTM based Graphical User Interface (GUI) programme called COLDD (CANDU On-Line Defected fuel Diagnostic) has been developed to provide detailed diagnostics of PHTS activities. COLDD is based on new techniques, a new empirical diffusion coefficient, new algorithms, and refinement of existing techniques. Several techniques are based on detailed mechanistic models that are presented in detail, while other techniques are based on empirical rules from experimental and commercial experience; the diversity of techniques are shown to be self-consistent. The techniques employed by COLDD are compared to techniques used internationally by other defected fuel diagnostic tools for non-CANDU type reactors. The ability for COLDD to perform successful defected fuel diagnostics is dependent on the quality of the data provided. A detailed sensitivity analysis is performed to determine key measurement parameters. Techniques are also developed to allow operators to perform robust error checking of data to ensure consistency. COLDD is validated against theoretical, experimental, and commercial reactor data, and shown to be stable and consistent in all cases. The plethora of analysis modes are shown to be self-consistent. The version of COLDD described in

  18. Analysis of the impact of random summing on passive assay of pebble bed reactor fuel using gamma-ray spectrometry

    NASA Astrophysics Data System (ADS)

    Chen, J.; Hawari, A. I.

    2007-08-01

    Pebble bed reactors (PBR) are characterized by multi-pass fuel systems in which spherical fuel pebbles are circulated through the core until they reach a proposed burnup limit. The fuel is assayed on-line to ensure that the burnup limit is not breached. However, random summing effects can impact the response of the burnup measurement system and result in distortions that degrade the accuracy of the assay results. Monte Carlo analysis was performed to estimate the magnitude and effect of random summing on the absolute and relative indicators that have been identified as usable in on-line assay. For a throughput rate of 10 5 counts/s and trapezoidal pulse shaping of the signals, the results show that absolute indicators suffer from severe distortions due to this effect. Relative indicators are found to be resistant to random summing with the deviation in the ratio of peak areas remaining less than 5-15% depending on pulse width.

  19. Computerized operating procedures for shearing and dissolution of segments from LWBR (Light Water Breeder Reactor) fuel rods

    SciTech Connect

    Osudar, J.; Deeken, P.G.; Graczyk, D.G.; Fagan, J.E.; Martino, F.J.; Parks, J.E.; Levitz, N.M.; Kessie, R.W.; Leddin, J.M.

    1987-05-01

    This report presents two detailed computerized operating procedures developed to assist and control the shearing and dissolution of irradiated fuel rods. The procedures were employed in the destructive analysis of end-of-life fuel rods from the Light Water Breeder Reactor (LWBR) that was designed by the Westinghouse Electric Corporation Bettis Atomic Power Laboratory. Seventeen entire fuel rods from the end-of-life core of the LWBR were sheared into 169 precisely characterized segments, and more than 150 of these segments were dissolved during execution of the LWBR Proof-of-Breeding (LWBR-POB) Analytical Support Project at Argonne National Laboratory. The procedures illustrate our approaches to process monitoring, data reduction, and quality assurance during the LWBR-POB work.

  20. Impact investigation of reactor fuel operating parameters on reactivity for use in burnup credit applications

    NASA Astrophysics Data System (ADS)

    Sloma, Tanya Noel

    When representing the behavior of commercial spent nuclear fuel (SNF), credit is sought for the reduced reactivity associated with the net depletion of fissile isotopes and the creation of neutron-absorbing isotopes, a process that begins when a commercial nuclear reactor is first operated at power. Burnup credit accounts for the reduced reactivity potential of a fuel assembly and varies with the fuel burnup, cooling time, and the initial enrichment of fissile material in the fuel. With regard to long-term SNF disposal and transportation, tremendous benefits, such as increased capacity, flexibility of design and system operations, and reduced overall costs, provide an incentive to seek burnup credit for criticality safety evaluations. The Nuclear Regulatory Commission issued Interim Staff Guidance 8, Revision 2 in 2002, endorsing burnup credit of actinide composition changes only; credit due to actinides encompasses approximately 30% of exiting pressurized water reactor SNF inventory and could potentially be increased to 90% if fission product credit were accepted. However, one significant issue for utilizing full burnup credit, compensating for actinide and fission product composition changes, is establishing a set of depletion parameters that produce an adequately conservative representation of the fuel's isotopic inventory. Depletion parameters can have a significant effect on the isotopic inventory of the fuel, and thus the residual reactivity. This research seeks to quantify the reactivity impact on a system from dominant depletion parameters (i.e., fuel temperature, moderator density, burnable poison rod, burnable poison rod history, and soluble boron concentration). Bounding depletion parameters were developed by statistical evaluation of a database containing reactor operating histories. The database was generated from summary reports of commercial reactor criticality data. Through depletion calculations, utilizing the SCALE 6 code package, several light

  1. Summary of the radiological assessment of the fuel cycle for a thorium-uranium carbide-fueled fast breeder reactor

    SciTech Connect

    Tennery, V.J.; Bomar, E.S.; Bond, W.D.; Meyer, H.R.; Morse, L.E.; Till, J.E.; Yalcintas, M.G.

    1980-01-01

    A large fraction of the potential fuel for nuclear power reactors employing fissionable materials exists as ores of thorium. In addition, certain characteristics of a fuel system based on breeding of the fissionable isotope {sup 233}U from thorium offer the possibility of a greater resistance to the diversion of fissionable material for the fabrication of nuclear weapons. This report consolidates into a single source the principal content of two previous reports which assess the radiological environmental impact of mining and milling of thorium ore and of the reprocessing and refabrication of spent FBR thorium-uranium carbide fuel.

  2. Analysis of fluid fuel flow to the neutron kinetics on molten salt reactor FUJI-12

    NASA Astrophysics Data System (ADS)

    Aji, Indarta Kuncoro; Waris, Abdul; Permana, Sidik

    2015-09-01

    Molten Salt Reactor is a reactor are operating with molten salt fuel flowing. This condition interpret that the neutron kinetics of this reactor is affected by the flow rate of the fuel. This research analyze effect by the alteration velocity of the fuel by MSR type Fuji-12, with fuel composition LiF-BeF2-ThF4-233UF4 respectively 71.78%-16%-11.86%-0.36%. Calculation process in this study is performed numerically by SOR and finite difference method use C programming language. Data of reactivity, neutron flux, and the macroscopic fission cross section for calculation process obtain from SRAC-CITATION (Standard thermal Reactor Analysis Code) and JENDL-4.0 data library. SRAC system designed and developed by JAEA (Japan Atomic Energy Agency). This study aims to observe the effect of the velocity of fuel salt to the power generated from neutron precursors at fourth year of reactor operate (last critical condition) with number of multiplication effective; 1.0155.

  3. Analysis of fluid fuel flow to the neutron kinetics on molten salt reactor FUJI-12

    SciTech Connect

    Aji, Indarta Kuncoro; Waris, Abdul Permana, Sidik

    2015-09-30

    Molten Salt Reactor is a reactor are operating with molten salt fuel flowing. This condition interpret that the neutron kinetics of this reactor is affected by the flow rate of the fuel. This research analyze effect by the alteration velocity of the fuel by MSR type Fuji-12, with fuel composition LiF-BeF{sub 2}-ThF{sub 4}-{sup 233}UF{sub 4} respectively 71.78%-16%-11.86%-0.36%. Calculation process in this study is performed numerically by SOR and finite difference method use C programming language. Data of reactivity, neutron flux, and the macroscopic fission cross section for calculation process obtain from SRAC-CITATION (Standard thermal Reactor Analysis Code) and JENDL-4.0 data library. SRAC system designed and developed by JAEA (Japan Atomic Energy Agency). This study aims to observe the effect of the velocity of fuel salt to the power generated from neutron precursors at fourth year of reactor operate (last critical condition) with number of multiplication effective; 1.0155.

  4. Analysis on burnup step effect for evaluating reactor criticality and fuel breeding ratio

    SciTech Connect

    Saputra, Geby; Purnama, Aditya Rizki; Permana, Sidik; Suzuki, Mitsutoshi

    2014-09-30

    Criticality condition of the reactors is one of the important factors for evaluating reactor operation and nuclear fuel breeding ratio is another factor to show nuclear fuel sustainability. This study analyzes the effect of burnup steps and cycle operation step for evaluating the criticality condition of the reactor as well as the performance of nuclear fuel breeding or breeding ratio (BR). Burnup step is performed based on a day step analysis which is varied from 10 days up to 800 days and for cycle operation from 1 cycle up to 8 cycles reactor operations. In addition, calculation efficiency based on the variation of computer processors to run the analysis in term of time (time efficiency in the calculation) have been also investigated. Optimization method for reactor design analysis which is used a large fast breeder reactor type as a reference case was performed by adopting an established reactor design code of JOINT-FR. The results show a criticality condition becomes higher for smaller burnup step (day) and for breeding ratio becomes less for smaller burnup step (day). Some nuclides contribute to make better criticality when smaller burnup step due to individul nuclide half-live. Calculation time for different burnup step shows a correlation with the time consuming requirement for more details step calculation, although the consuming time is not directly equivalent with the how many time the burnup time step is divided.

  5. High temperature gas-cooled reactor (HTGR) graphite pebble fuel: Review of technologies for reprocessing

    SciTech Connect

    Mcwilliams, A. J.

