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Sample records for nuclear reactor containment

  1. Collecting and recirculating condensate in a nuclear reactor containment

    DOEpatents

    Schultz, T.L.

    1993-10-19

    An arrangement passively cools a nuclear reactor in the event of an emergency, condensing and recycling vaporized cooling water. The reactor is surrounded by a containment structure and has a storage tank for cooling liquid, such as water, vented to the containment structure by a port. The storage tank preferably is located inside the containment structure and is thermally coupleable to the reactor, e.g. by a heat exchanger, such that water in the storage tank is boiled off to carry away heat energy. The water is released as a vapor (steam) and condenses on the cooler interior surfaces of the containment structure. The condensed water flows downwardly due to gravity and is collected and routed back to the storage tank. One or more gutters are disposed along the interior wall of the containment structure for collecting the condensate from the wall. Piping is provided for communicating the condensate from the gutters to the storage tank. 3 figures.

  2. Collecting and recirculating condensate in a nuclear reactor containment

    DOEpatents

    Schultz, Terry L.

    1993-01-01

    An arrangement passively cools a nuclear reactor in the event of an emergency, condensing and recycling vaporized cooling water. The reactor is surrounded by a containment structure and has a storage tank for cooling liquid, such as water, vented to the containment structure by a port. The storage tank preferably is located inside the containment structure and is thermally coupleable to the reactor, e.g. by a heat exchanger, such that water in the storage tank is boiled off to carry away heat energy. The water is released as a vapor (steam) and condenses on the cooler interior surfaces of the containment structure. The condensed water flows downwardly due to gravity and is collected and routed back to the storage tank. One or more gutters are disposed along the interior wall of the containment structure for collecting the condensate from the wall. Piping is provided for communicating the condensate from the gutters to the storage tank.

  3. Passive cooling system for nuclear reactor containment structure

    DOEpatents

    Gou, Perng-Fei; Wade, Gentry E.

    1989-01-01

    A passive cooling system for the contaminant structure of a nuclear reactor plant providing protection against overpressure within the containment attributable to inadvertent leakage or rupture of the system components. The cooling system utilizes natural convection for transferring heat imbalances and enables the discharge of irradiation free thermal energy to the atmosphere for heat disposal from the system.

  4. Natural circulating passive cooling system for nuclear reactor containment structure

    DOEpatents

    Gou, Perng-Fei; Wade, Gentry E.

    1990-01-01

    A passive cooling system for the contaminant structure of a nuclear reactor plant providing protection against overpressure within the containment attributable to inadvertent leakage or rupture of the system components. The cooling system utilizes natural convection for transferring heat imbalances and enables the discharge of irradiation free thermal energy to the atmosphere for heat disposal from the system.

  5. Neutron Spectra and Dose Equivalent Inside Nuclear Power Reactor Containment

    SciTech Connect

    Aldrich, J. M.

    1981-08-01

    This study was conducted to determine absorbed dose, dose-equivalent rates, and neutron spectra inside containment at nuclear power plants. We gratefully acknowledge funding support by the Nuclear Regulatory Commission. The purpose of this study is: 1) measure dose-equivalent rates with various commercial types of rem meters, such as the Snoopy and Rascal, and neutron absorbed dose rates with a tissue-equivalent proportional counter 2) determine neutron spectra using the multi sphere or Bonner sphere technique and a helium-3 spectrometer 3) compare several types of personnel neutron dosimeter responses such as NTA film, polycarbonates, TLD albedo, and a recently introduced proton recoil track etch dosimeter, and CR-39. These measurements were made inside containments of pressurized water reactors (PWRs) and outside containment penetrations of boiling water reactors (BWRs) operating at full power. The neutron spectral information, absorbed dose. and dose-equivalent measurements are needed for proper interpretation of instrument and personnel dosimeter responses.

  6. Evaluation of a hydrogen sensor for nuclear reactor containment monitoring

    SciTech Connect

    Hoffheins, B.S.; McKnight, T.E.; Lauf, R.J.; Smith, R.R.; James, R.E.

    1997-02-01

    Measurement of hydrogen concentration in containment atmospheres in nuclear plants is a key safety capability. Current technologies require extensive sampling systems and subsequent maintenance and calibration costs can be very expensive. A new hydrogen sensor has been developed that is small and potentially inexpensive to install and maintain. Its size and low power requirement make it suitable in distributed systems for pinpointing hydrogen buildup. This paper will address the first phase of a testing program conducted to evaluate this sensor for operation in reactor containments.

  7. Nuclear reactor containment structure with continuous ring tunnel at grade

    DOEpatents

    Seidensticker, Ralph W.; Knawa, Robert L.; Cerutti, Bernard C.; Snyder, Charles R.; Husen, William C.; Coyer, Robert G.

    1977-01-01

    A nuclear reactor containment structure which includes a reinforced concrete shell, a hemispherical top dome, a steel liner, and a reinforced-concrete base slab supporting the concrete shell is constructed with a substantial proportion thereof below grade in an excavation made in solid rock with the concrete poured in contact with the rock and also includes a continuous, hollow, reinforced-concrete ring tunnel surrounding the concrete shell with its top at grade level, with one wall integral with the reinforced concrete shell, and with at least the base of the ring tunnel poured in contact with the rock.

  8. Characterization of nuclear reactor containment penetrations. Final report

    SciTech Connect

    Shackelford, M.H.; Bump, T.R.; Seidensticker, R.W.

    1985-02-01

    This report concludes a preliminary report prepared by ANL for Sandia, published as NUREG/CR-3855, in June 1984. The preliminary report, NUREG/CR-3855, presented the results of a survey of nuclear reactor containment penetrations, covering the number of plants surveyed at that time (22 total). Since that time, an additional 26 plants have been included in the survey. This final report serves two purposes: (1) to add the summary data sheets and penetration details for the additional plants now included in the survey; and (2) to confirm, revise, or add to analyses and discussions presented in the first report which, of course, were based solely on the earlier sample of 22 plants. This final report follows the outline and format of the preliminary survey report. In general, changes and additions to the preliminary report are implied, rather than stated as such to avoid repeated reference to that report. If no changes have been made in a section the title of the section of the previous report is simply repeated followed by ''No Changes''. Some repetition is used for continuity and clarity.

  9. Characterization of nuclear reactor containment penetrations. Preliminary report

    SciTech Connect

    Bump, T.R.; Seidensticker, R.W.; Shackelford, M.A.; Gambhir, V.K.; McLennan, G.L.

    1984-06-01

    This report summarizes the survey work conducted by Argonne National Laboratory on the design and details of major penetrations in 22 nuclear power plants. The survey includes all containment types and materials in current use. It also includes details of all types of penetrations (except for electrical penetration assemblies and valves) and the seals and gaskets used in them. The report provides a test matrix for testing major penetrations and for testing seals and gaskets in order to evaluate their leakage potential under severe accident conditions.

  10. NUCLEAR REACTOR

    DOEpatents

    Young, G.

    1963-01-01

    This patent covers a power-producing nuclear reactor in which fuel rods of slightly enriched U are moderated by heavy water and cooled by liquid metal. The fuel rods arranged parallel to one another in a circle are contained in a large outer closed-end conduit that extends into a tank containing the heavy water. Liquid metal is introduced into the large conduit by a small inner conduit that extends within the circle of fuel rods to a point near the lower closed end of the outer conduit. (AEC) Production Reactors

  11. Code manual for CONTAIN 2.0: A computer code for nuclear reactor containment analysis

    SciTech Connect

    Murata, K.K.; Williams, D.C.; Griffith, R.O.; Gido, R.G.; Tadios, E.L.; Davis, F.J.; Martinez, G.M.; Washington, K.E.; Tills, J.

    1997-12-01

    The CONTAIN 2.0 computer code is an integrated analysis tool used for predicting the physical conditions, chemical compositions, and distributions of radiological materials inside a containment building following the release of material from the primary system in a light-water reactor accident. It can also predict the source term to the environment. CONTAIN 2.0 is intended to replace the earlier CONTAIN 1.12, which was released in 1991. The purpose of this Code Manual is to provide full documentation of the features and models in CONTAIN 2.0. Besides complete descriptions of the models, this Code Manual provides a complete description of the input and output from the code. CONTAIN 2.0 is a highly flexible and modular code that can run problems that are either quite simple or highly complex. An important aspect of CONTAIN is that the interactions among thermal-hydraulic phenomena, aerosol behavior, and fission product behavior are taken into account. The code includes atmospheric models for steam/air thermodynamics, intercell flows, condensation/evaporation on structures and aerosols, aerosol behavior, and gas combustion. It also includes models for reactor cavity phenomena such as core-concrete interactions and coolant pool boiling. Heat conduction in structures, fission product decay and transport, radioactive decay heating, and the thermal-hydraulic and fission product decontamination effects of engineered safety features are also modeled. To the extent possible, the best available models for severe accident phenomena have been incorporated into CONTAIN, but it is intrinsic to the nature of accident analysis that significant uncertainty exists regarding numerous phenomena. In those cases, sensitivity studies can be performed with CONTAIN by means of user-specified input parameters. Thus, the code can be viewed as a tool designed to assist the knowledge reactor safety analyst in evaluating the consequences of specific modeling assumptions.

  12. NUCLEAR REACTOR

    DOEpatents

    Christy, R.F.

    1958-07-15

    A nuclear reactor of the homogeneous liquid fuel type is described wherein the fissionable isotope is suspended or dissolved in a liquid moderator such as water. The reactor core is comprised essentially of a spherical vessel for containing the reactive composition surrounded by a reflector, preferably of beryllium oxide. The reactive composition may be an ordinary water solution of a soluble salt of uranium, the quantity of fissionable isotope in solution being sufficient to provide a critical mass in the vessel. The liquid fuel is stored in a tank of non-crtttcal geometry below the reactor vessel and outside of the reflector and is passed from the tank to the vessel through a pipe connecting the two by air pressure means. Neutron absorbing control and safety rods are operated within slots in the reflector adjacent to the vessel.

  13. Deposition of RuO 4 on various surfaces in a nuclear reactor containment

    NASA Astrophysics Data System (ADS)

    Holm, Joachim; Glänneskog, Henrik; Ekberg, Christian

    2009-07-01

    During a severe nuclear reactor accident with air ingress, ruthenium can be released from the nuclear fuel in the form of ruthenium tetroxide. Hence, it is important to investigate how the reactor containment is able to reduce the source term of ruthenium. The aim of this work was to investigate the deposition of gaseous ruthenium tetroxide on aluminium, copper and zinc, which all appear in relatively large amounts in reactor containment. The experiments show that ruthenium tetroxide is deposited on all the metal surfaces, especially on the copper and zinc surfaces. A large deposition of ruthenium tetroxide also appeared on the relatively inert glass surfaces in the experimental set-ups. The analyses of the different surfaces, with several analytical methods, showed that the form of deposited ruthenium was mainly ruthenium dioxide.

  14. Nuclear reactor melt-retention structure to mitigate direct containment heating

    DOEpatents

    Tutu, Narinder K.; Ginsberg, Theodore; Klages, John R.

    1991-01-01

    A light water nuclear reactor melt-retention structure to mitigate the extent of direct containment heating of the reactor containment building. The structure includes a retention chamber for retaining molten core material away from the upper regions of the reactor containment building when a severe accident causes the bottom of the pressure vessel of the reactor to fail and discharge such molten material under high pressure through the reactor cavity into the retention chamber. In combination with the melt-retention chamber there is provided a passageway that includes molten core droplet deflector vanes and has gas vent means in its upper surface, which means are operable to deflect molten core droplets into the retention chamber while allowing high pressure steam and gases to be vented into the upper regions of the containment building. A plurality of platforms are mounted within the passageway and the melt-retention structure to direct the flow of molten core material and help retain it within the melt-retention chamber. In addition, ribs are mounted at spaced positions on the floor of the melt-retention chamber, and grid means are positioned at the entrance side of the retention chamber. The grid means develop gas back pressure that helps separate the molten core droplets from discharged high pressure steam and gases, thereby forcing the steam and gases to vent into the upper regions of the reactor containment building.

  15. Foreign research reactor irradiated nuclear fuel inventories containing HEU and LEU of United States origin

    SciTech Connect

    Matos, J.E.

    1994-12-01

    This report provides estimates of foreign research reactor inventories of aluminum-based and TRIGA irradiated nuclear fuel elements containing highly enriched and low enriched uranium of United States origin that are anticipated in January 1996, January 2001, and January 2006. These fuels from 104 research reactors in 41 countries are the same aluminum-based and TRIGA fuels that were eligible for receipt under the Department of Energy`s Offsite Fuels Policy that was in effect in 1988. All fuel inventory and reactor data that were available as of December 1, 1994, have been included in the estimates of approximately 14,300 irradiated fuel elements in January 1996, 18,800 in January 2001, and 22,700 in January 2006.

  16. NUCLEAR REACTOR

    DOEpatents

    Sherman, J.; Sharbaugh, J.E.; Fauth, W.L. Jr.; Palladino, N.J.; DeHuff, P.G.

    1962-10-23

    A nuclear reactor incorporating seed and blanket assemblies is designed. Means are provided for obtaining samples of the coolant from the blanket assemblies and for varying the flow of coolant through the blanket assemblies. (AEC)

  17. Investigation of cable deterioration in the containment building of the Savannah River Nuclear Reactor

    SciTech Connect

    Gillen, K.T.; Clough, R.L.; Jones, L.H.

    1982-08-01

    This report describes an investigation of the deterioration of polyethylene and polyvinylchloride cable materials which occurred in the containment building of the Savannah River nuclear reactor located at Aiken, South Carolina. Radiation dosimetry and temperature mapping data of the containment area indicated that the maximum dose experienced by the cable materials was only 2.5 Mrad at an average operating temperature of 43/sup 0/C. Considering this relatively moderate environment, the amount of material degradation seemed surprising. To understand these findings, an experimental program was performed on the commercial polyethylene and polyvinylchloride materials used at the plant to investigate their degradation behavior under combined ..gamma..-radiation and elevated temperature conditions. It is established that the material deterioration at the plant resulted from radiation-induced oxidation and that the degradation rate can be correlated with local levels of radiation intensity in the containment area.

  18. NUCLEAR REACTOR

    DOEpatents

    Anderson, C.R.

    1962-07-24

    A fluidized bed nuclear reactor and a method of operating such a reactor are described. In the design means are provided for flowing a liquid moderator upwardly through the center of a bed of pellets of a nentron-fissionable material at such a rate as to obtain particulate fluidization while constraining the lower pontion of the bed into a conical shape. A smooth circulation of particles rising in the center and falling at the outside of the bed is thereby established. (AEC)

  19. NUCLEAR REACTOR

    DOEpatents

    Grebe, J.J.

    1959-07-14

    High temperature reactors which are uniquely adapted to serve as the heat source for nuclear pcwered rockets are described. The reactor is comprised essentially of an outer tubular heat resistant casing which provides the main coolant passageway to and away from the reactor core within the casing and in which the working fluid is preferably hydrogen or helium gas which is permitted to vaporize from a liquid storage tank. The reactor core has a generally spherical shape formed entirely of an active material comprised of fissile material and a moderator material which serves as a diluent. The active material is fabricated as a gas permeable porous material and is interlaced in a random manner with very small inter-connecting bores or capillary tubes through which the coolant gas may flow. The entire reactor is divided into successive sections along the direction of the temperature gradient or coolant flow, each section utilizing materials of construction which are most advantageous from a nuclear standpoint and which at the same time can withstand the operating temperature of that particular zone. This design results in a nuclear reactor characterized simultaneously by a minimum critiral size and mass and by the ability to heat a working fluid to an extremely high temperature.

  20. NUCLEAR REACTOR

    DOEpatents

    Breden, C.R.; Dietrich, J.R.

    1961-06-20

    A water-soluble non-volatile poison may be introduced into a reactor to nullify excess reactivity. The poison is removed by passing a side stream of the water containing the soluble poison to an evaporation chamber. The vapor phase is returned to the reactor to decrease the concentration of soluble poison and the liquid phase is returned to increase the concentration of soluble poison.

  1. NUCLEAR REACTOR

    DOEpatents

    Miller, H.I.; Smith, R.C.

    1958-01-21

    This patent relates to nuclear reactors of the type which use a liquid fuel, such as a solution of uranyl sulfate in ordinary water which acts as the moderator. The reactor is comprised of a spherical vessel having a diameter of about 12 inches substantially surrounded by a reflector of beryllium oxide. Conventionnl control rods and safety rods are operated in slots in the reflector outside the vessel to control the operation of the reactor. An additional means for increasing the safety factor of the reactor by raising the ratio of delayed neutrons to prompt neutrons, is provided and consists of a soluble sulfate salt of beryllium dissolved in the liquid fuel in the proper proportion to obtain the result desired.

  2. NUCLEAR REACTORS

    DOEpatents

    Long, E.; Ashby, J.W.

    1958-09-16

    ABS>A graphite moderator structure is presented for a nuclear reactor compriscd of an assembly of similarly orientated prismatic graphite blocks arranged on spaced longitudinal axes lying in common planes wherein the planes of the walls of the blocks are positioned so as to be twisted reintive to the planes of said axes so thatthe unlmpeded dtrect paths in direction wholly across the walls of the blocks are limited to the width of the blocks plus spacing between the blocks.

  3. XPS investigations of ruthenium deposited onto representative inner surfaces of nuclear reactor containment buildings

    NASA Astrophysics Data System (ADS)

    Mun, C.; Ehrhardt, J. J.; Lambert, J.; Madic, C.

    2007-07-01

    In the case of a hypothetical severe accident in a nuclear power plant, interactions of gaseous RuO 4 with reactor containment building surfaces (stainless steel and epoxy paint) could possibly lead to a black Ru-containing deposit on these surfaces. Some scenarios include the possibility of formation of highly radiotoxic RuO 4(g) by the interactions of these deposits with the oxidizing medium induced by air radiolysis, in the reactor containment building, and consequently dispersion of this species. Therefore, the accurate determination of the chemical nature of ruthenium in the deposits is of the high importance for safety studies. An experiment was designed to model the interactions of RuO 4(g) with samples of stainless steel and of steel covered with epoxy paint. Then, these deposits have been carefully characterised by scanning electron microscopy (SEM/EDS), electron probe microanalysis (EPMA) and X-ray photoelectron spectroscopy (XPS). The analysis by XPS of Ru deposits formed by interaction of RuO 4(g), revealed that the ruthenium is likely to be in the IV oxidation state, as the shapes of the Ru 3d core levels are very similar with those observed on the RuO 2· xH 2O reference powder sample. The analysis of O 1s peaks indicates a large component attributed to the hydroxyl functional groups. From these results, it was concluded that Ru was present on the surface of the deposits as an oxyhydroxide of Ru(IV). It has also to be pointed out that the presence of "pure" RuO 2, or of a thin layer of RuO 3 or Ru 2O 5, coming from the decomposition of RuO 4 on the surface of samples of stainless steel and epoxy paint, could be ruled out. These findings will be used for further investigations of the possible revolatilisation phenomena induced by ozone.

  4. Nuclear reactor

    DOEpatents

    Wade, Elman E.

    1979-01-01

    A nuclear reactor including two rotatable plugs and a positive top core holddown structure. The top core holddown structure is divided into two parts: a small core cover, and a large core cover. The small core cover, and the upper internals associated therewith, are attached to the small rotating plug, and the large core cover, with its associated upper internals, is attached to the large rotating plug. By so splitting the core holddown structures, under-the-plug refueling is accomplished without the necessity of enlarging the reactor pressure vessel to provide a storage space for the core holddown structure during refueling. Additionally, the small and large rotating plugs, and their associated core covers, are arranged such that the separation of the two core covers to permit rotation is accomplished without the installation of complex lifting mechanisms.

  5. NUCLEAR REACTOR

    DOEpatents

    Grebe, J.J.

    1959-12-15

    A reactor which is particularly adapted tu serve as a heat source for a nuclear powered alrcraft or rocket is described. The core of this reactor consists of a porous refractory modera;or body which is impregnated with fissionable nuclei. The core is designed so that its surface forms tapered inlet and outlet ducts which are separated by the porous moderator body. In operation a gaseous working fluid is circulated through the inlet ducts to the surface of the moderator, enters and passes through the porous body, and is heated therein. The hot gas emerges into the outlet ducts and is available to provide thrust. The principle advantage is that tremendous quantities of gas can be quickly heated without suffering an excessive pressure drop.

  6. NUCLEAR REACTORS

    DOEpatents

    Long, E.; Ashley, J.W.

    1958-12-16

    A graphite moderator structure is described for a gas-cooled nuclear reactor having a vertical orlentation wherein the structure is physically stable with regard to dlmensional changes due to Wigner growth properties of the graphite, and leakage of coolant gas along spaces in the structure is reduced. The structure is comprised of stacks of unlform right prismatic graphite blocks positioned in layers extending in the direction of the lengths of the blocks, the adjacent end faces of the blocks being separated by pairs of tiles. The blocks and tiles have central bores which are in alignment when assembled and are provided with cooperatlng keys and keyways for physical stability.

  7. Ultimate pressure capacity analysis of a post-tensioned reinforced concrete nuclear reactor containment building

    SciTech Connect

    Dooley, W.T.; Macek, R.W.; Sadik, S.

    1983-01-01

    This paper presents a three dimensional nonlinear structural analysis of a post-tensioned reinforced concrete nuclear reactor containment building. The objective of the analysis was to develop and demonstrate modeling techniques appropriate for determining the global ultimate internal pressure capacity of this type of containment. The structural model developed for the ADINA computer code employed nonlinear material models with truss, shell and three dimensional continuum elements to represent the major structural members (tendons, rebars, ties, concrete and liner). In addition a special uniaxial elastic plastic shell material model was developed to facilitate representation of the dome tendon lacing pattern. Effective use of symmetry permitted all three dimensional effects to be represented in a 30/sup 0/ (0.52 rad) segment of the structure. The analysis suggested that gross failure of the structure at a quasistatic internal pressure of 99 psi (0.68 MPa) may have been precipitated by the nonaxisymmetric three dimensional nature of the stress field in the concrete near the tendon anchorages.

  8. Accelerated development of Zr-containing new generation ferritic steels for advanced nuclear reactors

    SciTech Connect

    Tan, Lizhen; Yang, Ying; Sridharan, K.

    2015-12-01

    The mission of the Nuclear Energy Enabling Technologies (NEET) program is to develop crosscutting technologies for nuclear energy applications. Advanced structural materials with superior performance at elevated temperatures are always desired for nuclear reactors, which can improve reactor economics, safety margins, and design flexibility. They benefit not only new reactors, including advanced light water reactors (LWRs) and fast reactors such as the sodium-cooled fast reactor (SFR) that is primarily designed for management of high-level wastes, but also life extension of the existing fleet when component exchange is needed. Developing and utilizing the modern materials science tools (experimental, theoretical, and computational tools) is an important path to more efficient alloy development and process optimization. The ultimate goal of this project is, with the aid of computational modeling tools, to accelerate the development of Zr-bearing ferritic alloys that can be fabricated using conventional steelmaking methods. The new alloys are expected to have superior high-temperature creep performance and excellent radiation resistance as compared to Grade 91. The designed alloys were fabricated using arc-melting and drop-casting, followed by hot rolling and conventional heat treatments. Comprehensive experimental studies have been conducted on the developed alloys to evaluate their hardness, tensile properties, creep resistance, Charpy impact toughness, and aging resistance, as well as resistance to proton and heavy ion (Fe2+) irradiation.

  9. Nuclear reactor

    DOEpatents

    Pennell, William E.; Rowan, William J.

    1977-01-01

    A nuclear reactor in which the core components, including fuel-rod assemblies, control-rod assemblies, fertile rod-assemblies, and removable shielding assemblies, are supported by a plurality of separate inlet modular units. These units are referred to as inlet module units to distinguish them from the modules of the upper internals of the reactor. The modular units are supported, each removable independently of the others, in liners in the supporting structure for the lower internals of the reactor. The core assemblies are removably supported in integral receptacles or sockets of the modular units. The liners, units, sockets and assmblies have inlet openings for entry of the fluid. The modular units are each removably mounted in the liners with fluid seals interposed between the opening in the liner and inlet module into which the fluid enters and the upper and lower portion of the liner. Each assembly is similarly mounted in a corresponding receptacle with fluid seals interposed between the openings where the fluid enters and the lower portion of the receptacle or fitting closely in these regions. As fluid flows along each core assembly a pressure drop is produced along the fluid so that the fluid which emerges from each core assembly is at a lower pressure than the fluid which enters the core assembly. However because of the seals interposed in the mountings of the units and assemblies the pressures above and below the units and assemblies are balanced and the units are held in the liners and the assemblies are held in the receptacles by their weights as they have a higher specific gravity than the fluid. The low-pressure spaces between each module and its liner and between each core assembly and its module is vented to the low-pressure regions of the vessel to assure that fluid which leaks through the seals does not accumulate and destroy the hydraulic balance.

  10. Nuclear reactor

    DOEpatents

    Yant, Howard W.; Stinebiser, Karl W.; Anzur, Gregory C.

    1977-01-01

    A nuclear reactor, particularly a liquid-metal breeder reactor, whose upper internals include outlet modules for channeling the liquid-metal coolant from selected areas of the outlet of the core vertically to the outlet plenum. The modules are composed of a highly-refractory, high corrosion-resistant alloy, for example, INCONEL-718. Each module is disposed to confine and channel generally vertically the coolant emitted from a subplurality of core-component assemblies. Each module has a grid with openings, each opening disposed to receive the coolant from an assembly of the subplurality. The grid in addition serves as a holdown for the assemblies of the corresponding subplurality preventing their excessive ejection upwardly from the core. In the region directly over the core the outlet modules are of such peripheral form that they nest forming a continuum over the core-component assemblies whose outlet coolant they confine. Each subassembly includes a chimney which confines the coolant emitted by its corresponding subassemblies to generally vertical flow between the outlet of the core and the outlet plenum. Each subplurality of assemblies whose emitted coolant is confined by an outlet module includes assemblies which emit lower-temperature coolant, for example, a control-rod assembly, or fertile assemblies, and assemblies which emit coolant of substantially higher temperature, for example, fuel-rod assemblies. The coolants of different temperatures are mixed in the chimneys reducing the effect of stripping (hot-cold temperature fluctuations) on the remainder of the upper internals which are composed typically of AISI-304 or AISI-316 stainless steel.

  11. Nuclear reactor

    DOEpatents

    Thomson, Wallace B.

    2004-03-16

    A nuclear reactor comprising a cylindrical pressure vessel, an elongated annular core centrally disposed within and spaced from the pressure vessel, and a plurality of ducts disposed longitudinally of the pressure vessel about the periphery thereof, said core comprising an annular active portion, an annular reflector just inside the active portion, and an annular reflector just outside the active a portion, said annular active portion comprising rectangular slab, porous fuel elements radially disposed around the inner reflector and extending the length of the active portion, wedge-shaped, porous moderator elements disposed adjacent one face of each fuel element and extending the length of the fuel element, the fuel and moderator elements being oriented so that the fuel elements face each other and the moderator elements do likewise, adjacent moderator elements being spaced to provide air inlet channels, and adjacent fuel elements being spaced to provide air outlet channels which communicate with the interior of the peripheral ducts, and means for introducing air into the air inlet channels which passes through the porous moderator elements and porous fuel elements to the outlet channel.

  12. Nuclear reactor reflector

    DOEpatents

    Hopkins, Ronald J.; Land, John T.; Misvel, Michael C.

    1994-01-01

    A nuclear reactor reflector is disclosed that comprises a stack of reflector blocks with vertical water flow passages to cool the reflector. The interface between blocks is opposite support points for reactor fuel rods. Water flows between the reflector and the reactor barrel from passages in a bottom block. The top block contains a flange to limit this flow and the flange has a slot to receive an alignment pin that is welded to the barrel. The pin is held in the slot by two removable shims. Alignment bars extend the length of the stack in slots machined in each block when the stack is assembled.

  13. Nuclear reactor reflector

    DOEpatents

    Hopkins, R.J.; Land, J.T.; Misvel, M.C.

    1994-06-07

    A nuclear reactor reflector is disclosed that comprises a stack of reflector blocks with vertical water flow passages to cool the reflector. The interface between blocks is opposite support points for reactor fuel rods. Water flows between the reflector and the reactor barrel from passages in a bottom block. The top block contains a flange to limit this flow and the flange has a slot to receive an alignment pin that is welded to the barrel. The pin is held in the slot by two removable shims. Alignment bars extend the length of the stack in slots machined in each block when the stack is assembled. 12 figs.

  14. NUCLEAR REACTOR

    DOEpatents

    Moore, R.V.; Bowen, J.H.; Dent, K.H.

    1958-12-01

    A heterogeneous, natural uranium fueled, solid moderated, gas cooled reactor is described, in which the fuel elements are in the form of elongated rods and are dlsposed within vertical coolant channels ln the moderator symmetrically arranged as a regular lattice in groups. This reactor employs control rods which operate in vertical channels in the moderator so that each control rod is centered in one of the fuel element groups. The reactor is enclosed in a pressure vessel which ls provided with access holes at the top to facilitate loading and unloadlng of the fuel elements, control rods and control rod driving devices.

  15. Nuclear reactor control column

    SciTech Connect

    Bachovchin, D.M.

    1982-08-10

    The nuclear reactor control column comprises a column disposed within the nuclear reactor core having a variable cross-section hollow channel and containing balls whose vertical location is determined by the flow of the reactor coolant through the column. The control column is divided into three basic sections wherein each of the sections has a different cross-sectional area. The uppermost section of the control column has the greatest crosssectional area, the intermediate section of the control column has the smallest cross-sectional area, and the lowermost section of the control column has the intermediate cross-sectional area. In this manner, the area of the uppermost section can be established such that when the reactor coolant is flowing under normal conditions therethrough, the absorber balls will be lifted and suspended in a fluidized bed manner in the upper section. However, when the reactor coolant flow falls below a predetermined value, the absorber balls will fall through the intermediate section and into the lowermost section, thereby reducing the reactivity of the reactor core and shutting down the reactor.

  16. Nuclear reactor control column

    DOEpatents

    Bachovchin, Dennis M.

    1982-01-01

    The nuclear reactor control column comprises a column disposed within the nuclear reactor core having a variable cross-section hollow channel and containing balls whose vertical location is determined by the flow of the reactor coolant through the column. The control column is divided into three basic sections wherein each of the sections has a different cross-sectional area. The uppermost section of the control column has the greatest cross-sectional area, the intermediate section of the control column has the smallest cross-sectional area, and the lowermost section of the control column has the intermediate cross-sectional area. In this manner, the area of the uppermost section can be established such that when the reactor coolant is flowing under normal conditions therethrough, the absorber balls will be lifted and suspended in a fluidized bed manner in the upper section. However, when the reactor coolant flow falls below a predetermined value, the absorber balls will fall through the intermediate section and into the lowermost section, thereby reducing the reactivity of the reactor core and shutting down the reactor.

  17. BEACON/MOD: a computer program for thermal-hydraulic analysis of nuclear reactor containments - user's manual

    SciTech Connect

    Broadus, C.R.; Doyle, R.J.; James, S.W.; Lime, J.F.; Mings, W.J.

    1980-04-01

    The BEACON code is a best-estimate, advanced containment code designed to perform a best-estimate analysis of the flow of a mixture of air, water, and steam in a nuclear reactor containment system under loss-of-coolant accident conditions. The code can simulate two-component, two-phase fluid flow in complex geometries using a combination of two-dimensional, one-dimensional, and lumped-parameter representations for the various parts of the system. The current version of BEACON, which is designated BEACON/MOD3, contains mass and heat transfer models for wall film and wall conduction. It is suitable for the evaluation of short-term transients in dry-containment systems. This manual describes the models employed in BEACON/MOD3 and specifies code implementation requirements. It provides application information for input data preparation and for output data interpretation.

  18. NUCLEAR REACTOR

    DOEpatents

    Treshow, M.

    1958-08-19

    A neuclear reactor is described of the heterogeneous type and employing replaceable tubular fuel elements and heavy water as a coolant and moderator. A pluraltty of fuel tubesa having their axes parallel, extend through a tank type pressure vessel which contatns the liquid moderator. The fuel elements are disposed within the fuel tubes in the reaetive portion of the pressure vessel during normal operation and the fuel tubes have removable plug members at each end to permit charging and discharging of the fuel elements. The fuel elements are cylindrical strands of jacketed fissionable material having helical exterior ribs. A bundle of fuel elements are held within each fuel tube with their longitudinal axes parallel, the ribs serving to space them apart along their lengths. Coolant liquid is circulated through the fuel tubes between the spaced fuel elements. Suitable control rod and monitoring means are provided for controlling the reactor.

  19. NUCLEAR REACTORS

    DOEpatents

    Koch, L.J.; Rice, R.E. Jr.; Denst, A.A.; Rogers, A.J.; Novick, M.

    1961-12-01

    An active portion assembly for a fast neutron reactor is described wherein physical distortions resulting in adverse changes in the volume-to-mass ratio are minimized. A radially expandable locking device is disposed within a cylindrical tube within each fuel subassembly within the active portion assembly, and clamping devices expandable toward the center of the active portion assembly are disposed around the periphery thereof. (AEC)

  20. NUCLEAR REACTOR

    DOEpatents

    Grebe, J.J.

    1961-01-24

    A core structure for neutronic reactors adapted for the propulsion of aircraft and rockets is offered. The core is designed for cooling by gaseous media, and comprises a plurality of hollow tapered tubular segments of a porous moderating material impregniated with fissionable fuel nested about a common axis. Alternate ends of the segments are joined. In operation a coolant gas passes through the porous structure and is heated.

  1. Nuclear reactor fuel structure containing uranium alloy wires embedded in a metallic matrix plate

    DOEpatents

    Travelli, A.

    1985-10-25

    A flat or curved plate structure, to be used as fuel in a nuclear reactor, comprises elongated fissionable wires or strips embedded in a metallic continuous non-fissionable matrix plate. The wires or strips are made predominantly of a malleable uranium alloy, such as uranium silicide, uranium gallide or uranium germanide. The matrix plate is made predominantly of aluminum or an aluminum alloy. The wires or strips are located in a single row at the midsurface of the plate, parallel with one another and with the length dimension of the plate. The wires or strips are separated from each other, and from the surface of the plate, by sufficient thicknesses of matrix material, to provide structural integrity and effective fission product retention, under neutron irradiation. This construction makes it safely feasible to provide a high uranium density, so that the uranium enrichment with uranium 235 may be reduced below about 20%, to deter the reprocessing of the uranium for use in nuclear weapons.

  2. Nuclear reactor fuel structure containing uranium alloy wires embedded in a metallic matrix plate

    DOEpatents

    Travelli, Armando

    1988-01-01

    A flat or curved plate structure, to be used as fuel in a nuclear reactor, comprises elongated fissionable wires or strips embedded in a metallic continuous non-fissionable matrix plate. The wires or strips are made predominantly of a malleable uranium alloy, such as uranium silicide, uranium gallide or uranium germanide. The matrix plate is made predominantly of aluminum or an aluminum alloy. The wires or strips are located in a single row at the midsurface of the plate, parallel with one another and with the length dimension of the plate. The wires or strips are separated from each other, and from the surface of the plate, by sufficient thicknesses of matrix material, to provide structural integrity and effective fission product retention, under neutron irradiation. This construction makes it safely feasible to provide a high uranium density, so that the uranium enrichment with uranium 235 may be reduced below about 20%, to deter the reprocessing of the uranium for use in nuclear weapons.

  3. Reactor refueling containment system

    DOEpatents

    Gillett, J.E.; Meuschke, R.E.

    1995-05-02

    A method of refueling a nuclear reactor is disclosed whereby the drive mechanism is disengaged and removed by activating a jacking mechanism that raises the closure head. The area between the barrier plate and closure head is exhausted through the closure head penetrations. The closure head, upper drive mechanism, and bellows seal are lifted away and transported to a safe area. The barrier plate acts as the primary boundary and each drive and control rod penetration has an elastomer seal preventing excessive tritium gases from escaping. The individual instrumentation plugs are disengaged allowing the corresponding fuel assembly to be sealed and replaced. 2 figs.

  4. Reactor refueling containment system

    DOEpatents

    Gillett, James E.; Meuschke, Robert E.