    2015-09-08

    This report reviews literature on reprocessing high temperature gas-cooled reactor graphite fuel components. A basic review of the various fuel components used in the pebble bed type reactors is provided along with a survey of synthesis methods for the fabrication of the fuel components. Several disposal options are considered for the graphite pebble fuel elements including the storage of intact pebbles, volume reduction by separating the graphite from fuel kernels, and complete processing of the pebbles for waste storage. Existing methods for graphite removal are presented and generally consist of mechanical separation techniques such as crushing and grinding chemical techniques through the use of acid digestion and oxidation. Potential methods for reprocessing the graphite pebbles include improvements to existing methods and novel technologies that have not previously been investigated for nuclear graphite waste applications. The best overall method will be dependent on the desired final waste form and needs to factor in the technical efficiency, political concerns, cost, and implementation.

  6. End-of-life destructive examination of light water breeder reactor fuel rods (LWBR Development Program)

    SciTech Connect

    Richardson, K.D.

    1987-10-01

    Destructive examination of 12 representative Light Water Breeder Reactor fuel rods was performed following successful operation in the Shippingport Atomic Power Station for 29,047 effective full power hours, about five years. Light Water Breeder Reactor fuel rods were unique in that the thorium oxide and uranium-233 oxide fuel was contained within Zircaloy-4 cladding. Destructive examinations included analysis of released fission gas; chemical analysis of the fuel to determine depletion, iodine, and cesium levels; chemical analysis of the cladding to determine hydrogen, iodine, and cesium levels; metallographic examination of the cladding, fuel, and other rod components to determine microstructural features and cladding corrosion features; and tensile testing of the irradiated cladding to determine mechanical strength. The examinations confirmed that Light Water Breeder Reactor fuel rod performance was excellent. No evidence of fuel rod failure was observed, and the fuel operating temperature was low (below 2580/sup 0/F at which an increased percentage of fission gas is released). 21 refs., 80 figs., 20 tabs.

  7. Assessment of the use of extended burnup fuel in light water power reactors

    SciTech Connect

    Baker, D.A.; Bailey, W.J.; Beyer, C.E.; Bold, F.C.; Tawil, J.J.

    1988-02-01

    This study has been conducted by Pacific Northwest Laboratory for the US Nuclear Regulatory Commission to review the environmental and economic impacts associated with the use of extended burnup nuclear fuel in light water power reactors. It has been proposed that current batch average burnup levels of 33 GWd/t uranium be increased to above 50 GWd/t. The environmental effects of extending fuel burnup during normal operations and during accident events and the economic effects of cost changes on the fuel cycle are discussed in this report. The physical effects of extended burnup on the fuel and the fuel assembly are also presented as a basis for the environmental and economic assessments. Environmentally, this burnup increase would have no significant impact over that of normal burnup. Economically, the increased burnup would have favorable effects, consisting primarily of a reduction: (1) total fuel requirements; (2) reactor downtime for fuel replacement; (3) the number of fuel shipments to and from reactor sites; and (4) repository storage requirements. 61 refs., 4 figs., 27 tabs.

  8. Fuel Cycle System Analysis Implications of Sodium-Cooled Metal-Fueled Fast Reactor Transuranic Conversion Ratio

    SciTech Connect

    Steven J. Piet; Edward A. Hoffman; Samuel E. Bays; Gretchen E. Matthern; Jacob J. Jacobson; Ryan Clement; David W. Gerts

    2013-03-01

    If advanced fuel cycles are to include a large number of fast reactors (FRs), what should be the transuranic (TRU) conversion ratio (CR)? The nuclear energy era started with the assumption that they should be breeder reactors (CR > 1), but the full range of possible CRs eventually received attention. For example, during the recent U.S. Global Nuclear Energy Partnership program, the proposal was burner reactors (CR < 1). Yet, more recently, Massachusetts Institute of Technology's "Future of the Nuclear Fuel Cycle" proposed CR [approximately] 1. Meanwhile, the French company EDF remains focused on breeders. At least one of the reasons for the differences of approach is different fuel cycle objectives. To clarify matters, this paper analyzes the impact of TRU CR on many parameters relevant to fuel cycle systems and therefore spans a broad range of topic areas. The analyses are based on a FR physics parameter scan of TRU CR from 0 to [approximately]1.8 in a sodium-cooled metal-fueled FR (SMFR), in which the fuel from uranium-oxide-fueled light water reactors (LWRs) is recycled directly to FRs and FRs displace LWRs in the fleet. In this instance, the FRs are sodium cooled and metal fueled. Generally, it is assumed that all TRU elements are recycled, which maximizes uranium ore utilization for a given TRU CR and waste radiotoxicity reduction and is consistent with the assumption of used metal fuel separated by electrochemical means. In these analyses, the fuel burnup was constrained by imposing a neutron fluence limit to fuel cladding to the same constant value. This paper first presents static, time-independent measures of performance for the LWR [right arrow] FR fuel cycle, including mass, heat, gamma emission, radiotoxicity, and the two figures of merit for materials for weapon attractiveness developed by C. Bathke et al. No new fuel cycle will achieve a static equilibrium in the foreseeable future. Therefore, additional analyses are shown with dynamic, time

  9. Analysis of fuel options for the breakeven core configuration of the Advanced Recycling Reactor

    SciTech Connect

    Stauff, N.E.; Klim, T.K.; Taiwo, T.A.; Fiorina, C.; Franceschini, F.

    2013-07-01

    A trade-off study is performed to determine the impacts of various fuel forms on the core design and core physics characteristics of the sodium-cooled Toshiba- Westinghouse Advanced Recycling Reactor (ARR). The fuel forms include oxide, nitride, and metallic forms of U and Th. The ARR core configuration is redesigned with driver and blanket regions in order to achieve breakeven fissile breeding performance with the various fuel types. State-of-the-art core physics tools are used for the analyses. In addition, a quasi-static reactivity balance approach is used for a preliminary comparison of the inherent safety performances of the various fuel options. Thorium-fueled cores exhibit lower breeding ratios and require larger blankets compared to the U-fueled cores, which is detrimental to core compactness and increases reprocessing and manufacturing requirements. The Th cores also exhibit higher reactivity swings through each cycle, which penalizes reactivity control and increases the number of control rods required. On the other hand, using Th leads to drastic reductions in void and coolant expansion coefficients of reactivity, with the potential for enhancing inherent core safety. Among the U-fueled ARR cores, metallic and nitride fuels result in higher breeding ratios due to their higher heavy metal densities. On the other hand, oxide fuels provide a softer spectrum, which increases the Doppler effect and reduces the positive sodium void worth. A lower fuel temperature is obtained with the metallic and nitride fuels due to their higher thermal conductivities and compatibility with sodium bonds. This is especially beneficial from an inherent safety point of view since it facilitates the reactor cool-down during loss of power removal transients. The advantages in terms of inherent safety of nitride and metallic fuels are maintained when using Th fuel. However, there is a lower relative increase in heavy metal density and in breeding ratio going from oxide to metallic

  10. Fuel processing in integrated micro-structured heat-exchanger reactors

    NASA Astrophysics Data System (ADS)

    Kolb, G.; Schürer, J.; Tiemann, D.; Wichert, M.; Zapf, R.; Hessel, V.; Löwe, H.

    Micro-structured fuel processors are under development at IMM for different fuels such as methanol, ethanol, propane/butane (LPG), gasoline and diesel. The target application are mobile, portable and small scale stationary auxiliary power units (APU) based upon fuel cell technology. The key feature of the systems is an integrated plate heat-exchanger technology which allows for the thermal integration of several functions in a single device. Steam reforming may be coupled with catalytic combustion in separate flow paths of a heat-exchanger. Reactors and complete fuel processors are tested up to the size range of 5 kW power output of a corresponding fuel cell. On top of reactor and system prototyping and testing, catalyst coatings are under development at IMM for numerous reactions such as steam reforming of LPG, ethanol and methanol, catalytic combustion of LPG and methanol, and for CO clean-up reactions, namely water-gas shift, methanation and the preferential oxidation of carbon monoxide. These catalysts are investigated in specially developed testing reactors. In selected cases 1000 h stability testing is performed on catalyst coatings at weight hourly space velocities, which are sufficiently high to meet the demands of future fuel processing reactors.

  11. Comparative analysis of thorium and uranium fuel for transuranic recycle in a sodium cooled Fast Reactor

    SciTech Connect

    C. Fiorina; N. E. Stauff; F. Franceschini; M. T. Wenner; A. Stanculescu; T. K. Kim; A. Cammi; M. E. Ricotti; R. N. Hill; T. A. Taiwo; M. Salvatores

    2013-12-01

    The present paper compares the reactor physics and transmutation performance of sodium-cooled Fast Reactors (FRs) for TRansUranic (TRU) burning with thorium (Th) or uranium (U) as fertile materials. The 1000 MWt Toshiba-Westinghouse Advanced Recycling Reactor (ARR) conceptual core has been used as benchmark for the comparison. Both burner and breakeven configurations sustained or started with a TRU supply, and assuming full actinide homogeneous recycle strategy, have been developed. State-of-the-art core physics tools have been employed to establish fuel inventory and reactor physics performances for equilibrium and transition cycles. Results show that Th fosters large improvements in the reactivity coefficients associated with coolant expansion and voiding, which enhances safety margins and, for a burner design, can be traded for maximizing the TRU burning rate. A trade-off of Th compared to U is the significantly larger fuel inventory required to achieve a breakeven design, which entails additional blankets at the detriment of core compactness as well as fuel manufacturing and separation requirements. The gamma field generated by the progeny of U-232 in the U bred from Th challenges fuel handling and manufacturing, but in case of full recycle, the high contents of Am and Cm in the transmutation fuel impose remote fuel operations regardless of the presence of U-232.

  12. Enhanced CANDU6: Reactor and fuel cycle options - Natural uranium and beyond

    SciTech Connect

    Ovanes, M.; Chan, P. S. W.; Mao, J.; Alderson, N.; Hopwood, J. M.

    2012-07-01

    The Enhanced CANDU 6{sup R} (ECo{sup R}) is the updated version of the well established CANDU 6 family of units incorporating improved safety characteristics designed to meet or exceed Generation III nuclear power plant expectations. The EC6 retains the excellent neutron economy and fuel cycle flexibility that are inherent in the CANDU reactor design. The reference design is based on natural uranium fuel, but the EC6 is also able to utilize additional fuel options, including the use of Recovered Uranium (RU) and Thorium based fuels, without requiring major hardware upgrades to the existing control and safety systems. This paper outlines the major changes in the EC6 core design from the existing C6 design that significantly enhance the safety characteristics and operating efficiency of the reactor. The use of RU fuel as a transparent replacement fuel for the standard 37-el NU fuel, and several RU based advanced fuel designs that give significant improvements in fuel burnup and inherent safety characteristics are also discussed in the paper. In addition, the suitability of the EC6 to use MOX and related Pu-based fuels will also be discussed. (authors)

  13. Categorization of failed and damaged spent LWR (light-water reactor) fuel currently in storage

    SciTech Connect

    Bailey, W.J.

    1987-11-01

    The results of a study that was jointly sponsored by the US Department of Energy and the Electric Power Research Institute are described in this report. The purpose of the study was to (1) estimate the number of failed fuel assemblies and damaged fuel assemblies (i.e., ones that have sustained mechanical or chemical damage but with fuel rod cladding that is not breached) in storage, (2) categorize those fuel assemblies, and (3) prepare this report as an authoritative, illustrated source of information on such fuel. Among the more than 45,975 spent light-water reactor fuel assemblies currently in storage in the United States, it appears that there are nearly 5000 failed or damaged fuel assemblies. 78 refs., 23 figs., 19 tabs.

  14. Large-break LOCA, in-reactor fuel bundle Materials Test MT-6A

    SciTech Connect

    Wilson, C.L.; Hesson, G.M.; Pilger, J.P.; King, L.L.; Panisko, F.E.

    1993-09-01

    This is a report on one of a series of experiments to simulates a loss-of-coolant accident (LOCA) using full-length fuel rods for pressurized water reactors (PWR). The experiments were conducted by Pacific Northwest Laboratory (PNL) under the LOCA simulation Program sponsored by the US Nuclear Regulatory Commission (NRC). The major objective of this program was causing the maximum possible expansion of the cladding on the fuel rods from a short-term adiabatic temperature transient to 1200 K (1700 F) leading to the rupture of the cladding; and second, by reflooding the fuel rods to determine the rate at which the fuel bundle is cooled.

  15. Deep Burn Develpment of Transuranic Fuel for High-Temperature Helium-Cooled Reactors - July 2010

    SciTech Connect

    Snead, Lance Lewis; Besmann, Theodore M; Collins, Emory D; Bell, Gary L

    2010-08-01

    The DB Program Quarterly Progress Report for April - June 2010, ORNL/TM/2010/140, was distributed to program participants on August 4. This report discusses the following: (1) TRU (transuranic elements) HTR (high temperature helium-cooled reactor) Fuel Modeling - (a) Thermochemical Modeling, (b) 5.3 Radiation Damage and Properties; (2) TRU HTR Fuel Qualification - (a) TRU Kernel Development, (b) Coating Development, (c) ZrC Properties and Handbook; and (3) HTR Fuel Recycle - (a) Recycle Processes, (b) Graphite Recycle, (c) Pyrochemical Reprocessing - METROX (metal recovery from oxide fuel) Process Development.

  16. Public information circular for shipments of irradiated reactor fuel

    SciTech Connect

    1996-07-01

    This circular provides information on shipment of spent fuel subject to regulation by US NRC. It provides a brief description of spent fuel shipment safety and safeguards requirement of general interest, a summary of data for 1979-1995 highway and railway shipments, and a listing, by State, of recent highway and railway shipment routes. The enclosed route information reflects specific NRC approvals that have been granted in response to requests for shipments of spent fuel. This publication does not constitute authority for carriers or other persons to use the routes described to ship spent fuel, other categories of nuclear waste, or other materials.

  17. Development of Improved Models and Designs for Coated-Particle Gas Reactor Fuels (I-NERI Annual Report)

    SciTech Connect

    Petti, David Andrew; Maki, John Thomas; Languille, Alain; Martin, Philippe; Ballinger, Ronald

    2002-11-01

    The objective of this INERI project is to develop improved fuel behavior models for gas reactor coated particle fuels and to develop improved coated-particle fuel designs that can be used reliably at very high burnups and potentially in fast gas-cooled reactors. Thermomechanical, thermophysical, and physiochemical material properties data were compiled by both the US and the French and preliminary assessments conducted. Comparison between U.S. and European data revealed many similarities and a few important differences. In all cases, the data needed for accurate fuel performance modeling of coated particle fuel at high burnup were lacking. The development of the INEEL fuel performance model, PARFUME, continued from earlier efforts. The statistical model being used to simulate the detailed finite element calculations is being upgraded and improved to allow for changes in fuel design attributes (e.g. thickness of layers, dimensions of kernel) as well as changes in important material properties to increase the flexibility of the code. In addition, modeling of other potentially important failure modes such as debonding and asphericity was started. A paper on the status of the model was presented at the HTR-2002 meeting in Petten, Netherlands in April 2002, and a paper on the statistical method was submitted to the Journal of Nuclear Material in September 2002. Benchmarking of the model against Japanese and an older DRAGON irradiation are planned. Preliminary calculations of the stresses in a coated particle have been calculated by the CEA using the ATLAS finite element model. This model and the material properties and constitutive relationships will be incorporated into a more general software platform termed Pleiades. Pleiades will be able to analyze different fuel forms at different scales (from particle to fuel body) and also handle the statistical variability in coated particle fuel. Diffusion couple experiments to study Ag and Pd transport through SiC were

  18. Mixed oxide fuels testing in the advanced test reactor to support plutonium disposition

    SciTech Connect

    Ryskamp, J.M.; Sterbentz, J.W.; Chang, G.S.

    1995-09-01

    An intense worldwide effort is now under way to find means of reducing the stockpile of weapons-grade plutonium. One of the most attractive solutions would be to use WGPu as fuel in existing light water reactors (LWRs) in the form of mixed oxide (MOX) fuel - i.e., plutonia (PUO{sub 2}) mixed with urania (UO{sub 2}). Before U.S. reactors could be used for this purpose, their operating licenses would have to be amended. Numerous technical issues must be resolved before LWR operating licenses can be amended to allow the use of MOX fuel. These issues include the following: (1) MOX fuel fabrication process verification, (2) Whether and how to use burnable poisons to depress MOX fuel initial reactivity, which is higher than that of urania, (3) The effects of WGPu isotopic composition, (4) The feasibility of loading MOX fuel with plutonia content up to 7% by weight, (5) The effects of americium and gallium in WGPu, (6) Fission gas release from MOX fuel pellets made from WGPu, (7) Fuel/cladding gap closure, (8) The effects of power cycling and off-normal events on fuel integrity, (9) Development of radial distributions of burnup and fission products, (10) Power spiking near the interfaces of MOX and urania fuel assemblies, and (11) Fuel performance code validation. We have performed calculations to show that the use of hafnium shrouds can produce spectrum adjustments that will bring the flux spectrum in ATR test loops into a good approximation to the spectrum anticipated in a commercial LWR containing MOX fuel while allowing operation of the test fuel assemblies near their optimum values of linear heat generation rate. The ATR would be a nearly ideal test bed for developing data needed to support applications to license LWRs for operation with MOX fuel made from weapons-grade plutonium. The requirements for planning and implementing a test program in the ATR have been identified.

  19. Characterization of an Experimental Referee Broadened Specification (ERBS) aviation turbine fuel and ERBS fuel blends

    NASA Technical Reports Server (NTRS)

    Seng, G. T.

    1982-01-01

    Characterization data and comparisons of these data are presented for three individual lots of a research test fuel designated as an Experimental Referee Broadened Specification (ERBS) aviation turbine fuel. This research fuel, which is a blend of kerosene and hydrotreated catalytic gas oil, is a representation of a kerojet fuel with broadened properties. To lower the hydrogen content of the ERBS fuel, a blending stock, composed of xylene bottoms and hydrotreated catalytic gas oil, was developed and employed to produce two different ERBS fuel blends. The ERBS fuel blends and the blending stock were also characterized and the results for the blends are compared to those of the original ERBS fuel. The characterization results indicate that with the exception of the freezing point for ERBS lot 2, which was slightly high, the three lots, produced over a 2 year period, met all general fuel requirements. However, although the properties of the fuels were found to be fairly consistent, there were differences in composition. Similarly, all major requirements for the ERBS fuel blends were met or closely approached, and the properties of the blended fuels were found to generally reflect those expected for the proportions of ERBS fuel and blending stock used in their production.