    1995-01-01

    A method of refueling a nuclear reactor whereby the drive mechanism is disengaged and removed by activating a jacking mechanism that raises the closure head. The area between the barrier plate and closure head is exhausted through the closure head penetrations. The closure head, upper drive mechanism, and bellows seal are lifted away and transported to a safe area. The barrier plate acts as the primary boundary and each drive and control rod penetration has an elastomer seal preventing excessive tritium gases from escaping. The individual instrumentation plugs are disengaged allowing the corresponding fuel assembly to be sealed and replaced.

  5. Nuclear Reactors and Technology

    SciTech Connect

    Cason, D.L.; Hicks, S.C.

    1992-01-01

    This publication Nuclear Reactors and Technology (NRT) announces on a monthly basis the current worldwide information available from the open literature on nuclear reactors and technology, including all aspects of power reactors, components and accessories, fuel elements, control systems, and materials. This publication contains the abstracts of DOE reports, journal articles, conference papers, patents, theses, and monographs added to the Energy Science and Technology Database during the past month. Also included are US information obtained through acquisition programs or interagency agreements and international information obtained through the International Energy Agency`s Energy Technology Data Exchange or government-to-government agreements. The digests in NRT and other citations to information on nuclear reactors back to 1948 are available for online searching and retrieval on the Energy Science and Technology Database and Nuclear Science Abstracts (NSA) database. Current information, added daily to the Energy Science and Technology Database, is available to DOE and its contractors through the DOE Integrated Technical Information System. Customized profiles can be developed to provide current information to meet each user`s needs.

  6. NUCLEAR REACTOR CONTROL SYSTEM

    DOEpatents

    Epler, E.P.; Hanauer, S.H.; Oakes, L.C.

    1959-11-01

    A control system is described for a nuclear reactor using enriched uranium fuel of the type of the swimming pool and other heterogeneous nuclear reactors. Circuits are included for automatically removing and inserting the control rods during the course of normal operation. Appropriate safety circuits close down the nuclear reactor in the event of emergency.

  7. Heat dissipating nuclear reactor

    DOEpatents

    Hunsbedt, A.; Lazarus, J.D.

    1985-11-21

    Disclosed is a nuclear reactor containment adapted to retain and cool core debris in the unlikely event of a core meltdown and subsequent breach in the reactor vessel. The reactor vessel is seated in a cavity which has a thick metal sidewall that is integral with a thick metal basemat at the bottom of the cavity. The basemat extends beyond the perimeter of the cavity sidewall. Underneath the basemat is a porous bed with water pipes and steam pipes running into it. Water is introduced into the bed and converted into steam which is vented to the atmosphere. A plurality of metal pilings in the form of H-beams extend from the metal base plate downwardly and outwardly into the earth.

  8. Heat dissipating nuclear reactor

    DOEpatents

    Hunsbedt, Anstein; Lazarus, Jonathan D.

    1987-01-01

    Disclosed is a nuclear reactor containment adapted to retain and cool core debris in the unlikely event of a core meltdown and subsequent breach in the reactor vessel. The reactor vessel is seated in a cavity which has a thick metal sidewall that is integral with a thick metal basemat at the bottom of the cavity. The basemat extends beyond the perimeter of the cavity sidewall. Underneath the basemat is a porous bed with water pipes and steam pipes running into it. Water is introduced into the bed and converted into steam which is vented to the atmosphere. A plurality of metal pilings in the form of H-beams extends from the metal base plate downwardly and outwardly into the earth.

  9. The French nuclear power plant reactor building containment contributions of prestressing and concrete performances in reliability improvements and cost savings

    SciTech Connect

    Rouelle, P.; Roy, F.

    1998-12-31

    The Electricite de France`s N4 CHOOZ B nuclear power plant, two units of the world`s largest PWR model (1450 Mwe each), has earned the Electric Power International`s 1997 Powerplant Award. This lead NPP for EDF`s N4 series has been improved notably in terms of civil works. The presentation will focus on the Reactor Building`s inner containment wall which is one of the main civil structures on a technical and safety point of view. In order to take into account the necessary evolution of the concrete technical specification such as compressive strength low creep and shrinkage, the HSC/HPC has been used on the last N4 Civaux 2 NPP. As a result of the use of this type of professional concrete, the containment withstands an higher internal pressure related to severe accident and ensures higher level of leak-tightness, thus improving the overall safety of the NPP. On that occasion, a new type of prestressing has been tested locally through 55 C 15 S tendons using a new C 1500 FE Jack. These updated civil works techniques shall allow EDF to ensure a Reactor Containment lifespan for more than 50 years. The gains in terms of reliability and cost saving of these improved techniques will be developed hereafter.

  10. 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)

  11. Nuclear reactor building

    DOEpatents

    Gou, P.F.; Townsend, H.E.; Barbanti, G.

    1994-04-05

    A reactor building for enclosing a nuclear reactor includes a containment vessel having a wetwell disposed therein. The wetwell includes inner and outer walls, a floor, and a roof defining a wetwell pool and a suppression chamber disposed there above. The wetwell and containment vessel define a drywell surrounding the reactor. A plurality of vents are disposed in the wetwell pool in flow communication with the drywell for channeling into the wetwell pool steam released in the drywell from the reactor during a LOCA for example, for condensing the steam. A shell is disposed inside the wetwell and extends into the wetwell pool to define a dry gap devoid of wetwell water and disposed in flow communication with the suppression chamber. In a preferred embodiment, the wetwell roof is in the form of a slab disposed on spaced apart support beams which define there between an auxiliary chamber. The dry gap, and additionally the auxiliary chamber, provide increased volume to the suppression chamber for improving pressure margin. 4 figures.

  12. Nuclear reactor building

    DOEpatents

    Gou, Perng-Fei; Townsend, Harold E.; Barbanti, Giancarlo

    1994-01-01

    A reactor building for enclosing a nuclear reactor includes a containment vessel having a wetwell disposed therein. The wetwell includes inner and outer walls, a floor, and a roof defining a wetwell pool and a suppression chamber disposed thereabove. The wetwell and containment vessel define a drywell surrounding the reactor. A plurality of vents are disposed in the wetwell pool in flow communication with the drywell for channeling into the wetwell pool steam released in the drywell from the reactor during a LOCA for example, for condensing the steam. A shell is disposed inside the wetwell and extends into the wetwell pool to define a dry gap devoid of wetwell water and disposed in flow communication with the suppression chamber. In a preferred embodiment, the wetwell roof is in the form of a slab disposed on spaced apart support beams which define therebetween an auxiliary chamber. The dry gap, and additionally the auxiliary chamber, provide increased volume to the suppression chamber for improving pressure margin.

  13. HOMOGENEOUS NUCLEAR REACTOR

    DOEpatents

    Hammond, R.P.; Busey, H.M.

    1959-02-17

    Nuclear reactors of the homogeneous liquid fuel type are discussed. The reactor is comprised of an elongated closed vessel, vertically oriented, having a critical region at the bottom, a lower chimney structure extending from the critical region vertically upwardly and surrounded by heat exchanger coils, to a baffle region above which is located an upper chimney structure containing a catalyst functioning to recombine radiolyticallydissociated moderator gages. In operation the liquid fuel circulates solely by convection from the critical region upwardly through the lower chimney and then downwardly through the heat exchanger to return to the critical region. The gases formed by radiolytic- dissociation of the moderator are carried upwardly with the circulating liquid fuel and past the baffle into the region of the upper chimney where they are recombined by the catalyst and condensed, thence returning through the heat exchanger to the critical region.

  14. GAS COOLED NUCLEAR REACTORS

    DOEpatents

    Long, E.; Rodwell, W.

    1958-06-10

    A gas-cooled nuclear reactor consisting of a graphite reacting core and reflector structure supported in a containing vessel is described. A gas sealing means is included for sealing between the walls of the graphite structure and containing vessel to prevent the gas coolant by-passing the reacting core. The reacting core is a multi-sided right prismatic structure having a pair of parallel slots around its periphery. The containing vessel is cylindrical and has a rib on its internal surface which supports two continuous ring shaped flexible web members with their radially innermost ends in sealing engagement within the radially outermost portion of the slots. The core structure is supported on ball bearings. This design permits thermal expansion of the core stracture and vessel while maintainirg a peripheral seal between the tvo elements.

  15. Nuclear reactor sealing system

    DOEpatents

    McEdwards, James A.

    1983-01-01

    A liquid metal-cooled nuclear reactor sealing system. The nuclear reactor includes a vessel sealed at its upper end by a closure head. The closure head comprises at least two components, one of which is rotatable; and the two components define an annulus therebetween. The sealing system includes at least a first and second inflatable seal disposed in series in an upper portion of the annulus. The system further includes a dip seal extending into a body of insulation located adjacent a bottom portion of the closure head. The dip seal comprises a trough formed by a lower portion of one of the components, and a seal blade pendently supported from the other component and extending downwardly into the trough. A body of liquid metal is contained in the trough which submerges a portion of the seal blade. The seal blade is provided with at least one aperture located above the body of liquid metal for providing fluid communication between the annulus intermediate the dip seal and the inflatable seals, and a body of cover gas located inside the vessel. There also is provided means for introducing a purge gas into the annulus intermediate the inflatable seals and the seal blade. The purge gas is introduced in an amount sufficient to substantially reduce diffusion of radioactive cover gas or sodium vapor up to the inflatable seals. The purge gas mixes with the cover gas in the reactor vessel where it can be withdrawn from the vessel for treatment and recycle to the vessel.

  16. Nuclear Reactors and Technology; (USA)

    SciTech Connect

    Cason, D.L.; Hicks, S.C.

    1991-01-01

    Nuclear Reactors and Technology (NRT) announces on a monthly basis the current worldwide information available from the open literature on nuclear reactors and technology, including all aspects of power reactors, components and accessories, fuel elements, control systems, and materials. This publication contains the abstracts of DOE reports, journal articles, conference papers, patents, theses, and monographs added to the Energy Science and Technology Database (EDB) during the past month. Also included are US information obtained through acquisition programs or interagency agreements and international information obtained through the International Energy Agency's Energy Technology Data Exchange or government-to-government agreements. The digests in NRT and other citations to information on nuclear reactors back to 1948 are available for online searching and retrieval on EDB and Nuclear Science Abstracts (NSA) database. Current information, added daily to EDB, is available to DOE and its contractors through the DOE integrated Technical Information System. Customized profiles can be developed to provide current information to meet each user's needs.

  17. A 1055 ft/sec impact test of a two foot diameter model nuclear reactor containment system without fracture

    NASA Technical Reports Server (NTRS)

    Puthoff, R. L.

    1972-01-01

    A study to determine the feasibility of containing the fission products of a mobile reactor in the event of an impact is presented. The model simulated the reactor core, energy absorbing gamma shielding, neutron shielding and the containment vessel. It was impacted against an 18,000 pound reinforced concrete block at 1055 ft/sec. The model was significantly deformed and the concrete block demolished. No leaks were detected nor were any cracks observed in the model after impact.

  18. Corrosion in the aluminum containment tank at the Nuclear Center of Mexico TRIGA Mark III reactor

    SciTech Connect

    Mota, Juan Ramon

    1986-07-01

    The reactor developed a leak inside the exposure room discovered when it was opened for a routine inspection. This leak started to diminish immediately after it was found and disappeared completely in 2.5 months. The hydrostatic tests of the exposure room cooling water pipes and of the primary cooling system suction pipe proved that piping do not have leaks. A portion of the total volume of water was drained from the pool to conduct an inspection on the aluminum liner. Penetrant dye tests were initiated over welded Joints and walls. Welded Joints were all found to be in good condition but a total of 35 indications were reported on walls and concentrated on two main areas. A vacuum system was used to test for leakage. Seven indications were found to be perforations that crossed through the wall, fifteen indications did not cross through the wall but required repair and the rest were superficial irregularities. For the inspection of surfaces that remained covered by water, two methods were used. One was a television camera that was adapted to be used under water and hooked to a monitor and a videorecorder for close up inspection of the walls. The other consisted of submarine still color photography performed by divers. The evaluation of these inspections concluded that out of the 10 areas previously identified, only one presented the kind of problem that required repair. The last inspection performed was that using ultrasound techniques. Irregularities found did not require complete replacement of the aluminum liner. The repair procedures included the welding of aluminum plates over damaged areas and the injection of an effective insulating material (resin) to stop the corrosion mechanism.

  19. DEVELOPING AND QUANTIFYING PARAMETERS FOR CLOSURE WELDING OVERPACKS CONTAINING RESEARCH REACTOR SPENT NUCLEAR FUEL AT HANFORD

    SciTech Connect

    CANNELL GR

    2007-11-07

    Fluor engineers developed a Gas Tungsten Arc Welding (GTAW) technique and parameters, demonstrated requisite weld quality and successfully closure-welded packaged spent nuclear fuel (SNF) overpacks at the Hanford Site. This paper reviews weld development and qualification activities associated with the overpack closure-welding and provides a summary of the production campaign. The primary requirement of the closure weld is to provide leaktight confinement of the packaged material against release to the environment during interim storage (40-year design term). Required weld quality, in this case, was established through up-front development and qualification, and then verification of parameter compliance during production welding. This approach was implemented to allow for a simpler overpack design and more efficient production operations than possible with approaches using routine post-weld testing and nondestructive examination (NDE). . A series of welding trials were conducted to establish the desired welding technique and parameters. Qualification of the process included statistical evaluation and American Society of Mechanical Engineers (ASME) Section IX testing. In addition, pull testing with a weighted mockup, and thermal calculation/physical testing to identify the maximum temperature the packaged contents would be subject to during welding, was performed. Thirteen overpacks were successfully packaged and placed into interim storage. The closure-welding development activities (including pull testing and thermal analysis) provided the needed confidence that the packaged SNF overpacks could be safely handled and placed into interim storage, and remain leaktight for the duration of the storage term.

  20. Developing and Qualifying Parameters for Closure Welding Overpacks Containing Research Reactor Spent Nuclear Fuel at Hanford

    SciTech Connect

    Cannell, G.R.; Goldmann, L.H.; McCormack, R.L.

    2008-07-01

    Fluor engineers developed a Gas Tungsten Arc Welding (GTAW) technique and parameters, demonstrated requisite weld quality, and successfully closure-welded packaged spent nuclear fuel (SNF) overpacks at the Hanford Site. This paper reviews weld development and qualification activities associated with the overpack closure-welding and provides a summary of the production campaign. The primary requirement of the closure weld is to provide leak-tight confinement of the packaged material against release to the environment during interim storage (40-year design term). Required weld quality, in this case, was established through up-front development and qualification, and then verification of parameter compliance during production welding. This approach was implemented to allow for a simpler overpack design and more efficient production operations than possible with approaches using routine post-weld testing and nondestructive examination (NDE). A series of welding trials were conducted to establish the desired welding technique and parameters. Qualification of the process included statistical evaluation and American Society of Mechanical Engineers (ASME) Section IX testing. In addition, pull testing with a weighted mockup, and thermal calculation/physical testing to identify the maximum temperature the packaged contents would be subject to during welding, was performed. Thirteen overpacks were successfully packaged and placed into interim storage. The closure-welding development activities (including pull testing and thermal analysis) provided the needed confidence that the packaged SNF overpacks could be safely handled and placed into interim storage, and remain leak-tight for the duration of the storage term. (author)

  1. Nuclear reactor overflow line

    DOEpatents

    Severson, Wayne J.

    1976-01-01

    The overflow line for the reactor vessel of a liquid-metal-cooled nuclear reactor includes means for establishing and maintaining a continuous bleed flow of coolant amounting to 5 to 10% of the total coolant flow through the overflow line to prevent thermal shock to the overflow line when the reactor is restarted following a trip. Preferably a tube is disposed concentrically just inside the overflow line extending from a point just inside the reactor vessel to an overflow tank and a suction line is provided opening into the body of liquid metal in the reactor vessel and into the annulus between the overflow line and the inner tube.

  2. RADIATION FACILITY FOR NUCLEAR REACTORS

    DOEpatents

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

    1961-12-12

    A radiation facility is designed for irradiating samples in close proximity to the core of a nuclear reactor. The facility comprises essentially a tubular member extending through the biological shield of the reactor and containing a manipulatable rod having the sample carrier at its inner end, the carrier being longitudinally movable from a position in close proximity to the reactor core to a position between the inner and outer faces of the shield. Shield plugs are provided within the tubular member to prevent direct radiation from the core emanating therethrough. In this device, samples may be inserted or removed during normal operation of the reactor without exposing personnel to direct radiation from the reactor core. A storage chamber is also provided within the radiation facility to contain an irradiated sample during the period of time required to reduce the radioactivity enough to permit removal of the sample for external handling. (AEC)

  3. Nuclear reactor apparatus

    DOEpatents

    Wade, Elman E.

    1978-01-01

    A lifting, rotating and sealing apparatus for nuclear reactors utilizing rotating plugs above the nuclear reactor core. This apparatus permits rotation of the plugs to provide under the plug refueling of a nuclear core. It also provides a means by which positive top core holddown can be utilized. Both of these operations are accomplished by means of the apparatus lifting the top core holddown structure off the nuclear core while stationary, and maintaining this structure in its elevated position during plug rotation. During both of these operations, the interface between the rotating member and its supporting member is sealingly maintained.

  4. A 640 foot per second impact test of a two foot diameter model nuclear reactor containment system without fracture

    NASA Technical Reports Server (NTRS)

    Puthoff, R. L.

    1971-01-01

    An impact test was conducted on an 1142 pound 2 foot diameter sphere model. The purpose of this test was to determine the feasibility of containing the fission products of a mobile reactor in an impact. The model simulated the reactor core, energy absorbing gamma shielding, neutron shielding and the containment vessel. It was impacted against an 18,000 pound reinforced concrete block. The model was significantly deformed and the concrete block demolished. No leaks were detected nor cracks observed in the model after impact.

  5. NUCLEAR REACTOR CONTROL SYSTEM

    DOEpatents

    Howard, D.F.; Motta, E.E.

    1961-06-27

    A method for controlling the excess reactivity in a nuclear reactor throughout the core life while maintaining the neutron flux distribution at the desired level is described. The control unit embodies a container having two electrodes of different surface area immersed in an electrolytic solution of a good neutron sbsorbing metal ion such as boron, gadolinium, or cadmium. Initially, the neutron absorber is plated on the larger electrode to control the greater neutron flux of a freshly refueled core. As the fuel burns up, the excess reactivity decreases and the neutron absorber is then plated onto the smaller electrode so that the number of neutrons absorbed also decreases. The excess reactivity in the core may thus be maintained without the introduction of serious perturbations in the neutron flux distributibn.

  6. HOMOGENEOUS NUCLEAR POWER REACTOR

    DOEpatents

    King, L.D.P.

    1959-09-01

    A homogeneous nuclear power reactor utilizing forced circulation of the liquid fuel is described. The reactor does not require fuel handling outside of the reactor vessel during any normal operation including complete shutdown to room temperature, the reactor being selfregulating under extreme operating conditions and controlled by the thermal expansion of the liquid fuel. The liquid fuel utilized is a uranium, phosphoric acid, and water solution which requires no gus exhaust system or independent gas recombining system, thereby eliminating the handling of radioiytic gas.

  7. Nuclear reactor control

    SciTech Connect

    Ingham, R.V.

    1980-01-01

    A liquid metal cooled fast breeder nuclear reactor has power setback means for use in an emergency. On initiation of a trip-signal a control rod is injected into the core in two stages, firstly, by free fall to effect an immediate power-set back to a safe level and, secondly, by controlled insertion. Total shut-down of the reactor under all emergencies is avoided. 4 claims.

  8. Digital computer operation of a nuclear reactor

    DOEpatents

    Colley, Robert W.

    1984-01-01

    A method is described for the safe operation of a complex system such as a nuclear reactor using a digital computer. The computer is supplied with a data base containing a list of the safe state of the reactor and a list of operating instructions for achieving a safe state when the actual state of the reactor does not correspond to a listed safe state, the computer selects operating instructions to return the reactor to a safe state.

  9. Digital computer operation of a nuclear reactor

    DOEpatents

    Colley, R.W.

    1982-06-29

    A method is described for the safe operation of a complex system such as a nuclear reactor using a digital computer. The computer is supplied with a data base containing a list of the safe state of the reactor and a list of operating instructions for achieving a safe state when the actual state of the reactor does not correspond to a listed safe state, the computer selects operating instructions to return the reactor to a safe state.

  10. NUCLEAR REACTOR COMPENENT CLADDING MATERIAL

    DOEpatents

    Draley, J.E.; Ruther, W.E.

    1959-01-27

    Fuel elements and coolant tubes used in nuclear reactors of the heterogeneous, water-cooled type are described, wherein the coolant tubes extend through the moderator and are adapted to contain the fuel elements. The invention comprises forming the coolant tubes and the fuel element cladding material from an alloy of aluminum and nickel, or an alloy of aluminum, nickel, alloys are selected to prevent intergranular corrosion of these components by water at temperatures up to 35O deg C.

  11. 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.

  12. Heat dissipating nuclear reactor with metal liner

    DOEpatents

    Gluekler, E.L.; Hunsbedt, A.; Lazarus, J.D.

    1985-11-21

    A nuclear reactor containment including a reactor vessel disposed within a cavity with capability for complete inherent decay heat removal in the earth and surrounded by a cast steel containment member which surrounds the vessel is described in this disclosure. The member has a thick basemat in contact with metal pilings. The basemat rests on a bed of porous particulate material, into which water is fed to produce steam which is vented to the atmosphere. There is a gap between the reactor vessel and the steel containment member. The containment member holds any sodium or core debris escaping from the reactor vessel if the core melts and breaches the vessel.

  13. Heat dissipating nuclear reactor with metal liner

    DOEpatents

    Gluekler, Emil L.; Hunsbedt, Anstein; Lazarus, Jonathan D.

    1987-01-01

    Disclosed is a nuclear reactor containment including a reactor vessel disposed within a cavity with capability for complete inherent decay heat removal in the earth and surrounded by a cast steel containment member which surrounds the vessel. The member has a thick basemat in contact with metal pilings. The basemat rests on a bed of porous particulate material, into which water is fed to produce steam which is vented to the atmosphere. There is a gap between the reactor vessel and the steel containment member. The containment member holds any sodium or core debris escaping from the reactor vessel if the core melts and breaches the vessel.

  14. THERMAL NUCLEAR REACTOR

    DOEpatents

    Fenning, F.W.; Jackson, R.F.

    1957-09-24

    Nuclear reactors of the graphite moderated air cooled type in which canned slugs or rods of fissile material are employed are discussed. Such a reactor may be provided with a means for detecting dust particles in the exhausted air. The means employed are lengths of dust absorbent cord suspended in vertical holes in the shielding structure above each vertical coolant flow channel to hang in the path of the cooling air issuing from the channels, and associated spindles and drive motors for hauling the cords past detectors, such as Geiger counters, for inspecting the cords periodically. This design also enables detecting the individual channel in which a fault condition may have occurred.

  15. Advanced Non-Destructive Assessment Technology to Determine the Aging of Silicon Containing Materials for Generation IV Nuclear Reactors

    NASA Astrophysics Data System (ADS)

    Koenig, T. W.; Olson, D. L.; Mishra, B.; King, J. C.; Fletcher, J.; Gerstenberger, L.; Lawrence, S.; Martin, A.; Mejia, C.; Meyer, M. K.; Kennedy, R.; Hu, L.; Kohse, G.; Terry, J.

    2011-06-01

    To create an in-situ, real-time method of monitoring neutron damage within a nuclear reactor core, irradiated silicon carbide samples are examined to correlate measurable variations in the material properties with neutron fluence levels experienced by the silicon carbide (SiC) during the irradiation process. The reaction by which phosphorus doping via thermal neutrons occurs in the silicon carbide samples is known to increase electron carrier density. A number of techniques are used to probe the properties of the SiC, including ultrasonic and Hall coefficient measurements, as well as high frequency impedance analysis. Gamma spectroscopy is also used to examine residual radioactivity resulting from irradiation activation of elements in the samples. Hall coefficient measurements produce the expected trend of increasing carrier concentration with higher fluence levels, while high frequency impedance analysis shows an increase in sample impedance with increasing fluence.

  16. Shutdown system for a nuclear reactor

    DOEpatents

    Groh, E.F.; Olson, A.P.; Wade, D.C.; Robinson, B.W.

    1984-06-05

    An ultimate shutdown system is provided for termination of neutronic activity in a nuclear reactor. The shutdown system includes bead chains comprising spherical containers suspended on a flexible cable. The containers are comprised of mating hemispherical shells which provide a ruggedized enclosure for reactor poison material. The bead chains, normally suspended above the reactor core on storage spools, are released for downward travel upon command from an external reactor monitor. The chains are capable of horizontal movement, so as to flow around obstructions in the reactor during their downward motion. 8 figs.

  17. Shutdown system for a nuclear reactor

    DOEpatents

    Groh, Edward F.; Olson, Arne P.; Wade, David C.; Robinson, Bryan W.

    1984-01-01

    An ultimate shutdown system is provided for termination of neutronic activity in a nuclear reactor. The shutdown system includes bead chains comprising spherical containers suspended on a flexible cable. The containers are comprised of mating hemispherical shells which provide a ruggedized enclosure for reactor poison material. The bead chains, normally suspended above the reactor core on storage spools, are released for downward travel upon command from an external reactor monitor. The chains are capable of horizontal movement, so as to flow around obstructions in the reactor during their downward motion.

  18. 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)

  19. Nuclear reactor safety device

    DOEpatents

    Hutter, Ernest

    1986-01-01

    A safety device is disclosed for use in a nuclear reactor for axially repositioning a control rod with respect to the reactor core in the event of an upward thermal excursion. Such safety device comprises a laminated helical ribbon configured as a tube-like helical coil having contiguous helical turns with slidably abutting edges. The helical coil is disclosed as a portion of a drive member connected axially to the control rod. The laminated ribbon is formed of outer and inner laminae. The material of the outer lamina has a greater thermal coefficient of expansion than the material of the inner lamina. In the event of an upward thermal excursion, the laminated helical coil curls inwardly to a smaller diameter. Such inward curling causes the total length of the helical coil to increase by a substantial increment, so that the control rod is axially repositioned by a corresponding amount to reduce the power output of the reactor.

  20. Aerodynamic induced vibrations on reactor containment fan coolers

    SciTech Connect

    Jones, R. )

    1989-05-01

    Diagnosis of problems on reactor containment fan coolers is not always straightforward. This article describes the method of investigation employed at Tennessee Valley Authority's Sequoyah Nuclear Power Station to diagnose and correct aerodynamic induced vibrations on the coolers.

  1. Nuclear reactor safety device

    DOEpatents

    Hutter, E.

    1983-08-15

    A safety device is described for use in a nuclear reactor for axially repositioning a control rod with respect to the reactor core in the event of a thermal excursion. It comprises a laminated strip helically configured to form a tube, said tube being in operative relation to said control rod. The laminated strip is formed of at least two materials having different thermal coefficients of expansion, and is helically configured such that the material forming the outer lamina of the tube has a greater thermal coefficient of expansion than the material forming the inner lamina of said tube. In the event of a thermal excursion the laminated strip will tend to curl inwardly so that said tube will increase in length, whereby as said tube increases in length it exerts a force on said control rod to axially reposition said control rod with respect to said core.

  2. Nuclear reactor shutdown system

    DOEpatents

    Bhate, Suresh K.; Cooper, Martin H.; Riffe, Delmar R.; Kinney, Calvin L.

    1981-01-01

    An inherent shutdown system for a nuclear reactor having neutron absorbing rods affixed to an armature which is held in an upper position by a magnetic flux flowing through a Curie temperature material. The Curie temperature material is fixedly positioned about the exterior of an inner duct in an annular region through which reactor coolant flows. Elongated fuel rods extending from within the core upwardly toward the Curie temperature material are preferably disposed within the annular region. Upon abnormal conditions which result in high neutron flux and coolant temperature, the Curie material loses its magnetic permeability, breaking the magnetic flux path and allowing the armature and absorber rods to drop into the core, thus shutting down the fissioning reaction. The armature and absorber rods are retrieved by lowering the housing for the electromagnet forming coils which create a magnetic flux path which includes the inner duct wall. The coil housing then is raised, resetting the armature.

  3. Nuclear Reactor Kinetics and Control.

    SciTech Connect

    JEFFERY,; LEWINS, D.

    2009-07-27

    Version 00 Dr. J.D. Lewins has now released the following legacy book for free distribution: Nuclear Reactor Kinetics and Control, Pergamon Press, London, 275 pages, 1978. 1. Introductory Review 2. Neutron and Precursor Equations 3. Elementary Solutions of the Kinetics Equations at Low Power 4. Linear Reactor Process Dynamics with Feedback 5. Power Reactor Control Systems 6. Fluctuations and Reactor Noise 7. Safety and Reliability 8. Non Linear Systems; Stability and Control 9. Analogue Computing Addendum: Jay Basken and Jeffery D. Lewins: Power Series Solution of the Reactor Kinetics Equations, Nuclear Science and Engineering: 122, 407-436 (1996) (authorized for distribution with the book: courtesy of the American Nuclear Society)

  4. Nuclear reactor control apparatus

    DOEpatents

    Sridhar, Bettadapur N.

    1983-11-01

    Nuclear reactor core safety rod release apparatus comprises a control rod having a detent notch in the form of an annular peripheral recess at its upper end, a control rod support tube for raising and lowering the control rod under normal conditions, latches pivotally mounted on the control support tube with free ends thereof normally disposed in the recess in the control rod, and cam means for pivoting the latches out of the recess in the control rod when a scram condition occurs. One embodiment of the invention comprises an additional magnetically-operated latch for releasing the control rod under two different conditions, one involving seismic shock.

  5. 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.

  6. NUCLEAR REACTOR CORE DESIGN

    DOEpatents

    Mahlmeister, J.E.; Peck, W.S.; Haberer, W.V.; Williams, A.C.

    1960-03-22

    An improved core design for a sodium-cooled, graphitemoderated nuclear reactor is described. The improved reactor core comprises a number of blocks of moderator material, each block being in the shape of a regular prism. A number of channels, extending the length of each block, are disposed around the periphery. When several blocks are placed in contact to form the reactor core, the channels in adjacent blocks correspond with each other to form closed conduits extending the length of the core. Fuel element clusters are disposed in these closed conduits, and liquid coolant is forced through the annulus between the fuel cluster and the inner surface of the conduit. In a preferred embodiment of the invention, the moderator blocks are in the form of hexagonal prisms with longitudinal channels cut into the corners of the hexagon. The main advantage of an "edge-loaded" moderator block is that fewer thermal neutrons are absorbed by the moderator cladding, as compared with a conventional centrally loaded moderator block.

  7. REACTOR BASE, SOUTHEAST CORNER. INTERIOR WILL CONTAIN REACTOR TANK, COOLING ...

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

    REACTOR BASE, SOUTHEAST CORNER. INTERIOR WILL CONTAIN REACTOR TANK, COOLING WATER PIPES, COOLING AIR DUCTS, AND SHIELDING. INL NEGATIVE NO. 776. Unknown Photographer, 10/1950 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

  8. EMERGENCY SHUTDOWN FOR NUCLEAR REACTORS

    DOEpatents

    Paget, J.A.; Koutz, S.L.; Stone, R.S.; Stewart, H.B.

    1963-12-24

    An emergency shutdown or scram apparatus for use in a nuclear reactor that includes a neutron absorber suspended from a temperature responsive substance that is selected to fail at a preselected temperature in excess of the normal reactor operating temperature, whereby the neutron absorber is released and allowed to fall under gravity to a preselected position within the reactor core is presented. (AEC)

  9. Nuclear reactor construction with bottom supported reactor vessel

    DOEpatents

    Sharbaugh, John E.

    1987-01-01

    An improved liquid metal nuclear reactor construction has a reactor core and a generally cylindrical reactor vessel for holding a large pool of low pressure liquid metal coolant and housing the core within the pool. The reactor vessel has an open top end, a closed flat bottom end wall and a continuous cylindrical closed side wall interconnecting the top end and bottom end wall. The reactor also has a generally cylindrical concrete containment structure surrounding the reactor vessel and being formed by a cylindrical side wall spaced outwardly from the reactor vessel side wall and a flat base mat spaced below the reactor vessel bottom end wall. A central support pedestal is anchored to the containment structure base mat and extends upwardly therefrom to the reactor vessel and upwardly therefrom to the reactor core so as to support the bottom end wall of the reactor vessel and the lower end of the reactor core in spaced apart relationship above the containment structure base mat. Also, an annular reinforced support structure is disposed in the reactor vessel on the bottom end wall thereof and extends about the lower end of the core so as to support the periphery thereof. In addition, an annular support ring having a plurality of inward radially extending linear members is disposed between the containment structure base mat and the bottom end of the reactor vessel wall and is connected to and supports the reactor vessel at its bottom end on the containment structure base mat so as to allow the reactor vessel to expand radially but substantially prevent any lateral motions that might be imposed by the occurrence of a seismic event. The reactor construction also includes a bed of insulating material in sand-like granular form, preferably being high density magnesium oxide particles, disposed between the containment structure base mat and the bottom end wall of the reactor vessel and uniformly supporting the reactor vessel at its bottom end wall on the containment

  10. Nuclear Reactors. Revised.

    ERIC Educational Resources Information Center

    Hogerton, John F.

    This publication is one of a series of information booklets for the general public published by the United States Atomic Energy Commission. Among the topics discussed are: How Reactors Work; Reactor Design; Research, Teaching, and Materials Testing; Reactors (Research, Teaching and Materials); Production Reactors; Reactors for Electric Power…

  11. Nuclear reactor control apparatus

    DOEpatents

    Sridhar, Bettadapur N.

    1983-10-25

    Nuclear reactor safety rod release apparatus comprises a ring which carries detents normally positioned in an annular recess in outer side of the rod, the ring being held against the lower end of a drive shaft by magnetic force exerted by a solenoid carried by the drive shaft. When the solenoid is de-energized, the detent-carrying ring drops until the detents contact a cam surface associated with the lower end of the drive shaft, at which point the detents are cammed out of the recess in the safety rod to release the rod from the drive shaft. In preferred embodiments of the invention, an additional latch is provided to release a lower portion of a safety rod under conditions that may interfere with movement of the entire rod.

  12. Nuclear reactor control

    DOEpatents

    Cawley, William E.; Warnick, Robert F.

    1982-01-01

    1. In a nuclear reactor incorporating a plurality of columns of tubular fuel elements disposed in horizontal tubes in a mass of graphite wherein water flows through the tubes to cool the fuel elements, the improvement comprising at least one control column disposed in a horizontal tube including fewer fuel elements than in a normal column of fuel elements and tubular control elements disposed at both ends of said control column, and means for varying the horizontal displacement of the control column comprising a winch at the upstream end of the control column and a cable extending through the fuel and control elements and attached to the element at the downstream end of the column.

  13. Fast reactors and nuclear nonproliferation

    SciTech Connect

    Avrorin, E.N.; Rachkov, V.I.; Chebeskov, A.N.

    2013-07-01

    Problems are discussed with regard to nuclear fuel cycle resistance in fast reactors to nuclear proliferation risk due to the potential for use in military programs of the knowledge, technologies and materials gained from peaceful nuclear power applications. Advantages are addressed for fast reactors in the creation of a more reliable mode of nonproliferation in the closed nuclear fuel cycle in comparison with the existing fully open and partially closed fuel cycles of thermal reactors. Advantages and shortcomings are also discussed from the point of view of nonproliferation from the start with fast reactors using plutonium of thermal reactor spent fuel and enriched uranium fuel to the gradual transition using their own plutonium as fuel. (authors)

  14. Reactivity control assembly for nuclear reactor. [LMFBR

    DOEpatents

    Bollinger, L.R.

    1982-03-17

    This invention, which resulted from a contact with the United States Department of Energy, relates to a control mechanism for a nuclear reactor and, more particularly, to an assembly for selectively shifting different numbers of reactivity modifying rods into and out of the core of a nuclear reactor. It has been proposed heretofore to control the reactivity of a breeder reactor by varying the depth of insertion of control rods (e.g., rods containing a fertile material such as ThO/sub 2/) in the core of the reactor, thereby varying the amount of neutron-thermalizing coolant and the amount of neutron-capturing material in the core. This invention relates to a mechanism which can advantageously be used in this type of reactor control system.