  20. Status of the NGNP fuel experiment AGR-2 irradiated in the advanced test reactor

    SciTech Connect

    S. Blaine Grover; David A. Petti

    2014-05-01

    The United States Department of Energy's Next Generation Nuclear Plant (NGNP) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating up to seven separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States, and will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of at least six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also undergo on-line fission product monitoring to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and was completed in November 2009. The second experiment (AGR-2), which utilized the same experiment design as well as control and monitoring systems as AGR-1, started irradiation in June 2010 and is currently scheduled to be completed in April 2013. The design of this experiment and sup

  1. Fission Product Monitoring of TRISO Coated Fuel For The Advanced Gas Reactor -1 Experiment

    SciTech Connect

    Dawn M. Scates; John K. Hartwell; John b. Walter

    2010-10-01

    The US Department of Energy has embarked on a series of tests of TRISO-coated particle reactor fuel intended for use in the Very High Temperature Reactor (VHTR) as part of the Advanced Gas Reactor (AGR) program. The AGR-1 TRISO fuel experiment, currently underway, is the first in a series of eight fuel tests planned for irradiation in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The AGR-1 experiment reached a peak compact averaged burn up of 9% FIMA with no known TRISO fuel particle failures in March 2008. The burnup goal for the majority of the fuel compacts is to have a compact averaged burnup greater than 18% FIMA and a minimum compact averaged burnup of 14% FIMA. At the INL the TRISO fuel in the AGR-1 experiment is closely monitored while it is being irradiated in the ATR. The effluent monitoring system used for the AGR-1 fuel is the Fission Product Monitoring System (FPMS). The FPMS is a valuable tool that provides near real-time data indicative of the AGR-1 test fuel performance and incorporates both high-purity germanium (HPGe) gamma-ray spectrometers and sodium iodide [NaI(Tl)] scintillation detector-based gross radiation monitors. To quantify the fuel performance, release-to-birth ratios (R/B’s) of radioactive fission gases are computed. The gamma-ray spectra acquired by the AGR-1 FPMS are analyzed and used to determine the released activities of specific fission gases, while a dedicated detector provides near-real time count rate information. Isotopic build up and depletion calculations provide the associated isotopic birth rates. This paper highlights the features of the FPMS, encompassing the equipment, methods and measures that enable the calculation of the release-to-birth ratios. Some preliminary results from the AGR-1 experiment are also presented.

  2. Status of the NGNP Fuel Experiment AGR-2 Irradiated in the Advanced Test Reactor

    SciTech Connect

    Blaine Grover

    2012-10-01

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating up to seven separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States, and will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of at least six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and was completed in November 2009. The second experiment (AGR-2), which utilized the same experiment design as well as control and monitoring systems as AGR-1, started irradiation in June 2010 and is currently scheduled to be completed in April 2013. The design of this experiment and support systems will be briefly discussed, followed by the progress and status of the experiment to date.

  3. Fission Product Monitoring of TRISO Coated Fuel For The Advanced Gas Reactor -1 Experiment

    SciTech Connect

    Dawn M. Scates; John K Hartwell; John B. Walter

    2008-09-01

    The US Department of Energy has embarked on a series of tests of TRISO-coated particle reactor fuel intended for use in the Very High Temperature Reactor (VHTR) as part of the Advanced Gas Reactor (AGR) program. The AGR-1 TRISO fuel experiment, currently underway, is the first in a series of eight fuel tests planned for irradiation in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The AGR-1 experiment reached a peak compact averaged burn up of 9% FIMA with no known TRISO fuel particle failures in March 2008. The burnup goal for the majority of the fuel compacts is to have a compact averaged burnup greater than 18% FIMA and a minimum compact averaged burnup of 14% FIMA. At the INL the TRISO fuel in the AGR-1 experiment is closely monitored while it is being irradiated in the ATR. The effluent monitoring system used for the AGR-1 fuel is the Fission Product Monitoring System (FPMS). The FPMS is a valuable tool that provides near real-time data indicative of the AGR-1 test fuel performance and incorporates both high-purity germanium (HPGe) gamma-ray spectrometers and sodium iodide [NaI(Tl)] scintillation detector-based gross radiation monitors. To quantify the fuel performance, release-to-birth ratios (R/B’s) of radioactive fission gases are computed. The gamma-ray spectra acquired by the AGR-1 FPMS are analyzed and used to determine the released activities of specific fission gases, while a dedicated detector provides near-real time count rate information. Isotopic build up and depletion calculations provide the associated isotopic birth rates. This paper highlights the features of the FPMS, encompassing the equipment, methods and measures that enable the calculation of the release-to-birth ratios. Some preliminary results from the AGR-1 experiment are also presented.

  4. Survey of Worldwide Light Water Reactor Experience with Mixed Uranium-Plutonium Oxide Fuel

    SciTech Connect

    Cowell, B.S.; Fisher, S.E.

    1999-02-01

    The US and the Former Soviet Union (FSU) have recently declared quantities of weapons materials, including weapons-grade (WG) plutonium, excess to strategic requirements. One of the leading candidates for the disposition of excess WG plutonium is irradiation in light water reactors (LWRs) as mixed uranium-plutonium oxide (MOX) fuel. A description of the MOX fuel fabrication techniques in worldwide use is presented. A comprehensive examination of the domestic MOX experience in US reactors obtained during the 1960s, 1970s, and early 1980s is also presented. This experience is described by manufacturer and is also categorized by the reactor facility that irradiated the MOX fuel. A limited summary of the international experience with MOX fuels is also presented. A review of MOX fuel and its performance is conducted in view of the special considerations associated with the disposition of WG plutonium. Based on the available information, it appears that adoption of foreign commercial MOX technology from one of the successful MOX fuel vendors will minimize the technical risks to the overall mission. The conclusion is made that the existing MOX fuel experience base suggests that disposition of excess weapons plutonium through irradiation in LWRs is a technically attractive option.

  5. Nuclear reactors using fine-particulate fuel for primary power in space

    SciTech Connect

    Botts, T.E.; Powell, J.R.; Usher, J.L.; Horn, F.L.

    1982-01-01

    Large future power requirements in space, include power beaming to earth, airplanes, and solar-powered satellites in eclipse; industrial processing; and space colonies. The Rotating Bed Nuclear Reactor (RBR) and Fixed Bed Reactor (FBR) are multi-megawatt power systems which are light, compact and suited to operation in space. Both are cavity reactors, with an annular fuel region (e.g., a bed of 500 ..mu.. HTGR fuel particulates made of UC with ceramic coating) surrounded by a reflector that moderates fast neutrons from the /sup 235/U fuel. A porous metal drum holds the fuel. In the RBR, rotation of the drum allows the particulate fuel bed to fluidize as cooling gas passes through. In the FBR, an inner porous carbon drum holds the packed fuel bed, which is not fluidized. The RBR and FBR have many important features for space nuclear power: very high power density (up to thousands of MW(th)/m/sup 3/ of fuel); very small size and weight, excellent thermal shock and fatigue resistance; short start/stop times (sec); high gas outlet temperatures (to 3000/sup 0/K), good neutron economy, low critical mass; and simple/reliable construction.

  6. Options Study Documenting the Fast Reactor Fuels Innovative Design Activity

    SciTech Connect

    Jon Carmack; Kemal Pasamehmetoglu

    2010-07-01

    This document provides presentation and general analysis of innovative design concepts submitted to the FCRD Advanced Fuels Campaign by nine national laboratory teams as part of the Innovative Transmutation Fuels Concepts Call for Proposals issued on October 15, 2009 (Appendix A). Twenty one whitepapers were received and evaluated by an independent technical review committee.

  7. Radionuclide characterization of reactor decommissioning waste and neutron-activated metals

    SciTech Connect

    Robertson, D.E.; Thomas, C.W.; Wynhoff, N.L.; Haggard, D.L.

    1993-06-01

    This study is providing the NRC and licensees with a more comprehensive data base for regulatory assessment of the radiological factors associated with reactor decommissioning and disposal of wastes generated during these activities. The objectives of this study are being accomplished during the actual decommissioning of Shippingport Station and the detailed analysis of neutron-activated materials from commercial reactors. The radiological characterization studies of Shippingport decommissioning materials have now been completed, and analyses of dismantled piping and scabbled concrete have shown that neutron activation products, dominated by {sup 60}Co, comprised the residual radionuclide inventory. Waste classification assessment have shown that all decommissioning materials (except reactor pressure vessel internals) could be disposed of as Class A waste. Spent fuel disassembly hardware from the Shippingport Core-3 was analyzed for long-lived activation products. {sup 94}Nb and {sup 63}Ni concentrations in Inconel-X750 and stainless steel components exceeded their Class C limits. Measurements and assessments of {sup 14}C in spent fuel disassembly hardware from three commercial nuclear power stations showed that this radionuclide never exceeded the Class C limit for all components. However, the {sup 63}Ni and {sup 94}Nb concentrations in some of these materials did exceed the Class C limits. These measurements are providing the basis for an assessment of the disposal options for these types of highly radioactive materials. Work is continuing on radiological characterization of spent PWR and BWR control rod assemblies. Three control rods, including a BWR cruciform control rod blade, a PWR control rod cluster assembly, and a PWR burnable poison rod assembly, have been characterized for their long-lived activation product concentrations and distribution by direct assay methods. These spent control rods could all be classified as Class C low-level waste.

  8. Radionuclide characterization of reactor decommissioning waste and neutron-activated metals

    SciTech Connect

    Robertson, D.E.; Thomas, C.W.; Wynhoff, N.L.; Haggard, D.L.