  15. Autonomous Control of Space Nuclear Reactors

    NASA Technical Reports Server (NTRS)

    Merk, John

    2013-01-01

    Nuclear reactors to support future robotic and manned missions impose new and innovative technological requirements for their control and protection instrumentation. Long-duration surface missions necessitate reliable autonomous operation, and manned missions impose added requirements for failsafe reactor protection. There is a need for an advanced instrumentation and control system for space-nuclear reactors that addresses both aspects of autonomous operation and safety. The Reactor Instrumentation and Control System (RICS) consists of two functionally independent systems: the Reactor Protection System (RPS) and the Supervision and Control System (SCS). Through these two systems, the RICS both supervises and controls a nuclear reactor during normal operational states, as well as monitors the operation of the reactor and, upon sensing a system anomaly, automatically takes the appropriate actions to prevent an unsafe or potentially unsafe condition from occurring. The RPS encompasses all electrical and mechanical devices and circuitry, from sensors to actuation device output terminals. The SCS contains a comprehensive data acquisition system to measure continuously different groups of variables consisting of primary measurement elements, transmitters, or conditioning modules. These reactor control variables can be categorized into two groups: those directly related to the behavior of the core (known as nuclear variables) and those related to secondary systems (known as process variables). Reliable closed-loop reactor control is achieved by processing the acquired variables and actuating the appropriate device drivers to maintain the reactor in a safe operating state. The SCS must prevent a deviation from the reactor nominal conditions by managing limitation functions in order to avoid RPS actions. The RICS has four identical redundancies that comply with physical separation, electrical isolation, and functional independence. This architecture complies with the

  16. Nuclear Reactor Kinetics and Control.

    Energy Science and Technology Software Center (ESTSC)

    2009-07-27

    Version 00 Dr. J.D. Lewins has now released the following legacy book for free distribution: Nuclear Reactor Kinetics and Control, Pergamon Press, London, 275 pages, 1978. 1. Introductory Review 2. Neutron and Precursor Equations 3. Elementary Solutions of the Kinetics Equations at Low Power 4. Linear Reactor Process Dynamics with Feedback 5. Power Reactor Control Systems 6. Fluctuations and Reactor Noise 7. Safety and Reliability 8. Non Linear Systems; Stability and Control 9. Analogue Computingmore » Addendum: Jay Basken and Jeffery D. Lewins: Power Series Solution of the Reactor Kinetics Equations, Nuclear Science and Engineering: 122, 407-436 (1996) (authorized for distribution with the book: courtesy of the American Nuclear Society)« less

  17. Nuclear reactor I

    DOEpatents

    Ference, Edward W.; Houtman, John L.; Waldby, Robert N.

    1977-01-01

    A nuclear reactor, particularly a liquid-metal breeder reactor whose upper internals include provision for channeling the liquid metal flowing from the core-component assemblies to the outlet plenum in vertical paths in direction generally along the direction of the respective assemblies. The metal is channeled by chimneys, each secured to, and extending from, a grid through whose openings the metal emitted by a plurality of core-component assemblies encompassed by the grid flows. To reduce the stresses resulting from structural interaction, or the transmissive of thermal strains due to large temperature differences in the liquid metal emitted from neighboring core-component assemblies, throughout the chimneys and the other components of the upper internals, the grids and the chimneys are supported from the heat plate and the core barrel by support columns (double portal support) which are secured to the head plate at the top and to a member, which supports the grids and is keyed to the core barrel, at the bottom. In addition to being restrained from lateral flow by the chimneys, the liquid metal is also restrained from flowing laterally by a peripheral seal around the top of the core. This seal limits the flow rate of liquid metal, which may be sharply cooled during a scram, to the outlet nozzles. The chimneys and the grids are formed of a highly-refractory, high corrosion-resistant nickel-chromium-iron alloy which can withstand the stresses produced by temperature differences in the liquid metal. The chimneys are supported by pairs of plates, each pair held together by hollow stubs coaxial with, and encircling, the chimneys. The plates and stubs are a welded structure but, in the interest of economy, are composed of stainless steel which is not weld compatible with the refractory metal. The chimneys and stubs are secured together by shells of another nickel-chromium-iron alloy which is weld compatible with, and is welded to, the stubs and has about the same

  18. Nuclear Reactor Engineering Analysis Laboratory

    SciTech Connect

    Carlos Chavez-Mercado; Jaime B. Morales-Sandoval; Benjamin E. Zayas-Perez

    1998-12-31

    The Nuclear Reactor Engineering Analysis Laboratory (NREAL) is a sophisticated computer system with state-of-the-art analytical tools and technology for analysis of light water reactors. Multiple application software tools can be activated to carry out different analyses and studies such as nuclear fuel reload evaluation, safety operation margin measurement, transient and severe accident analysis, nuclear reactor instability, operator training, normal and emergency procedures optimization, and human factors engineering studies. An advanced graphic interface, driven through touch-sensitive screens, provides the means to interact with specialized software and nuclear codes. The interface allows the visualization and control of all observable variables in a nuclear power plant (NPP), as well as a selected set of nonobservable or not directly controllable variables from conventional control panels.

  19. An 810 ft/sec soil impact test of a 2-foot diameter model nuclear reactor containment system

    NASA Technical Reports Server (NTRS)

    Puthoff, R. L.

    1972-01-01

    A soil impact test was conducted on a 880-pound 2-foot diameter sphere model. The impact area consisted of back filled desert earth and rock. The impact generated a crater 5 feet in diameter by 5 feet deep. It buried itself a total of 15 feet - as measured to the bottom of the model. After impact the containment vessel was pressure checked. No leaks were detected nor cracks observed.

  20. 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.

  1. Nuclear reactors for space power

    SciTech Connect

    Buden, D.

    1985-02-01

    The growth in power demands for spacecraft, especially outer planet missions, is driving the development of space nuclear power systems. Nuclear reactors could also be used to process lunar materials to take advantage of order of magnitude lower fuel requirements to move construction components off the moon instead of the earth. Larger, more powerful broadcast satellites which lower the GEO station space demand could use nuclear power, as could navigational systems, orbital transfer vehicles and a manned Mars mission. The SP-100 design is currently undergoing parametric evaluation before engineering studies begin. Safety concerns are concentrated on preventing fissioning until the reactor is on-orbit and keeping the active or discarded reactor out of the atmosphere until the radioactivity has decayed to levels defined by international standards.

  2. NUCLEAR REACTOR UNLOADING APPARATUS

    DOEpatents

    Leverett, M.C.; Howe, J.P.

    1959-01-20

    An unloading device is described for a heterogeneous reactor of the type wherein the fuel elements are in the form of cylindrical slugs and are disposed in horizontal coolant tubes which traverse the reactor core, coolant fluid being circulated through the tubes. The coolant tubes have at least two inwardly protruding ribs from their lower surfaces to support the slugs in spaced relationship to the inside walls of the tubes. The unloading device consists of a ribbon-like extractor member insertable into the coolant tubes in the space between the ribs and adapted to slide under the fuel slugs thereby raising them off of the ribs and forming a slideway for removing them from the reactor. The fuel slugs are ejected by being forced out of the tubes by incoming new fuel slugs or by a push rod insentable through the inlet end of the fuel tubes.

  3. Nuclear reactor downcomer flow deflector

    DOEpatents

    Gilmore, Charles B.; Altman, David A.; Singleton, Norman R.

    2011-02-15

    A nuclear reactor having a coolant flow deflector secured to a reactor core barrel in line with a coolant inlet nozzle. The flow deflector redirects incoming coolant down an annulus between the core barrel and the reactor vessel. The deflector has a main body with a front side facing the fluid inlet nozzle and a rear side facing the core barrel. The rear side of the main body has at least one protrusion secured to the core barrel so that a gap exists between the rear side of the main body adjacent the protrusion and the core barrel. Preferably, the protrusion is a relief that circumscribes the rear side of the main body.

  4. NUCLEAR REACTOR COOLANT

    DOEpatents

    Colichman, E.L.

    1959-10-20

    The formation of new reactor coolants which suppress polymerization resulting from pyrolytic and radiation decomposition is described. The coolants consist of polyphenyls and condensed ring compounds having from two to about four carbon rings and from 0.1 to about 5% of beryllium or magnesium dispersed in the hydrocarbon.

  5. NUCLEAR REACTOR COOLANT

    DOEpatents

    Colichman, E.L.

    1959-10-20

    The formation of new reactor coolants which suppress polymerization resulting from pyrolitic and radiation decomposition is described. The coolants consist of polyphenyls and condensed ring compounds having from two to about four carbon rings and from 0.1 to about 10% of an alkall metal dispersed in the hydrocarbon.

  6. Fluid sampling system for a nuclear reactor

    DOEpatents

    Lau, L.K.; Alper, N.I.

    1994-11-22

    A system of extracting fluid samples, either liquid or gas, from the interior of a nuclear reactor containment utilizes a jet pump. To extract the sample fluid, a nonradioactive motive fluid is forced through the inlet and discharge ports of a jet pump located outside the containment, creating a suction that draws the sample fluid from the containment through a sample conduit connected to the pump suction port. The mixture of motive fluid and sample fluid is discharged through a return conduit to the interior of the containment. The jet pump and means for removing a portion of the sample fluid from the sample conduit can be located in a shielded sample grab station located next to the containment. A non-nuclear grade active pump can be located outside the grab sampling station and the containment to pump the nonradioactive motive fluid through the jet pump. 1 fig.

  7. Fluid sampling system for a nuclear reactor

    DOEpatents

    Lau, Louis K.; Alper, Naum I.

    1994-01-01

    A system of extracting fluid samples, either liquid or gas, from the interior of a nuclear reactor containment utilizes a jet pump. To extract the sample fluid, a nonradioactive motive fluid is forced through the inlet and discharge ports of a jet pump located outside the containment, creating a suction that draws the sample fluid from the containment through a sample conduit connected to the pump suction port. The mixture of motive fluid and sample fluid is discharged through a return conduit to the interior of the containment. The jet pump and means for removing a portion of the sample fluid from the sample conduit can be located in a shielded sample grab station located next to the containment. A non-nuclear grade active pump can be located outside the grab sampling station and the containment to pump the nonradioactive motive fluid through the jet pump.

  8. Permanent seal ring for a nuclear reactor cavity

    SciTech Connect

    Hankinson, M.F.; Marshall, J.R.

    1988-05-31

    A nuclear reactor containment arrangement is described including: a. a reactor vessel which thermally expands and contracts during cyclic operation of the reactor and which has a peripheral wall; b. a containment wall spaced apart from and surrounding the peripheral wall of the reactor vessel and defining an annular thermal expansion gap therebetween for accommodating thermal expansion; and c. an annular ring seal which sealingly engages and is affixed to and extends between the peripheral wall of the reactor vessel and the containment wall.

  9. Heat pipe nuclear reactor for space power

    NASA Technical Reports Server (NTRS)

    Koening, D. R.

    1976-01-01

    A heat-pipe-cooled nuclear reactor has been designed to provide 3.2 MWth to an out-of-core thermionic conversion system. The reactor is a fast reactor designed to operate at a nominal heat-pipe temperature of 1675 K. Each reactor fuel element consists of a hexagonal molybdenum block which is bonded along its axis to one end of a molybdenum/lithium-vapor heat pipe. The block is perforated with an array of longitudinal holes which are loaded with UO2 pellets. The heat pipe transfers heat directly to a string of six thermionic converters which are bonded along the other end of the heat pipe. An assembly of 90 such fuel elements forms a hexagonal core. The core is surrounded by a thermal radiation shield, a thin thermal neutron absorber, and a BeO reflector containing boron-loaded control drums.

  10. Passive heat transfer means for nuclear reactors

    DOEpatents

    Burelbach, James P.

    1984-01-01

    An improved passive cooling arrangement is disclosed for maintaining adjacent or related components of a nuclear reactor within specified temperature differences. Specifically, heat pipes are operatively interposed between the components, with the vaporizing section of the heat pipe proximate the hot component operable to cool it and the primary condensing section of the heat pipe proximate the other and cooler component operable to heat it. Each heat pipe further has a secondary condensing section that is located outwardly beyond the reactor confinement and in a secondary heat sink, such as air ambient the containment, that is cooler than the other reactor component. Means such as shrouding normally isolated the secondary condensing section from effective heat transfer with the heat sink, but a sensor responds to overheat conditions of the reactor to open the shrouding, which thereby increases the cooling capacity of the heat pipe. By having many such heat pipes, an emergency passive cooling system is defined that is operative without electrical power.

  11. 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.

  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. METHOD OF OPERATING NUCLEAR REACTORS

    DOEpatents

    Untermyer, S.

    1958-10-14

    A method is presented for obtaining enhanced utilization of natural uranium in heavy water moderated nuclear reactors by charging the reactor with an equal number of fuel elements formed of natural uranium and of fuel elements formed of uranium depleted in U/sup 235/ to the extent that the combination will just support a chain reaction. The reactor is operated until the rate of burnup of plutonium equals its rate of production, the fuel elements are processed to recover plutonium, the depleted uranium is discarded, and the remaining uranium is formed into fuel elements. These fuel elements are charged into a reactor along with an equal number of fuel elements formed of uranium depleted in U/sup 235/ to the extent that the combination will just support a chain reaction, and reuse of the uranium is continued as aforesaid until it wlll no longer support a chain reaction when combined with an equal quantity of natural uranium.

  14. Nuclear reactor effluent monitoring

    SciTech Connect

    Minns, J.L.; Essig, T.H.

    1993-12-31

    Radiological environmental monitoring and effluent monitoring at nuclear power plants is important both for normal operations, as well as in the event of an accident. During normal operations, environmental monitoring verifies the effectiveness of in-plant measures for controlling the release of radioactive materials in the plant. Following an accident, it would be an additional mechanism for estimating doses to members of the general public. This paper identifies the U.S. Nuclear Regulatory Commission (NRC) regulatory basis for requiring radiological environmental and effluent monitoring, licensee conditions for effluent and environmental monitoring, NRC independent oversight activities, and NRC`s program results.

  15. NUCLEAR REACTOR FUEL ELEMENT

    DOEpatents

    Anderson, W.F.; Tellefson, D.R.; Shimazaki, T.T.

    1962-04-10

    A plate type fuel element which is particularly useful for organic cooled reactors is described. Generally, the fuel element comprises a plurality of fissionable fuel bearing plates held in spaced relationship by a frame in which the plates are slidably mounted in grooves. Clearance is provided in the grooves to allow the plates to expand laterally. The plates may be rigidly interconnected but are floatingly supported at their ends within the frame to allow for longi-tudinal expansion. Thus, this fuel element is able to withstand large temperature differentials without great structural stresses. (AEC)

  16. Heat-generating nuclear reactor

    SciTech Connect

    Dupuy, G.; Fajeau, M.; Labrousse, M.; Lerouge, B.; Minguet, J.

    1981-01-20

    A reactor vessel filled with coolant fluid is divided by a wall into an upper region and a lower region which contains the reactor core, part of the coolant fluid in the upper region being injected into the lower region. The injection flow rate is regulated as a function of the variations in pressure in the lower region by means of a baffle-plate container which communicates with a leak-tight chamber and with a storage reservoir, a flow of fluid from the chamber to the reservoir being established only at the time of a reduction in the rate of injection into the container. The reactor can be employed for the production of hot water which is passed through a heat exchanger and supplied to a heating installation.

  17. Current Abstracts Nuclear Reactors and Technology

    SciTech Connect

    Bales, J.D.; Hicks, S.C.

    1993-01-01

    This publication Nuclear Reactors and Technology (NRT) announces on a monthly basis the current worldwide information available from the open literature on nuclear reactors and technology, including all aspects of power reactors, components and accessories, fuel elements, control systems, and materials. This publication contains the abstracts of DOE reports, journal articles, conference papers, patents, theses, and monographs added to the Energy Science and Technology Database during the past month. Also included are US information obtained through acquisition programs or interagency agreements and international information obtained through acquisition programs or interagency agreements and international information obtained through the International Energy Agency`s Energy Technology Data Exchange or government-to-government agreements. The digests in NRT and other citations to information on nuclear reactors back to 1948 are available for online searching and retrieval on the Energy Science and Technology Database and Nuclear Science Abstracts (NSA) database. Current information, added daily to the Energy Science and Technology Database, is available to DOE and its contractors through the DOE Integrated Technical Information System. Customized profiles can be developed to provide current information to meet each user`s needs.

  18. 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.

  19. Horizontal baffle for nuclear reactors

    DOEpatents

    Rylatt, John A.

    1978-01-01

    A horizontal baffle disposed in the annulus defined between the core barrel and the thermal liner of a nuclear reactor thereby physically separating the outlet region of the core from the annular area below the horizontal baffle. The horizontal baffle prevents hot coolant that has passed through the reactor core from thermally damaging apparatus located in the annulus below the horizontal baffle by utilizing the thermally induced bowing of the horizontal baffle to enhance sealing while accommodating lateral motion of the baffle base plate.

  20. Nuclear reactors and the nuclear fuel cycle

    SciTech Connect

    Pearlman, H

    1989-11-01

    According to the author, the first sustained nuclear fission chain reaction was not at the University of Chicago, but at the Oklo site in the African country of Gabon. Proof of this phenomenon is provided by mass spectrometric and analytical chemical measurements by French scientists. The U.S. experience in developing power-producing reactors and their related fuel and fuel cycles is discussed.

  1. Containment system for supercritical water oxidation reactor

    SciTech Connect

    Chastagner, P.

    1991-12-31

    This invention is comprised of a system for containment of a supercritical water oxidation reactor in the event of a rupture of the reactor. The system includes a containment for housing the reaction vessel and a communicating chamber for holding a volume of coolant, such as water. The coolant is recirculated and sprayed to entrain and cool any reactants that might have escaped from the reaction vessel. Baffles at the entrance to the chamber prevent the sprayed coolant from contacting the reaction vessel. An impact-absorbing layer is positioned between the vessel and the containment to at least partially absorb momentum of any fragments propelled by the rupturing vessel. Remote, quick-disconnecting fittings exterior to the containment, in cooperation with shut-off valves, enable the vessel to be isolated and the system safely taken off-line. Normally-closed orifices throughout the containment and chamber enable decontamination of interior surfaces when necessary.

  2. Containment system for supercritical water oxidation reactor

    DOEpatents

    Chastagner, Philippe

    1994-01-01

    A system for containment of a supercritical water oxidation reactor in the event of a rupture of the reactor. The system includes a containment for housing the reaction vessel and a communicating chamber for holding a volume of coolant, such as water. The coolant is recirculated and sprayed to entrain and cool any reactants that might have escaped from the reaction vessel. Baffles at the entrance to the chamber prevent the sprayed coolant from contacting the reaction vessel. An impact-absorbing layer is positioned between the vessel and the containment to at least partially absorb momentum of any fragments propelled by the rupturing vessel. Remote, quick-disconnecting fittings exterior to the containment, in cooperation with shut-off valves, enable the vessel to be isolated and the system safely taken off-line. Normally-closed orifices throughout the containment and chamber enable decontamination of interior surfaces when necessary.

  3. Containment system for supercritical water oxidation reactor

    DOEpatents

    Chastagner, P.

    1994-07-05

    A system is described for containment of a supercritical water oxidation reactor in the event of a rupture of the reactor. The system includes a containment for housing the reaction vessel and a communicating chamber for holding a volume of coolant, such as water. The coolant is recirculated and sprayed to entrain and cool any reactants that might have escaped from the reaction vessel. Baffles at the entrance to the chamber prevent the sprayed coolant from contacting the reaction vessel. An impact-absorbing layer is positioned between the vessel and the containment to at least partially absorb momentum of any fragments propelled by the rupturing vessel. Remote, quick-disconnecting fittings exterior to the containment, in cooperation with shut-off valves, enable the vessel to be isolated and the system safely taken off-line. Normally-closed orifices throughout the containment and chamber enable decontamination of interior surfaces when necessary. 2 figures.

  4. Evaluated Neutron Nuclear Data for Reactor Physics Calculations.

    Energy Science and Technology Software Center (ESTSC)

    1988-09-15

    Version 00 The data file KEDAK contains the evaluated neutron nuclear data for a number of materials important for the reactor physics, specific physical experiments, burn up calculations, shielding and other applications.

  5. Weld monitor and failure detector for nuclear reactor system

    DOEpatents

    Sutton, Jr., Harry G.

    1987-01-01

    Critical but inaccessible welds in a nuclear reactor system are monitored throughout the life of the reactor by providing small aperture means projecting completely through the reactor vessel wall and also through the weld or welds to be monitored. The aperture means is normally sealed from the atmosphere within the reactor. Any incipient failure or cracking of the weld will cause the environment contained within the reactor to pass into the aperture means and thence to the outer surface of the reactor vessel where its presence is readily detected.

  6. Propellant actuated nuclear reactor steam depressurization valve

    DOEpatents

    Ehrke, Alan C.; Knepp, John B.; Skoda, George I.

    1992-01-01

    A nuclear fission reactor combined with a propellant actuated depressurization and/or water injection valve is disclosed. The depressurization valve releases pressure from a water cooled, steam producing nuclear reactor when required to insure the safety of the reactor. Depressurization of the reactor pressure vessel enables gravity feeding of supplementary coolant water through the water injection valve to the reactor pressure vessel to prevent damage to the fuel core.

  7. Nuclear reactors built, being built, or planned 1992

    SciTech Connect

    Not Available

    1993-07-01

    Nuclear Reactors Built, Being Built, or Planned contains unclassified information about facilities built, being built, or planned in the United States for domestic use or export as of December 31, 1992. The Office of Scientific and Technical Information, US Department of Energy, gathers this information annually from Washington headquarters and field offices of DOE from the US Nuclear Regulatory Commission (NRC); from the US reactor manufacturers who are the principal nuclear contractors for foreign reactor locations; from US and foreign embassies; and from foreign governmental nuclear departments. Information is presented on five parts: Civilian, Production, Military, Export and Critical Assembly.

  8. Nuclear reactors built, being built, or planned 1993

    SciTech Connect

    Not Available

    1993-08-01

    Nuclear Reactors Built, Being Built, or Planned contains unclassified information about facilities built, being built, or planned in the United States for domestic use or export as of December 31, 1993. The Office of Scientific and Technical Information, US Department of Energy, gathers this information annually from Washington headquarters and field offices of DOE; from the US Nuclear Regulatory Commission (NRC); from the US reactor manufacturers who are the principal nuclear embassies; and from foreign governmental nuclear departments. The book consists of three divisions, as follows: (1) a commercial reactor locator map and tables of the characteristic and statistical data that follow; a table of abbreviations; (2) tables of data for reactors operating, being built, or planned; and (3) tables of data for reactors that have been shut down permanently or dismantled. The reactors are subdivided into the following parts: civilian, production, military, export, and critical assembly.

  9. Nuclear reactors built, being built, or planned, 1991

    SciTech Connect

    Simpson, B.

    1992-07-01

    This document contains unclassified information about facilities built, being built, or planned in the United States for domestic use or export as of December 31, 1991. The book is divided into three major sections: Section 1 consists of a reactor locator map and reactor tables; Section 2 includes nuclear reactors that are operating, being built, or planned; and Section 3 includes reactors that have been shut down permanently or dismantled. Sections 2 and 3 contain the following classification of reactors: Civilian, Production, Military, Export, and Critical Assembly. Export reactor refers to a reactor for which the principal nuclear contractor is an American company -- working either independently or in cooperation with a foreign company (Part 4, in each section). Critical assembly refers to an assembly of fuel and assembly of fuel and moderator that requires an external source of neutrons to initiate and maintain fission. A critical assembly is used for experimental measurements (Part 5).

  10. Reactors for nuclear electric propulsion

    SciTech Connect

    Buden, D.; Angelo, J.A. Jr.

    1981-01-01

    Propulsion is the key to space exploitation and power is the key to propulsion. This paper examines the role of nuclear fission reactors as the primary power source for high specific impulse electric propulsion systems for space missions of the 1980s and 1990s. Particular mission applications include transfer to and a reusable orbital transfer vehicle from low-Earth orbit to geosynchronous orbit, outer planet exploration and reconnaissance missions, and as a versatile space tug supporting lunar resource development. Nuclear electric propulsion is examined as an indispensable component in space activities of the next two decades.

  11. Flow duct for nuclear reactors

    DOEpatents

    Straalsund, Jerry L.

    1978-01-01

    Improved liquid sodium flow ducts for nuclear reactors are described wherein the improvement comprises varying the wall thickness of each of the walls of a polygonal tubular duct structure so that each of the walls is of reduced cross-section along the longitudinal center line and of a greater cross-section along wall junctions with the other walls to form the polygonal tubular configuration.

  12. Designed porosity materials in nuclear reactor components

    DOEpatents

    Yacout, A. M.; Pellin, Michael J.; Stan, Marius

    2016-09-06

    A nuclear fuel pellet with a porous substrate, such as a carbon or tungsten aerogel, on which at least one layer of a fuel containing material is deposited via atomic layer deposition, and wherein the layer deposition is controlled to prevent agglomeration of defects. Further, a method of fabricating a nuclear fuel pellet, wherein the method features the steps of selecting a porous substrate, depositing at least one layer of a fuel containing material, and terminating the deposition when the desired porosity is achieved. Also provided is a nuclear reactor fuel cladding made of a porous substrate, such as silicon carbide aerogel or silicon carbide cloth, upon which layers of silicon carbide are deposited.

  13. Nuclear reactor pressure vessel support system

    DOEpatents

    Sepelak, George R.

    1978-01-01

    A support system for nuclear reactor pressure vessels which can withstand all possible combinations of stresses caused by a postulated core disrupting accident during reactor operation. The nuclear reactor pressure vessel is provided with a flange around the upper periphery thereof, and the flange includes an annular vertical extension formed integral therewith. A support ring is positioned atop of the support ledge and the flange vertical extension, and is bolted to both members. The plug riser is secured to the flange vertical extension and to the top of a radially outwardly extension of the rotatable plug. This system eliminates one joint through which fluids contained in the vessel could escape by making the fluid flow path through the joint between the flange and the support ring follow the same path through which fluid could escape through the plug risers. In this manner, the sealing means to prohibit the escape of contained fluids through the plug risers can also prohibit the escape of contained fluid through the securing joint.

  14. Underground reactor containments: An option for the future?

    SciTech Connect

    Forsberg, C.W.; Kress, T.

    1997-02-18

    Changing world conditions and changing technologies suggest that serious consideration should be given to siting of nuclear power plants underground. Underground siting is not a new concept. Multiple research reactors, several weapons production reactors, and one power reactor have been built underground. What is new are the technologies and incentives that may now make underground siting a preferred option. The conditions and technologies, along with their implications, are discussed herein. Underground containments can be constructed in mined cavities or pits that are then backfilled with thick layers of rock and soil. Conventional above-ground containments resist assaults and accidents because of the strength of their construction materials and the effectiveness of their safety features that are engineered to reduce loads. However, underground containments can provide even more resistance to assaults and accidents because of the inertia of the mass of materials over the reactor. High-technology weapons or some internal accidents can cause existing strong-material containments to fail, but only very-high energy releases can move large inertial masses associated with underground containments. New methods of isolation may provide a higher confidence in isolation that is independent of operator action.

  15. METHOD AND APPARATUS FOR CONTROL OF A NUCLEAR REACTOR

    DOEpatents

    Cawley, W.E.

    1962-12-11

    A method and apparatus are described for controlling an overmoderated nuclear reactor containing columns of fuel elements aligned in a plurality of coolant tubes in a stream of coolant water. The invention includes means for adjusting the distance between halves of the fuel element column to vary the relative proportion of fuel and moderator at the center of the reactor. (AEC)

  16. Nuclear Reactor Safety: a current awareness bulletin

    SciTech Connect

    Cunningham, D.C.

    1985-01-15

    Nuclear Reactor Safety announces on a semimonthly basis the current worldwide information available on all safety-related aspects of fission reactors, including: accident analysis, safety systems, radiation protection, decommissioning and dismantling, and security measures.

  17. Reactivity control assembly for nuclear reactor

    DOEpatents

    Bollinger, Lawrence R.

    1984-01-01

    Reactivity control assembly for nuclear reactor comprises supports stacked above reactor core for holding control rods. Couplers associated with the supports and a vertically movable drive shaft have lugs at their lower ends for engagement with the supports.

  18. The Angra Project: Monitoring Nuclear Reactors with Antineutrino Detectors

    NASA Astrophysics Data System (ADS)

    Anjos, J. C.; Barbosa, A. F.; Bezerra, T. J. C.; Chimenti, P.; Gonzalez, L. F. G.; Kemp, E.; de Oliveira, M. A. Leigui; Lima, H. P.; Lima, R. M.; Nunokawa, H.

    2010-03-01

    We present the status of the Angra Neutrino project, describing the development of an antineutrino detector aimed at monitoring nuclear reactor activity. The experiment will take place at the Brazilian nuclear power plant located in Angra dos Reis. The Angra II reactor, with 4 GW of thermal power, will be used as a source of antineutrinos. A water Cherenkov detector will be placed above ground in a commercial container outside the reactor containment, about 30 m from the reactor core. With a detector of one ton scale a few thousand antineutrino interactions per day are expected. We intend, in a first step, to use the measured neutrino event rate to monitor the on—off status and the thermal power delivered by the reactor. In addition to the safeguards issues the project will provide an alternative tool to have an independent measurement of the reactor power.

  19. Models of iodine behavior in reactor containments

    SciTech Connect

    Weber, C.F.; Beahm, E.C.; Kress, T.S.

    1992-10-01

    Models are developed for many phenomena of interest concerning iodine behavior in reactor containments during severe accidents. Processes include speciation in both gas and liquid phases, reactions with surfaces, airborne aerosols, and other materials, and gas-liquid interface behavior. Although some models are largely empirical formulations, every effort has been made to construct mechanistic and rigorous descriptions of relevant chemical processes. All are based on actual experimental data generated at the Oak Ridge National Laboratory (ORNL) or elsewhere, and, hence, considerable data evaluation and parameter estimation are contained in this study. No application or encoding is attempted, but each model is stated in terms of rate processes, with the intention of allowing mechanistic simulation. Taken together, this collection of models represents a best estimate iodine behavior and transport in reactor accidents.

  20. STEAM STIRRED HOMOGENEOUS NUCLEAR REACTOR

    DOEpatents

    Busey, H.M.

    1958-06-01

    A homogeneous nuclear reactor utilizing a selfcirculating liquid fuel is described. The reactor vessel is in the form of a vertically disposed tubular member having the lower end closed by the tube walls and the upper end closed by a removal fianged assembly. A spherical reaction shell is located in the lower end of the vessel and spaced from the inside walls. The reaction shell is perforated on its lower surface and is provided with a bundle of small-diameter tubes extending vertically upward from its top central portion. The reactor vessel is surrounded in the region of the reaction shell by a neutron reflector. The liquid fuel, which may be a solution of enriched uranyl sulfate in ordinary or heavy water, is mainiained at a level within the reactor vessel of approximately the top of the tubes. The heat of the reaction which is created in the critical region within the spherical reaction shell forms steam bubbles which more upwardly through the tubes. The upward movement of these bubbles results in the forcing of the liquid fuel out of the top of these tubes, from where the fuel passes downwardly in the space between the tubes and the vessel wall where it is cooled by heat exchangers. The fuel then re-enters the critical region in the reaction shell through the perforations in the bottom. The upper portion of the reactor vessel is provided with baffles to prevent the liquid fuel from splashing into this region which is also provided with a recombiner apparatus for recombining the radiolytically dissociated moderator vapor and a control means.

  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. 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.

  3. Simulated nuclear reactor fuel assembly

    DOEpatents

    Berta, V.T.

    1993-04-06

    An apparatus for electrically simulating a nuclear reactor fuel assembly. It includes a heater assembly having a top end and a bottom end and a plurality of concentric heater tubes having electrical circuitry connected to a power source, and radially spaced from each other. An outer target tube and an inner target tube is concentric with the heater tubes and with each other, and the outer target tube surrounds and is radially spaced from the heater tubes. The inner target tube is surrounded by and radially spaced from the heater tubes and outer target tube. The top of the assembly is generally open to allow for the electrical power connection to the heater tubes, and the bottom of the assembly includes means for completing the electrical circuitry in the heater tubes to provide electrical resistance heating to simulate the power profile in a nuclear reactor. The embedded conductor elements in each heater tube is split into two halves for a substantial portion of its length and provided with electrical isolation such that each half of the conductor is joined at one end and is not joined at the other end.

  4. Simulated nuclear reactor fuel assembly

    DOEpatents

    Berta, Victor T.

    1993-01-01

    An apparatus for electrically simulating a nuclear reactor fuel assembly. It includes a heater assembly having a top end and a bottom end and a plurality of concentric heater tubes having electrical circuitry connected to a power source, and radially spaced from each other. An outer target tube and an inner target tube is concentric with the heater tubes and with each other, and the outer target tube surrounds and is radially spaced from the heater tubes. The inner target tube is surrounded by and radially spaced from the heater tubes and outer target tube. The top of the assembly is generally open to allow for the electrical power connection to the heater tubes, and the bottom of the assembly includes means for completing the electrical circuitry in the heater tubes to provide electrical resistance heating to simulate the power profile in a nuclear reactor. The embedded conductor elements in each heater tube is split into two halves for a substantial portion of its length and provided with electrical isolation such that each half of the conductor is joined at one end and is not joined at the other end.

  5. Licensed reactor nuclear safety criteria applicable to DOE reactors

    SciTech Connect

    Not Available

    1991-04-01

    The Department of Energy (DOE) Order DOE 5480.6, Safety of Department of Energy-Owned Nuclear Reactors, establishes reactor safety requirements to assure that reactors are sited, designed, constructed, modified, operated, maintained, and decommissioned in a manner that adequately protects health and safety and is in accordance with uniform standards, guides, and codes which are consistent with those applied to comparable licensed reactors. This document identifies nuclear safety criteria applied to NRC (Nuclear Regulatory Commission) licensed reactors. The titles of the chapters and sections of USNRC Regulatory Guide 1.70, Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants, Rev. 3, are used as the format for compiling the NRC criteria applied to the various areas of nuclear safety addressed in a safety analysis report for a nuclear reactor. In each section the criteria are compiled in four groups: (1) Code of Federal Regulations, (2) US NRC Regulatory Guides, SRP Branch Technical Positions and Appendices, (3) Codes and Standards, and (4) Supplemental Information. The degree of application of these criteria to a DOE-owned reactor, consistent with their application to comparable licensed reactors, must be determined by the DOE and DOE contractor.

  6. Licensed reactor nuclear safety criteria applicable to DOE reactors

    SciTech Connect

    Not Available

    1993-11-01

    This document is a compilation and source list of nuclear safety criteria that the Nuclear Regulatory Commission (NRC) applies to licensed reactors; it can be used by DOE and DOE contractors to identify NRC criteria to be evaluated for application to the DOE reactors under their cognizance. The criteria listed are those that are applied to the areas of nuclear safety addressed in the safety analysis report of a licensed reactor. They are derived from federal regulations, USNRC regulatory guides, Standard Review Plan (SRP) branch technical positions and appendices, and industry codes and standards.

  7. 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

  8. Nuclear reactors built, being built, or planned 1996

    SciTech Connect

    1997-08-01

    This publication contains unclassified information about facilities, built, being built, or planned in the United States for domestic use or export as of December 31, 1996. The Office of Scientific and Technical Information, U.S. Department of Energy, gathers this information annually from Washington headquarters, and field offices of DOE; from the U.S. Nuclear Regulatory Commission (NRC); from the U. S. reactor manufacturers who are the principal nuclear contractors for foreign reactor locations; from U.S. and foreign embassies; and from foreign governmental nuclear departments. The book consists of three divisions, as follows: (1) a commercial reactor locator map and tables of the characteristic and statistical data that follow; a table of abbreviations; (2) tables of data for reactors operating, being built, or planned; and (3) tables of data for reactors that have been shut down permanently or dismantled.