    1993-06-01

    This study is providing the NRC and licensees with a more comprehensive data base for regulatory assessment of the radiological factors associated with reactor decommissioning and disposal of wastes generated during these activities. The objectives of this study are being accomplished during the actual decommissioning of Shippingport Station and the detailed analysis of neutron-activated materials from commercial reactors. The radiological characterization studies of Shippingport decommissioning materials have now been completed, and analyses of dismantled piping and scabbled concrete have shown that neutron activation products, dominated by Co-60, comprised the residual radionuclide inventory. Waste classification assessment have shown that all decommissioning materials (except reactor pressure vessel internals) could be disposed of as Class A waste. Spent fuel disassembly hardware from the Shippingport Core-3 was analyzed for long-lived activation products. Nb-94 and Ni-63 concentrations in Inconel-X750 and stainless steel components exceeded their Class C limits. Measurements and assessments of C-14 in spent fuel disassembly hardware from three commercial nuclear power stations showed that this radionuclide never exceeded the Class C limit for al components. However, the Ni-63 and Nb-94 concentrations in some of these materials did exceed the Class C limits. These measurements are providing the basis for an assessment of the disposal options for these types of highly radioactive materials. Work is continuing on radiological characterization of spent PWR and BWR control rod assemblies. Three control rods, including a BWR cruciform control rod blade, a PWR control rod cluster assembly, and a PWR burnable poison rod assembly, have been characterized for their long-lived activation product concentrations and distribution by direct assay methods. These spent control rods could all be classified as Class C low-level waste. These rods are presently being sampled.

  9. Fuel supply of nuclear power industry with the introduction of fast reactors

    NASA Astrophysics Data System (ADS)

    Muraviev, E. V.

    2014-12-01

    The results of studies conducted for the validation of the updated development strategy for nuclear power industry in Russia in the 21st century are presented. Scenarios with different options for the reprocessing of spent fuel of thermal reactors and large-scale growth of nuclear power industry based on fast reactors of inherent safety with a breeding ratio of ˜1 in a closed nuclear fuel cycle are considered. The possibility of enhanced fuel breeding in fast reactors is also taken into account in the analysis. The potential to establish a large-scale nuclear power industry that covers 100% of the increase in electric power requirements in Russia is demonstrated. This power industry may be built by the end of the century through the introduction of fast reactors (replacing thermal ones) with a gross uranium consumption of up to ˜1 million t and the termination of uranium mining even if the reprocessing of spent fuel of thermal reactors is stopped or suffers a long-term delay.

  10. Helium Leak Detection of Vessels in Fuel Transfer Cell (FTC) of Prototype Fast Breeder Reactor (PFBR)

    NASA Astrophysics Data System (ADS)

    Dutta, N. G.

    2012-11-01

    Bharatiya Nabhikiya Vidyut Nigam (BHAVINI) is engaged in construction of 500MW Prototype Fast Breeder Reactor (PFBR) at Kalpak am, Chennai. In this very important and prestigious national programme Special Product Division (SPD) of M/s Kay Bouvet Engg.pvt. ltd. (M/s KBEPL) Satara is contributing in a major way by supplying many important sub-assemblies like- Under Water trolley (UWT), Airlocks (PAL, EAL) Container and Storage Rack (CSR) Vessels in Fuel Transfer Cell (FTC) etc for PFBR. SPD of KBEPL caters to the requirements of Government departments like - Department of Atomic Energy (DAE), BARC, Defense, and Government undertakings like NPCIL, BHAVINI, BHEL etc. and other precision Heavy Engg. Industries. SPD is equipped with large size Horizontal Boring Machines, Vertical Boring Machines, Planno milling, Vertical Turret Lathe (VTL) & Radial drilling Machine, different types of welding machines etc. PFBR is 500 MWE sodium cooled pool type reactor in which energy is produced by fissions of mixed oxides of Uranium and Plutonium pellets by fast neutrons and it also breeds uranium by conversion of thorium, put along with fuel rod in the reactor. In the long run, the breeder reactor produces more fuel then it consumes. India has taken the lead to go ahead with Fast Breeder Reactor Programme to produce electricity primarily because India has large reserve of Thorium. To use Thorium as further fuel in future, thorium has to be converted in Uranium by PFBR Technology.

  11. Status of Fuel Development and Manufacturing for Space Nuclear Reactors at BWX Technologies

    SciTech Connect

    Carmack, W.J.; Husser, D.L.; Mohr, T.C.; Richardson, W.C.

    2004-02-04

    New advanced nuclear space propulsion systems will soon seek a high temperature, stable fuel form. BWX Technologies Inc (BWXT) has a long history of fuel manufacturing. UO2, UCO, and UCx have been fabricated at BWXT for various US and international programs. Recent efforts at BWXT have focused on establishing the manufacturing techniques and analysis capabilities needed to provide a high quality, high power, compact nuclear reactor for use in space nuclear powered missions. To support the production of a space nuclear reactor, uranium nitride has recently been manufactured by BWXT. In addition, analytical chemistry and analysis techniques have been developed to provide verification and qualification of the uranium nitride production process. The fabrication of a space nuclear reactor will require the ability to place an unclad fuel form into a clad structure for assembly into a reactor core configuration. To this end, BWX Technologies has reestablished its capability for machining, GTA welding, and EB welding of refractory metals. Specifically, BWX Technologies has demonstrated GTA welding of niobium flat plate and EB welding of niobium and Nb-1Zr tubing. In performing these demonstration activities, BWX Technologies has established the necessary infrastructure to manufacture UO2, UCx, or UNx fuel, components, and complete reactor assemblies in support of space nuclear programs.

  12. Status of Fuel Development and Manufacturing for Space Nuclear Reactors at BWX Technologies

    NASA Astrophysics Data System (ADS)

    Carmack, W. J.; Husser, D. L.; Mohr, T. C.; Richardson, W. C.

    2004-02-01

    New advanced nuclear space propulsion systems will soon seek a high temperature, stable fuel form. BWX Technologies Inc (BWXT) has a long history of fuel manufacturing. UO2, UCO, and UCx have been fabricated at BWXT for various US and international programs. Recent efforts at BWXT have focused on establishing the manufacturing techniques and analysis capabilities needed to provide a high quality, high power, compact nuclear reactor for use in space nuclear powered missions. To support the production of a space nuclear reactor, uranium nitride has recently been manufactured by BWXT. In addition, analytical chemistry and analysis techniques have been developed to provide verification and qualification of the uranium nitride production process. The fabrication of a space nuclear reactor will require the ability to place an unclad fuel form into a clad structure for assembly into a reactor core configuration. To this end, BWX Technologies has reestablished its capability for machining, GTA welding, and EB welding of refractory metals. Specifically, BWX Technologies has demonstrated GTA welding of niobium flat plate and EB welding of niobium and Nb-1Zr tubing. In performing these demonstration activities, BWX Technologies has established the necessary infrastructure to manufacture UO2, UCx, or UNx fuel, components, and complete reactor assemblies in support of space nuclear programs.

  13. Further evaluations of the toxicity of irradiated advanced heavy water reactor fuels.

    PubMed

    Edwards, Geoffrey W R; Priest, Nicholas D

    2014-11-01

    The neutron economy and online refueling capability of heavy water moderated reactors enable them to use many different fuel types, such as low enriched uranium, plutonium mixed with uranium, or plutonium and/or U mixed with thorium, in addition to their traditional natural uranium fuel. However, the toxicity and radiological protection methods for fuels other than natural uranium are not well established. A previous paper by the current authors compared the composition and toxicity of irradiated natural uranium to that of three potential advanced heavy water fuels not containing plutonium, and this work uses the same method to compare irradiated natural uranium to three other fuels that do contain plutonium in their initial composition. All three of the new fuels are assumed to incorporate plutonium isotopes characteristic of those that would be recovered from light water reactor fuel via reprocessing. The first fuel investigated is a homogeneous thorium-plutonium fuel designed for a once-through fuel cycle without reprocessing. The second fuel is a heterogeneous thorium-plutonium-U bundle, with graded enrichments of U in different parts of a single fuel assembly. This fuel is assumed to be part of a recycling scenario in which U from previously irradiated fuel is recovered. The third fuel is one in which plutonium and Am are mixed with natural uranium. Each of these fuels, because of the presence of plutonium in the initial composition, is determined to be considerably more radiotoxic than is standard natural uranium. Canadian nuclear safety regulations require that techniques be available for the measurement of 1 mSv of committed effective dose after exposure to irradiated fuel. For natural uranium fuel, the isotope Pu is a significant contributor to the committed effective dose after exposure, and thermal ionization mass spectrometry is sensitive enough that the amount of Pu excreted in urine is sufficient to estimate internal doses, from all isotopes, as low

  14. Further evaluations of the toxicity of irradiated advanced heavy water reactor fuels.

    PubMed

    Edwards, Geoffrey W R; Priest, Nicholas D

    2014-11-01

    The neutron economy and online refueling capability of heavy water moderated reactors enable them to use many different fuel types, such as low enriched uranium, plutonium mixed with uranium, or plutonium and/or U mixed with thorium, in addition to their traditional natural uranium fuel. However, the toxicity and radiological protection methods for fuels other than natural uranium are not well established. A previous paper by the current authors compared the composition and toxicity of irradiated natural uranium to that of three potential advanced heavy water fuels not containing plutonium, and this work uses the same method to compare irradiated natural uranium to three other fuels that do contain plutonium in their initial composition. All three of the new fuels are assumed to incorporate plutonium isotopes characteristic of those that would be recovered from light water reactor fuel via reprocessing. The first fuel investigated is a homogeneous thorium-plutonium fuel designed for a once-through fuel cycle without reprocessing. The second fuel is a heterogeneous thorium-plutonium-U bundle, with graded enrichments of U in different parts of a single fuel assembly. This fuel is assumed to be part of a recycling scenario in which U from previously irradiated fuel is recovered. The third fuel is one in which plutonium and Am are mixed with natural uranium. Each of these fuels, because of the presence of plutonium in the initial composition, is determined to be considerably more radiotoxic than is standard natural uranium. Canadian nuclear safety regulations require that techniques be available for the measurement of 1 mSv of committed effective dose after exposure to irradiated fuel. For natural uranium fuel, the isotope Pu is a significant contributor to the committed effective dose after exposure, and thermal ionization mass spectrometry is sensitive enough that the amount of Pu excreted in urine is sufficient to estimate internal doses, from all isotopes, as low

  15. Subcritical Noise Analysis Measurements with Fresh and Spent Research Reactor Fuels Elements

    SciTech Connect

    Valentine, T.E.; Mihalczo, J.T.; Kryter, R.C.; Miller, V.C.

    1999-02-01

    The verification of the subcriticality is of utmost importance for the safe transportation and storage of nuclear reactor fuels. Transportation containers and storage facilities are designed such that nuclear fuels remain in a subcritical state. Such designs often involve excess conservatism because of the lack of relevant experimental data to verify the accuracy of Monte Carlo codes used in nuclear criticality safety analyses. A joint experimental research program between Oak Ridge National Laboratory, Westinghouse Safety Management Solutions, Inc., and the University of Missouri was initiated to obtain measured quantities that could be directly related to the subcriticality of simple arrays of Missouri University Research Reactor (MURR) fuel elements. A series of measurement were performed to assess the reactivity of materials such as BORAL, stainless steel, aluminum, and lead that are typically used in the construction of shipping casks. These materials were positioned between the fuel elements. In addition, a limited number of measurements were performed with configurations of fresh and spent (irradiated) fuel elements to ascertain the reactivity of the spent fuel elements. In these experiments, fresh fuel elements were replaced by spent fuel elements such that the subcritical reactivity change could be measured. The results of these measurements were used by Westinghouse Safety Management Solutions to determine the subcriticality of MURR fuel elements isolated by absorbing materials. The measurements were interpreted using the MCNP-DSP Monte Carlo code to obtain the subcritical neutron multiplication factor k(sub eff), and the bias in K(sub eff) that are used in criticality safety analyses.

  16. Analysis of steam explosions in plate-type, uranium-aluminum fuel test reactors

    SciTech Connect

    Taleyarkhan, R.P. )

    1989-01-01

    The concern over steam explosions in nuclear reactors can be traced to prompt critical nuclear excursions in aluminum-clad/fueled test reactors, as well as to explosive events in aluminum, pulp, and paper industries. The Reactor Safety Study prompted an extensive analytical and experimental effort for over a decade. This has led to significant improvements in their understanding of the steam explosion issue for commercial light water reactors. However, little progress has been made toward applying the lessons learned from this effort to the understanding and modeling of steam explosion phenomena in aluminum-clad/fueled research and test reactors. The purposes of this paper are to (a) provide a preliminary analysis of the destructive events in test reactors, based on current understandings of steam explosions; (b) provide a proposed approach for determining the likelihood of a steam explosion event under scenarios in which molten U-Al fuel drops into a water-filled cavity; and (c) present a benchmarking study conducted to estimate peak pressure pulse magnitudes.

  17. On0Line Fuel Failure Monitor for Fuel Testing and Monitoring of Gas Cooled Very High Temperature Reactor

    SciTech Connect

    Ayman I. Hawari; Mohamed A. Bourham

    2010-04-22

    IVery High Temperature Reactors (VHTR) utilize the TRISO microsphere as the fundamental fuel unit in the core. The TRISO microsphere (~ 1- mm diameter) is composed of a UO2 kernel surrounded by a porous pyrolytic graphite buffer, an inner pyrolytic graphite layer, a silicon carbide (SiC) coating, and an outer pyrolytic graphite layer. The U-235 enrichment of the fuel is expected to range from 4% – 10% (higher enrichments are also being considered). The layer/coating system that surrounds the UO2 kernel acts as the containment and main barrier against the environmental release of radioactivity. To understand better the behavior of this fuel under in-core conditions (e.g., high temperature, intense fast neutron flux, etc.), the US Department of Energy (DOE) is launching a fuel testing program that will take place at the Advanced Test Reactor (ATR) located at Idaho National Laboratory (INL). During this project North Carolina State University (NCSU) researchers will collaborate with INL staff for establishing an optimized system for fuel monitoring for the ATR tests. In addition, it is expected that the developed system and methods will be of general use for fuel failure monitoring in gas cooled VHTRs.

  18. The benefits of a fast reactor closed fuel cycle in the UK

    SciTech Connect

    Gregg, R.; Hesketh, K.

    2013-07-01

    The work has shown that starting a fast reactor closed fuel cycle in the UK, requires virtually all of Britain's existing and future PWR spent fuel to be reprocessed, in order to obtain the plutonium needed. The existing UK Pu stockpile is sufficient to initially support only a modest SFR 'closed' fleet assuming spent fuel can be reprocessed shortly after discharge (i.e. after two years cooling). For a substantial fast reactor fleet, most Pu will have to originate from reprocessing future spent PWR fuel. Therefore, the maximum fast reactor fleet size will be limited by the preceding PWR fleet size, so scenarios involving fast reactors still require significant quantities of uranium ore indirectly. However, once a fast reactor fuel cycle has been established, the very substantial quantities of uranium tails in the UK would ensure there is sufficient material for several centuries. Both the short and long term impacts on a repository have been considered in this work. Over the short term, the decay heat emanating from the HLW and spent fuel will limit the density of waste within a repository. For scenarios involving fast reactors, the only significant heat bearing actinide content will be present in the final cores, resulting in a 50% overall reduction in decay energy deposited within the repository when compared with an equivalent open fuel cycle. Over the longer term, radiological dose becomes more important. Total radiotoxicity (normalised by electricity generated) is lower for scenarios with Pu recycle after 2000 years. Scenarios involving fast reactors have the lowest radiotoxicity since the quantities of certain actinides (Np, Pu and Am) eventually stabilise. However, total radiotoxicity as a measure of radiological risk does not account for differences in radionuclide mobility once in repository. Radiological dose is dominated by a small number of fission products so is therefore not affected significantly by reactor type or recycling strategy (since the

  19. Irradiation performance of (Th,Pu)O2 fuel under Pressurized Water Reactor conditions

    NASA Astrophysics Data System (ADS)

    Boer, B.; Lemehov, S.; Wéber, M.; Parthoens, Y.; Gysemans, M.; McGinley, J.; Somers, J.; Verwerft, M.

    2016-04-01

    This paper examines the in-pile safety performance of (Th,Pu)O2 fuel pins under simulated Pressurized Water Reactor (PWR) conditions. Both sol-gel and SOLMAS produced (Th,Pu)O2 fuels at enrichments of 7.9% and 12.8% in Pu/HM have been irradiated at SCK·CEN. The irradiation has been performed under PWR conditions (155 bar, 300 °C) in a dedicated loop of the BR-2 reactor. The loop is instrumented with flow and temperature monitors at inlet and outlet, which allow for an accurate measurement of the deposited enthalpy.

  20. Fission product release phenomena during core melt accidents in metal fueled heavy water reactors

    SciTech Connect

    Ellison, P G; Hyder, M L; Monson, P R; Randolph, H W; Hagrman, D L; McClure, P R; Leonard, M T

    1990-01-01

    The phenomena that determine fission product release rates from a core melting accident in a metal-fueled, heavy water reactor are described in this paper. This information is obtained from the analysis of the current metal fuel experimental data base and from the results of analytical calculations. Experimental programs in place at the Savannah River Site are described that will provide information to resolve uncertainties in the data base. The results of the experiments will be incorporated into new severe accident computer codes recently developed for this reactor design. 47 refs., 4 figs.

  1. Characterization of Suspect Fuel Rod Pieces from the 105 K West Basin

    SciTech Connect

    Delegard, Calvin H.; Schmidt, Andrew J.; Pool, Karl N.; Thornton, Brenda M.

    2006-09-15

    This report provides physical and radiochemical characterization results from examinations and laboratory analyses performed on ~0.55-inch diameter rod pieces found in the 105 K West (KW) Basin that were suspected to be from nuclear reactor fuel. The characterization results will be used to establish the technical basis for adding this material to the contents of one of the final Multi-Canister Overpacks (MCOs) that will be loaded out of the KW Basin in late FY2006 or at a later time depending on project priorities. Fifteen fuel rod pieces were found during the clean out of the KW Basin. Based on lack of specific credentials, documentation, or obvious serial numbers, none of the items could be positively identified nor could their sources or compositions be described. Item weights and dimensions measured in the KW Basin indicated densities consistent with the suspect fuel rods containing uranium dioxide (UO2), uranium metal, or being empty. Extensive review of the Hanford Site technical literature led to the postulation that these pieces likely were irradiated test fuel prepared to support of the development of the Hanford “New Production Reactor,” later called N Reactor. To obtain definitive data on the composition of the suspect fuel, 4 representative fuel rod pieces, with densities corresponding to oxide fuel were selected from the 15 items, and shipped from the KW Basin to the Pacific Northwest National Laboratory’s (PNNL) Radiological Processing Laboratory (RPL; also known at the 325 Building) for examinations and characterization. The three fuel rod that were characterized appear to contain slightly irradiated UO2 fuel, originally of natural enrichment, with zirconium cladding. The uranium-235 isotopic concentrations decreased by the irradiation and become slightly lower than the natural enrichment of 0.72% to range from 0.67 to 0.71 atom%. The plutonium concentrations, ranged from about 200 to 470 grams per metric ton of uranium and ranged in Plutonium

  2. New concept of designing Pu and MA containing fuel for fast reactors

    NASA Astrophysics Data System (ADS)

    Savchenko, A. M.; Konovalov, I. I.; Vatulin, A. V.; Glagovsky, E. M.