  9. Non-equilibrium radiation nuclear reactor

    NASA Technical Reports Server (NTRS)

    Thom, K.; Schneider, R. T. (Inventor)

    1978-01-01

    An externally moderated thermal nuclear reactor is disclosed which is designed to provide output power in the form of electromagnetic radiation. The reactor is a gaseous fueled nuclear cavity reactor device which can operate over wide ranges of temperature and pressure, and which includes the capability of processing and recycling waste products such as long-lived transuranium actinides. The primary output of the device may be in the form of coherent radiation, so that the reactor may be utilized as a self-critical nuclear pumped laser.

  10. Nuclear reactor control room construction

    DOEpatents

    Lamuro, Robert C.; Orr, Richard

    1993-01-01

    A control room 10 for a nuclear plant is disclosed. In the control room, objects 12, 20, 22, 26, 30 are no less than four inches from walls 10.2. A ceiling 32 contains cooling fins 35 that extend downwards toward the floor from metal plates 34. A concrete slab 33 is poured over the plates. Studs 36 are welded to the plates and are encased in the concrete.

  11. Nuclear reactor control room construction

    DOEpatents

    Lamuro, R.C.; Orr, R.

    1993-11-16

    A control room for a nuclear plant is disclosed. In the control room, objects labelled 12, 20, 22, 26, 30 in the drawing are no less than four inches from walls labelled 10.2. A ceiling contains cooling fins that extend downwards toward the floor from metal plates. A concrete slab is poured over the plates. Studs are welded to the plates and are encased in the concrete. 6 figures.

  12. Dielectric Heaters for Testing Spacecraft Nuclear Reactors

    NASA Technical Reports Server (NTRS)

    Sims, William Herbert; Bitteker, Leo; Godfroy, Thomas

    2006-01-01

    A document proposes the development of radio-frequency-(RF)-driven dielectric heaters for non-nuclear thermal testing of the cores of nuclear-fission reactors for spacecraft. Like the electrical-resistance heaters used heretofore for such testing, the dielectric heaters would be inserted in the reactors in place of nuclear fuel rods. A typical heater according to the proposal would consist of a rod of lossy dielectric material sized and shaped like a fuel rod and containing an electrically conductive rod along its center line. Exploiting the dielectric loss mechanism that is usually considered a nuisance in other applications, an RF signal, typically at a frequency .50 MHz and an amplitude between 2 and 5 kV, would be applied to the central conductor to heat the dielectric material. The main advantage of the proposal is that the wiring needed for the RF dielectric heating would be simpler and easier to fabricate than is the wiring needed for resistance heating. In some applications, it might be possible to eliminate all heater wiring and, instead, beam the RF heating power into the dielectric rods from external antennas.

  13. Nuclear reactor internals alignment configuration

    DOEpatents

    Gilmore, Charles B.; Singleton, Norman R.

    2009-11-10

    An alignment system that employs jacking block assemblies and alignment posts around the periphery of the top plate of a nuclear reactor lower internals core shroud to align an upper core plate with the lower internals and the core shroud with the core barrel. The distal ends of the alignment posts are chamfered and are closely received within notches machined in the upper core plate at spaced locations around the outer circumference of the upper core plate. The jacking block assemblies are used to center the core shroud in the core barrel and the alignment posts assure the proper orientation of the upper core plate. The alignment posts may alternately be formed in the upper core plate and the notches may be formed in top plate.

  14. 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)

  15. 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.

  16. Nuclear reactor loss of coolant protection system

    SciTech Connect

    Loose, R.A.

    1986-03-18

    A pressurized water reactor system is described of a nuclear power plant having a water storage tank for providing emergency coolant water and means provided external to the containment vessel, for use in the event of a primary loss of coolant situation, to circulate emergency water as a coolant by withdrawal through a wall of the containment vessel and return the same back through the wall of the containment vessel and passing the water through a heat exchange means prior to use as a coolant for the reactor core. The improvement described here consists of: an enslosure, the interior of which is sealed to the atmosphere, positioned adjacent to and exterior of a wall of the containment vessel; an inlet conduit, enclosed within a sealed outer casing, communicating between the interior of the containment vessel and the interior of the enclosure; an exhaust conduit, enclosed within a sealed outer casing, communicating between the interior of the enclosure and the interior of the containment vessel; a rupture disc on the inlet conduit within the enclosure, such that failure of the exhaust conduit within the enclosure will produce an increase of the pressure within the enclosure and above a predetermined pressure will fracture the rupture disc, and will circulate the coolant within the enclosure; and means within the interior of the enclosure for pumping coolant from the interior of the containment vessel through the inlet conduit, and back to the interior of the containment vessel through the exhaust conduit; whereby if either of the conduits should fail, coolant will be collected within the enclosure and sealed to the atmosphere.

  17. Liquid metal cooled nuclear reactor plant system

    DOEpatents

    Hunsbedt, Anstein; Boardman, Charles E.

    1993-01-01

    A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting for fuel decay during reactor shutdown, or heat produced during a mishap. The reactor system is enhanced with sealing means for excluding external air from contact with the liquid metal coolant leaking from the reactor vessel during an accident. The invention also includes a silo structure which resists attack by leaking liquid metal coolant, and an added unique cooling means.

  18. Daddy, What's a Nuclear Reactor?

    SciTech Connect

    Reisenweaver, Dennis W.

    2008-01-15

    No matter what we think of the nuclear industry, it is part of mankind's heritage. The decommissioning process is slowly making facilities associated with this industry disappear and not enough is being done to preserve the information for future generations. This paper provides some food for thought and provides a possible way forward. Industrial archaeology is an ever expanding branch of archaeology that is dedicated to preserving, interpreting and documenting our industrial past and heritage. Normally it begins with analyzing an old building or ruins and trying to determine what was done, how it was done and what changes might have occurred during its operation. We have a unique opportunity to document all of these issues and provide them before the nuclear facility disappears. Entombment is an acceptable decommissioning strategy; however we would have to change our concept of entombment. It is proposed that a number of nuclear facilities be entombed or preserved for future generations to appreciate. This would include a number of different types of facilities such as different types of nuclear power and research reactors, a reprocessing plant, part of an enrichment plant and a fuel manufacturing plant. One of the main issues that would require resolution would be that of maintaining information of the location of the buried facility and the information about its operation and structure, and passing this information on to future generations. This can be done, but a system would have to be established prior to burial of the facility so that no information would be lost. In general, our current set of requirements and laws may need to be re-examined and modified to take into account these new situations. As an alternative, and to compliment the above proposal, it is recommended that a study and documentation of the nuclear industry be considered as part of twentieth century industrial archaeology. This study should not only include the power and fuel cycle

  19. Nuclear reactors built, being built, or planned, 1994

    SciTech Connect

    1995-07-01

    This document contains unclassified information about facilities built, being built, or planned in the United States for domestic use or export as of December 31, 1994. The Office of Scientific and Technical Information, US Department of Energy, gathers this information annually from Washington headquarters and field offices of DOE; from the US Nuclear Regulatory Commission (NRC); from the US reactor manufacturers who are the principal nuclear contractors for foreign reactor locations; from US and foreign embassies; and from foreign governmental nuclear departments. The book consists of three divisions, as follows: a commercial reactor locator map and tables of the characteristic and statistical data that follow; a table of abbreviations; tables of data for reactors operating, being built, or planned; and tables of data for reactors that have been shut down permanently or dismantled. The reactors are subdivided into the following parts: Civilian, Production, Military, Export, and Critical Assembly. Export reactor refers to a reactor for which the principal nuclear contractor is a US company -- working either independently or in cooperation with a foreign company (Part 4). Critical assembly refers to an assembly of fuel and moderator that requires an external source of neutrons to initiate and maintain fission. A critical assembly is used for experimental measurements (Part 5).

  20. Nuclear reactors built, being built, or planned: 1995

    SciTech Connect

    1996-08-01

    This report contains unclassified information about facilities built, being built, or planned in the US for domestic use or export as of December 31, 1995. The Office of Scientific and Technical Information, US Department of Energy, gathers this information annually from Washington headquarters and field offices of DOE; from the US Nuclear Regulatory Commission (NRC); from the US reactor manufacturers who are the principal nuclear contractors for foreign reactor locations; from US and foreign embassies; and from foreign governmental nuclear departments. The book consists of three divisions, as follows: (1) a commercial reactor locator map and tables of the characteristic and statistical data that follow; a table of abbreviations; (2) tables of data for reactors operating, being built, or planned; and (3) tables of data for reactors that have been shut down permanently or dismantled. The reactors are subdivided into the following parts: Civilian, Production, Military, Export, and Critical Assembly. Export reactor refers to a reactor for which the principal nuclear contractor is a US company--working either independently or in cooperation with a foreign company (Part 4). Critical assembly refers to an assembly of fuel and moderator that requires an external source of neutrons to initiate and maintain fission. A critical assembly is used for experimental measurements (Part 5).

  1. 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)

  2. Post impact behavior of mobile reactor core containment systems

    NASA Technical Reports Server (NTRS)

    Puthoff, R. L.; Parker, W. G.; Vanbibber, L. E.

    1972-01-01

    The reactor core containment vessel temperatures after impact, and the design variables that affect the post impact survival of the system are analyzed. The heat transfer analysis includes conduction, radiation, and convection in addition to the core material heats of fusion and vaporization under partially burial conditions. Also, included is the fact that fission products vaporize and transport radially outward and condense outward and condense on cooler surfaces, resulting in a moving heat source. A computer program entitled Executive Subroutines for Afterheat Temperature Analysis (ESATA) was written to consider this complex heat transfer analysis. Seven cases were calculated of a reactor power system capable of delivering up to 300 MW of thermal power to a nuclear airplane.

  3. Gas-cooled nuclear reactor

    DOEpatents

    Peinado, Charles O.; Koutz, Stanley L.

    1985-01-01

    A gas-cooled nuclear reactor includes a central core located in the lower portion of a prestressed concrete reactor vessel. Primary coolant gas flows upward through the core and into four overlying heat-exchangers wherein stream is generated. During normal operation, the return flow of coolant is between the core and the vessel sidewall to a pair of motor-driven circulators located at about the bottom of the concrete pressure vessel. The circulators repressurize the gas coolant and return it back to the core through passageways in the underlying core structure. If during emergency conditions the primary circulators are no longer functioning, the decay heat is effectively removed from the core by means of natural convection circulation. The hot gas rising through the core exits the top of the shroud of the heat-exchangers and flows radially outward to the sidewall of the concrete pressure vessel. A metal liner covers the entire inside concrete surfaces of the concrete pressure vessel, and cooling tubes are welded to the exterior or concrete side of the metal liner. The gas coolant is in direct contact with the interior surface of the metal liner and transfers its heat through the metal liner to the liquid coolant flowing through the cooling tubes. The cooler gas is more dense and creates a downward convection flow in the region between the core and the sidewall until it reaches the bottom of the concrete pressure vessel when it flows radially inward and up into the core for another pass. Water is forced to flow through the cooling tubes to absorb heat from the core at a sufficient rate to remove enough of the decay heat created in the core to prevent overheating of the core or the vessel.

  4. Neutron transport analysis for nuclear reactor design

    DOEpatents

    Vujic, J.L.

    1993-11-30

    Replacing regular mesh-dependent ray tracing modules in a collision/transfer probability (CTP) code with a ray tracing module based upon combinatorial geometry of a modified geometrical module (GMC) provides a general geometry transfer theory code in two dimensions (2D) for analyzing nuclear reactor design and control. The primary modification of the GMC module involves generation of a fixed inner frame and a rotating outer frame, where the inner frame contains all reactor regions of interest, e.g., part of a reactor assembly, an assembly, or several assemblies, and the outer frame, with a set of parallel equidistant rays (lines) attached to it, rotates around the inner frame. The modified GMC module allows for determining for each parallel ray (line), the intersections with zone boundaries, the path length between the intersections, the total number of zones on a track, the zone and medium numbers, and the intersections with the outer surface, which parameters may be used in the CTP code to calculate collision/transfer probability and cross-section values. 28 figures.

  5. Neutron transport analysis for nuclear reactor design

    DOEpatents

    Vujic, Jasmina L.

    1993-01-01

    Replacing regular mesh-dependent ray tracing modules in a collision/transfer probability (CTP) code with a ray tracing module based upon combinatorial geometry of a modified geometrical module (GMC) provides a general geometry transfer theory code in two dimensions (2D) for analyzing nuclear reactor design and control. The primary modification of the GMC module involves generation of a fixed inner frame and a rotating outer frame, where the inner frame contains all reactor regions of interest, e.g., part of a reactor assembly, an assembly, or several assemblies, and the outer frame, with a set of parallel equidistant rays (lines) attached to it, rotates around the inner frame. The modified GMC module allows for determining for each parallel ray (line), the intersections with zone boundaries, the path length between the intersections, the total number of zones on a track, the zone and medium numbers, and the intersections with the outer surface, which parameters may be used in the CTP code to calculate collision/transfer probability and cross-section values.

  6. 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.

  7. Nuclear Archeology for CANDU Power Reactors

    SciTech Connect

    Broadhead, Bryan L

    2011-01-01

    The goal of this work is the development of so-called 'nuclear archeology' techniques to predict the irradiation history of both fuel-related and non-fuel-related materials irradiated in the CANDU (CANada Deuterium Uranium) family of nuclear reactors. In this application to CANDU-type reactors, two different scenarios for the collection of the appropriate data for use in these procedures will be assumed: the first scenario is the removal of the pressure tubes, calandria tubes, or fuel cladding and destructive analysis of the activation products contained in these structural materials; the second scenario is the nondestructive analysis (NDA) of the same hardware items via high-resolution gamma ray scans. There are obvious advantages and disadvantages for each approach; however, the NDA approach is the central focus of this work because of its simplicity and lack of invasiveness. The use of these techniques along with a previously developed inverse capability is expected to allow for the prediction of average flux levels and irradiation time, and the total fluence for samples where the values of selected isotopes can be measured.

  8. Fast-acting nuclear reactor control device

    DOEpatents

    Kotlyar, Oleg M.; West, Phillip B.

    1993-01-01

    A fast-acting nuclear reactor control device for moving and positioning a fety control rod to desired positions within the core of the reactor between a run position in which the safety control rod is outside the reactor core, and a shutdown position in which the rod is fully inserted in the reactor core. The device employs a hydraulic pump/motor, an electric gear motor, and solenoid valve to drive the safety control rod into the reactor core through the entire stroke of the safety control rod. An overrunning clutch allows the safety control rod to freely travel toward a safe position in the event of a partial drive system failure.

  9. Nuclear propulsion apparatus with alternate reactor segments

    DOEpatents

    Szekely, Thomas

    1979-04-03

    1. Nuclear propulsion apparatus comprising: A. means for compressing incoming air; B. nuclear fission reactor means for heating said air; C. means for expanding a portion of the heated air to drive said compressing means; D. said nuclear fission reactor means being divided into a plurality of radially extending segments; E. means for directing a portion of the compressed air for heating through alternate segments of said reactor means and another portion of the compressed air for heating through the remaining segments of said reactor means; and F. means for further expanding the heated air from said drive means and the remaining heated air from said reactor means through nozzle means to effect reactive thrust on said apparatus.

  10. Nuclear reactor cooling system decontamination reagent regeneration

    DOEpatents

    Anstine, Larry D.; James, Dean B.; Melaika, Edward A.; Peterson, Jr., John P.

    1985-01-01

    An improved method for decontaminating the coolant system of water-cooled nuclear power reactors and for regenerating the decontamination solution. A small amount of one or more weak-acid organic complexing agents is added to the reactor coolant, and the pH is adjusted to form a decontamination solution which is circulated throughout the coolant system to dissolve metal oxides from the interior surfaces and complex the resulting metal ions and radionuclide ions. The coolant containing the complexed metal ions and radionuclide ions is passed through a strong-base anion exchange resin bed which has been presaturated with a solution containing the complexing agents in the same ratio and having the same pH as the decontamination solution. As the decontamination solution passes through the resin bed, metal-complexed anions are exchanged for the metal-ion-free anions on the bed, while metal-ion-free anions in the solution pass through the bed, thus removing the metal ions and regenerating the decontamination solution.

  11. NUCLEAR DATABASES FOR REACTOR APPLICATIONS.

    SciTech Connect

    PRITYCHENKO, B.; ARCILLA, R.; BURROWS, T.; HERMAN, M.W.; MUGHABGHAB, S.; OBLOZINSKY, P.; ROCHMAN, D.; SONZOGNI, A.A.; TULI, J.; WINCHELL, D.F.

    2006-06-05

    The National Nuclear Data Center (NNDC): An overview of nuclear databases, related products, nuclear data Web services and publications. The NNDC collects, evaluates, and disseminates nuclear physics data for basic research and applied nuclear technologies. The NNDC maintains and contributes to the nuclear reaction (ENDF, CSISRS) and nuclear structure databases along with several others databases (CapGam, MIRD, IRDF-2002) and provides coordination for the Cross Section Evaluation Working Group (CSEWG) and the US Nuclear Data Program (USNDP). The Center produces several publications and codes such as Atlas of Neutron Resonances, Nuclear Wallet Cards booklets and develops codes, such as nuclear reaction model code Empire.

  12. 10 CFR 50.44 - Combustible gas control for nuclear power reactors.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 10 Energy 1 2010-01-01 2010-01-01 false Combustible gas control for nuclear power reactors. 50.44 Section 50.44 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION... for nuclear power reactors. (a) Definitions—(1) Inerted atmosphere means a containment atmosphere...

  13. 10 CFR 50.44 - Combustible gas control for nuclear power reactors.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 10 Energy 1 2011-01-01 2011-01-01 false Combustible gas control for nuclear power reactors. 50.44 Section 50.44 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION... for nuclear power reactors. (a) Definitions—(1) Inerted atmosphere means a containment atmosphere...

  14. Removal of hydrogen bubbles from nuclear reactors

    NASA Technical Reports Server (NTRS)

    Jenkins, R. V.

    1980-01-01

    Method proposed for removing large hydrogen bubbles from nuclear environment uses, in its simplest form, hollow spheres of palladium or platinum. Methods would result in hydrogen bubble being reduced in size without letting more radioactivity outside reactor.

  15. Nuclear reactor shield including magnesium oxide

    DOEpatents

    Rouse, Carl A.; Simnad, Massoud T.

    1981-01-01

    An improvement in nuclear reactor shielding of a type used in reactor applications involving significant amounts of fast neutron flux, the reactor shielding including means providing structural support, neutron moderator material, neutron absorber material and other components as described below, wherein at least a portion of the neutron moderator material is magnesium in the form of magnesium oxide either alone or in combination with other moderator materials such as graphite and iron.

  16. Fission control system for nuclear reactor

    DOEpatents

    Conley, G.H.; Estes, G.P.

    Control system for nuclear reactor comprises a first set of reactivity modifying rods fixed in a reactor core with their upper ends stepped in height across the core, and a second set of reactivity modifying rods movable vertically within the reactor core and having their lower ends stepped to correspond with the stepped arrangement of the first set of rods, pairs of the rods of the first and second sets being in coaxial alignment.

  17. Reactivity Transients in Nuclear Research Reactors

    Energy Science and Technology Software Center (ESTSC)

    2015-01-01

    Version 01 AIREMOD-RR is a point kinetics code which can simulate fast transients in nuclear research reactor cores. It can also be used for theoretical reactor dynamics studies. It is used for research reactor kinetic analysis and provides a point neutron kinetic capability. The thermal hydraulic behavior is governed by a one-dimensional heat balance equation. The calculations are restricted to a single equivalent unit cell which consists of fuel, clad and coolant.

  18. 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.

  19. MODERATOR ELEMENTS FOR UNIFORM POWER NUCLEAR REACTOR

    DOEpatents

    Balent, R.

    1963-03-12

    This patent describes a method of obtaining a flatter flux and more uniform power generation across the core of a nuclear reactor. The method comprises using moderator elements having differing moderating strength. The elements have an increasing amount of the better moderating material as a function of radial and/or axial distance from the reactor core center. (AEC)

  20. Nuclear electric propulsion reactor control systems status

    NASA Technical Reports Server (NTRS)

    Ferg, D. A.

    1973-01-01

    The thermionic reactor control system design studies conducted over the past several years for a nuclear electric propulsion system are described and summarized. The relevant reactor control system studies are discussed in qualitative terms, pointing out the significant advantages and disadvantages including the impact that the various control systems would have on the nuclear electric propulsion system design. A recommendation for the reference control system is made, and a program for future work leading to an engineering model is described.

  1. Nuclear reactor for breeding U.sup.233

    DOEpatents

    Bohanan, Charles S.; Jones, David H.; Raab, Jr., Harry F.; Radkowsky, Alvin

    1976-01-01

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

  2. 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.

  3. Post impact behavior of mobile reactor core containment systems.

    NASA Technical Reports Server (NTRS)

    Puthoff, R. L.; Parker, W. G.; Van Bibber, L. E.

    1972-01-01

    In the future, nuclear assemblies containing fission products will be transported at high speeds. An example is a reactor supplying power to a large subsonic airplane. In this case an accident can occur resulting in a ground impact at speeds up to 1000 ft/sec. This paper analyzes the containment vessel temperatures after impact and attempts to understand the design variables that affect the post impact survival of the system. The heat transfer analysis includes conduction, radiation, and convection in addition to the core material heats of fusion and vaporization under partial-burial conditions. Also, included is the fact that fission products vaporize and transport radially outward and condense on cooler surfaces, resulting in a moving heat source.

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

    DOEpatents

    Roake, W.E.; Evans, E.A.; Brite, D.W.

    1960-06-21

    A method of preparing a fuel element for a nuclear reactor is given in which an internally and externally cooled fuel element consisting of two coaxial tubes having a plurality of integral radial ribs extending between the tubes and containing a powdered fuel material is isostatically pressed to form external coolant channels and compact the powder simultaneously.

  5. Nuclear Power from Fission Reactors. An Introduction.

    ERIC Educational Resources Information Center

    Department of Energy, Washington, DC. Technical Information Center.

    The purpose of this booklet is to provide a basic understanding of nuclear fission energy and different fission reaction concepts. Topics discussed are: energy use and production, current uses of fuels, oil and gas consumption, alternative energy sources, fossil fuel plants, nuclear plants, boiling water and pressurized water reactors, the light…

  6. 78 FR 71675 - Update of the Office of Nuclear Reactor Regulation's Electronic Operating Reactor Correspondence

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-11-29

    ... COMMISSION Update of the Office of Nuclear Reactor Regulation's Electronic Operating Reactor Correspondence... public of a slight change in the manner of distribution of publicly available operating reactor licensing... Division of Operating Reactor Licensing began transmitting correspondence to addressees and...

  7. Fission product scrubbing system for a nuclear reactor

    SciTech Connect

    Leach, D.S.

    1986-09-09

    A fission product scrubbing system is described for a nuclear reactor including a containment building defining a containment space for accommodating reactor components, comprising (a) means defining a water tank in the containment building; (b) a dividing wall extending into the water tank for separating the water tank into a first and a second compartment; (c) means defining a collection plenum normally hermetically sealed from the containment space and the environment externally of the containment building; (d) means defining a communication passage in the dividing wall underneath the water level in the first and second compartments for maintaining communication between the water stored in the first and second compartments; (e) a standpipe extending from the containment space into the second compartment; (f) a vent pipe extending from the collection plenum into the environment externally of the containment building; and (g) a rupture disc mounted in the vent pipe for normally blocking communication between the collection plenum and the environment.

  8. Completely automated nuclear reactors for long-term operation

    SciTech Connect

    Teller, E.; Ishikawa, M.; Wood, L.

    1996-01-01

    The authors discuss new types of nuclear fission reactors optimized for the generation of high-temperature heat for exceedingly safe, economic, and long-duration electricity production in large, long-lived central power stations. These reactors are quite different in design, implementation and operation from conventional light-water-cooled and -moderated reactors (LWRs) currently in widespread use, which were scaled-up from submarine nuclear propulsion reactors. They feature an inexpensive initial fuel loading which lasts the entire 30-year design life of the power-plant. The reactor contains a core comprised of a nuclear ignitor and a nuclear burn-wave propagating region comprised of natural thorium or uranium, a pressure shell for coolant transport purposes, and automatic emergency heat-dumping means to obviate concerns regarding loss-of-coolant accidents during the plant`s operational and post-operational life. These reactors are proposed to be situated in suitable environments at {approximately}100 meter depths underground, and their operation is completely automatic, with no moving parts and no human access during or after its operational lifetime, in order to avoid both error and misuse. The power plant`s heat engine and electrical generator subsystems are located above-ground.

  9. Thermionic reactors for space nuclear power

    NASA Technical Reports Server (NTRS)

    Homeyer, W. G.; Merrill, M. H.; Holland, J. W.; Fisher, C. R.; Allen, D. T.

    1985-01-01

    Thermionic reactor designs for a variety of space power applications spanning the range from 5 kWe to 3 MWe are described. In all of these reactors, nuclear heat is converted directly to electrical energy in thermionic fuel elements (TFEs). A circulating reactor coolant carries heat from the core of TFEs directly to a heat rejection radiator system. The recent design of a thermionic reactor to meet the SP-100 requirements is emphasized. Design studies of reactors at other power levels show that the same TFE can be used over a broad range in power, and that design modifications can extend the range to many megawatts. The design of the SP-100 TFE is similar to that of TFEs operated successfully in test reactors, but with design improvements to extend the operating lifetime to seven years.

  10. STEAM GENERATOR FOR NUCLEAR REACTOR

    DOEpatents

    Kinyon, B.W.; Whitman, G.D.

    1963-07-16

    The steam generator described for use in reactor powergenerating systems employs a series of concentric tubes providing annular passage of steam and water and includes a unique arrangement for separating the steam from the water. (AEC)

  11. Erosion-corrosion entrainment of iron-containing compounds as a source of deposits in steam generators used at nuclear power plants equipped with VVER reactors

    NASA Astrophysics Data System (ADS)

    Tomarov, G. V.; Shipkov, A. A.

    2011-03-01

    The main stages and processes through which deposits are generated, migrate, and precipitate in the metal-secondary coolant system of power units at nuclear power plants are analyzed and determined. It is shown that substances produced by the mechanism of general erosion-corrosion are the main source of the ionic-colloid form of iron, which is the main component of deposits in a steam generator. Ways for controlling the formation of deposits in a nuclear power plant's steam generator are proposed together with methods for estimating their efficiency.

  12. 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.

  13. Nuclear energy: salvation or suicide. [Contains glossary

    SciTech Connect

    Collins, C.C.

    1984-01-01

    A collection of 700 editorials and feature articles collected from 125 US newspapers addresses the dominant areas of concern about nuclear power: plant safety, radioactive wastes, proliferation, and cost. The editorial debates present the pros and cons of Three Mile Island and other accidents, ocean dumping, evacuation plans, radioactive waste transport and storage, nuclear fuel processing, the Karen Silkwood case, and breeder reactors. The appendix raises the question of the future for fission and the possibility of nuclear fusion as an alternative. There is a subject index and a glossary of basic terms.

  14. Space nuclear reactor shielding optimization studies

    NASA Astrophysics Data System (ADS)

    Jimenez, Richard D.; El-Genk, Mohamed S.

    The Institute for Space Nuclear Reactor Studies is investigating optimal techniques for shielding spacecraft (payload) electronics from the combined radiation effects of the SP-100 system nuclear reactor core and the natural space environment. The academic challenge of this research includes the investigation of the combined influences of radiation from the space environment and the radiations from the reactor power source. The technical application includes a series of shielding mass penalty tradeoffs for the SP-100 Program concept between the reactor core shield and the additional shielding of the spacecraft enclosure. These mass penalty tradeoffs are being conducted for several space flight orbits of future interest to the space military and civilian communities. It was shown that several potential mission orbits may pose environmental radiation dosages which are more severe than the SP-100 specification of core escape neutron and gamma ray particle fluences incident on the spacecraft.

  15. 10 CFR 1.43 - Office of Nuclear Reactor Regulation.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... 10 Energy 1 2013-01-01 2013-01-01 false Office of Nuclear Reactor Regulation. 1.43 Section 1.43 Energy NUCLEAR REGULATORY COMMISSION STATEMENT OF ORGANIZATION AND GENERAL INFORMATION Headquarters Program Offices § 1.43 Office of Nuclear Reactor Regulation. The Office of Nuclear Reactor Regulation— (a) Develops, promulgates and...

  16. Cooling system for a nuclear reactor

    DOEpatents

    Amtmann, Hans H.

    1982-01-01

    A cooling system for a gas-cooled nuclear reactor is disclosed which includes at least one primary cooling loop adapted to pass coolant gas from the reactor core and an associated steam generator through a duct system having a main circulator therein, and at least one auxiliary cooling loop having communication with the reactor core and adapted to selectively pass coolant gas through an auxiliary heat exchanger and circulator. The main and auxiliary circulators are installed in a common vertical cavity in the reactor vessel, and a common return duct communicates with the reactor core and intersects the common cavity at a junction at which is located a flow diverter valve operative to effect coolant flow through either the primary or auxiliary cooling loops.

  17. Nuclear Power Plant Containment Pressure Boundary Research

    SciTech Connect

    Cherry, J.L.; Chokshi, N.C.; Costello, J.F.; Ellingwood, B.R.; Naus, D.J.

    1999-09-15

    Research to address aging of the containment pressure boundary in light-water reactor plants is summarized. This research is aimed at understanding the significant factors relating occurrence of corrosion, efficacy of inspection, and structural capacity reduction of steel containment and liners of concrete containment. This understanding will lead to improvements in risk-informed regulatory decision making. Containment pressure boundary components are described and potential aging factors identified. Quantitative tools for condition assessments of aging structures to maintain an acceptable level of reliability over the service life of the plant are discussed. Finally, the impact of aging (i.e., loss of shell thickness due to corrosion) on steel containment fragility for a pressurized water reactor ice-condenser plant is presented.

  18. 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... 10 Energy 2 2010-01-01 2010-01-01 false Advance notification of shipment of irradiated reactor... notification of shipment of irradiated reactor fuel and nuclear waste. (a) As specified in paragraphs (b),...

  19. Cancer incidence among Finnish nuclear reactor workers.

    PubMed

    Auvinen, Anssi; Pukkala, Eero; Hyvönen, Hannu; Hakama, Matti; Rytömaa, Tapio

    2002-07-01

    Because of their well-documented exposures to repeated low doses of ionizing radiation, nuclear reactor workers offer an opportunity to assess cancer risk from low-dose radiation. A cohort of all 15,619 Finnish nuclear reactor workers was established through dose-monitoring records. A questionnaire survey revealed no substantial differences in consumption of tobacco or alcohol between different exposure groups nor between nuclear power company employees and contract workers. In the follow-up for cancer incidence, no clear excess in cancer incidence was observed overall, nor was any observed in any of the specific cancer types studied. There was little evidence for an association between cancer incidence and cumulative radiation dose, but the statistical power was limited. More precise estimates will be available from an international collaborative study of nuclear industry workers, including our cohort. PMID:12134527

  20. Nuclear reactor alignment plate configuration

    DOEpatents

    Altman, David A; Forsyth, David R; Smith, Richard E; Singleton, Norman R

    2014-01-28

    An alignment plate that is attached to a core barrel of a pressurized water reactor and fits within slots within a top plate of a lower core shroud and upper core plate to maintain lateral alignment of the reactor internals. The alignment plate is connected to the core barrel through two vertically-spaced dowel pins that extend from the outside surface of the core barrel through a reinforcement pad and into corresponding holes in the alignment plate. Additionally, threaded fasteners are inserted around the perimeter of the reinforcement pad and into the alignment plate to further secure the alignment plate to the core barrel. A fillet weld also is deposited around the perimeter of the reinforcement pad. To accomodate thermal growth between the alignment plate and the core barrel, a gap is left above, below and at both sides of one of the dowel pins in the alignment plate holes through with the dowel pins pass.

  1. Composite construction for nuclear fuel containers

    DOEpatents

    Cheng, Bo-Ching [Fremont, CA; Rosenbaum, Herman S [Fremont, CA; Armijo, Joseph S [Saratoga, CA

    1987-01-01

    An improved method for producing nuclear fuel containers of a composite construction having components providing therein a barrier system for resisting destructive action by volatile fission products or impurities and also interdiffusion of metal constituents, and the product thereof. The composite nuclear fuel containers of the method comprise a casing of zirconium or alloy thereof with a layer of copper overlying an oxidized surface portion of the zirconium or alloy thereof.

  2. Composite construction for nuclear fuel containers

    DOEpatents

    Cheng, B. C.; Rosenbaum, H. S.; Armijo, J. S.

    1987-04-21

    Disclosed is an improved method for producing nuclear fuel containers of a composite construction having components providing therein a barrier system for resisting destructive action by volatile fission products or impurities and also interdiffusion of metal constituents, and the product thereof. The composite nuclear fuel containers of the method comprise a casing of zirconium or alloy thereof with a layer of copper overlying an oxidized surface portion of the zirconium or alloy thereof. 1 fig.

  3. Nuclear Data and the Oklo Natural Nuclear Reactors

    NASA Astrophysics Data System (ADS)

    Gould, C. R.; Sharapov, E. I.; Sonzogni, A. A.

    2014-04-01

    Data from the Oklo natural nuclear reactors have enabled some of the most sensitive terrestrial tests of time variation of dimensionless fundamental constants. The constraints on variation of αEM, the fine structure constant are particular good, but depend on the reliability of the nuclear data, and on the reliability of the modeling of the reactor environment. We briefly review the history of these tests and discuss our recent work in 1) attempting to better bound the temperatures at which the reactors operated, 2) investigating whether the γ-ray fluxes in the reactors could have contributed to changing lutetium isotopic abundances and 3) determining whether lanthanum isotopic data could provide an alternate estimate of the neutron fluence.

  4. Technological Transfer from Research Nuclear Reactors to New Generation Nuclear Power Reactors

    SciTech Connect

    Radulescu, Laura; Pavelescu, Margarit

    2010-01-21

    The goal of this paper is the analysis of the technological transfer role in the nuclear field, with particular emphasis on nuclear reactors domain. The presentation is sustained by historical arguments. In this frame, it is very important to start with the achievements of the first nuclear systems, for instant those with natural uranium as fuel and heavy water as moderator, following in time through the history until the New Generation Nuclear Power Reactors.Starting with 1940, the accelerated development of the industry has implied the increase of the global demand for energy. In this respect, the nuclear energy could play an important role, being essentially an unlimited source of energy. However, the nuclear option faces the challenges of increasingly demanding safety requirements, economic competitiveness and public acceptance. Worldwide, a significant amount of experience has been accumulated during development, licensing, construction, and operation of nuclear power reactors. The experience gained is a strong basis for further improvements. Actually, the nuclear programs of many countries are addressing the development of advanced reactors, which are intended to have better economics, higher reliability, improved safety, and proliferation-resistant characteristics in order to overcome the current concerns about nuclear power. Advanced reactors, now under development, may help to meet the demand for energy power of both developed and developing countries as well as for district heating, desalination and for process heat.The paper gives historical examples that illustrate the steps pursued from first research nuclear reactors to present advanced power reactors. Emphasis was laid upon the fact that the progress is due to the great discoveries of the nuclear scientists using the technological transfer.

  5. The siting of UK nuclear reactors.

    PubMed

    Grimston, Malcolm; Nuttall, William J; Vaughan, Geoff

    2014-06-01

    Choosing a suitable site for a nuclear power station requires the consideration and balancing of several factors. Some 'physical' site characteristics, such as the local climate and the potential for seismic activity, will be generic to all reactors designs, while others, such as the availability of cooling water, the area of land required and geological conditions capable of sustaining the weight of the reactor and other buildings will to an extent be dependent on the particular design of reactor chosen (or alternatively the reactor design chosen may to an extent be dependent on the characteristics of an available site). However, one particularly interesting tension is a human and demographic one. On the one hand it is beneficial to place nuclear stations close to centres of population, to reduce transmission losses and other costs (including to the local environment) of transporting electricity over large distances from generator to consumer. On the other it is advantageous to place nuclear stations some distance away from such population centres in order to minimise the potential human consequences of a major release of radioactive materials in the (extremely unlikely) event of a major nuclear accident, not only in terms of direct exposure but also concerning the management of emergency planning, notably evacuation.This paper considers the emergence of policies aimed at managing this tension in the UK. In the first phase of nuclear development (roughly speaking 1945-1965) there was a highly cautious attitude, with installations being placed in remote rural locations with very low population density. The second phase (1965-1985) saw a more relaxed approach, allowing the development of AGR nuclear power stations (which with concrete pressure vessels were regarded as significantly safer) closer to population centres (in 'semi-urban' locations, notably at Hartlepool and Heysham). In the third phase (1985-2005) there was very little new nuclear development, Sizewell

  6. Nuclear reactor shutdown control rod assembly

    DOEpatents

    Bilibin, Konstantin

    1988-01-01

    A temperature responsive, self-actuated nuclear reactor shutdown control rod assembly 10. The upper end 18 of a lower drive line 17 fits within the lower end of an upper drive line 12. The lower end (not shown) of the lower drive line 17 is connected to a neutron absorber. During normal temperature conditions the lower drive line 17 is supported by detent means 22,26. When an overtemperature condition occurs thermal actuation means 34 urges ring 26 upwardly sufficiently to allow balls 22 to move radially outwardly thereby allowing lower drive line 17 to move downwardly toward the core of the nuclear reactor resulting in automatic reduction of the reactor powder.