    2009-03-01

    New type of metal base fuel element is suggested for fast reactors. Basic approach to fuel element development - separated operations of fabricating uranium meat fuel element and introducing into it Pu or MA dioxides powder, that results in minimizing dust forming operations in fuel element fabrication. According to new fuel element design a framework fuel element having a porous uranium alloy meat is filled with standard PuO 2 powder of <50 μm fractions prepared by pyrochemical or other methods. In this way a high uranium content fuel meat metallurgically bonded to cladding forms a heat conducting framework, pores of which contain PuO 2 powder. Framework fuel element having porous meat is fabricated by capillary impregnation method with the use of Zr eutectic matrix alloys, which provides metallurgical bond between fuel and cladding and protects it from interaction. As compared to MOX fuel the new one features high thermal conductivity, higher uranium content, hence, high conversion ratio does not interact with fuel cladding and is more environmentally clean. Its principle advantage is a simple production process that is easily realized remotely, feasibility of involving high background Pu and MA isotopes into closed nuclear fuel cycle at the minimal influence on environment.

  3. Environmental Assessment of Urgent-Relief Acceptance of Foreign Research Reactor Spent Nuclear Fuel

    SciTech Connect

    Not Available

    1994-04-01

    The Department of Energy has completed the Environmental Assessment (EA) of Urgent-Relief Acceptance of Foreign Research Reactor Spent Nuclear Fuel and issued a Finding of No Significant Impact (FONSI) for the proposed action. The EA and FONSI are enclosed for your information. The Department has decided to accept a limited number of spent nuclear fuel elements (409 elements) containing uranium that was enriched in the United States from eight research reactors in Austria, Denmark, Germany, Greece, the Netherlands, Sweden, and Switzerland. This action is necessary to maintain the viability of a major US nuclear weapons nonproliferation program to limit or eliminate the use of highly enriched uranium in civil programs. The purpose of the EA is to maintain the cooperation of the foreign research reactor operators with the nonproliferation program while a more extensive Environmental Impact Statement (EIS) is prepared on a proposed broader policy involving the acceptance of up to 15,000 foreign research reactor spent fuel elements over a 10 to 15 year period. Based on an evaluation of transport by commercial container liner or chartered vessel, five eastern seaboard ports, and truck and train modes of transporting the spent fuel overland to the Savannah River Sits, the Department has concluded that no significant impact would result from any combination of port and made of transport. In addition, no significant impacts were found from interim storage of spent fuel at the Savannah River Site.

  4. Using graphitic foam as the bonding material in metal fuel pins for sodium fast reactors

    NASA Astrophysics Data System (ADS)

    Karahan, Aydın; Kazimi, Mujid S.

    2013-10-01

    The study evaluates the possible use of graphite foam as the bonding material between U-Pu-Zr metallic fuel and steel clad for sodium fast reactor applications using FEAST-METAL fuel performance code. Furthermore, the applicability of FEAST-METAL to the advanced fuel designs is demonstrated. Replacing the sodium bond with a chemically stable foam material would eliminate fuel clad metallurgical interactions, and allow for fuel swelling under low external stress. Hence, a significant improvement is expected for the steady state and transient performance. FEAST-METAL was used to assess the thermo-mechanical behavior of the new fuel form and a reference metallic fuel pin. Nearly unity conversion ratio, 75% smear density U-15Pu-6Zr metallic fuel pin with sodium bond, and T91 cladding was selected as a reference case. It was found that operating the reference case at high clad temperatures (600-660 °C) results in (1) excessive clad wastage formation/clad thinning due to lanthanide migration and formation of brittle phases at clad inner surface, and (2) excessive clad hoop strain at the upper axial section due mainly to the occurrence of thermal creep. The combination of these two factors may lead to cladding breach. The work concludes that replacing the sodium bond with 80% porous graphite foam and reducing the fuel smear density to 70%, it is likely that the fuel clad metallurgical interaction would be eliminated while the fuel swelling is allowed without excessive fuel clad mechanical interaction. The suggested design appears as an alternative for a high performance metallic fuel design for sodium fast reactors.

  5. Spark Plasma Sintering of Fuel Cermets for Nuclear Reactor Applications

    SciTech Connect

    Yang Zhong; Robert C. O'Brien; Steven D. Howe; Nathan D. Jerred; Kristopher Schwinn; Laura Sudderth; Joshua Hundley

    2011-11-01

    The feasibility of the fabrication of tungsten based nuclear fuel cermets via Spark Plasma Sintering (SPS) is investigated in this work. CeO2 is used to simulate fuel loadings of UO2 or Mixed-Oxide (MOX) fuels within tungsten-based cermets due to the similar properties of these materials. This study shows that after a short time sintering, greater than 90 % density can be achieved, which is suitable to possess good strength as well as the ability to contain fission products. The mechanical properties and the densities of the samples are also investigated as functions of the applied pressures during the sintering.

  6. Lattice cell and full core physics of internally cooled annular fuel in heavy water moderated reactors

    SciTech Connect

    Armstrong, J.; Hamilton, H.; Hyland, B.

    2013-07-01

    A program is underway at Atomic Energy of Canada Limited (AECL) to develop a new fuel bundle concept to enable greater burnups for PT-HWR (pressure tube heavy water reactor) cores. One option that AECL is investigating is an internally cooled annular fuel (ICAF) element concept. ICAF contains annular cylindrical pellets with cladding on the inner and outer diameters. Coolant flows along the outside of the element and through the centre. With such a concept, the maximum fuel temperature as a function of linear element rating is significantly reduced compared to conventional, solid-rod type fuel. The preliminary ICAF bundle concept considered in this study contains 24 half-metre long internally cooled annular fuel elements and one non-fuelled centre pin. The introduction of the non-fuelled centre pin reduces the coolant void reactivity (CVR), which is the increase in reactivity that occurs on voiding the coolant in accident scenarios. Lattice cell and full core physics calculations of the preliminary ICAF fuel bundle concept have been performed for medium burnups of approximately 18 GWd/tU using WIMS-AECL and reactor fuel simulation program (RFSP). The results will be used to assist in concept configuration optimization. The effects of radial and axial core power distributions, linear element power ratings, refuelling rates and operational power ramps have been analyzed. The results suggest that burnups of greater than 18 GWd/tU can be achieved in current reactor designs. At approximately 18 GWd/tU, expected maximum linear element ratings in a PT-HWR with online-refuelling are approximately 90 kW/m. These conditions would be prohibitive for solid-rod fuel, but may be possible in ICAF fuel given the reduced maximum fuel temperature as a function of linear element rating. (authors)

  7. A physical description of fission product behavior fuels for advanced power reactors.

    SciTech Connect

    Kaganas, G.; Rest, J.; Nuclear Engineering Division; Florida International Univ.

    2007-10-18

    The Global Nuclear Energy Partnership (GNEP) is considering a list of reactors and nuclear fuels as part of its chartered initiative. Because many of the candidate materials have not been explored experimentally under the conditions of interest, and in order to economize on program costs, analytical support in the form of combined first principle and mechanistic modeling is highly desirable. The present work is a compilation of mechanistic models developed in order to describe the fission product behavior of irradiated nuclear fuel. The mechanistic nature of the model development allows for the possibility of describing a range of nuclear fuels under varying operating conditions. Key sources include the FASTGRASS code with an application to UO{sub 2} power reactor fuel and the Dispersion Analysis Research Tool (DART ) with an application to uranium-silicide and uranium-molybdenum research reactor fuel. Described behavior mechanisms are divided into subdivisions treating fundamental materials processes under normal operation as well as the effect of transient heating conditions on these processes. Model topics discussed include intra- and intergranular gas-atom and bubble diffusion, bubble nucleation and growth, gas-atom re-solution, fuel swelling and ?scion gas release. In addition, the effect of an evolving microstructure on these processes (e.g., irradiation-induced recrystallization) is considered. The uranium-alloy fuel, U-xPu-Zr, is investigated and behavior mechanisms are proposed for swelling in the {alpha}-, intermediate- and {gamma}-uranium zones of this fuel. The work reviews the FASTGRASS kinetic/mechanistic description of volatile ?scion products and, separately, the basis for the DART calculation of bubble behavior in amorphous fuels. Development areas and applications for physical nuclear fuel models are identified.

  8. Fuel element design for the enhanced destruction of plutonium in a nuclear reactor

    SciTech Connect

    Crawford, Douglas C.; Porter, Douglas L.; Hayes, Steven L.; Hill, Robert N.

    1997-12-01

    A uranium-free fuel for a fast nuclear reactor comprising an alloy of Pu, Zr and Hf, wherein Hf is present in an amount less than about 10% by weight of the alloy. The fuel may be in the form of a Pu alloy surrounded by a Zr-Hf alloy or an alloy of Pu-Zr-Hf or a combination of both.