  7. Damper mechanism for nuclear reactor control elements

    DOEpatents

    Taft, William Elwood

    1976-01-01

    A damper mechanism which provides a nuclear reactor control element decelerating function at the end of the scram stroke. The total damping function is produced by the combination of two assemblies, which operate in sequence. First, a tapered dashram assembly decelerates the control element to a lower velocity, after which a spring hydraulic damper assembly takes over to complete the final damping.

  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. Synfuel production in nuclear reactors

    DOEpatents

    Henning, C.D.

    Apparatus and method for producing synthetic fuels and synthetic fuel components by using a neutron source as the energy source, such as a fusion reactor. Neutron absorbers are disposed inside a reaction pipe and are heated by capturing neutrons from the neutron source. Synthetic fuel feedstock is then placed into contact with the heated neutron absorbers. The feedstock is heated and dissociates into its constituent synfuel components, or alternatively is at least preheated sufficiently to use in a subsequent electrolysis process to produce synthetic fuels and synthetic fuel components.

  10. Five Lectures on Nuclear Reactors Presented at Cal Tech

    DOE R&D Accomplishments Database

    Weinberg, Alvin M.

    1956-02-10

    The basic issues involved in the physics and engineering of nuclear reactors are summarized. Topics discussed include theory of reactor design, technical problems in power reactors, physical problems in nuclear power production, and future developments in nuclear power. (C.H.)

  11. DOE SPENT NUCLEAR FUEL DISPOSAL CONTAINER

    SciTech Connect

    F. Habashi

    1998-06-26

    The DOE Spent Nuclear Fuel Disposal Container (SNF DC) supports the confinement and isolation of waste within the Engineered Barrier System of the Mined Geologic Disposal System (MGDS). Disposal containers are loaded and sealed in the surface waste handling facilities, transferred to the underground through the access mains, and emplaced in emplacement drifts. The DOE Spent Nuclear Fuel Disposal Container provides long term confinement of DOE SNF waste, and withstands the loading, transfer, emplacement, and retrieval loads and environments. The DOE SNF Disposal Containers provide containment of waste for a designated period of time, and limit radionuclide release thereafter. The disposal containers maintain the waste in a designated configuration, withstand maximum handling and rockfall loads, limit the individual waste canister temperatures after emplacement. The disposal containers also limit the introduction of moderator into the disposal container during the criticality control period, resist corrosion in the expected repository environment, and provide complete or limited containment of waste in the event of an accident. Multiple disposal container designs may be needed to accommodate the expected range of DOE Spent Nuclear Fuel. The disposal container will include outer and inner barrier walls and outer and inner barrier lids. Exterior labels will identify the disposal container and contents. Differing metal barriers will support the design philosophy of defense in depth. The use of materials with different failure mechanisms prevents a single mode failure from breaching the waste package. The corrosion-resistant inner barrier and inner barrier lid will be constructed of a high-nickel alloy and the corrosion-allowance outer barrier and outer barrier lid will be made of carbon steel. The DOE Spent Nuclear Fuel Disposal Containers interface with the emplacement drift environment by transferring heat from the waste to the external environment and by protecting

  12. 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)

  13. Simulation of a marine nuclear reactor

    SciTech Connect

    Kusunoki, Tsuyoshi; Kyouya, Masahiko; Kobayashi, Hideo; Ochiai, Masaaki

    1995-02-01

    A Nuclear-powered ship Engineering Simulation SYstem (NESSY) has been developed by the Japan Atomic Energy Research Institute as an advanced design tool for research and development of future marine reactors. A marine reactor must respond to changing loads and to the ship`s motions because of the ship`s maneuvering and its presence in a marine environment. The NESSY has combined programs for the reactor plant behavior calculations and the ship`s motion calculations. Thus, it can simulate reactor power fluctuations caused by changing loads and the ship`s motions. It can also simulate the behavior of water in the pressurizer and steam generators. This water sloshes in response to the ship`s motions. The performance of NESSY has been verified by comparing the simulation calculations with the measured data obtained by experiments performed using the nuclear ship Mutsu. The effects of changing loads and the ship`s motions on the reactor behavior can be accurately simulated by NESSY.

  14. 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)

  15. Antineutrinos from nuclear reactors: recent oscillation measurements

    NASA Astrophysics Data System (ADS)

    Dwyer, D. A.

    2015-02-01

    Nuclear reactors are the most intense man-made source of antineutrinos, providing a useful tool for the study of these particles. Oscillation due to the neutrino mixing angle {{θ }13} is revealed by the disappearance of reactor {{\\bar{ν }}e} over ˜km distances. Use of additional identical detectors located near nuclear reactors reduce systematic uncertainties related to reactor {{\\bar{ν }}e} emission and detector efficiency, significantly improving the sensitivity of oscillation measurements. The Double Chooz, RENO, and Daya Bay experiments set out in search of {{θ }13} using these techniques. All three experiments have recently observed reactor {{\\bar{ν }}e} disappearance, and have estimated values for {{θ }13} of 9.3◦ ± 2.1°, 9.2◦ ± 0.9°, and 8.7◦ ± 0.4° respectively. The energy-dependence of {{\\bar{ν }}e} disappearance has also allowed measurement of the effective neutrino mass difference, \\mid Δ mee2\\mid ≈ \\mid Δ m312\\mid . Comparison with \\mid Δ mμ μ 2\\mid ≈ \\mid Δ m322\\mid from accelerator {{ν }μ } measurements supports the three-flavor model of neutrino oscillation. The current generation of reactor {{\\bar{ν }}e} experiments are expected to reach ˜3% precision in both {{θ }13} and \\mid Δ mee2\\mid . Precise knowledge of these parameters aids interpretation of planned {{ν }μ } measurements, and allows future experiments to probe the neutrino mass hierarchy and possible CP-violation in neutrino oscillation. Absolute measurements of the energy spectra of {{\\bar{ν }}e} deviate from existing models of reactor emission, particularly in the range of 5-7 MeV.

  16. Distributed computing and nuclear reactor analysis

    SciTech Connect

    Brown, F.B.; Derstine, K.L.; Blomquist, R.N.

    1994-03-01

    Large-scale scientific and engineering calculations for nuclear reactor analysis can now be carried out effectively in a distributed computing environment, at costs far lower than for traditional mainframes. The distributed computing environment must include support for traditional system services, such as a queuing system for batch work, reliable filesystem backups, and parallel processing capabilities for large jobs. All ANL computer codes for reactor analysis have been adapted successfully to a distributed system based on workstations and X-terminals. Distributed parallel processing has been demonstrated to be effective for long-running Monte Carlo calculations.

  17. Acoustic transducer for nuclear reactor monitoring

    DOEpatents

    Ahlgren, Frederic F.; Scott, Paul F.

    1977-01-01

    A transducer to monitor a parameter and produce an acoustic signal from which the monitored parameter can be recovered. The transducer comprises a modified Galton whistle which emits a narrow band acoustic signal having a frequency dependent upon the parameter being monitored, such as the temperature of the cooling media of a nuclear reactor. Multiple locations within a reactor are monitored simultaneously by a remote acoustic receiver by providing a plurality of transducers each designed so that the acoustic signal it emits has a frequency distinct from the frequencies of signals emitted by the other transducers, whereby each signal can be unambiguously related to a particular transducer.

  18. Secondary coolant circuit for nuclear-reactors

    SciTech Connect

    Brachet, A.

    1981-10-06

    A secondary coolant circuit for a nuclear-reactor of the liquid metal cooled type is described. The circuit comprises at least one intermediate exchanger mounted in the vessel of said reactor, Also included is a steam-generator for the exchange of calories between the secondary liquid-metal flowing through said secondary circuit and water-steam, at least one pump for circulating said secondary sodium and one tank for storing said secondary liquid-metal andrecovering those products generated by a possible liquid-metal-water reaction in said steam-generator.

  19. Reactor design for nuclear electric propulsion

    NASA Technical Reports Server (NTRS)

    Koenig, D. R.; Ranken, W. A.

    1979-01-01

    The paper analyzes the consequences of heat pipe failures, that resulted in modifications to the basic design of a heat-pipe cooled, fast spectrum nuclear reactor and led to consideration of an entirely different core design. The new design features an integral laminated core configuration consisting of alternating layers of UO2 and molybdenum sheets that span the diameter of the core. Design characteristics are presented and compared for two reactors. A conceptual design for a heat exchanger between the core and the thermionic converter assembly is described. This heat exchanger would provide design and fabrication decoupling of these two assemblies.

  20. Rodded shutdown system for a nuclear reactor

    DOEpatents

    Golden, Martin P.; Govi, Aldo R.

    1978-01-01

    A top mounted nuclear reactor diverse rodded shutdown system utilizing gas fed into a pressure bearing bellows region sealed at the upper extremity to an armature. The armature is attached to a neutron absorber assembly by a series of shafts and connecting means. The armature is held in an uppermost position by an electromagnet assembly or by pressurized gas in a second embodiment. Deenergizing the electromagnet assembly, or venting the pressurized gas, causes the armature to fall by the force of gravity, thereby lowering the attached absorber assembly into the reactor core.

  1. 78 FR 28896 - Design Limits and Loading Combinations for Metal Primary Reactor Containment System Components

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-05-16

    ... COMMISSION Design Limits and Loading Combinations for Metal Primary Reactor Containment System Components... Combinations for Metal Primary Reactor Containment System Components,'' in which there are no substantive... loading combinations for metal primary reactor containment system components. ADDRESSES: Please refer...

  2. Nuclear reactor operators: Training and qualifications. (Latest citations from the NTIS bibliographic database). Published Search

    SciTech Connect

    Not Available

    1994-04-01

    The bibliography contains citations concerning the development of methods for nuclear reactor operator training and qualifications. Programs, simulators, and emergency instructions developed for training and evaluating operations personnel in safe and efficient operation, and maintenance of various types of nuclear reactors are examined. Operator examination guides, simulator experiments, and retraining programs are discussed. (Contains 250 citations and includes a subject term index and title list.)

  3. Westinghouse Small Modular Reactor nuclear steam supply system design

    SciTech Connect

    Memmott, M. J.; Harkness, A. W.; Van Wyk, J.

    2012-07-01

    The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (>225 MWe) integral pressurized water reactor (iPWR), in which all of the components typically associated with the nuclear steam supply system (NSSS) of a nuclear power plant are incorporated within a single reactor pressure vessel. This paper is the first in a series of four papers which describe the design and functionality of the Westinghouse SMR. Also described in this series are the key drivers influencing the design of the Westinghouse SMR and the unique passive safety features of the Westinghouse SMR. Several critical motivators contributed to the development and integration of the Westinghouse SMR design. These design driving motivators dictated the final configuration of the Westinghouse SMR to varying degrees, depending on the specific features under consideration. These design drivers include safety, economics, AP1000{sup R} reactor expertise and experience, research and development requirements, functionality of systems and components, size of the systems and vessels, simplicity of design, and licensing requirements. The Westinghouse SMR NSSS consists of an integral reactor vessel within a compact containment vessel. The core is located in the bottom of the reactor vessel and is composed of 89 modified Westinghouse 17x17 Robust Fuel Assemblies (RFA). These modified fuel assemblies have an active core length of only 2.4 m (8 ft) long, and the entirety of the core is encompassed by a radial reflector. The Westinghouse SMR core operates on a 24 month fuel cycle. The reactor vessel is approximately 24.4 m (80 ft) long and 3.7 m (12 ft) in diameter in order to facilitate standard rail shipping to the site. The reactor vessel houses hot and cold leg channels to facilitate coolant flow, control rod drive mechanisms (CRDM), instrumentation and cabling, an intermediate flange to separate flow and instrumentation and facilitate simpler refueling, a pressurizer, a straight tube, recirculating steam

  4. CONTROL MEANS FOR A NUCLEAR REACTOR

    DOEpatents

    Teitel, R.J.

    1961-09-01

    A control means is described for a reactor which employs a liquid fuel consisting of a fissile isotope in a liquid bismuth solvent. The liquid fuel is contained in a plurality of tubular vessels. Control is effected by inserting plungers in the vessels to displace the liquid fuel and provide a critical or non- critical fuel configuration as desired.

  5. Particle bed reactor nuclear rocket concept

    NASA Technical Reports Server (NTRS)

    Ludewig, Hans

    1991-01-01

    The particle bed reactor nuclear rocket concept consists of fuel particles (in this case (U,Zr)C with an outer coat of zirconium carbide). These particles are packed in an annular bed surrounded by two frits (porous tubes) forming a fuel element; the outer one being a cold frit, the inner one being a hot frit. The fuel element are cooled by hydrogen passing in through the moderator. These elements are assembled in a reactor assembly in a hexagonal pattern. The reactor can be either reflected or not, depending on the design, and either 19 or 37 elements, are used. Propellant enters in the top, passes through the moderator fuel element and out through the nozzle. Beryllium used for the moderator in this particular design to withstand the high radiation exposure implied by the long run times.

  6. Method for automatically scramming a nuclear reactor

    DOEpatents

    Ougouag, Abderrafi M.; Schultz, Richard R.; Terry, William K.

    2005-12-27

    An automatically scramming nuclear reactor system. One embodiment comprises a core having a coolant inlet end and a coolant outlet end. A cooling system operatively associated with the core provides coolant to the coolant inlet end and removes heated coolant from the coolant outlet end, thus maintaining a pressure differential therebetween during a normal operating condition of the nuclear reactor system. A guide tube is positioned within the core with a first end of the guide tube in fluid communication with the coolant inlet end of the core, and a second end of the guide tube in fluid communication with the coolant outlet end of the core. A control element is positioned within the guide tube and is movable therein between upper and lower positions, and automatically falls under the action of gravity to the lower position when the pressure differential drops below a safe pressure differential.

  7. Power conditioning for space nuclear reactor systems

    NASA Technical Reports Server (NTRS)

    Berman, Baruch

    1987-01-01

    This paper addresses the power conditioning subsystem for both Stirling and Brayton conversion of space nuclear reactor systems. Included are the requirements summary, trade results related to subsystem implementation, subsystem description, voltage level versus weight, efficiency and operational integrity, components selection, and shielding considerations. The discussion is supported by pertinent circuit and block diagrams. Summary conclusions and recommendations derived from the above studies are included.

  8. 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)

  9. Some views on nuclear reactor safety

    SciTech Connect

    Tanguy, P.Y.

    1995-04-01

    This document is the text of a speech given by Pierre Y. Tanguy (Electricite de France) at the 22nd Water Reactor Safety Meeting held in Bethesda, MD in 1994. He describes the EDF nuclear program in broad terms and proceeds to discuss operational safety results with EDF plants. The speaker also outlines actions to enhance safety planned for the future, and he briefly mentions French cooperation with the Chinese on the Daya Bay project.

  10. Oklo reactors and implications for nuclear science

    NASA Astrophysics Data System (ADS)

    Davis, E. D.; Gould, C. R.; Sharapov, E. I.

    2014-04-01

    We summarize the nuclear physics interests in the Oklo natural nuclear reactors, focusing particularly on developments over the past two decades. Modeling of the reactors has become increasingly sophisticated, employing Monte Carlo simulations with realistic geometries and materials that can generate both the thermal and epithermal fractions. The water content and the temperatures of the reactors have been uncertain parameters. We discuss recent work pointing to lower temperatures than earlier assumed. Nuclear cross-sections are input to all Oklo modeling and we discuss a parameter, the 175Lu ground state cross-section for thermal neutron capture leading to the isomer 176mLu, that warrants further investigation. Studies of the time dependence of dimensionless fundamental constants have been a driver for much of the recent work on Oklo. We critically review neutron resonance energy shifts and their dependence on the fine structure constant α and the ratio Xq = mq/Λ (where mq is the average of the u and d current quark masses and Λ is the mass scale of quantum chromodynamics (QCD)). We suggest a formula for the combined sensitivity to α and Xq that exhibits the dependence on proton number Z and mass number A, potentially allowing quantum electrodynamic (QED) and QCD effects to be disentangled if a broader range of isotopic abundance data becomes available.

  11. Study Gives Good Odds on Nuclear Reactor Safety

    ERIC Educational Resources Information Center

    Russell, Cristine

    1974-01-01

    Summarized is data from a recent study on nuclear reactor safety completed by Norman C. Rasmussen and others. Non-nuclear events are about 10,000 times more likely to produce large accidents than nuclear plants. (RH)

  12. Passive heat-transfer means for nuclear reactors. [LMFBR

    DOEpatents

    Burelbach, J.P.

    1982-06-10

    An improved passive cooling arrangement is disclosed for maintaining adjacent or related components of a nuclear reactor within specified temperature differences. Specifically, heat pipes are operatively interposed between the components, with the vaporizing section of the heat pipe proximate the hot component operable to cool it and the primary condensing section of the heat pipe proximate the other and cooler component operable to heat it. Each heat pipe further has a secondary condensing section that is located outwardly beyond the reactor confinement and in a secondary heat sink, such as air ambient the containment, that is cooler than the other reactor component. By having many such heat pipes, an emergency passive cooling system is defined that is operative without electrical power.

  13. 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.

  14. Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR)

    SciTech Connect

    Gilbert, B.G.; Richards, R.E.; Reece, W.J.; Gertman, D.I.

    1992-10-01

    This Reference Guide contains instructions on how to install and use Version 3.5 of the NRC-sponsored Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR). The NUCLARR data management system is contained in compressed files on the floppy diskettes that accompany this Reference Guide. NUCLARR is comprised of hardware component failure data (HCFD) and human error probability (HEP) data, both of which are available via a user-friendly, menu driven retrieval system. The data may be saved to a file in a format compatible with IRRAS 3.0 and commercially available statistical packages, or used to formulate log-plots and reports of data retrieval and aggregation findings.

  15. Systems and methods for dismantling a nuclear reactor

    DOEpatents

    Heim, Robert R; Adams, Scott Ryan; Cole, Matthew Denver; Kirby, William E; Linnebur, Paul Damon

    2014-10-28

    Systems and methods for dismantling a nuclear reactor are described. In one aspect the system includes a remotely controlled heavy manipulator ("manipulator") operatively coupled to a support structure, and a control station in a non-contaminated portion of a workspace. The support structure provides the manipulator with top down access into a bioshield of a nuclear reactor. At least one computing device in the control station provides remote control to perform operations including: (a) dismantling, using the manipulator, a graphite moderator, concrete walls, and a ceiling of the bioshield, the manipulator being provided with automated access to all internal portions of the bioshield; (b) loading, using the manipulator, contaminated graphite blocks from the graphite core and other components from the bioshield into one or more waste containers; and (c) dispersing, using the manipulator, dust suppression and contamination fixing spray to contaminated matter.

  16. Thermoacoustic Thermometry for Nuclear Reactor Monitoring

    SciTech Connect

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

    2013-06-01

    On Friday, March 11, 2011, at 2:46pm (Japan Standard Trme), the Tohoku region on the east coast of northern Japan experi­enced what would become known as the largest earthquake in the country's history at magnitude 9.0 on the Richter scale. The Fukushima Daiichi nuclear power plant suffered exten­sive and irreversible damage. Six operating units were at the site, each with a boiling water reactor. When the earthquake struck, three of the six reactors were operating and the others were in a periodic inspection outage phase. In one reactor, all of the fuel had been relocated to a spent fuel pool in the reactor building. The seismic acceleration caused by the earthquake brought the three operating units to an automatic shutdown. Since there was damage to the power transmission lines, the emergency diesel generators (EDG) were automat­ically started to ensure continued cooling of the reactors and spent fuel pools. The situation was under control until the tsunami hit about forty-five minutes later with a maximum wave height of approximately 15 meters, which was three times taller than the sea wall of 5m. The influx of water submerged the EDGs, the electrical switchgear, and dc batteries, resulting in the total loss of power to five of the six reactors. The flooding also resulted in the loss of instrumentation that would have other­ wise been used to monitor and control the emergency. The ugly aftermath included high radiation exposure to operators at the nuclear power plants and early contamina­tion of food supplies and water within several restricted areas in Japan, where high radiation levels have rendered them un­safe for human habitation. While the rest of the story will remain a tragic history, it is this part of the series of unfortunate events that has inspired our research. It has indubitably highlighted the need for a novel sensor and instrumentation system that can withstand similar or worse conditions to avoid future catastrophe and assume damage

  17. BN-800 advanced nuclear power plant with fast reactor

    SciTech Connect

    Shishkin, A.N.; Kuzavkov, N.G.; Sobolev, V.A.; Shestakov, G.V.; Bagdasarov, Yu.E.; Kochetkov, L.A.; Matveyev, V.I.; Poplavsky, V.M.

    1993-12-31

    Bn-800 reactor plant with fast reactor and sodium coolant in the primary and secondary circuits is designed for operation as part of the power units in the Yuzhno-Uralskaya nuclear power plant scheduled to be constructed in Chelyabinsk region and as part unit 4 in the Beloyarskaya nuclear power plant. Reactor operations are described.

  18. Lessons from Fukushima for Improving the Safety of Nuclear Reactors

    NASA Astrophysics Data System (ADS)

    Lyman, Edwin

    2012-02-01

    The March 2011 accident at the Fukushima Daiichi nuclear power plant has revealed serious vulnerabilities in the design, operation and regulation of nuclear power plants. While some aspects of the accident were plant- and site-specific, others have implications that are broadly applicable to the current generation of nuclear plants in operation around the world. Although many of the details of the accident progression and public health consequences are still unclear, there are a number of lessons that can already be drawn. The accident demonstrated the need at nuclear plants for robust, highly reliable backup power sources capable of functioning for many days in the event of a complete loss of primary off-site and on-site electrical power. It highlighted the importance of detailed planning for severe accident management that realistically evaluates the capabilities of personnel to carry out mitigation operations under extremely hazardous conditions. It showed how emergency plans rooted in the assumption that only one reactor at a multi-unit site would be likely to experience a crisis fail miserably in the event of an accident affecting multiple reactor units simultaneously. It revealed that alternate water injection following a severe accident could be needed for weeks or months, generating large volumes of contaminated water that must be contained. And it reinforced the grim lesson of Chernobyl: that a nuclear reactor accident could lead to widespread radioactive contamination with profound implications for public health, the economy and the environment. While many nations have re-examined their policies regarding nuclear power safety in the months following the accident, it remains to be seen to what extent the world will take the lessons of Fukushima seriously and make meaningful changes in time to avert another, and potentially even worse, nuclear catastrophe.

  19. Advanced nuclear reactor public opinion project

    SciTech Connect

    Benson, B.

    1991-07-25

    This Interim Report summarizes the findings of our first twenty in-depth interviews in the Advanced Nuclear Reactor Public Opinion Project. We interviewed 6 industry trade association officials, 3 industry attorneys, 6 environmentalists/nuclear critics, 3 state officials, and 3 independent analysts. In addition, we have had numerous shorter discussions with various individuals concerned about nuclear power. The report is organized into the four categories proposed at our April, 1991, Advisory Group meeting: safety, cost-benefit analysis, science education, and communications. Within each category, some change of focus from that of the Advisory Group has been required, to reflect the findings of our interviews. This report limits itself to describing our findings. An accompanying memo draws some tentative conclusions.

  20. Worldwide advanced nuclear power reactors with passive and inherent safety: What, why, how, and who

    SciTech Connect

    Forsberg, C.W.; Reich, W.J.

    1991-09-01

    The political controversy over nuclear power, the accidents at Three Mile Island (TMI) and Chernobyl, international competition, concerns about the carbon dioxide greenhouse effect and technical breakthroughs have resulted in a segment of the nuclear industry examining power reactor concepts with PRIME safety characteristics. PRIME is an acronym for Passive safety, Resilience, Inherent safety, Malevolence resistance, and Extended time after initiation of an accident for external help. The basic ideal of PRIME is to develop power reactors in which operator error, internal sabotage, or external assault do not cause a significant release of radioactivity to the environment. Several PRIME reactor concepts are being considered. In each case, an existing, proven power reactor technology is combined with radical innovations in selected plant components and in the safety philosophy. The Process Inherent Ultimate Safety (PIUS) reactor is a modified pressurized-water reactor, the Modular High Temperature Gas-Cooled Reactor (MHTGR) is a modified gas-cooled reactor, and the Advanced CANDU Project is a modified heavy-water reactor. In addition to the reactor concepts, there is parallel work on super containments. The objective is the development of a passive box'' that can contain radioactivity in the event of any type of accident. This report briefly examines: why a segment of the nuclear power community is taking this new direction, how it differs from earlier directions, and what technical options are being considered. A more detailed description of which countries and reactor vendors have undertaken activities follows. 41 refs.

  1. NEW EMPLOYEES ON THE JOB - DONALD E HEGBERG OF THE NUCLEAR REACTOR DIVISION DISCUSSES NUCLEAR ROCKET

    NASA Technical Reports Server (NTRS)

    1963-01-01

    NEW EMPLOYEES ON THE JOB - DONALD E HEGBERG OF THE NUCLEAR REACTOR DIVISION DISCUSSES NUCLEAR ROCKET FUEL ELEMENT EXPERIMENT WITH CHARLES L YOUNGER - THE DISCUSSION IS PREPATORY TO CONDUCTING THE EXPERIMENT AT THE PLUM BROOK STATION REACTOR FACILITY

  2. Nuclear reactor insulation and preheat system

    DOEpatents

    Wampole, Nevin C.

    1978-01-01

    An insulation and preheat system for preselected components of a fluid cooled nuclear reactor. A gas tight barrier or compartment of thermal insulation surrounds the selected components and includes devices to heat the internal atmosphere of the compartment. An external surface of the compartment or enclosure is cooled, such as by a circulating fluid. The heating devices provide for preheating of the components, as well as maintenance of a temperature sufficient to ensure that the reactor coolant fluid will not solidify during shutdown. The external cooling limits the heat transferred to other plant structures, such as supporting concrete and steel. The barrier is spaced far enough from the surrounded components so as to allow access for remote or manual inspection, maintenance, and repair.

  3. Closure head for a nuclear reactor

    DOEpatents

    Wade, Elman E.

    1980-01-01

    A closure head for a nuclear reactor includes a stationary outer ring integral with the reactor vessel with a first rotatable plug disposed within the stationary outer ring and supported from the stationary outer ring by a bearing assembly. A sealing system is associated with the bearing assembly to seal the annulus defined between the first rotatable plug and the stationary outer ring. The sealing system comprises tubular seal elements disposed in the annulus with load springs contacting the tubular seal elements so as to force the tubular seal elements against the annulus in a manner to seal the annulus. The sealing system also comprises a sealing fluid which is pumped through the annulus and over the tubular seal elements causing the load springs to compress thereby reducing the friction between the tubular seal elements and the rotatable components while maintaining a gas-tight seal therebetween.

  4. Eugene Wigner, The First Nuclear Reactor Engineer

    NASA Astrophysics Data System (ADS)

    Weinberg, Alvin M.

    2002-04-01

    All physicists recognize Eugene Wigner as a theoretical physicist of the very first rank. Yet Wigner's only advanced degree was in Chemical Engineering. His physics was largely self-taught. During WWII, Wigner brilliantly returned to his original occupation as an engineer. He led the small team of theoretical physicists and engineers who designed, in remarkable detail, the original graphite-moderated, water-cooled Hanford reactor, which produced the Pu239 of the Trinity and Nagasaki bombs. With his unparalleled understanding of chain reactors (matched only by Fermi) and his skill and liking for engineering, Wigner can properly be called the Founder of Nuclear Engineering. The evidence for this is demonstrated by a summary of his 37 Patents on various chain reacting systems.

  5. MEANS FOR CONTROLLING A NUCLEAR REACTOR

    DOEpatents

    Wilson, V.C.; Overbeck, W.P.; Slotin, L.; Froman, D.K.

    1957-12-17

    This patent relates to nuclear reactors of the type using a solid neutron absorbing material as a means for controlling the reproduction ratio of the system and thereby the power output. Elongated rods of neutron absorbing material, such as boron steel for example, are adapted to be inserted and removed from the core of tae reactor by electronic motors and suitable drive means. The motors and drive means are controlled by means responsive to the neutron density, such as ionization chambers. The control system is designed to be responsive also to the rate of change in neutron density to automatically maintain the total power output at a substantially constant predetermined value. A safety rod means responsive to neutron density is also provided for keeping the power output below a predetermined maximum value at all times.

  6. 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.

  7. 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.

  8. Nuclear reactor flow control method and apparatus

    DOEpatents

    Church, John P.

    1993-01-01

    Method and apparatus for improving coolant flow in a nuclear reactor during accident as well as nominal conditions. The reactor has a plurality of fuel elements in sleeves and a plenum above the fuel and through which the sleeves penetrate. Holes are provided in the sleeve so that coolant from the plenum can enter the sleeve and cool the fuel. The number and size of the holes are varied from sleeve to sleeve with the number and size of holes being greater for sleeves toward the center of the core and less for sleeves toward the periphery of the core. Preferably the holes are all the same diameter and arranged in rows and columns, the rows starting from the bottom of every sleeve and fewer rows in peripheral sleeves and more rows in the central sleeves.

  9. Nuclear reactor flow control method and apparatus

    DOEpatents

    Church, J.P.

    1993-03-30

    Method and apparatus for improving coolant flow in a nuclear reactor during accident as well as nominal conditions. The reactor has a plurality of fuel elements in sleeves and a plenum above the fuel and through which the sleeves penetrate. Holes are provided in the sleeve so that coolant from the plenum can enter the sleeve and cool the fuel. The number and size of the holes are varied from sleeve to sleeve with the number and size of holes being greater for sleeves toward the center of the core and less for sleeves toward the periphery of the core. Preferably the holes are all the same diameter and arranged in rows and columns, the rows starting from the bottom of every sleeve and fewer rows in peripheral sleeves and more rows in the central sleeves.

  10. 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

  11. Predicting the Lifetimes of Nuclear Waste Containers

    NASA Astrophysics Data System (ADS)

    King, Fraser

    2014-03-01

    As for many aspects of the disposal of nuclear waste, the greatest challenge we have in the study of container materials is the prediction of the long-term performance over periods of tens to hundreds of thousands of years. Various methods have been used for predicting the lifetime of containers for the disposal of high-level waste or spent fuel in deep geological repositories. Both mechanical and corrosion-related failure mechanisms need to be considered, although until recently the interactions of mechanical and corrosion degradation modes have not been considered in detail. Failure from mechanical degradation modes has tended to be treated through suitable container design. In comparison, the inevitable loss of container integrity due to corrosion has been treated by developing specific corrosion models. The most important aspect, however, is to be able to justify the long-term predictions by demonstrating a mechanistic understanding of the various degradation modes.

  12. Cladding and Duct Materials for Advanced Nuclear Recycle Reactors

    SciTech Connect

    Allen, Todd R.; Busby, Jeremy T; Klueh, Ronald L; Maloy, S; Toloczko, M

    2008-01-01

    The expanded use of nuclear energy without risk of nuclear weapons proliferation and with safe nuclear waste disposal is a primary goal of the Global Nuclear Energy Partnership (GNEP). To achieve that goal the GNEP is exploring advanced technologies for recycling spent nuclear fuel that do not separate pure plutonium, and advanced reactors that consume transuranic elements from recycled spent fuel. The GNEP s objectives will place high demands on reactor clad and structural materials. This article discusses the materials requirements of the GNEP s advanced nuclear recycle reactors program.

  13. 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.

  14. Measuring Neutrino Oscillations with Nuclear Reactors

    SciTech Connect

    McKeown, R. D.

    2007-10-26

    Since the first direct observations of antineutrino events by Reines and Cowan in the 1950's, nuclear reactors have been an important tool in the study of neutrino properties. More recently, the study of neutrino oscillations has been a very active area of research. The pioneering observation of oscillations by the KamLAND experiment has provided crucial information on the neutrino mixing matrix. New experiments to study the remaining unknown mixing angle are currently under development. These recent studies and potential future developments will be discussed.

  15. Liquid metal pump for nuclear reactors

    DOEpatents

    Allen, H.G.; Maloney, J.R.

    1975-10-01

    A pump for use in pumping high temperature liquids at high pressures, particularly liquid metals used to cool nuclear reactors is described. It is of the type in which the rotor is submerged in a sump but is fed by an inlet duct which bypasses the sump. A chamber, kept full of fluid, surrounds the pump casing into which fluid is bled from the pump discharge and from which fluid is fed to the rotor bearings and hence to the sump. This equalizes pressure inside and outside the pump casing and reduces or eliminates the thermal shock to the bearings and sump tank.

  16. Nuclear reactor accidents: Chernobyl, TMI, and Windscale. (Latest citations from Pollution abstracts). Published Search

    SciTech Connect

    1995-12-01

    The bibliography contains citations concerning studies and measurements of the radiological consequences of nuclear reactor accidents. The citations cover specifically the Chernobyl reactor in the USSR, the Three Mile Island (TMI) reactor in the US, and the Windscale reactor in the UK. Included are detection and monitoring of the fallout; the resultant runoff into rivers, lakes, and the sea; the radiation effects on people; and the transfrontier radioactive contamination of the environment. (Contains 50-250 citations and includes a subject term index and title list.) (Copyright NERAC, Inc. 1995)

  17. Nuclear reactor accidents: Chernobyl, TMI, and windscale. (Latest citations from Pollution Abstracts). Published Search

    SciTech Connect

    Not Available

    1994-11-01

    The bibliography contains citations concerning studies and measurements of the radiological consequences of nuclear reactor accidents. The citations cover specifically the Chernobyl reactor in the USSR, the Three Mile Island (TMI) reactor in the US, and the Windscale reactor in the UK. Included are detection and monitoring of the fallout; the resultant runoff into rivers, lakes, and the sea; the radiation effects on people; and the transfrontier radioactive contamination of the environment. (Contains 250 citations and includes a subject term index and title list.)

  18. The nuclear reactor strategy between fast breeder reactors and advanced pressurized water reactors

    SciTech Connect

    Seifritz, W.

    1983-11-01

    A nuclear reactor strategy between fast breeder reactors (FBRs) and advanced pressurized water reactors (APWRs) is being studied. The principal idea of this strategy is that the discharged plutonium from light water reactors (LWRs) provides the inventories of the FBRs and the high-converter APWRs, whereby the LWRs are installed according to the derivative of a logistical S curve. Special emphasis is given to the dynamics of reaching an asymptotic symbiosis between FBRs and APWRs. The main conclusion is that if a symbiotic APWR-FBR family with an asymptotic total power level in the terawatt range is to exist in about half a century from now, we need a large number of FBRs already in an early phase.

  19. Analysis of nuclear reactor instability phenomena

    SciTech Connect

    Lahey, R.T. Jr.

    1993-01-01

    The phenomena known as density-wave instability often occurs in phase change systems, such as boiling water nuclear reactors (BWRS). Our current understanding of density-wave oscillations is in fairly good shape for linear phenomena (eg, the onset of instabilities) but is not very advanced for non-linear phenomena [Lahey and Podowski, 1989]. In particular, limit cycle and chaotic instability modes are not well understood in boiling systems such as current and advanced generation BWRs (eg, SBWR). In particular, the SBWR relies on natural circulation and is thus inherently prone to problems with density-wave instabilities. The purpose of this research is to develop a quantitative understanding of nonlinear nuclear-coupled density-wave instability phenomena in BWRS. This research builds on the work of Achard et al [1985] and Clausse et al [1991] who showed, respectively, that Hopf bifurcations and chaotic oscillations may occur in boiling systems.

  20. Request for Naval Reactors Comment on Proposed Prometheus Space Flight Nuclear Reactor High Tier Reactor Safety Requirements and for Naval Reactors Approval to Transmit These Requirements to JPL

    SciTech Connect

    D. Kokkinos

    2005-04-28

    The purpose of this letter is to request Naval Reactors comments on the nuclear reactor high tier requirements for the PROMETHEUS space flight reactor design, pre-launch operations, launch, ascent, operation, and disposal, and to request Naval Reactors approval to transmit these requirements to Jet Propulsion Laboratory to ensure consistency between the reactor safety requirements and the spacecraft safety requirements. The proposed PROMETHEUS nuclear reactor high tier safety requirements are consistent with the long standing safety culture of the Naval Reactors Program and its commitment to protecting the health and safety of the public and the environment. In addition, the philosophy on which these requirements are based is consistent with the Nuclear Safety Policy Working Group recommendations on space nuclear propulsion safety (Reference 1), DOE Nuclear Safety Criteria and Specifications for Space Nuclear Reactors (Reference 2), the Nuclear Space Power Safety and Facility Guidelines Study of the Applied Physics Laboratory.

  1. Safety analysis of irradiated nuclear fuel transportation container

    SciTech Connect

    Uspuras, E.; Rimkevicius, S.

    2007-07-01

    Ignalina NPP comprises two Units with RBMK-1500 reactors. After the Unit 1 of the Ignalina Nuclear Power Plant was shut down in 2004, approximately 1000 fuel assemblies from Unit were available for further reuse in Unit 2. The fuel-transportation container, vehicle, protection shaft and other necessary equipment were designed in order to implement the process for on-site transportation of Unit 1 fuel for reuse in the Unit 2. The Safety Analysis Report (SAR) was developed to demonstrate that the proposed set of equipment performs all functions and assures the required level of safety for both normal operation and accident conditions. The purpose of this paper is to introduce the content and main results of SAR, focusing attention on the container used to transport spent fuel assemblies from Unit I on Unit 2. In the SAR, the structural integrity, thermal, radiological and nuclear safety calculations are performed to assess the acceptance of the proposed set of equipment. The safety analysis demonstrated that the proposed nuclear fuel transportation container and other equipment are in compliance with functional, design and regulatory requirements and assure the required safety level. (authors)

  2. Electrochemistry of Water-Cooled Nuclear Reactors

    SciTech Connect

    Macdonald, Dgiby; Urquidi-Macdonald, Mirna; Pitt, Jonathan

    2006-08-08

    This project developed a comprehensive mathematical and simulation model for calculating thermal hydraulic, electrochemical, and corrosion parameters, viz. temperature, fluid flow velocity, pH, corrosion potential, hydrogen injection, oxygen contamination, stress corrosion cracking, crack growth rate, and other important quantities in the coolant circuits of water-cooled nuclear power plants, including both Boiling Water Reactors (BWRs) and Pressurized Water Reactors (PWRs). The model is being used to assess the three major operational problems in Pressurized Water Reactors (PWR), which include mass transport, activity transport, and the axial offset anomaly, and provide a powerful tool for predicting the accumulation of SCC damage in BWR primary coolant circuits as a function of operating history. Another achievement of the project is the development of a simulation tool to serve both as a training tool for plant operators and as an engineering test-bed to evaluate new equipment and operating strategies (normal operation, cold shut down and others). The development and implementation of the model allows us to estimate the activity transport or "radiation fields" around the primary loop and the vessel, as a function of the operating parameters and the water chemistry.

  3. Minimizing or eliminating refueling of nuclear reactor

    DOEpatents

    Doncals, Richard A.; Paik, Nam-Chin; Andre, Sandra V.; Porter, Charles A.; Rathbun, Roy W.; Schwallie, Ambrose L.; Petras, Diane S.

    1989-01-01

    Demand for refueling of a liquid metal fast nuclear reactor having a life of 30 years is eliminated or reduced to intervals of at least 10 years by operating the reactor at a low linear-power density, typically 2.5 kw/ft of fuel rod, rather than 7.5 or 15 kw/ft, which is the prior art practice. So that power of the same magnitude as for prior art reactors is produced, the volume of the core is increased. In addition, the height of the core and it diameter are dimensioned so that the ratio of the height to the diameter approximates 1 to the extent practicable considering the requirement of control and that the pressure drop in the coolant shall not be excessive. The surface area of a cylinder of given volume is a minimum if the ratio of the height to the diameter is 1. By minimizing the surface area, the leakage of neutrons is reduced. By reducing the linear-power density, increasing core volume, reducing fissile enrichment and optimizing core geometry, internal-core breeding of fissionable fuel is substantially enhanced. As a result, core operational life, limited by control worth requirements and fuel burnup capability, is extended up to 30 years of continuous power operation.

  4. Laboratory instrumentation modernization at the WPI Nuclear Reactor Facility

    SciTech Connect

    Not Available

    1995-01-01

    With partial funding from the Department of Energy (DOE) University Reactor Instrumentation Program several laboratory instruments utilized by students and researchers at the WPI Nuclear Reactor Facility have been upgraded or replaced. Designed and built by General Electric in 1959, the open pool nuclear training reactor at WPI was one of the first such facilities in the nation located on a university campus. Devoted to undergraduate use, the reactor and its related facilities have been since used to train two generations of nuclear engineers and scientists for the nuclear industry. The low power output of the reactor and an ergonomic facility design make it an ideal tool for undergraduate nuclear engineering education and other training. The reactor, its control system, and the associate laboratory equipment are all located in the same room. Over the years, several important milestones have taken place at the WPI reactor. In 1969, the reactor power level was upgraded from 1 kW to 10 kW. The reactor`s Nuclear Regulatory Commission operating license was renewed for 20 years in 1983. In 1988, under DOE Grant No. DE-FG07-86ER75271, the reactor was converted to low-enriched uranium fuel. In 1992, again with partial funding from DOE (Grant No. DE-FG02-90ER12982), the original control console was replaced.

  5. 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.

  6. Nuclear-safety criteria and specifications for space nuclear reactors

    SciTech Connect

    Not Available

    1982-08-01

    The policy of the United States for all US nuclear power sources in space is to ensure that the probability of release of radioactive material and the amounts released are such that an undue risk is not presented, considering the benefits of the mission. The objective of this document is to provide safety criteria which a mission/reactor designer can use to help ensure that the design is acceptable from a radiological safety standpoint. These criteria encompass mission design, reactor design, and radiological impact limitation requirements for safety, and the documentation required. They do not address terrestrial operations, occupational safety or system reliability except where the systems are important for radiological safety. Specific safety specifications based on these criteria shall also be generated and made part of contractual requirements.

  7. Ancient metallurgy and nuclear waste containment

    SciTech Connect

    Goodway, M.

    1993-12-31

    Archaeological artifacts of glass, ceramic, and metal provide examples of long term durability and as such have been surveyed by the nuclear agencies of several countries as a possible guide to choices of materials for the containment of nuclear waste. In the case of metals evaluation is difficult because of the loss of many artifacts to recycling and corrosion processes, as well as by uncertainty as to the environmental history under which the remainder survived. More recently the study of ancient metallurgy has expanded to included other materials associated with metals processing. It is suggested that an impermeable ceramic composite used in ancient metals processing installations should be reproduced and tested for its resistance to radiation damage. This material was synthesized more than two millennia ago and has a proven record of durability. These installations have had no maintenance but are intact, some still holding water.

  8. Nuclear reactor decommissioning. (Latest citations from the NTIS bibliographic database). Published Search

    SciTech Connect

    1997-10-01

    The bibliography contains citations concerning nuclear power and research reactor decommissioning and decontamination plans, costs, and safety standards. References discuss the design and evaluation of protective confinement, entombment, and dismantling systems. Topics include decommissioning regulations and rules, public and occupational radiation exposure estimates, comparative evaluation, and reactor performance under high neutron flux conditions. Waste packaging and disposal, environmental compliance, and public opinion are examined. (Contains 50-250 citations and includes a subject term index and title list.) (Copyright NERAC, Inc. 1995)

  9. Characteristics of potential repository wastes: Volume 4, Appendix 4A, Nuclear reactors at educational institutions of the United States; Appendix 4B, Data sheets for nuclear reactors at educational institutions; Appendix 4C, Supplemental data for Fort St. Vrain spent fuel; Appendix 4D, Supplemental data for Peach Bottom 1 spent fuel; Appendix 4E, Supplemental data for Fast Flux Test Facility

    SciTech Connect

    Not Available

    1992-07-01

    Volume 4 contains the following appendices: nuclear reactors at educational institutions in the United States; data sheets for nuclear reactors at educational institutions in the United States(operational reactors and shut-down reactors); supplemental data for Fort St. Vrain spent fuel; supplemental data for Peach Bottom 1 spent fuel; and supplemental data for Fast Flux Test Facility.

  10. Determination of parameters of a nuclear reactor through noise measurements

    DOEpatents

    Cohn, C.E.

    1975-07-15

    A method of measuring parameters of a nuclear reactor by noise measurements is described. Noise signals are developed by the detectors placed in the reactor core. The polarity coincidence between the noise signals is used to develop quantities from which various parameters of the reactor can be calculated. (auth)

  11. SELF-REGULATING BOILING-WATER NUCLEAR REACTORS

    DOEpatents

    Ransohoff, J.A.; Plawchan, J.D.

    1960-08-16

    A boiling-water reactor was designed which comprises a pressure vessel containing a mass of water, a reactor core submerged within the water, a reflector tank disposed within the reactor, the reflector tank being open at the top to the interior of the pressure vessel, and a surge tank connected to the reflector tank. In operation the reflector level changes as a function of the pressure witoin the reactor so that the reactivity of the reactor is automatically controlled.

  12. Nuclear waste disposal utilizing a gaseous core reactor

    NASA Technical Reports Server (NTRS)

    Paternoster, R. R.

    1975-01-01

    The feasibility of a gaseous core nuclear reactor designed to produce power to also reduce the national inventories of long-lived reactor waste products through nuclear transmutation was examined. Neutron-induced transmutation of radioactive wastes is shown to be an effective means of shortening the apparent half life.

  13. Coupled IVPs to Investigate a Nuclear Reactor Poison Burn Up

    SciTech Connect

    Faghihi, F.

    2009-09-09

    A set of coupled IVPs that describe the change rate of an important poison, in a nuclear reactor, has been written herein. Specifically, in this article, we have focused on the samarium-149 (as a poison) burnup in a desired pressurized water nuclear reactor and its concentration are given using our MATLAB-linked 'solver'.

  14. Method for producing heat treated composite nuclear fuel containers

    SciTech Connect

    Rosenbaum, H.S.

    1993-06-29

    A method is described for producing composite constructed nuclear fuel containers for service in water cooled nuclear fission reactors comprising a tubular zirconium alloy casing having a protective lining of a zirconium metal covering the inside surface of the zirconium alloy tubular casing and being metallurgically bonded thereto, consisting essentially of the steps of: heat treating a large diameter zirconium alloy tube comprising a beta-quench treatment of heating the zirconium alloy to a temperature sufficient to recrystallize the zirconium alloy to its beta phase of at least about 970 C and then rapidly cooling the thus heated and recrystallized zirconium alloy tube stock, heat treating a large diameter zirconium metal hollow liner stock comprising a beta-quench treatment of heating the zirconium metal to a temperature sufficient to recrystallize the zirconium metal to its beta phase of at least about 900 C and then rapidly cooling the thus heated and recrystallized zirconium metal hollow liner stock, assembling the heat treated zirconium alloy tube stock and heat treated lining stock with the hollow zirconium metal lining stock inserted inclose fitting contact within the tube stock, metallurgically bonding the tube and liner stocks providing a composite tubular stock unit with the tube stock surrounding the lining stock, and reducing the circumference of the assembled composite tubular stock unit down to a size suitable for service as a lined tubular container for nuclear fuel in a series of progressive reductions in circumference applied in sequence.

  15. 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.

  16. Space Nuclear Reactor Electric Power. (Latest citations from the Aerospace database). Published Search

    SciTech Connect

    Not Available

    1994-05-01

    The bibliography contains citations concerning studies and conceptual designs of nuclear space power reactors to generate electric power for space missions. The citations cover the technology, safety aspects, and policy considerations. (Contains 250 citations and includes a subject term index and title list.)

  17. 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.

  18. Control rod for a nuclear reactor

    DOEpatents

    Roman, Walter G.; Sutton, Jr., Harry G.

    1979-01-01

    A control rod assembly for a nuclear reactor is disclosed having a remotely disengageable coupling between the control rod and the control rod drive shaft. The coupling is actuated by first lowering then raising the drive shaft. The described motion causes axial repositioning of a pin in a grooved rotatable cylinder, each being attached to different parts of the drive shaft which are axially movable relative to each other. In one embodiment, the relative axial motion of the parts of the drive shaft is used either to couple or to uncouple the connection by forcing resilient members attached to the drive shaft into or out of shouldered engagement, respectively, with an indentation formed in the control rod.

  19. Fuel Element for a Nuclear Reactor

    DOEpatents

    Duffy, Jr., J. G.

    1961-05-30

    A lattice-type fissionable fuel structure for a nuclear reactor is offered. The fissionable material is formed into a plurality of rod-like bodies each encased in a fluid-tight jacket. A plurality of spaced longitudinal fins are mounted on the exterior of and extend radially from each jacket, and a portion of the fins extends radially beyond the remainder of the fins. A collar of short lengih for each body is mounted on the extended fins for spacing the bodies, and adjacent bodies abut each other through these collars. Should distortion of the bodies take place, collapse of the outer fins is limited by the shorter fins thereby insuring some coolant flow therethrough at all times.

  20. FUEL ELEMENT FOR A NUCLEAR REACTOR

    DOEpatents

    Duffy, J.G. Jr.

    1961-05-30

    A lattice type fissionable fuel structure for a nuclear reactor is described. The fissionable material is formed into a plurality of rod-llke bodies with each encased in a fluid-tight jacket. A plurality of spaced longitudinal fins are mounted on the exterior and extend radially from each jacket, with a portion of the fins extending radially beyond the remainder of the fins. A collar of short length for each body is mounted on the extended fins for spacing the bodies, and adjacent bodies abut each other through these collars. Should distortion of the bodies take place, coilapse of the outer fins is limited by the shorter flns, thereby insuring some coolant flow at all times. (AEC)

  1. SPRING DRIVEN ACTUATING MECHANISM FOR NUCLEAR REACTOR CONTROL

    DOEpatents

    Bevilacqua, F.; Uecker, D.F.; Groh, E.F.

    1962-01-23

    l962. rod in a nuclear reactor to shut it down. The control rod or an extension thereof is wound on a drum as it is withdrawn from the reactor. When an emergency occurs requiring the reactor to be shut down, the drum is released so as to be free to rotate, and the tendency of the control rod or its extension coiled on the drum to straighten itself is used for quickly returning the control rod to the reactor. (AEC)

  2. Uncertainties in the Anti-neutrino Production at Nuclear Reactors

    SciTech Connect

    Djurcic, Zelimir; Detwiler, Jason A.; Piepke, Andreas; Foster Jr., Vince R.; Miller, Lester; Gratta, Giorgio

    2008-08-06

    Anti-neutrino emission rates from nuclear reactors are determined from thermal power measurements and fission rate calculations. The uncertainties in these quantities for commercial power plants and their impact on the calculated interaction rates in {bar {nu}}{sub e} detectors is examined. We discuss reactor-to-reactor correlations between the leading uncertainties, and their relevance to reactor {bar {nu}}{sub e} experiments.

  3. Structural integrity of nuclear reactor pressure vessels

    NASA Astrophysics Data System (ADS)

    Knott, John F.

    2013-09-01

    The paper starts from concerns expressed by Sir Alan Cottrell, in the early 1970s, related to the safety of the pressurized water reactor (PWR) proposed at that time for the next phase of electrical power generation. It proceeds to describe the design and operation of nuclear generation plant and gives details of the manufacture of PWR reactor pressure vessels (RPVs). Attention is paid to stress-relief cracking and under-clad cracking, experienced with early RPVs, explaining the mechanisms for these forms of cracking and the means taken to avoid them. Particular note is made of the contribution of non-destructive inspection to structural integrity. Factors affecting brittle fracture in RPV steels are described: in particular, effects of neutron irradiation. The use of fracture mechanics to assess defect tolerance is explained, together with the failure assessment diagram embodied in the R6 procedure. There is discussion of the Master Curve and how it incorporates effects of irradiation on fracture toughness. Dangers associated with extrapolation of data to low probabilities are illustrated. The treatment of fatigue-crack growth is described, in the context of transients that may be experienced in the operation of PWR plant. Detailed attention is paid to the thermal shock associated with a large loss-of-coolant accident. The final section reviews the arguments advanced to justify 'Incredibility of Failure' and how these are incorporated in assessments of the integrity of existing plant and proposed 'new build' PWR pressure vessels.

  4. Flow instability in particle-bed nuclear reactors

    NASA Technical Reports Server (NTRS)

    Kerrebrock, Jack L.

    1993-01-01

    The particle-bed core offers mitigation of some of the problems of solid-core nuclear rocket reactors. Dividing the fuel elements into small spherical particles contained in a cylindrical bed through which the propellant flows radially, may reduce the thermal stress in the fuel elements, allowing higher propellant temperatures to be reached. The high temperature regions of the reactor are confined to the interior of cylindrical fuel assemblies, so most of the reactor can be relatively cool. This enables the use of structural and moderating materials which reduce the minimum critical size and mass of the reactor. One of the unresolved questions about this concept is whether the flow through the particle-bed will be well behaved, or will be subject to destructive flow instabilities. Most of the recent analyses of the stability of the particle-bed reactor have been extensions of the approach of Bussard and Delauer, where the bed is essentially treated as an array of parallel passages, so that the mass flow is continuous from inlet to outlet through any one passage. A more general three dimensional model of the bed is adopted, in which the fluid has mobility in three dimensions. Comparison of results of the earlier approach to the present one shows that the former does not accurately represent the stability at low Re. The more complete model presented should be capable of meeting this deficiency while accurately representing the effects of the cold and hot frits, and of heat conduction and radiation in the particle-bed. It can be extended to apply to the cylindrical geometry of particle-bed reactors without difficulty. From the exemplary calculations which were carried out, it can be concluded that a particle-bed without a cold frit would be subject to instability if operated at the high temperatures desired for nuclear rockets, and at power densities below about 4 megawatts per liter. Since the desired power density is about 40 megawatts per liter, it can be concluded

  5. 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.

  6. Nuclear waste; Can we contain it

    SciTech Connect

    King, F.; Ikeda, B.M.; Shoesmith, D.W.

    1992-04-01

    This paper reports that the safe disposal of nuclear waste requires that the waste be isolated from the environment until radioactive decay has reduced its toxicity to innocuous levels. The disposal of such wastes deep in stable geological formations has been extensively researched since the late 1970s and is now the preferred option internationally. In all of the proposed disposal concepts, the natural barrier of the geological formation is supplemented by a series of engineered barriers each of which retards the transport of radionuclides to the environment. The geological formations being considered usually fall into one of three general categories: crystalline rock (Canada, Sweden, Switzerland, United Kingdom, United States); salt deposits (United States, Germany); and sedimentary deposits, such as clay or seabed sediments (Belgium, United Kingdom, United States), illustrates the Canadian disposal concept based on disposal in igneous rock in the Canadian Shield. The waste will consist of either used fuel bundles or immobilized reprocessed material. In the multibarrier approach the principal engineered component, and the only absolute barrier, is a metallic container enclosing the waste. The required period of containment will influence the choice of material and the thickness of the container.

  7. Thermodynamic consequences of hydrogen combustion within a containment of pressurized water reactor

    NASA Astrophysics Data System (ADS)

    Bury, Tomasz

    2011-12-01

    Gaseous hydrogen may be generated in a nuclear reactor system as an effect of the core overheating. This creates a risk of its uncontrolled combustion which may have a destructive consequences, as it could be observed during the Fukushima nuclear power plant accident. Favorable conditions for hydrogen production occur during heavy loss-of-coolant accidents. The author used an own computer code, called HEPCAL, of the lumped parameter type to realize a set of simulations of a large scale loss-of-coolant accidents scenarios within containment of second generation pressurized water reactor. Some simulations resulted in high pressure peaks, seemed to be irrational. A more detailed analysis and comparison with Three Mile Island and Fukushima accidents consequences allowed for withdrawing interesting conclusions.

  8. A brief history of design studies on innovative nuclear reactors

    SciTech Connect

    Sekimoto, Hiroshi

    2014-09-30

    In a short period after the success of CP1, many types of nuclear reactors were proposed and investigated. However, soon only a small number of reactors were selected for practical use. Around 1970, only LWRs with small number of CANDUs were operated in the western world, and FBRs were under development. It was about the time when Apollo moon landing was accomplished. However, at the same time, the future of human being was widely considered pessimistic and Limits to Growth was published. In the end of 1970’s the TMI accident occurred and many nuclear reactor contracts were cancelled in USA and any more contracts had not been concluded until recent years. From the reflection of this accident, many Inherent Safe Reactors (ISRs) were proposed, though none of them were constructed. A common idea of ISRs is smallness of their size. Tokyo Institute of Technology (TokyoTech) held a symposium on small reactors, SR/TIT, in 1991, where many types of small ISRs were presented. Recently small reactors attract interest again. The most ideas employed in these reactors were the same discussed in SR/TIT. In 1980’s the radioactive wastes from fuel cycle became a severe problem around the world. In TokyoTech, this issue was discussed mainly from the viewpoint of nuclear transmutations. The neutron economy became inevitable for these innovative nuclear reactors especially small long-life reactors and transmutation reactors.

  9. Proposal of Space Reactor for Nuclear Electric Propulsion System

    NASA Astrophysics Data System (ADS)

    Nagata, Hidetaka; Nishiyama, Takaaki; Nakashima, Hideki

    Currently, the solar battery, the chemical cell, and the RI battery are used for the energy source in space. However, it is difficult for them to satisfy requirements for deep space explorations. Therefore, other electric power sources which can stably produce high electric energy output, regardless of distance from the sun, are necessary to execute such missions. Then, we here propose small nuclear reactors as power sources for deep space exploration, and consider a conceptual design of a small nuclear reactor for Nuclear Electric Propulsion System. It is found from nuclear analyses that the Gas-Cooled reactor could not meet the design requirement imposed on the core mass. On the other hand, a light water reactor is found to be a promising alternative to the Gas-Cooled reactor.

  10. Systems and methods for processing irradiation targets through a nuclear reactor

    DOEpatents

    Dayal, Yogeshwar; Saito, Earl F.; Berger, John F.; Brittingham, Martin W.; Morales, Stephen K.; Hare, Jeffrey M.

    2016-05-03

    Apparatuses and methods produce radioisotopes in instrumentation tubes of operating commercial nuclear reactors. Irradiation targets may be inserted and removed from instrumentation tubes during operation and converted to radioisotopes otherwise unavailable during operation of commercial nuclear reactors. Example apparatuses may continuously insert, remove, and store irradiation targets to be converted to useable radioisotopes or other desired materials at several different origin and termination points accessible outside an access barrier such as a containment building, drywell wall, or other access restriction preventing access to instrumentation tubes during operation of the nuclear plant.

  11. Assessment of nuclear reactor concepts for low power space applications

    NASA Technical Reports Server (NTRS)

    Klein, Andrew C.; Gedeon, Stephen R.; Morey, Dennis C.

    1988-01-01

    The results of a preliminary small reactor concepts feasibility and safety evaluation designed to provide a first order validation of the nuclear feasibility and safety of six small reactor concepts are given. These small reactor concepts have potential space applications for missions in the 1 to 20 kWe power output range. It was concluded that low power concepts are available from the U.S. nuclear industry that have the potential for meeting both the operational and launch safety space mission requirements. However, each design has its uncertainties, and further work is required. The reactor concepts must be mated to a power conversion technology that can offer safe and reliable operation.

  12. The necessity of nuclear reactors for targeted radionuclide therapies.

    PubMed

    Krijger, Gerard C; Ponsard, Bernard; Harfensteller, Mark; Wolterbeek, Hubert T; Nijsen, Johannes W F

    2013-07-01

    Nuclear medicine has been contributing towards personalized therapies. Nuclear reactors are required for the working horses of both diagnosis and treatment, i.e., Tc-99m and I-131. In fact, reactors will remain necessary to fulfill the demand for a variety of radionuclides and are essential in the expanding field of targeted radionuclide therapies for cancer. However, the main reactors involved in the global supply are ageing and expected to shut down before 2025. Therefore, the fields of (nuclear) medicine, nuclear industry and politics share a global responsibility, faced with the task to secure future access to suitable nuclear reactors. At the same time, alternative production routes should be industrialized. For this, a coordinating entity should be put into place. PMID:23731577

  13. A Wireless Monitoring System for Cracks on the Surface of Reactor Containment Buildings.

    PubMed

    Zhou, Jianguo; Xu, Yaming; Zhang, Tao

    2016-01-01

    Structural health monitoring with wireless sensor networks has been increasingly popular in recent years because of the convenience. In this paper, a real-time monitoring system for cracks on the surface of reactor containment buildings is presented. Customized wireless sensor networks platforms are designed and implemented with sensors especially for crack monitoring, which include crackmeters and temperature detectors. Software protocols like route discovery, time synchronization and data transfer are developed to satisfy the requirements of the monitoring system and stay simple at the same time. Simulation tests have been made to evaluate the performance of the system before full scale deployment. The real-life deployment of the crack monitoring system is carried out on the surface of reactor containment building in Daya Bay Nuclear Power Station during the in-service pressure test with 30 wireless sensor nodes. PMID:27314357

  14. A Wireless Monitoring System for Cracks on the Surface of Reactor Containment Buildings

    PubMed Central

    Zhou, Jianguo; Xu, Yaming; Zhang, Tao

    2016-01-01

    Structural health monitoring with wireless sensor networks has been increasingly popular in recent years because of the convenience. In this paper, a real-time monitoring system for cracks on the surface of reactor containment buildings is presented. Customized wireless sensor networks platforms are designed and implemented with sensors especially for crack monitoring, which include crackmeters and temperature detectors. Software protocols like route discovery, time synchronization and data transfer are developed to satisfy the requirements of the monitoring system and stay simple at the same time. Simulation tests have been made to evaluate the performance of the system before full scale deployment. The real-life deployment of the crack monitoring system is carried out on the surface of reactor containment building in Daya Bay Nuclear Power Station during the in-service pressure test with 30 wireless sensor nodes. PMID:27314357

  15. 78 FR 64028 - Decommissioning of Nuclear Power Reactors

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-10-25

    ... Register on February 14, 2012 (77 FR 8902), for a 60-day public comment period. The public comment period... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION Decommissioning of Nuclear Power Reactors AGENCY: Nuclear Regulatory Commission....

  16. Role of nuclear reactors in future military satellites

    SciTech Connect

    Buden, D.; Angelo, J.A. Jr.

    1982-01-01

    Future military capabilities will be profoundly influenced by emerging Shuttle Era space technology. Regardless of the specific direction or content of tomorrow's military space program, it is clear that advanced space transportation systems, orbital support facilities, and large-capacity power subsystems will be needed to create the generally larger, more sophisticated military space systems of the future. This paper explores the critical role that space nuclear reactors should play in America's future space program and reviews the current state of nuclear reactor power plant technology. Space nuclear reactor technologies have the potential of satisfying power requirements ranging from 10 kW/sub (e)/ to 100 MW/sub (e)/.

  17. The behavior of fission products during nuclear rocket reactor tests

    SciTech Connect

    Bokor, P.C.; Kirk, W.L.; Bohl, R.J.

    1991-01-01

    The experience base regarding fission product behavior developed during the Rover program, the nuclear rocket development program of 1955--1972, will be useful in planning a renewed nuclear rocket program. During the Rover program, 20 reactors were tested at the Nuclear Rocket Development Station in Nevada. Nineteen of these discharged effluent directly into the atmosphere; the last reactor tested, a non-flight-prototypic, fuel-element-testing reactor called the Nuclear Furnace (NF-1) was connected to an effluent cleanup system that removed fission products before the hydrogen coolant (propellant) was discharged to the atmosphere. In general, we are able to increase both test duration and fuel temperature during the test series. Therefore fission product data from the later part of the program are more interesting and more applicable to future reactors. We have collected fission product retention (and release) data reported in both formal and informal publications for six of the later reactor tests; five of these were Los Alamos reactors that were firsts of a kind in configuration or operating conditions. We have also, with the cooperation of Westinghouse, included fission product data from the NRX-A6 reactor, the final member of series of developmental reactors with the same basic geometry, but with significant design and fabrication improvements as the series continued. Table 1 lists the six selected reactors and the test parameters for each.

  18. Indirect passive cooling system for liquid metal cooled nuclear reactors

    DOEpatents

    Hunsbedt, Anstein; Boardman, Charles E.

    1990-01-01

    A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting from fuel decay during reactor shutdown. The passive cooling system comprises a plurality of partitions surrounding the reactor vessel in spaced apart relation forming intermediate areas for circulating heat transferring fluid which remove and carry away heat from the reactor vessel. The passive cooling system includes a closed primary fluid circuit through the partitions surrounding the reactor vessel and a partially adjoining secondary open fluid circuit for carrying transferred heat out into the atmosphere.

  19. Passive cooling safety system for liquid metal cooled nuclear reactors

    DOEpatents

    Hunsbedt, Anstein; Boardman, Charles E.; Hui, Marvin M.; Berglund, Robert C.

    1991-01-01

    A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting from fuel decay during reactor shutdown. The passive cooling system comprises a plurality of partitions surrounding the reactor vessel in spaced apart relation forming intermediate areas for circulating heat transferring fluid which remove and carry away heat from the reactor vessel. The passive cooling system includes a closed primary fluid circuit through the partitions surrounding the reactor vessel and a partially adjoining secondary open fluid circuit for carrying transferred heat out into the atmosphere.

  20. Nuclear safety as applied to space power reactor systems

    SciTech Connect

    Cummings, G.E.

    1987-01-01

    Current space nuclear power reactor safety issues are discussed with respect to the unique characteristics of these reactors. An approach to achieving adequate safety and a perception of safety is outlined. This approach calls for a carefully conceived safety program which makes uses of lessons learned from previous terrestrial power reactor development programs. This approach includes use of risk analyses, passive safety design features, and analyses/experiments to understand and control off-design conditions. The point is made that some recent accidents concerning terrestrial power reactors do not imply that space power reactors cannot be operated safety.

  1. Inspection of Nuclear Power Plant Containment Structures

    SciTech Connect

    Graves, H.L.; Naus, D.J.; Norris, W.E.

    1998-12-01

    Safety-related nuclear power plant (NPP) structures are designed to withstand loadings from a number of low-probability external and interval events, such as earthquakes, tornadoes, and loss-of-coolant accidents. Loadings incurred during normal plant operation therefore generally are not significant enough to cause appreciable degradation. However, these structures are susceptible to aging by various processes depending on the operating environment and service conditions. The effects of these processes may accumulate within these structures over time to cause failure under design conditions, or lead to costly repair. In the late 1980s and early 1990s several occurrences of degradation of NPP structures were discovered at various facilities (e.g., corrosion of pressure boundary components, freeze- thaw damage of concrete, and larger than anticipated loss of prestressing force). Despite these degradation occurrences and a trend for an increasing rate of occurrence, in-service inspection of the safety-related structures continued to be performed in a somewhat cursory manner. Starting in 1991, the U.S. Nuclear Regulatory Commission (USNRC) published the first of several new requirements to help ensure that adequate in-service inspection of these structures is performed. Current regulatory in-service inspection requirements are reviewed and a summary of degradation experience presented. Nondestructive examination techniques commonly used to inspect the NPP steel and concrete structures to identify and quantify the amount of damage present are reviewed. Finally, areas where nondestructive evaluation techniques require development (i.e., inaccessible portions of the containment pressure boundary, and thick heavily reinforced concrete sections are discussed.

  2. Nuclear reactor operator training for disadvantaged Americans

    SciTech Connect

    Farrar, J.P.; Mulder, R.U.

    1992-12-01

    The Nuclear Engineering and Engineering Physics Department of the University of Virginia was awarded a grant by the US Department of Energy in 1984 to establish and administer a reactor operator training program for disadvantaged Americans. Stipends were provided by the US DOE for five trainees with the anticipation that four other educational facilities would participate in the program. Sub-contracts were awarded to four other Universities: Massachusetts Institute of Technology, The University of Missouri at Columbia, Oregon State University, and The State University of New York at Buffalo. The initial two year program was very successful and the grant was renewed in late 1986 for another two years. MIT declined to participate in the second program and was replaced by Ohio State University. U.VA. was notified in September, 1987 that new funding would no longer be provided for this program after December, 1987. U.VA. requested and was granted a no cost extention for the program through December, 1990, since sufficient funds remained in the initial grant to pursue the program further. DOE subsequently approved a no cost extension through November, 1992.

  3. Nuclear reactor cooling system decontamination reagent regeneration. [PWR; BWR

    DOEpatents

    Anstine, L.D.; James, D.B.; Melaika, E.A.; Peterson, J.P. Jr.

    1980-06-06

    An improved method for decontaminating the coolant system of water-cooled nuclear power reactors and for regenerating the decontamination solution is described. A small amount of one or more weak-acid organic complexing agents is added to the reactor coolant, and the pH is adjusted to form a decontamination solution which is circulated throughout the coolant system to dissolve metal oxides from the interior surfaces and complex the resulting metal ions and radionuclide ions. The coolant containing the complexed metal ions and radionuclide ions is passed through a strong-base anion exchange resin bed which has been presaturated with a solution containing the complexing agents in the same ratio and having the same pH as the decontamination solution. As the decontamination solution passes through the resin bed, metal-complexed anions are exchanged for the metal-ion-free anions on the bed, while metal-ion-free anions in the solution pass through the bed, thus removing the metal ions and regenerating the decontamination solution.

  4. 78 FR 35990 - All Operating Boiling-Water Reactor Licensees With Mark I And Mark II Containments; Docket Nos...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-06-14

    ... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION All Operating Boiling-Water Reactor Licensees With Mark I And Mark II Containments; Docket Nos. (As Shown In Attachment 1), License Nos. (As Shown In Attachment 1), EA-13-109; Order Modifying Licenses With Regard to Reliable Hardened...

  5. Secrecy, Simultaneous Discovery, and the Theory of Nuclear Reactors

    ERIC Educational Resources Information Center

    Weart, Spencer

    1977-01-01

    Discusses the simultaneous discovery of the four-factor formula in various countries, the influence of secrecy in preventing the sharing of discovery, and the resultant direction in the development of nuclear reactor theory. (SL)

  6. Solid0Core Heat-Pipe Nuclear Batterly Type Reactor

    SciTech Connect

    Ehud Greenspan

    2008-09-30

    This project was devoted to a preliminary assessment of the feasibility of designing an Encapsulated Nuclear Heat Source (ENHS) reactor to have a solid core from which heat is removed by liquid-metal heat pipes (HP).

  7. Uncertainties in the anti-neutrino production at nuclear reactors

    NASA Astrophysics Data System (ADS)

    Djurcic, Z.; Detwiler, J. A.; Piepke, A.; Foster, V. R.; Miller, L.; Gratta, G.

    2009-04-01

    Anti-neutrino emission rates from nuclear reactors are determined from thermal power measurements and fission rate calculations. The uncertainties in these quantities for commercial power plants and their impact on the calculated interaction rates in \\bar{\

  8. A new concept of nuclear fission reactors safety

    SciTech Connect

    Petrov, Y.V.

    1993-12-31

    To develop safe nuclear energy production acceptable to the society it is proposed to use in the future strongly subcritical reactors (k=0.96-0.97) driven by proton or deuteron accelerators. The accelerator with the current of 40mA and particle energy {approximately}0.8 GeV/nucleon will provide 2 GW (th.) reactor power in fast reactor with metallic U-Pu fuel. The design, control and parameters of such a system are discussed.