  9. Nondestructive verification with minimal movement of irradiated light-water-reactor fuel assemblies

    SciTech Connect

    Phillips, J.R.; Bosler, G.E.; Halbig, J.K.; Klosterbuer, S.F.; Menlove, H.O.

    1982-10-01

    Nondestructive verification of irradiated light-water reactor fuel assemblies can be performed rapidly and precisely by measuring their gross gamma-ray and neutron signatures. A portable system measured fuel assemblies with exposures ranging from 18.4 to 40.6 GWd/tU and with cooling times ranging from 1575 to 2638 days. Differences in the measured results for side or corner measurements are discussed. 25 figures, 20 tables.

  10. Flowsheet for shear/leach processing of N Reactor fuel at PUREX

    SciTech Connect

    Enghusen, M.B.

    1995-04-13

    This document was originally prepared to support the restart of the PUREX plant using a new Shear/Leach head end process. However, the PUREX facility was shutdown and processing of the remaining N Reactor fuel is no longer considered an alternative for fuel disposition. This document is being issued for reference only to document the activities which were investigated to incorporate the shear/leach process in the PUREX plant.

  11. Worker exposure for at-reactor management of spent nuclear fuel.

    PubMed

    Weck, Philippe F

    2013-09-01

    The radiological impact on workers associated with spent nuclear fuel dry storage operations at reactor sites is discussed. The resulting doses to workers exposed to external radiation include the dose during dry storage system loading, unloading and handling activities, the dose associated with independent spent fuel storage installation (ISFSI) operations, maintenance and surveillance activities, and the dose associated with additional ISFSI construction. Comprehensive dose estimates are reported based on previous radiation surveys.

  12. Fuel element design for the enhanced destruction of plutonium in a nuclear reactor

    DOEpatents

    Crawford, Douglas C.; Porter, Douglas L.; Hayes, Steven L.; Hill, Robert N.

    1999-01-01

    A uranium-free fuel for a fast nuclear reactor comprising an alloy of Pu, Zr and Hf, wherein Hf is present in an amount less than about 10% by weight of the alloy. The fuel may be in the form of a Pu alloy surrounded by a Zr--Hf alloy or an alloy of Pu--Zr--Hf or a combination of both.

  13. Fuel element design for the enhanced destruction of plutonium in a nuclear reactor

    DOEpatents

    Crawford, D.C.; Porter, D.L.; Hayes, S.L.; Hill, R.N.

    1999-03-23

    A uranium-free fuel for a fast nuclear reactor comprising an alloy of Pu, Zr and Hf, wherein Hf is present in an amount less than about 10% by weight of the alloy. The fuel may be in the form of a Pu alloy surrounded by a Zr--Hf alloy or an alloy of Pu--Zr--Hf or a combination of both. 7 figs.

  14. Worker exposure for at-reactor management of spent nuclear fuel.

    PubMed

    Weck, Philippe F

    2013-09-01

    The radiological impact on workers associated with spent nuclear fuel dry storage operations at reactor sites is discussed. The resulting doses to workers exposed to external radiation include the dose during dry storage system loading, unloading and handling activities, the dose associated with independent spent fuel storage installation (ISFSI) operations, maintenance and surveillance activities, and the dose associated with additional ISFSI construction. Comprehensive dose estimates are reported based on previous radiation surveys. PMID:23564883

  15. K-Basin spent nuclear fuel characterization data report

    SciTech Connect

    Abrefah, J.; Gray, W.J.; Ketner, G.L.; Marschman, S.C.; Pyecha, T.D.; Thornton, T.A.

    1995-11-01

    The spent nuclear fuel (SNF) project characterization activities will be furnishing technical data on SNF stored at the K Basins in support of a pathway for placement of a ``stabilized`` form of SNF into an interim storage facility. This report summarizes the results so far of visual inspection of the fuel samples, physical characterization (e.g., weight and immersion density measurements), metallographic examinations, and controlled atmosphere furnace testing of three fuel samples shipped from the KW Basin to the Postirradiation Testing Laboratory (PTL). Data on sludge material collected by filtering the single fuel element canister (SFEC) water are also discussed in this report.

  16. Fuel development activities of the US RERTR Program. [Reduced Enrichment Research and Test Reactor

    SciTech Connect

    Snelgrove, J.L.; Domagala, R.F.; Wiencek, T.C.; Copeland, G.L.

    1983-01-01

    Progress in the development and irradiation testing of high-density fuels for use with low-enriched uranium in research and test reactors is reported. Swelling and blister-threshold temperature data obtained from the examination of miniature fuel plates containing UAl/sub x/, U/sub 3/O/sub 8/, U/sub 3/Si/sub 2/, or U/sub 3/Si dispersed in an aluminum matrix are presented. Combined with the results of metallurgical examinations, these data show that these four fuel types will perform adequately to full burnup of the /sup 235/U contained in the low-enriched fuel. The exothermic reaction of the uranium-silicide fuels with aluminum has been found to occur at about the same temperature as the melting of the aluminum matrix and cladding and to be essentially quenched by the melting endotherm. A new series of miniature fuel plate irradiations is also discussed.

  17. NUCLEAR REACTOR FUEL ELEMENT AND METHOD OF MANUFACTURE

    DOEpatents

    Brooks, H.

    1960-04-26

    A description is given for a fuel element comprising a body of uranium metal or an uranium compound dispersed in a matrix material made from magnesium, calcium, or barium and a stainless steel jacket enclosing the body.

  18. Proliferation resistance for fast reactors and related fuel cycles: issues and impacts

    SciTech Connect

    Pilat, Joseph F

    2010-01-01

    The prospects for a dramatic growth in nuclear power may depend to a significant degree on the effectiveness of, and the resources devoted to, plans to develop and implement technologies and approaches that strengthen proliferation resistance and nuclear materials accountability. The challenges for fast reactors and related fuel cycles are especially critical. They are being explored in the Generation IV Tnternational Forum (GIF) and the Tnternational Atomic Energy Agency's (IAEA's) International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) initiative, as well as by many states that are looking to these systems for the efficient lise of uranium resources and long-term energy security. How do any proliferation risks they may pose compare to other reactors, both existing and under development, and their fuel cycles? Can they be designed with intrinsic (technological) features to make these systems more proliferation resistant? What roles can extrinsic (institutional) features play in proliferation resistance? What are the anticipated safeguards requirements, and will new technologies and approaches need to be developed? How can safeguards be facilitated by the design process? These and other questions require a rethinking of proliferation resistance and the prospects for new technologies and other intrinsic and extrinsic features being developed that are responsive to specific issues for fast reactors and related fuel cycles and to the broader threat environment in which these systems will have to operate. There are no technologies that can wholly eliminate the risk of proliferation by a determined state, but technology and design can playa role in reducing state threats and perhaps in eliminating non-state threats. There will be a significant role for extrinsic factors, especially the various measures - from safeguards and physical protection to export controls - embodied in the international nuclear nonproliferation regime. This paper will offer

  19. Improved Prediction of the Temperature Feedback in TRISO-Fueled Reactors

    SciTech Connect

    Javier Ortensi; Abderrafi M. Ougouag

    2009-08-01

    The Doppler feedback mechanism is a major contributor to the passive safety of gas-cooled, graphite-moderated high temperature reactors that use fuel based on Tristructural-Isotropic coated particles. It follows that the correct prediction of the magnitude and time-dependence of this feedback effect is essential to the conduct of safety analyses for these reactors. We present a fuel conduction model for obtaining better estimates of the temperature feedback during moderate and fast transients. The fuel model has been incorporated in the CYNOD-THERMIX-KONVEK suite of coupled codes as a single TRISO particle within each calculation cell. The heat generation rate is scaled down from the neutronic solution and a Dirichlet boundary condition is imposed as the bulk graphite temperature from the thermal-hydraulic solution. This simplified approach yields similar results to those obtained with more complex methods, requiring multi-TRISO calculations within one control volume, but with much less computational effort. An analysis of the hypothetical total control ejection in the PBMR-400 design verifies the performance of the code during fast transients. In addition, the analysis of the earthquake-initiated event in the PBMR-400 design verifies the performance of the code during slow transients. These events clearly depict the improvement in the predictions of the fuel temperature, and consequently, of the power escalations. In addition, a brief study of the potential effects of particle layer de-bonding on the transient behavior of high temperature reactors is included. Although the formation of a gap occurs under special conditions its consequences on the dynamic behavior of the reactor should be analyzed. The presence of a gap in the fuel can cause some unusual reactor behavior during fast transients, but still the reactor shuts down due to the strong feedback effects.

  20. Configuration of a molten chloride fast reactor on a thorium fuel cycle to current nuclear fuel cycle concerns

    SciTech Connect

    Ottewitte, E.H.

    1982-01-01

    Current concerns about the nuclear fuel cycle seem to center on waste management, non-proliferation, and optimum fuel utilization (including use of thorium). This thesis attempts to design a fast molten-salt reactor on the thorium fuel cycle to address these concerns and then analyzes its potential performance. The result features (1) A simplified easy-to-replace skewed-tube geometry for the core. (2) A very hard neutron spectrum which allows the useful consumption of all the actinides (no actinide waste). (3) Reduced proliferation risks on the equilibrium cycle compared to conventional fuel cycles because of the absence of carcinogenic, chemically-separable plutonium and the presence of /sup 232/U which gives a tell-tale signal and is hazardous to work with. (4) A breeding gain in the neighborhood of 0.3.