  9. Organic matter and containment of uranium and fissiogenic isotopes at the Oklo natural reactors

    USGS Publications Warehouse

    Nagy, B.; Gauthier-Lafaye, F.; Holliger, P.; Davis, D.W.; Mossman, D.J.; Leventhal, J.S.; Rigali, M.J.; Parnell, J.

    1991-01-01

    SOME of the Precambrian natural fission reactors at Oklo in Gabon contain abundant organic matter1,2, part of which was liquefied at the time of criticality and subsequently converted to a graphitic solid3,4. The liquid organic matter helps to reduce U(VI) to U(IV) from aqueous solutions, resulting in the precipitation of uraninite5. It is known that in the prevailing reactor environments, precipitated uraninite grains incorporated fission products. We report here observations which show that these uraninite crystals were held immobile within the resolidified, graphitic bitumen. Unlike water-soluble (humic) organic matter, the graphitic bituminous organics at Oklo thus enhanced radionu-clide containment. Uraninite encased in solid graphitic matter in the organic-rich reactor zones lost virtually no fissiogenic lan-thanide isotopes. The first major episode of uranium and lead migration was caused by the intrusion of a swarm of adjacent dolerite dykes about 1,100 Myr after the reactors went critical. Our results from Oklo imply that the use of organic, hydrophobic solids such as graphitic bitumen as a means of immobilizing radionuclides in pretreated nuclear waste warrants further investigation. ?? 1991 Nature Publishing Group.

  10. Investigation of the basic reactor physics characteristics of the Dalat Nuclear Research Reactor

    SciTech Connect

    Huy, N.Q.; Thong, H.V.; Khang, N.P.

    1994-12-31

    The Dalat nuclear research reactor was reconstructed from the TRIGA Mark II reactor, built in 1963 with a nominal power of 250 kW, and reached its planned nominal power of 500 kW for the first time in February 1984. The Dalat reactor has some characteristics distinct from the former TRIGA reactor. Investigation of its characteristics is carried out by the determination of the reactor physics parameters. This paper represents the experimental results obtained for the effective fraction of the delayed photoneutrons, the extraneous neutron source left after the reactor is shut down, the lowest power levels of reactor critical states, the relative axial and radial distributions of thermal neutrons, the safe positive reactivity inserted into the reactor at a deep subcritical state, the reactivity temperature coefficient of water, the temperature on the surface of the fuel elements, etc.

  11. 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.

  12. Nuclear Design of the HOMER-15 Mars Surface Fission Reactor

    SciTech Connect

    Poston, David I.

    2002-07-01

    The next generation of robotic missions to Mars will most likely require robust power sources in the range of 3 to 20 kWe. Fission systems are well suited to provide safe, reliable, and economic power within this range. The goal of this study is to design a compact, low-mass fission system that meets Mars surface power requirements, while maintaining a high level of safety and reliability at a relatively low cost. The Heat pipe Power System (HPS) is one possible approach for producing near-term, low-cost, space fission power. The goal of the HPS project is to devise an attractive space fission system that can be developed quickly and affordably. The primary ways of doing this are by using existing technology and by designing the system for inexpensive testing. If the system can be designed to allow highly prototypic testing with electrical heating, then an exhaustive test program can be carried out quickly and inexpensively, and thorough testing of the actual flight unit can be performed - which is a major benefit to reliability. Over the past 4 years, three small HPS proof-of-concept technology demonstrations have been conducted, and each has been highly successful. The Heat pipe-Operated Mars Exploration Reactor (HOMER) is a derivative of the HPS designed especially for producing power on the surface of Mars. The HOMER-15 is a 15-kWt reactor that couples with a 3-kWe Stirling engine power system. The reactor contains stainless-steel (SS)-clad uranium nitride (UN) fuel pins that are structurally and thermally bonded to SS/sodium heat pipes. Fission energy is conducted from the fuel pins to the heat pipes, which then carry the heat to the Stirling engine. This paper describes conceptual design and nuclear performance the HOMER-15 reactor. (author)

  13. Liquid metal cooled nuclear reactors with passive cooling system

    DOEpatents

    Hunsbedt, Anstein; Fanning, Alan W.

    1991-01-01

    A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting from fuel decay during reactor shutdown. The passive cooling system comprises a plurality of cooling medium flow circuits which cooperate to remove and carry heat away from the fuel core upon loss of the normal cooling flow circuit to areas external thereto.

  14. Thermionic reactor power conditioner design for nuclear electric propulsion.

    NASA Technical Reports Server (NTRS)

    Jacobsen, A. S.; Tasca, D. M.

    1971-01-01

    Consideration of the effects of various thermionic reactor parameters and requirements upon spacecraft power conditioning design. A basic spacecraft is defined using nuclear electric propulsion, requiring approximately 120 kWe. The interrelationships of reactor operating characteristics and power conditioning requirements are discussed and evaluated, and the effects on power conditioner design and performance are presented.

  15. Radioactive fallout from the Chernobyl nuclear reactor accident

    SciTech Connect

    Beiriger, J.M.; Failor, R.A.; Marsh, K.V.; Shaw, G.E.

    1987-08-01

    This report describes the detection of fallout in the United States from the Chernobyl nuclear reactor accident. As part of its environmental surveillance program, Lawrence Livermore National Laboratory maintained detectors for gamma-emitting radionuclides. Following the reactor accident, additional air filters were set out. Several uncommon isotopes were detected at the time the plume passed into the US. (TEM)

  16. Supplying the nuclear arsenal: American production reactors, 1942--1992

    SciTech Connect

    Carlisle, R.P.; Zenzen, J.M.

    1996-01-01

    Although the history of commercial-power nuclear reactors is well known, the story of the government reactors that produce weapons-grade plutonium and tritium has been shrouded in secrecy. In the first detailed look at the origin and development of these production reactors, the authors describe a fifty-year government effort no less complex, expensive, and technologically demanding than the Polaris or Apollo programs--yet one about which most Americans know virtually nothing. The book describes the evolution of the early reactors, the atomic weapons establishment that surrounded them, and the sometimes bitter struggles between business and political constituencies for their share of 'nuclear pork.' They show how, since the 1980s, aging production reactors have increased the risk of radioactive contamination of the atmosphere and water table. And they describe how the Department of Energy mounted a massive effort to find the right design for a new generation of reactors, only to abandon that effort with the end of the Cold War. Today, all American production reactors remain closed. Due to short half-life, the nation's supply of tritium, crucial to modern weapons, is rapidly dwindling. As countries like Iraq and North Korea threaten to join the nuclear club, the authors contend, the United States needs to revitalize tritium production capacity in order to maintain a viable nuclear deterrent. Meanwhile, as slowly decaying artifacts of the Cold War, the closed production reactors at Hanford, Washington, and Savannah River, South Carolina, loom ominously over the landscape.

  17. OVERVIEW OF NUCLEAR PHYSICS LABORATORY (IMMEDIATELY EAST OF SPSE REACTOR ...

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

    OVERVIEW OF NUCLEAR PHYSICS LABORATORY (IMMEDIATELY EAST OF SP-SE REACTOR ROOM), LEVEL -15’, LOOKING SOUTHWEST. NOTE SLIDING STEEL PLATE DOOR BETWEEN LABORATORY AND REACTOR ROOM - Physics Assembly Laboratory, Area A/M, Savannah River Site, Aiken, Aiken County, SC

  18. Nuclear rocket engine design based on the particle bed reactor

    SciTech Connect

    Powell, J.R.; Ludewig, H.; Mughabghab, S.; Lazareth, O. Jr.; Schmidt, E.; Maise, G. )

    1993-01-01

    A nuclear thermal rocket (NTR) engine design based on the particle-bed reactor (PBR) concept is described in this paper. This engine is designed to satisfy a mission to Mars and thus must develop a thrust of [approximately]1.75 (6) N. This requirement is satisfied if the reactor generates 2000 MW of power.

  19. 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.

  20. MPP: A modular library of models of nuclear reactor components

    SciTech Connect

    Abdalla, M.A.; Guimaraes, L.; Ugolini, D. ); March-Leuba, C.; Nypaver, D.J. ); Ford, C.E. )

    1992-01-01

    This paper presents the Modular Power Plant (MPP) library and its application to simulate the Advanced Liquid Metal Reactor. The MPP library is being developed as part of the Advanced Controls Program of the Oak Ridge National Laboratory. The general purpose of the library is to provide a set of modular models of components needed to simulate nuclear power plants. To give the MPP models modularity characteristics, each model is developed as a stand-alone system. The MPP contains 28 models coded in either the Advanced Continuous Simulation Language (ACSL), or the Generalized Object-Oriented Simulation Environment (GOOSE). The MPP development is parallel to the GOOSE development, and we are currently translating the MPP components from ACSL to GOOSE.

  1. MPP: A modular library of models of nuclear reactor components

    SciTech Connect

    Abdalla, M.A.; Guimaraes, L.; Ugolini, D.; March-Leuba, C.; Nypaver, D.J.; Ford, C.E.

    1992-05-01

    This paper presents the Modular Power Plant (MPP) library and its application to simulate the Advanced Liquid Metal Reactor. The MPP library is being developed as part of the Advanced Controls Program of the Oak Ridge National Laboratory. The general purpose of the library is to provide a set of modular models of components needed to simulate nuclear power plants. To give the MPP models modularity characteristics, each model is developed as a stand-alone system. The MPP contains 28 models coded in either the Advanced Continuous Simulation Language (ACSL), or the Generalized Object-Oriented Simulation Environment (GOOSE). The MPP development is parallel to the GOOSE development, and we are currently translating the MPP components from ACSL to GOOSE.

  2. Comparative assessment of nuclear fuel cycles. Light-water reactor once-through, classical fast breeder reactor, and symbiotic fast breeder reactor cycles

    SciTech Connect

    Hardie, R.W.; Barrett, R.J.; Freiwald, J.G.

    1980-06-01

    The object of the Alternative Nuclear Fuel Cycle Study is to perform comparative assessments of nuclear power systems. There are two important features of this study. First, this evaluation attempts to encompass the complete, integrated fuel cycle from mining of uranium ore to disposal of waste rather than isolated components. Second, it compares several aspects of each cycle - energy use, economics, technological status, proliferation, public safety, and commercial potential - instead of concentrating on one or two assessment areas. This report presents assessment results for three fuel cycles. These are the light-water reactor once-through cycle, the fast breeder reactor on the classical plutonium cycle, and the fast breeder reactor on a symbiotic cycle using plutonium and /sup 233/U as fissile fuels. The report also contains a description of the methodology used in this assessment. Subsequent reports will present results for additional fuel cycles.

  3. Instrumentation availability during severe accidents for a boiling water reactor with a Mark I containment

    SciTech Connect

    Arcieri, W.C.; Hanson, D.J. )

    1992-02-01

    In support of the US Nuclear Regulatory Commission Accident Management Research Program, the availability of instruments to supply accident management information during a broad range of severe accidents is evaluated for a Boiling Water Reactor with a Mark I containment. Results from this evaluation include: (1) the identification of plant conditions that would impact instrument performance and information needs during severe accidents; (2) the definition of envelopes of parameters that would be important in assessing the performance of plant instrumentation for a broad range of severe accident sequences; and (3) assessment of the availability of plant instrumentation during severe accidents.

  4. A Wide Range Neutron Detector for Space Nuclear Reactor Applications

    SciTech Connect

    Nassif, Eduardo; Sismonda, Miguel; Matatagui, Emilio; Pretorius, Stephan

    2007-01-30

    We propose here a versatile and innovative solution for monitoring and controlling a space-based nuclear reactor that is based on technology already proved in ground based reactors. A Wide Range Neutron Detector (WRND) allows for a reduction in the complexity of space based nuclear instrumentation and control systems. A ground model, predecessor of the proposed system, has been installed and is operating at the OPAL (Open Pool Advanced Light Water Research Reactor) in Australia, providing long term functional data. A space compatible Engineering Qualification Model of the WRND has been developed, manufactured and verified satisfactorily by analysis, and is currently under environmental testing.

  5. Soviet space nuclear reactor incidents - Perception versus reality

    NASA Technical Reports Server (NTRS)

    Bennett, Gary L.

    1992-01-01

    Since the Soviet Union reportedly began flying nuclear power sources in 1965 it has had four publicly known accidents involving space reactors, two publicly known accidents involving radioisotope power sources and one close call with a space reactor (Cosmos 1900). The reactor accidents, particularly Cosmos 954 and Cosmos 1402, indicated that the Soviets had adopted burnup as their reentry philosophy which is consistent with the U.S. philosophy from the 1960s and 1970s. While quantitative risk analyses have shown that the Soviet accidents have not posed a serious risk to the world's population, concerns still remain about Soviet space nuclear safety practices.

  6. Guidelines for preparing criticality safety evaluations at Department of Energy non-reactor nuclear facilities

    SciTech Connect

    Not Available

    1993-11-01

    This document contains guidelines that should be followed when preparing Criticality Safety Evaluations that will be used to demonstrate the safety of operations performed at DOE non-reactor nuclear facilities. Adherence to these guidelines will provide consistency and uniformity in criticality safety evaluations (CSEs) across the complex and will document compliance with the requirements of DOE Order 5480.24.

  7. Pellet bed reactor for nuclear propelled vehicles: Part 1: Reactor technology

    NASA Technical Reports Server (NTRS)

    El-Genk, Mohamed S.

    1991-01-01

    The pellet bed reactor (PBR) for nuclear propelled vehicles is briefly discussed. Much of the information is given in viewgraph form. Viewgraphs include information on the layout for a Mars mission using a PBR nuclear thermal rocket, the rocket reactor layout, the fuel pellet design, materials compatibility, fuel microspheres, microsphere coating, melting points in quasibinary systems, stress analysis of microspheres, safety features, and advantages of the PBR concept.

  8. Inherently safe reactors and a second nuclear era.

    PubMed

    Weinberg, A M; Spiewak, I

    1984-06-29

    The Swedish PIUS reactor and the German-American small modular high-temperature gas-cooled reactor are inherently safe-that is, their safety relies not upon intervention of humans or of electromechanical devices but on immutable principles of physics and chemistry. A second nuclear era may require commercialization and deployment of such inherently safe reactors, even though existing light-water reactors appear to be as safe as other well-accepted sources of central electricity, particularly hydroelectric dams. PMID:17793362

  9. Reactor design and integration into a nuclear electric spacecraft

    NASA Technical Reports Server (NTRS)

    Phillips, W. M.; Koenig, D. R.

    1978-01-01

    One of the well-defined applications for nuclear power in space is nuclear electric propulsion (NEP). Mission studies have identified the optimum power level (400 kWe). A single Shuttle launch requirement and science-package integration have added additional constraints to the design. A reactor design which will meet these constraints has been studied. The reactor employs 90 fuel elements, each heat pipe cooled. Reactor control is obtained with BeO/B4C drums in a BeO reflector. The balance of the spacecraft is shielded from the reactor with LiH. Power conditioning and reactor control drum drives are located behind the LiH with the power conditioning. Launch safety, mechanical design and integration with the power conversion subsystem are discussed.

  10. Nuclear design of a very-low-activation fusion reactor

    SciTech Connect

    Cheng, E.T.; Hopkins, G.R.

    1983-06-01

    An investigation was conducted to study the nuclear design aspects of using very-low-activation materials, such as SiC, MgO, and aluminum for fusion-reactor first wall, blanket, and shield applications. In addition to the advantage of very-low radioactive inventory, it was found that the very-low-activation fusion reactor can also offer an adequate tritium-breeding ratio and substantial amount of blanket nuclear heating as a conventional-material-structured reactor does. The most-stringent design constraint found in a very-low-activation fusion reactor is the limited space available in the inboard region of a tokamak concept for shielding to protect the superconducting toroidal field coil. A reference design was developed which mitigates the constraint by adopting a removable tungsten shield design that retains the inboard dimensions and gives the same shield performance as the reference STARFIRE tokamak reactor design.

  11. Pellet bed reactor concepts for nuclear propulsion applications

    NASA Astrophysics Data System (ADS)

    El-Genk, Mohamed S.; Morley, Nicholas J.; Pelaccio, Dennis G.; Juhasz, Albert

    1994-11-01

    Pellet bed reactor (PeBR) concepts have been developed for nuclear thermal and nuclear electric propulsion, and bimodal applications. This annular core, fast spectrum reactor offers many desirable design and safety features. These features include high-power density, small reactor size, full retention of fission products, passive decay heat removal, redundancy in reactor control, negative temperature reactivity feedback, ground testing of the fully assembled reactor using electric heating and nonnuclear fuel elements, and the option of fueling on the launch pad or fueling and refueling in orbit. In addition to these features, the concepts for nuclear electric propulsion and for bimodal power and thermal propulsion have no single point failure. The average power density in the reactor for nuclear thermal propulsion ranges from 2.2 to 3.3 MW/I and for a 15-MWe nuclear electric propulsion system the total power system specific mass is about 3.3 kg/kWe. The bimodal-PeBR system concepts offer specific impulse in excess of 650 s, tens of Newtons of thrust, and total system specific power ranging from 11 to 21.9 We/kg at the 10- and 40-kWe levels, respectively.

  12. Nuclear Technology Series. Course 7: Reactor Operations.

    ERIC Educational Resources Information Center

    Center for Occupational Research and Development, Inc., Waco, TX.

    This technical specialty course is one of thirty-five courses designed for use by two-year postsecondary institutions in five nuclear technician curriculum areas: (1) radiation protection technician, (2) nuclear instrumentation and control technician, (3) nuclear materials processing technician, (4) nuclear quality-assurance/quality-control…

  13. Nuclear Technology Series. Course 12: Reactor Physics.

    ERIC Educational Resources Information Center

    Center for Occupational Research and Development, Inc., Waco, TX.

    This technical specialty course is one of thirty-five courses designed for use by two-year postsecondary institutions in five nuclear technician curriculum areas: (1) radiation protection technician, (2) nuclear instrumentation and control technician, (3) nuclear materials processing technician, (4) nuclear quality-assurance/quality-control…

  14. Nuclear Technology Series. Course 8: Reactor Safety.

    ERIC Educational Resources Information Center

    Center for Occupational Research and Development, Inc., Waco, TX.

    This technical specialty course is one of thirty-five courses designed for use by two-year postsecondary institutians in five nuclear technician curriculum areas: (1) radiation protection technician, (2) nuclear instrumentation and control technician, (3) nuclear materials processing technician, (4) nuclear quality-assurance/quality-control…

  15. PROCESS FOR COOLING A NUCLEAR REACTOR

    DOEpatents

    Borst, L.B.

    1962-12-11

    This patent relates to the operation of a reactor cooled by liquid sulfur dioxide. According to the invention the pressure on the sulfur dioxide in the reactor is maintained at least at the critical pressure of the sulfur dioxide. Heating the sulfur dioxide to its critical temperature results in vaporization of the sulfur dioxide without boiling. (AEC)

  16. Generating unstructured nuclear reactor core meshes in parallel

    DOE PAGESBeta

    Jain, Rajeev; Tautges, Timothy J.

    2014-10-24

    Recent advances in supercomputers and parallel solver techniques have enabled users to run large simulations problems using millions of processors. Techniques for multiphysics nuclear reactor core simulations are under active development in several countries. Most of these techniques require large unstructured meshes that can be hard to generate in a standalone desktop computers because of high memory requirements, limited processing power, and other complexities. We have previously reported on a hierarchical lattice-based approach for generating reactor core meshes. Here, we describe efforts to exploit coarse-grained parallelism during reactor assembly and reactor core mesh generation processes. We highlight several reactor coremore » examples including a very high temperature reactor, a full-core model of the Korean MONJU reactor, a ¼ pressurized water reactor core, the fast reactor Experimental Breeder Reactor-II core with a XX09 assembly, and an advanced breeder test reactor core. The times required to generate large mesh models, along with speedups obtained from running these problems in parallel, are reported. A graphical user interface to the tools described here has also been developed.« less

  17. Generating unstructured nuclear reactor core meshes in parallel

    SciTech Connect

    Jain, Rajeev; Tautges, Timothy J.

    2014-10-24

    Recent advances in supercomputers and parallel solver techniques have enabled users to run large simulations problems using millions of processors. Techniques for multiphysics nuclear reactor core simulations are under active development in several countries. Most of these techniques require large unstructured meshes that can be hard to generate in a standalone desktop computers because of high memory requirements, limited processing power, and other complexities. We have previously reported on a hierarchical lattice-based approach for generating reactor core meshes. Here, we describe efforts to exploit coarse-grained parallelism during reactor assembly and reactor core mesh generation processes. We highlight several reactor core examples including a very high temperature reactor, a full-core model of the Korean MONJU reactor, a ¼ pressurized water reactor core, the fast reactor Experimental Breeder Reactor-II core with a XX09 assembly, and an advanced breeder test reactor core. The times required to generate large mesh models, along with speedups obtained from running these problems in parallel, are reported. A graphical user interface to the tools described here has also been developed.

  18. 10 CFR 1.43 - Office of Nuclear Reactor Regulation.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... produced at facilities licensed under 10 CFR parts 50, 52, and 54; (3) Operators of such facilities; (4... 10 Energy 1 2012-01-01 2012-01-01 false Office of Nuclear Reactor Regulation. 1.43 Section 1.43 Energy NUCLEAR REGULATORY COMMISSION STATEMENT OF ORGANIZATION AND GENERAL INFORMATION...

  19. 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.

  20. 10 CFR 1.43 - Office of Nuclear Reactor Regulation.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... produced at facilities licensed under 10 CFR parts 50, 52, and 54; (3) Operators of such facilities; (4... 10 Energy 1 2014-01-01 2014-01-01 false Office of Nuclear Reactor Regulation. 1.43 Section 1.43 Energy NUCLEAR REGULATORY COMMISSION STATEMENT OF ORGANIZATION AND GENERAL INFORMATION...

  1. 10 CFR 1.43 - Office of Nuclear Reactor Regulation.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... produced at facilities licensed under 10 CFR parts 50, 52, and 54; (3) Operators of such facilities; (4... 10 Energy 1 2011-01-01 2011-01-01 false Office of Nuclear Reactor Regulation. 1.43 Section 1.43 Energy NUCLEAR REGULATORY COMMISSION STATEMENT OF ORGANIZATION AND GENERAL INFORMATION...

  2. 10 CFR 1.43 - Office of Nuclear Reactor Regulation.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... produced at facilities licensed under 10 CFR parts 50, 52, and 54; (3) Operators of such facilities; (4... 10 Energy 1 2010-01-01 2010-01-01 false Office of Nuclear Reactor Regulation. 1.43 Section 1.43 Energy NUCLEAR REGULATORY COMMISSION STATEMENT OF ORGANIZATION AND GENERAL INFORMATION...

  3. Regulatory Concerns on the In-Containment Water Storage System of the Korean Next Generation Reactor

    SciTech Connect

    Ahn, Hyung-Joon; Lee, Jae-Hun; Bang, Young-Seok; Kim, Hho-Jung

    2002-07-15

    The in-containment water storage system (IWSS) is a newly adopted system in the design of the Korean Next Generation Reactor (KNGR). It consists of the in-containment refueling water storage tank, holdup volume tank, and cavity flooding system (CFS). The IWSS has the function of steam condensation and heat sink for the steam release from the pressurizer and provides cooling water to the safety injection system and containment spray system in an accident condition and to the CFS in a severe accident condition. With the progress of the KNGR design, the Korea Institute of Nuclear Safety has been developing Safety and Regulatory Requirements and Guidances for safety review of the KNGR. In this paper, regarding the IWSS of the KNGR, the major contents of the General Safety Criteria, Specific Safety Requirements, Safety Regulatory Guides, and Safety Review Procedures were introduced, and the safety review items that have to be reviewed in-depth from the regulatory viewpoint were also identified.

  4. PRELIMINARY NUCLEAR CRITICALITY NUCLEAR SAFETY EVLAUATION FOR THE CONTAINER SURVEILLANCE AND STORAGE CAPABILITY PROJECT

    SciTech Connect

    Low, M; Matthew02 Miller, M; Thomas Reilly, T

    2007-04-30

    Washington Safety Management Solutions (WSMS) provides criticality safety services to Washington Savannah River Company (WSRC) at the Savannah River Site. One activity at SRS is the Container Surveillance and Storage Capability (CSSC) Project, which will perform surveillances on 3013 containers (hereafter referred to as 3013s) to verify that they meet the Department of Energy (DOE) Standard (STD) 3013 for plutonium storage. The project will handle quantities of material that are greater than ANS/ANSI-8.1 single parameter mass limits, and thus required a Nuclear Criticality Safety Evaluation (NCSE). The WSMS methodology for conducting an NCSE is outlined in the WSMS methods manual. The WSMS methods manual currently follows the requirements of DOE-O-420.1B, DOE-STD-3007-2007, and the Washington Savannah River Company (WSRC) SCD-3 manual. DOE-STD-3007-2007 describes how a NCSE should be performed, while DOE-O-420.1B outlines the requirements for a Criticality Safety Program (CSP). The WSRC SCD-3 manual implements DOE requirements and ANS standards. NCSEs do not address the Nuclear Criticality Safety (NCS) of non-reactor nuclear facilities that may be affected by overt or covert activities of sabotage, espionage, terrorism or other security malevolence. Events which are beyond the Design Basis Accidents (DBAs) are outside the scope of a double contingency analysis.

  5. Temperature actuated shutdown assembly for a nuclear reactor

    DOEpatents

    Sowa, Edmund S.

    1976-01-01

    Three identical bimetallic disks, each shaped as a spherical cap with its convex side composed of a layer of metal such as molybdenum and its concave side composed of a metal of a relatively higher coefficient of thermal expansion such as stainless steel, are retained within flanges attached to three sides of an inner hexagonal tube containing a neutron absorber to be inserted into a nuclear reactor core. Each disk holds a metal ball against its normally convex side so that the ball projects partially through a hole in the tube located concentrically with the center of each disk; at a predetermined temperature an imbalance of thermally induced stresses in at least one of the disks will cause its convex side to become concave and its concave side to become convex, thus pulling the ball from the hole in which it is located. The absorber has a conical bottom supported by the three balls and is small enough in relation to the internal dimensions of the tube to allow it to slip toward the removed ball or balls, thus clearing the unremoved balls or ball so that it will fall into the reactor core.

  6. Evaluating Russian space nuclear reactor technology for United States applications

    SciTech Connect

    Polansky, G.F.; Schmidt, G.L.; Voss, S.S.; Reynolds, E.L.

    1994-08-01

    Space nuclear power and nuclear electric propulsion are considered important technologies for planetary exploration, as well as selected earth orbit applications. The Nuclear Electric Propulsion Space Test Program (NEPSTP) was intended to provide an early flight demonstration of these technologies at relatively low cost through extensive use of existing Russian technology. The key element of Russian technology employed in the program was the Topaz II reactor. Refocusing of the activities of the Ballistic Missile Defense Organization (BMDO), combined with budgetary pressures, forced the cancellation of the NEPSTP at the end of the 1993 fiscal year. The NEPSTP was faced with many unique flight qualification issues. In general, the launch of a spacecraft employing a nuclear reactor power system complicates many spacecraft qualification activities. However, the NEPSTP activities were further complicated because the reactor power system was a Russian design. Therefore, this program considered not only the unique flight qualification issues associated with space nuclear power, but also with differences between Russian and United States flight qualification procedures. This paper presents an overview of the NEPSTP. The program goals, the proposed mission, the spacecraft, and the Topaz II space nuclear power system are described. The subject of flight qualification is examined and the inherent difficulties of qualifying a space reactor are described. The differences between United States and Russian flight qualification procedures are explored. A plan is then described that was developed to determine an appropriate flight qualification program for the Topaz II reactor to support a possible NEPSTP launch.

  7. New reactor technology: safety improvements in nuclear power systems.

    PubMed

    Corradini, M L

    2007-11-01

    Almost 450 nuclear power plants are currently operating throughout the world and supplying about 17% of the world's electricity. These plants perform safely, reliably, and have no free-release of byproducts to the environment. Given the current rate of growth in electricity demand and the ever growing concerns for the environment, nuclear power can only satisfy the need for electricity and other energy-intensive products if it can demonstrate (1) enhanced safety and system reliability, (2) minimal environmental impact via sustainable system designs, and (3) competitive economics. The U.S. Department of Energy with the international community has begun research on the next generation of nuclear energy systems that can be made available to the market by 2030 or earlier, and that can offer significant advances toward these challenging goals; in particular, six candidate reactor system designs have been identified. These future nuclear power systems will require advances in materials, reactor physics, as well as thermal-hydraulics to realize their full potential. However, all of these designs must demonstrate enhanced safety above and beyond current light water reactor systems if the next generation of nuclear power plants is to grow in number far beyond the current population. This paper reviews the advanced Generation-IV reactor systems and the key safety phenomena that must be considered to guarantee that enhanced safety can be assured in future nuclear reactor systems. PMID:18049233

  8. SCW Pressure-Channel Nuclear Reactor Some Design Features

    NASA Astrophysics Data System (ADS)

    Pioro, Igor L.; Khan, Mosin; Hopps, Victory; Jacobs, Chris; Patkunam, Ruban; Gopaul, Sandeep; Bakan, Kurtulus

    Concepts of nuclear reactors cooled with water at supercritical pressures were studied as early as the 1950s and 1960s in the USA and Russia. After a 30-year break, the idea of developing nuclear reactors cooled with SuperCritical Water (SCW) became attractive again as the ultimate development path for water cooling. The main objectives of using SCW in nuclear reactors are: 1) to increase the thermal efficiency of modern Nuclear Power Plants (NPPs) from 30-35% to about 45-48%, and 2) to decrease capital and operational costs and hence decrease electrical energy costs (˜1000 US/kW or even less). SCW NPPs will have much higher operating parameters compared to modern NPPs (pressure about 25 MPa and outlet temperature up to 625°C), and a simplified flow circuit, in which steam generators, steam dryers, steam separators, etc., can be eliminated. Also, higher SCW temperatures allow direct thermo-chemical production of hydrogen at low cost, due to increased reaction rates. Pressure-tube or pressure-channel SCW nuclear reactor concepts are being developed in Canada and Russia for some time. Some design features of the Canadian concept related to fuel channels are discussed in this paper. The main conclusion is that the development of SCW pressure-tube nuclear reactors is feasible and significant benefits can be expected over other thermal-energy systems.

  9. 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.

  10. Nuclear waste storage container with metal matrix

    DOEpatents

    Sump, Kenneth R.

    1978-01-01

    The invention relates to a storage container for high-level waste having a metal matrix for the high-level waste, thereby providing greater impact strength for the waste container and increasing heat transfer properties.

  11. Technology, Safety and Costs of Decommissioning Nuclear Reactors At Multiple-Reactor Stations

    SciTech Connect

    Wittenbrock, N. G.

    1982-01-01

    Safety and cost information is developed for the conceptual decommissioning of large (1175-MWe) pressurized water reactors (PWRs) and large (1155-MWe) boiling water reactors {BWRs) at multiple-reactor stations. Three decommissioning alternatives are studied: DECON (immediate decontamination), SAFSTOR (safe storage followed by deferred decontamination), and ENTOMB (entombment). Safety and costs of decommissioning are estimated by determining the impact of probable features of multiple-reactor-station operation that are considered to be unavailable at a single-reactor station, and applying these estimated impacts to the decommissioning costs and radiation doses estimated in previous PWR and BWR decommissioning studies. The multiple-reactor-station features analyzed are: the use of interim onsite nuclear waste storage with later removal to an offsite nuclear waste disposal facility, the use of permanent onsite nuclear waste disposal, the dedication of the site to nuclear power generation, and the provision of centralized services. Five scenarios for decommissioning reactors at a multiple-reactor station are investigated. The number of reactors on a site is assumed to be either four or ten; nuclear waste disposal is varied between immediate offsite disposal, interim onsite storage, and immediate onsite disposal. It is assumed that the decommissioned reactors are not replaced in one scenario but are replaced in the other scenarios. Centralized service facilities are provided in two scenarios but are not provided in the other three. Decommissioning of a PWR or a BWR at a multiple-reactor station probably will be less costly and result in lower radiation doses than decommissioning an identical reactor at a single-reactor station. Regardless of whether the light water reactor being decommissioned is at a single- or multiple-reactor station: • the estimated occupational radiation dose for decommissioning an LWR is lowest for SAFSTOR and highest for DECON • the estimated

  12. 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.

  13. Risk assessment methodologies for nuclear weapons compared to risk assessment methodologies for nuclear reactors

    SciTech Connect

    Benjamin, A.S.

    1994-12-31

    There are major differences between the safety principles for nuclear weapons and for nuclear reactors. For example, a principal concern for nuclear weapons is to prevent electrical energy from reaching the nuclear package during accidents produced by crashes, fires, and other hazards, whereas the foremost concern for nuclear reactors is to maintain coolant around the core in the event of certain system failures. Not surprisingly, new methods have had to be developed to assess the risk from nuclear weapons. These include fault tree transformations that accommodate time dependencies, thermal and structural analysis techniques that are fast and unconditionally stable, and Monte-Carlo-based sampling methods that incorporate intelligent searching. This paper provides an overview of the new methods for nuclear weapons, compares them with existing methods for nuclear reactors, identifies some of their dual-use characteristics, and discusses ongoing developmental activities.

  14. Multiscale Methods for Nuclear Reactor Analysis

    NASA Astrophysics Data System (ADS)

    Collins, Benjamin S.

    The ability to accurately predict local pin powers in nuclear reactors is necessary to understand the mechanisms that cause fuel pin failure during steady state and transient operation. In the research presented here, methods are developed to improve the local solution using high order methods with boundary conditions from a low order global solution. Several different core configurations were tested to determine the improvement in the local pin powers compared to the standard techniques, that use diffusion theory and pin power reconstruction (PPR). Two different multiscale methods were developed and analyzed; the post-refinement multiscale method and the embedded multiscale method. The post-refinement multiscale methods use the global solution to determine boundary conditions for the local solution. The local solution is solved using either a fixed boundary source or an albedo boundary condition; this solution is "post-refinement" and thus has no impact on the global solution. The embedded multiscale method allows the local solver to change the global solution to provide an improved global and local solution. The post-refinement multiscale method is assessed using three core designs. When the local solution has more energy groups, the fixed source method has some difficulties near the interface: however the albedo method works well for all cases. In order to remedy the issue with boundary condition errors for the fixed source method, a buffer region is used to act as a filter, which decreases the sensitivity of the solution to the boundary condition. Both the albedo and fixed source methods benefit from the use of a buffer region. Unlike the post-refinement method, the embedded multiscale method alters the global solution. The ability to change the global solution allows for refinement in areas where the errors in the few group nodal diffusion are typically large. The embedded method is shown to improve the global solution when it is applied to a MOX/LEU assembly

  15. Nuclear data requirements for fission reactor neutronics calculations.

    SciTech Connect

    Finck, P.

    1998-06-29

    The paper discusses current European nuclear data measurement and evaluation requirements for fission reactor technology applications and problems involved in meeting the requirements. Reference is made to the NEA High Priority Nuclear Data Request List and to the production of the new JEFF-3 library of evaluated nuclear data. There are requirements for both differential (or basic) nuclear data measurements and for different types of integral measurement critical facility measurements and isotopic sample irradiation measurements. Cross-section adjustment procedures are being used to take into account the simpler types of integral measurement, and to define accuracy needs for evaluated nuclear data.

  16. 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.

  17. Glassy materials investigated for nuclear reactor applications

    NASA Technical Reports Server (NTRS)

    Lynch, E. D.

    1968-01-01

    Studies determine the feasibility of preparing fuel-bearing glasses and glasses bearing neutron-absorbing materials for use as crystalline fuel and control rods for reactors. Properties investigated were devitrification resistance, urania solubility, and density.

  18. Development of an Inspection System for the Reactor Vessel/Containment Vessel of the PRISM and SAFR Liquid Metal Reactors

    SciTech Connect

    1989-02-01

    The integrity of the reactor vessel is of utmost importance in both the PRISM and SAFR concepts. The reactor vessel operates at elevated temperatures and contains molten liquid sodium. To ensure safe operation of the reactor, a periodic, visual inspection of the walls of the containment vessel is required by ASME specifications. This inspection would be conducted during a time when the reactor is shut down for refueling or maintenance. Nuclear Systems Associates, Inc. (NSA) was issued a PRDA contract by the Department of Energy to design, develop, and test a Closed Circuit Television (CCTV) camera system. The purpose of the system is to inspect the welds and wall surface of the Reactor Vessel/Container Vessel for both the PRISM and SAFR type reactors. The system was designed to function at the reactor's normal shutdown temperature, and provide a clear indication of flaws in the wall's weld seams and any cracks that might develop. The project was performed in three phases. The first phase concentrated the efforts on producing a compact camera system with the required resolution, self -contained lighting, and remote control focus and viewing angle. The proposed camera was then tested in a vessel mock-up and found to perform to required specifications at room (cold) temperatures. Simulated flaws, cracks, and a sodium leak were observed with required clarity on both a commercial and blackened stainless steel surfaces. The camera was tested with a single clear glass dome, a single coated glass dome, and a dual-glass dome covering the camera lens and mirror. The second phase of the project was conducted in two parts. The first part involved testing the vessel mock-up at elevated temperatures to verify that the required temperatures can be obtained. The mock-up was constructed with imbedded heaters and both control and indicating thermocouples. Stable operating temperatures over 400°F were achieved. During the second part of this phase, the camera was inserted into the

  19. Nuclear-reactor accidents: Chernobyl, TMI, and Windscale. January 1974-September 1988 (Citations from Pollution Abstracts). Report for January 1974-September 1988

    SciTech Connect

    Not Available

    1988-11-01

    This bibliography contains citations concerning studies and measurements of the radiological consequences of nuclear-reactor accidents. The citations cover specifically the Chernobyl reactor in the USSR, the Three Mile Island (TMI) reactor in the US, and the Windscale reactor in the UK. Included are detection and monitoring of the fallout, the resultant runoff into rivers, lakes, and the sea, the radiation effects on people, and the transfrontier radioactive contamination of the environment. (Contains 105 citations fully indexed and including a title list.)

  20. 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

  1. Second generation Research Reactor Fuel Container (RRFC-II).

    SciTech Connect

    Abhold, M. E.; Baker, M. C.; Bourret, S. C.; Harker, W. C.; Pelowitz, D. G.; Polk, P. J.

    2001-01-01

    The second generation Research Reactor Fuel Counter (RRFC-II) has been developed to measure the remaining {sup 235}U content in foreign spent Material Test Reactor (MTR)-type fuel being returned to the Westinghouse Savannah River Site (WSRS) for interim storage and subsequent disposal. The fuel to be measured started as fresh fuel nominally with 93% enriched Uraniuin alloyed with A1 clad in Al. The fuel was irradiated to levels of up to 65% burnup. The RRFC-II, which will be located in the L-Basin spent fuel pool, is intended to assay the {sup 235}U content using a combination of passive neutron coincidence counting, active neutron coincidence counting, and active-multiplicity analysis. Measurements will be done underwater, eliminating the need for costly and hazardous handling operations of spent fuel out of water. The underwater portion of the RRFC-II consists of a watertight stainless steel housing containing neutron and gamma detectors and a scanning active neutron source. The portion of the system that resides above water consists of data-processing electronics; electromechanical drive electronics; a computer to control the operation of the counter, to collect, and to analyze data; and a touch screen interface located at the equipment rack. The RRFC-II is an improved version of the Los Alamos-designed RRFC already installed in the SRS Receipts Basin for Offsite Fuel. The RRFC-II has been fabricated and is scheduled for installation in late FY 2001 pending acceptance testing by Savannah River Site personnel.

  2. Two-billion-year-old nuclear reactors: Nature goes fission

    SciTech Connect

    Curtis, D.B.

    1992-12-31

    Once it was thought that the isotopic composition of natural uranium was invariant. It was thus surprising in 1972 when French scientists observed small but significant deficiencies of the minor isotope {sup 235}U in uranium ore. Subsequent investigations traced the isotopically anomalous material to the Oklo mine in the African Republic of Gabon. In the mine, cubic-dekametre-sized pods of rock were found to contain extraordinary concentrations of uranium, as much as 65%, with as little as half the normal isotopic abundance of {sup 235}U. In these rocks, neodymium was found to be deficient in the premordial isotope {sup 142}Nd and enriched in the fission-produced isotopes {sup 143-150}Nd. The presence of fission products was unambiguous evidence that the {sup 235}U deficiencies were the result of sustained nuclear fission. Within the heart of the natural reactors, the fission densities were on the order of 10{sup 20} fissions/cm{sup 3}, producing hundreds of megajoules of energy and tens of microwatts of power per gram of rock. Nature had forestalled man`s great discovery of energy production by nuclear fission.

  3. Nuclear Ribonucleoprotein Complexes Containing Polyadenylate from Mouse Ascites Cells

    PubMed Central

    Quinlan, Thomas J.; Billings, Peter B.; Martin, Terence E.

    1974-01-01

    Nuclear poly(A)-containing RNA of mouse ascites cells can be extracted in the form of 15-17S ribonucleoprotein complexes under conditions in which the bulk of the heterogeneous nuclear RNA is released as 30S complexes. The poly(A)-containing fraction of nuclear extracts has been resolved into two distinct components, 15 and 17 S; neither contains the two polypeptides of 30S ribonucleoprotein. The 17S particle contains approximately six polypeptide species of molecular masses 17,000-30,000 daltons. The 15S complex has four distinct polypeptides of higher molecular mass, including a prominent 80,000-dalton species. PMID:4368966

  4. Nuclear safety criteria and specifications for space nuclear reactors

    SciTech Connect

    Not Available

    1982-08-01

    The purpose of this document is to define safety criteria which must be met to implement US safety policy for space fission reactors. These criteria provide the bases for decisions on the acceptability of specific mission and reactor design proposals. (JDH)

  5. Dual annular rotating "windowed" nuclear reflector reactor control system

    DOEpatents

    Jacox, Michael G.; Drexler, Robert L.; Hunt, Robert N. M.; Lake, James A.

    1994-01-01

    A nuclear reactor control system is provided in a nuclear reactor having a core operating in the fast neutron energy spectrum where criticality control is achieved by neutron leakage. The control system includes dual annular, rotatable reflector rings. There are two reflector rings: an inner reflector ring and an outer reflector ring. The reflectors are concentrically assembled, surround the reactor core, and each reflector ring includes a plurality of openings. The openings in each ring are capable of being aligned or non-aligned with each other. Independent driving means for each of the annular reflector rings is provided so that reactor criticality can be initiated and controlled by rotation of either reflector ring such that the extent of alignment of the openings in each ring controls the reflection of neutrons from the core.

  6. Simulation of the control board of an experimental nuclear reactor

    SciTech Connect

    Mackieh, A.; Cilingir, C.; Alten, S.

    1995-12-31

    This study is performed as a part of a bigger project to analyze human factors in operations of an experimental nuclear reactor. In this context, the control board of the 10-kW university training reactor (UTR-10) located at Iowa State University, Ames, Iowa, was simulated. The software was developed in the ergonomics laboratory of the Middle East Technical University (METU) by using an object-oriented programming language (Visual Basic for IBM-compatible personal computers).

  7. Nuclear Technology Series. Nuclear Reactor (Plant) Operator Trainee. A Suggested Program Planning Guide. Revised June 80.

    ERIC Educational Resources Information Center

    Center for Occupational Research and Development, Inc., Waco, TX.

    This program planning guide for a two-year postsecondary nuclear reactor (plant) operator trainee program is designed for use with courses 1-16 of thirty-five in the Nuclear Technology Series. The purpose of the guide is to describe the nuclear power field and its job categories for specialists, technicians and operators; and to assist planners,…

  8. NASA-Lewis in-pile high temperature reactor meltdown containment experiment

    NASA Astrophysics Data System (ADS)

    Rom, Frank E.; Finnegan, Patrick M.; Gumto, Klaus H.

    1995-01-01

    The purpose of the in-pile meltdown experiment was to demonstrate the feasibility of a proposed concept for preventing melt-through of the containment vessel during and after the meltdown of a nuclear reactor. The concept was modeled and tested in the NASA Plum Brook Reactor Facility using fission heating to simulate fission product heating. In the concept a 38 mm (1.5 in) thick depleted UO2 pellet liner was placed on the inner surface of the stainless steel containment vessel. The high density, high melting point, and poor thermal conductivity of the UO2 pellet liner prevents molten materials from reaching the containment vessel walls. In turn, this results in a substantial increase in temperature of the molten pool. The high pool temperature causes the heat generating fission products to more quickly vaporize and condense throughout cooler regions within the containment volume. (At its melting point UO2 vaporizes at the rate of about 0.14 mm/s (0.006 in/s). Most fission products and their compounds vaporize even more rapidly.) The net effect is to reduce by orders of magnitude the heat flux through the containment vessel wall. A detailed physical description of the meltdown and containment process as derived from interpretation of the experimental measurements is presented. After meltdown, temperatures of the Mo-UO2 core exceeded 3393 K (5647 °F), which surpassed the melting point of UO2 by 503 K (905 °F) and Mo by 313 K (563 °F). Temperatures at the midpoint of the depleted UO2 liner never exceeded 1300 K (1934 °F). No molten material penetrated as much as halfway through the UO2 barrier. The peak surface temperature of the stainless steel containment vessel never exceeded 928 K (1210 °F).

  9. Nuclear Resonance Fluorescence and Isotopic Mapping of Containers

    SciTech Connect

    Johnson, M S; McNabb, D P

    2008-08-08

    National security programs have expressed interest in developing systems to isotopically map shipping containers, fuel assemblies, and waste barrels for various materials including special nuclear material (SNM). Current radiographic systems offer little more than an ambiguous density silhouette of a container's contents. In this paper we will present a system being developed at LLNL to isotopically map containers using the nuclear resonance fluorescence (NRF) method. Recent experimental measurements on NRF strengths in SNM are discussed.

  10. Raytheon explores thorium for next generation nuclear reactor

    SciTech Connect

    Crawford, M.

    1994-03-08

    Few new orders for nuclear power plants have been placed anywhere in the world in the last 20 years, but that is not discouraging Raytheon Engineers Constructors from making plans to explore new light water reactor technologies for commercial markets. The Lexington, Mass.-based company, which has extensive experience in nuclear power engineering and construction, has a vision for the light water reactor of the future - one that is based on the use of thorium-232, an element that decays over several steps to uranium-233. The use of thorium and a small amount of uranium that is 20 percent enriched is seen as providing operational, environmental, and safety advantages over reactors using the standard fuel mixture of uranium-238 and enriched uranium-235. According to Raytheon, the system could improve the economics of some reactors' operations by reducing fuel costs and lowering related waste volumes. At the same time, reactor safety could be improved by simpler control rod systems and the absence from reactor coolant of corrosive boric acid, which is used to slow neutrons in order to enhance reactions. Using thorium is also attractive because more of the fuel is burned up by the reactor, an estimated 12 percent as compared to about 4 percent for U-235. However, the technology's greatest attraction may well be its implications for nuclear proliferation. Growing plutonium inventories embedded in spent fuel rods from light water reactors have sparked concern worldwide. But according to Raytheon, using a thorium-based fuel core would alleviate this concern because it would produce only small quantities of plutonium. A thorium-based fuel system would produce 12 kilograms of plutonium over a decade versus 2,235 kilograms for an equivalent reactor operating with conventional uranium fuel.

  11. 10 CFR 73.58 - Safety/security interface requirements for nuclear power reactors.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 false Safety/security interface requirements for nuclear power reactors. 73.58 Section 73.58 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PHYSICAL PROTECTION OF... requirements for nuclear power reactors. (a) Each operating nuclear power reactor licensee with a...

  12. 10 CFR 73.58 - Safety/security interface requirements for nuclear power reactors.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 10 Energy 2 2011-01-01 2011-01-01 false Safety/security interface requirements for nuclear power reactors. 73.58 Section 73.58 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PHYSICAL PROTECTION OF... requirements for nuclear power reactors. (a) Each operating nuclear power reactor licensee with a...

  13. A Compact Nuclear Fusion Reactor for Space Flights

    NASA Astrophysics Data System (ADS)

    Nastoyashchiy, Anatoly F.

    2006-05-01

    A small-scale nuclear fusion reactor is suggested based on the concepts of plasma confinement (with a high pressure gas) which have been patented by the author. The reactor considered can be used as a power setup in space flights. Among the advantages of this reactor is the use of a D3He fuel mixture which at burning gives main reactor products — charged particles. The energy balance considerably improves, as synchrotron radiation turn out "captured" in the plasma volume, and dangerous, in the case of classical magnetic confinement, instabilities in the direct current magnetic field configuration proposed do not exist. As a result, the reactor sizes are quite suitable (of the order of several meters). A possibility of making reactive thrust due to employment of ejection of multiply charged ions formed at injection of pellets from some adequate substance into the hot plasma center is considered.

  14. Analysis of Nuclear Reactor Background Radiation for Neutrino Experiments

    NASA Astrophysics Data System (ADS)

    Leblanc, Ricky; Blackmon, J. C.; Rasco, B. C.; Mumm, H. P.; mTC; NuLat Collaboration

    2015-10-01

    Prior measurements of reactor antineutrinos have found a lower flux than expected. Precision measurements of antineutrino energy spectra are important for understanding the anomaly, reactor safeguards, and nuclear nonproliferation. Antineutrino detector designs rely on good characterization of gamma-ray and neutron backgrounds near the reactor core. To study the gamma-ray background at the NIST research reactor, spectra were collected using a 6.25 cm diameter × 5.5 cm germanium detector. We analyzed the measured spectra using simulations of the detector response using the GEANT4 toolkit to determine background fluxes and build a background model that will be used to understand shielding requirements and the impact of backgrounds on potential short-baseline reactor antineutrino studies at NIST. This work supported by the National Science Foundation and LSU.

  15. Shielding considerations for advanced space nuclear reactor systems

    SciTech Connect

    Angelo, J.P. Jr.; Buden, D.

    1982-01-01

    To meet the anticipated future space power needs, the Los Alamos National Laboratory is developing components for a compact, 100 kW/sub e/-class heat pipe nuclear reactor. The reactor uses uranium dioxide (UO/sub 2/) as its fuel, and is designed to operate around 1500 k. Heat pipes are used to remove thermal energy from the core without the use of pumps or compressors. The reactor heat pipes transfer mal energy to thermoelectric conversion elements that are advanced versions of the converters used on the enormously successful Voyager missions to the outer planets. Advanced versions of this heat pipe reactor could also be used to provide megawatt-level power plants. The paper reviews the status of this advanced heat pipe reactor and explores the radiation environments and shielding requirements for representative manned and unmanned applications.

  16. Low exchange element for nuclear reactor

    DOEpatents

    Brogli, Rudolf H.; Shamasunder, Bangalore I.; Seth, Shivaji S.

    1985-01-01

    A flow exchange element is presented which lowers temperature gradients in fuel elements and reduces maximum local temperature within high temperature gas-cooled reactors. The flow exchange element is inserted within a column of fuel elements where it serves to redirect coolant flow. Coolant which has been flowing in a hotter region of the column is redirected to a cooler region, and coolant which has been flowing in the cooler region of the column is redirected to the hotter region. The safety, efficiency, and longevity of the high temperature gas-cooled reactor is thereby enhanced.

  17. Spectral structure of electron antineutrinos from nuclear reactors.

    PubMed

    Dwyer, D A; Langford, T J

    2015-01-01

    Recent measurements of the positron energy spectrum obtained from inverse beta decay interactions of reactor electron antineutrinos show an excess in the 4 to 6 MeV region relative to current predictions. First-principles calculations of fission and beta decay processes within a typical pressurized water reactor core identify prominent fission daughter isotopes as a possible origin for this excess. These calculations also predict percent-level substructures in the antineutrino spectrum due to Coulomb effects in beta decay. Precise measurement of these substructures can elucidate the nuclear processes occurring within reactors. These substructures can be a systematic issue for measurements utilizing the detailed spectral shape. PMID:25615462

  18. Small nuclear reactors for power and process heat

    SciTech Connect

    Olds, F.C.

    1984-11-01

    There are a number of small reactor programs around the world, with a sizable variety of designs ranging from 0.2 MWt heat-only units to 440-MWe agro-nuclear complexes. Some are in operation, some are proposed. There are light and heavy water reactors, a liquid metal cooled unit with integral hot reprocessing/refabrication, and an HTGR built with 100-MWe modules. There are several inherently safe designs. Shop fabrication is emphasized for cost and quality control, and close-in siting is practiced here and there. The feasibility of using small reactors differs markedly, depending on type of service and on country-specific factors.

  19. Emergency heat removal system for a nuclear reactor

    DOEpatents

    Dunckel, Thomas L.

    1976-01-01

    A heat removal system for nuclear reactors serving as a supplement to an Emergency Core Cooling System (ECCS) during a Loss of Coolant Accident (LOCA) comprises a plurality of heat pipes having one end in heat transfer relationship with either the reactor pressure vessel, the core support grid structure or other in-core components and the opposite end located in heat transfer relationship with a heat exchanger having heat transfer fluid therein. The heat exchanger is located external to the pressure vessel whereby excessive core heat is transferred from the above reactor components and dissipated within the heat exchanger fluid.

  20. Strengthening the nuclear-reactor fuel cycle against proliferation

    SciTech Connect

    Travelli, A.; Snelgrove, J.; Persiani, P.

    1992-12-31

    Argonne National Laboratory (ANL) conducts several research programs that serve to reduce the risks of fissile-material diversion from the nuclear-reactor fuel cycle. The objectives are to provide economical and efficient neutron or power generation with the minimum of inherent risks, and to further minimize risks by utilizing sophisticated techniques to detect attempts at material diversion. This paper will discuss the Reduced Enrichment Research and Test Reactor (RERTR) Program, the Isotope Correlation Technique (ICT), and Proliferation-Resistant Closed-Cycle Reactors. The first two are sponsored by the DOE Office of Arms Control and Nonproliferation.

  1. Method of determining a content of a nuclear waste container

    DOEpatents

    Bernardi, Richard T.; Entwistle, David

    2003-04-22

    A method and apparatus are provided for identifying contents of a nuclear waste container. The method includes the steps of forming an image of the contents of the container using digital radiography, visually comparing contents of the image with expected contents of the container and performing computer tomography on the container when the visual inspection reveals an inconsistency between the contents of the image and the expected contents of the container.

  2. Support vector machines for nuclear reactor state estimation

    SciTech Connect

    Zavaljevski, N.; Gross, K. C.

    2000-02-14

    Validation of nuclear power reactor signals is often performed by comparing signal prototypes with the actual reactor signals. The signal prototypes are often computed based on empirical data. The implementation of an estimation algorithm which can make predictions on limited data is an important issue. A new machine learning algorithm called support vector machines (SVMS) recently developed by Vladimir Vapnik and his coworkers enables a high level of generalization with finite high-dimensional data. The improved generalization in comparison with standard methods like neural networks is due mainly to the following characteristics of the method. The input data space is transformed into a high-dimensional feature space using a kernel function, and the learning problem is formulated as a convex quadratic programming problem with a unique solution. In this paper the authors have applied the SVM method for data-based state estimation in nuclear power reactors. In particular, they implemented and tested kernels developed at Argonne National Laboratory for the Multivariate State Estimation Technique (MSET), a nonlinear, nonparametric estimation technique with a wide range of applications in nuclear reactors. The methodology has been applied to three data sets from experimental and commercial nuclear power reactor applications. The results are promising. The combination of MSET kernels with the SVM method has better noise reduction and generalization properties than the standard MSET algorithm.

  3. Robotic system for remote maintenance of a pulsed nuclear reactor

    SciTech Connect

    Thunborg, S.

    1986-01-01

    Guidelines recently established for occupational radiation exposure specify that exposure should be as low as reasonably achievable. In conformance with these guidelines, SNL has developed a remote maintenance robot (RMR) system for use in the Sandia Pulse Reactor III (SPR III) facility. The RMR should reduce occupational radiation exposure by a factor of 4 and decrease reactor downtime. Other goals include developing a technology base for a more advanced pulse reactor and for the nuclear fuel cycle programs of the US Department of Energy and US Nuclear Regulatory Commission. The RMR has five major subsystems: (a) a chain-driven cart to bring the system into the reactor room; (b) a Puma 560 robot to perform dextrous operations; (c) a programmable turntable to orient the robot to any of the reactor's four sides; (d) a programmable overhead hoist for lifting components weighing up to 400 lb onto or off of the reactor; and (e) a supervisory control console for the system operator. Figure 1 is a schematic diagram of the turntable, hoist, and robot system in position around the SPR III reactor.

  4. Nuclear Safeguards Considerations For The Pebble Bed Modular Reactor (PBMR)

    SciTech Connect

    Phillip Casey Durst; David Beddingfield; Brian Boyer; Robert Bean; Michael Collins; Michael Ehinger; David Hanks; David L. Moses; Lee Refalo

    2009-10-01

    High temperature reactors (HTRs) have been considered since the 1940s, and have been constructed and demonstrated in the United Kingdom (Dragon), United States (Peach Bottom and Fort Saint Vrain), Japan (HTTR), Germany (AVR and THTR-300), and have been the subject of conceptual studies in Russia (VGM). The attraction to these reactors is that they can use a variety of reactor fuels, including abundant thorium, which upon reprocessing of the spent fuel can produce fissile U-233. Hence, they could extend the stocks of available uranium, provided the fuel is reprocessed. Another attractive attribute is that HTRs typically operate at a much higher temperature than conventional light water reactors (LWRs), because of the use of pyrolytic carbon and silicon carbide coated (TRISO) fuel particles embedded in ceramic graphite. Rather than simply discharge most of the unused heat from the working fluid in the power plant to the environment, engineers have been designing reactors for 40 years to recover this heat and make it available for district heating or chemical conversion plants. Demonstrating high-temperature nuclear energy conversion was the purpose behind Fort Saint Vrain in the United States, THTR-300 in Germany, HTTR in Japan, and HTR-10 and HTR-PM, being built in China. This resulted in nuclear reactors at least 30% or more thermodynamically efficient than conventional LWRs, especially if the waste heat can be effectively utilized in chemical processing plants. A modern variant of high temperature reactors is the Pebble Bed Modular Reactor (PBMR). Originally developed in the United States and Germany, it is now being redesigned and marketed by the Republic of South Africa and China. The team examined historical high temperature and high temperature gas reactors (HTR and HTGR) and reviewed safeguards considerations for this reactor. The following is a preliminary report on this topic prepared under the ASA-100 Advanced Safeguards Project in support of the NNSA Next

  5. Nuclear reactor descriptions for space power systems analysis

    NASA Technical Reports Server (NTRS)

    Mccauley, E. W.; Brown, N. J.

    1972-01-01

    For the small, high performance reactors required for space electric applications, adequate neutronic analysis is of crucial importance, but in terms of computational time consumed, nuclear calculations probably yield the least amount of detail for mission analysis study. It has been found possible, after generation of only a few designs of a reactor family in elaborate thermomechanical and nuclear detail to use simple curve fitting techniques to assure desired neutronic performance while still performing the thermomechanical analysis in explicit detail. The resulting speed-up in computation time permits a broad detailed examination of constraints by the mission analyst.

  6. NUCLEAR REACTOR SLUG PROVIDED WITH THERMOCOUPLE

    DOEpatents

    Kanne, W.R.

    1958-10-14

    A temperature measuring apparatus is described for use in a reactor. In this invention a cylindrlcal fuel slug is provided with an axial bore in which is disposed a thermocouple. The lead wires extend to a remote indicating device which indicates the temperature in the fuel element measured by the thermocouple.

  7. Numerical simulation of gas dynamics and heat exchange tasks in fuel assemblies of the nuclear reactors

    SciTech Connect

    Zhuchenko, S. V.

    2014-11-12

    This report presents a PC-based program for solution gas dynamics and heat exchange mathematical tasks in fuel assemblies of the fast-neutron nuclear reactors. A fuel assembly consisting of bulk heat-generating elements, which are integrated together by the system of supply and pressure manifolds, is examined. Spherical heat-generating microelements, which contain nuclear fuel, are pulled into the heat-generating elements. Gaseous coolant proceed from supply manifolds to heat-generating elements, where it withdraws the nuclear reaction heat and assembles in pressure manifolds.

  8. Metallic Fast Reactor Fuel Fabrication for Global Nuclear Energy Partnership

    SciTech Connect

    Douglas E. Burkes; Randall S. Fielding; Douglas L. Porter

    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. Decommissioning of the secondary containment of the steam generating heavy water reactor at UKAEA-Winfrith

    SciTech Connect

    Miller, Keith; Cornell, Rowland; Parkinson, Steve; McIntyre, Kevin; Staples, Andy

    2007-07-01

    Available in abstract form only. Full text of publication follows: The Winfrith SGHWR was a prototype nuclear power plant operated for 23 years by the United Kingdom Atomic Energy Authority (UKAEA) until 1990 when it was shut down permanently. The current Stage 1 decommissioning contract is part of a multi-stage strategy. It involves the removal of all the ancillary plant and equipment in the secondary containment and non-containment areas ahead of a series of contracts for the decommissioning of the primary containment, the reactor core and demolition of the building and all remaining facilities. As an outcome of a competitive tending process, the Stage 1 decommissioning contract was awarded to NUKEM with operations commencing in April 2005. The decommissioning processes involved with these plant items will be described with some emphasis of the establishment of multiple work-fronts for the production, recovery, treatment and disposal of mainly tritium-contaminated waste arising from its contact with the direct cycle reactor coolant. The means of size reduction of a variety of large, heavy and complex items of plant made from a range of materials will also be described with some emphasis on the control of fumes during hot cutting operations and establishing effective containments within a larger secondary containment structure. Disposal of these wastes in a timely and cost-effective manner is a major challenge facing the decommissioning team and has required the development of a highly efficient means of packing the resultant materials into mainly one-third height ISO containers for disposal as LLW. Details of the quantities of LLW and exempt wastes handled during this process will be given with a commentary about the difficulty in segregating these two waste streams efficiently. (authors)

  10. Validation of CONTAIN-LMR code for accident analysis of sodium-cooled fast reactor containments

    SciTech Connect

    Gordeev, S.; Hering, W.; Schikorr, M.; Stieglitz, R.

    2012-07-01

    CONTAIN-LMR 1 is an analytical tool for the containment performance of sodium cooled fast reactors. In this code, the modelling for the sodium fire is included: the oxygen diffusion model for the sodium pool fire, and the liquid droplet model for the sodium spray fire. CONTAIN-LMR is also able to model the interaction of liquid sodium with concrete structure. It may be applicable to different concrete compositions. Testing and validation of these models will help to qualify the simulation results. Three experiments with sodium performed in the FAUNA facility at FZK have been used for the validation of CONTAIN-LMR. For pool fire tests, calculations have been performed with two models. The first model consists of one gas cell representing the volume of the burn compartment. The volume of the second model is subdivided into 32 coupled gas cells. The agreement between calculations and experimental data is acceptable. The detailed pool fire model shows less deviation from experiments. In the spray fire, the direct heating from the sodium burning in the media is dominant. Therefore, single cell modeling is enough to describe the phenomena. Calculation results have reasonable agreement with experimental data. Limitations of the implemented spray model can cause the overestimation of predicted pressure and temperature in the cell atmosphere. The ability of the CONTAIN-LMR to simulate the sodium pool fire accompanied by sodium-concrete reactions was tested using the experimental study of sodium-concrete interactions for construction concrete as well as for shielding concrete. The model provides a reasonably good representation of chemical processes during sodium-concrete interaction. The comparison of time-temperature profiles of sodium and concrete shows, that the model requires modifications for predictions of the test results. (authors)

  11. Investigation of Containment Flooding Strategy for Mark-III Nuclear Power Plant with MAAP4

    SciTech Connect

    Su Weinian; Wang, S.-J.; Chiang, S.-C

    2005-06-15

    Containment flooding is an important strategy for severe accident management of a conventional boiling water reactor (BWR) system. The purpose of this work is to investigate the containment flooding strategy of the Mark-III system after a reactor pressure vessel (RPV) breach. The Kuosheng Power Plant is a typical BWR-6 nuclear power plant (NPP) with Mark-III containment. The Severe Accident Management Guideline (SAMG) of the Kuosheng NPP has been developed based on the BWR Owners Group (BWROG) Emergency Procedure and Severe Accident Guidelines, Rev. 2. Therefore, the Kuosheng NPP is selected as the plant for study, and the MAAP4 code is chosen as the tool for analysis. A postulated specific station blackout sequence for the Kuosheng NPP is cited as a reference case for this analysis. Because of the design features of Mark-III containment, the debris in the reactor cavity may not be submerged after an RPV breach when one follows the containment flooding strategy as suggested in the BWROG generic guideline, and the containment integrity could be challenged eventually. A more specific containment flooding strategy with drywell venting after an RPV breach is investigated, and a more stable plant condition is achieved with this strategy. Accordingly, the containment flooding strategy after an RPV breach will be modified for the Kuosheng SAMG, and these results are applicable to typical Mark-III plants with drywell vent path.

  12. University Reactor Conversion Lessons Learned Workshop for Texas A&M University Nuclear Science Center Reactor

    SciTech Connect

    Eric C. Woolstenhulme; Dana M. Meyer

    2007-04-01

    The objectives of this meeting were to capture the observations, insights, issues, concerns, and ideas of those involved in the Texas A&M University Nuclear Science Center (TAMU NSC) TRIGA Reactor Conversion so that future efforts can be conducted with greater effectiveness, efficiency, and with fewer challenges. This workshop was held in conjunction with a similar workshop for the University of Florida Reactor Conversion. Some of the generic lessons from that workshop are included in this report for completeness.

  13. 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.

  14. Global risk of radioactive fallout after major nuclear reactor accidents

    NASA Astrophysics Data System (ADS)

    Lelieveld, J.; Kunkel, D.; Lawrence, M. G.

    2012-05-01

    Major reactor accidents of nuclear power plants are rare, yet the consequences are catastrophic. But what is meant by "rare"? And what can be learned from the Chernobyl and Fukushima incidents? Here we assess the cumulative, global risk of exposure to radioactivity due to atmospheric dispersion of gases and particles following severe nuclear accidents (the most severe ones on the International Nuclear Event Scale, INES 7), using particulate 137Cs and gaseous 131I as proxies for the fallout. Our results indicate that previously the occurrence of INES 7 major accidents and the risks of radioactive contamination have been underestimated. Using a global model of the atmosphere we compute that on average, in the event of a major reactor accident of any nuclear power plant worldwide, more than 90% of emitted 137Cs would be transported beyond 50 km and about 50% beyond 1000 km distance before being deposited. This corroborates that such accidents have large-scale and trans-boundary impacts. Although the emission strengths and atmospheric removal processes of 137Cs and 131I are quite different, the radioactive contamination patterns over land and the human exposure due to deposition are computed to be similar. High human exposure risks occur around reactors in densely populated regions, notably in West Europe and South Asia, where a major reactor accident can subject around 30 million people to radioactive contamination. The recent decision by Germany to phase out its nuclear reactors will reduce the national risk, though a large risk will still remain from the reactors in neighbouring countries.

  15. Seal welded cast iron nuclear waste container

    DOEpatents

    Filippi, Arthur M.; Sprecace, Richard P.

    1987-01-01

    This invention identifies methods and articles designed to circumvent metallurgical problems associated with hermetically closing an all cast iron nuclear waste package by welding. It involves welding nickel-carbon alloy inserts which are bonded to the mating plug and main body components of the package. The welding inserts might be bonded in place during casting of the package components. When the waste package closure weld is made, the most severe thermal effects of the process are restricted to the nickel-carbon insert material which is far better able to accommodate them than is cast iron. Use of nickel-carbon weld inserts should eliminate any need for pre-weld and post-weld heat treatments which are a problem to apply to nuclear waste packages. Although the waste package closure weld approach described results in a dissimilar metal combination, the relative surface area of nickel-to-iron, their electrochemical relationship, and the presence of graphite in both materials will act to prevent any galvanic corrosion problem.

  16. Development of Improved Burnable Poisons for Commercial Nuclear Power Reactors

    SciTech Connect

    Renier, J.A.

    2002-04-17

    Burnable poisons are used in all modern nuclear reactors to permit higher loading of fuel without the necessity of an overly large control rod system. This not only permits a longer core life but can also be used to level the power distribution. Commercial nuclear reactors commonly use B{sub 4}C in separate non-fueled rods and more recently, zirconium boride coatings on the fuel pellets or gadolinium oxide mixed with the fuel. Although the advantages are great, there are problems with using these materials. Boron, which is an effective neutron absorber, transmutes to lithium and helium upon absorption of a neutron. Helium is insoluble and is eventually released to the interior of the fuel rod, where it produces an internal pressure. When sufficiently high, this pressure stress could cause separation of the cladding from the fuel, causing overly high centerline temperatures. Gadolinium has several very strongly absorbing isotopes, but not all have large cross sections and result in residual burnable poison reactivity worth at the end of the fuel life. Even if the amount of this residual absorber is small and the penalty in operation small, the cost of this penalty, even if only several days, can be very high. The objective of this investigation was to study the performance of single isotopes in order to reduce the residual negative reactivity left over at the end of the fuel cycle. Since the behavior of burnable poisons can be strongly influenced by their configuration, four forms for the absorbers were studied: homogeneously mixed with the fuel, mixed with only the outer one-third of the fuel pellet, coated on the perimeter of the fuel pellets, and alloyed with the cladding. In addition, the numbers of fuel rods containing burnable poison were chosen as 8, 16, 64, and 104. Other configurations were chosen for a few special cases. An enrichment of 4.5 wt% {sup 235}U was chosen for most cases for study in order to achieve a 4-year fuel cycle. A standard pressurized

  17. Production capabilities in US nuclear reactors for medical radioisotopes

    SciTech Connect

    Mirzadeh, S.; Callahan, A.P.; Knapp, F.F. Jr.; Schenter, R.E.

    1992-11-01

    The availability of reactor-produced radioisotopes in the United States for use in medical research and nuclear medicine has traditionally depended on facilities which are an integral part of the US national laboratories and a few reactors at universities. One exception is the reactor in Sterling Forest, New York, originally operated as part of the Cintichem (Union Carbide) system, which is currently in the process of permanent shutdown. Since there are no industry-run reactors in the US, the national laboratories and universities thus play a critical role in providing reactor-produced radioisotopes for medical research and clinical use. The goal of this survey is to provide a comprehensive summary of these production capabilities. With the temporary shutdown of the Oak Ridge National Laboratory (ORNL) High Flux Isotope Reactor (HFIR) in November 1986, the radioisotopes required for DOE-supported radionuclide generators were made available at the Brookhaven National Laboratory (BNL) High Flux Beam Reactor (HFBR). In March 1988, however, the HFBR was temporarily shut down which forced investigators to look at other reactors for production of the radioisotopes. During this period the Missouri University Research Reactor (MURR) played an important role in providing these services. The HFIR resumed routine operation in July 1990 at 85 MW power, and the HFBR resumed operation in June 1991, at 30 MW power. At the time of the HFBR shutdown, there was no available comprehensive overview which could provide information on status of the reactors operating in the US and their capabilities for radioisotope production. The obvious need for a useful overview was thus the impetus for preparing this survey, which would provide an up-to-date summary of those reactors available in the US at both the DOE-funded national laboratories and at US universities where service irradiations are currently or expected to be conducted.

  18. Experiments to investigate direct containment heating phenomena with scaled models of the Surry Nuclear Power Plant

    SciTech Connect

    Blanchat, T.K.; Allen, M.D.; Pilch, M.M.; Nichols, R.T.

    1994-06-01

    The Containment Technology Test Facility (CTTF) and the Surtsey Test Facility at Sandia National Laboratories are used to perform scaled experiments that simulate High Pressure Melt Ejection accidents in a nuclear power plant (NPP). These experiments are designed to investigate the effects of direct containment heating (DCH) phenomena on the containment load. High-temperature, chemically reactive melt (thermite) is ejected by high-pressure steam into a scale model of a reactor cavity. Debris is entrained by the steam blowdown into a containment model where specific phenomena, such as the effect of subcompartment structures, prototypic air/steam/hydrogen atmospheres, and hydrogen generation and combustion, can be studied. Four Integral Effects Tests (IETs) have been performed with scale models of the Surry NPP to investigate DCH phenomena. The 1/61{sup th} scale Integral Effects Tests (IET-9, IET-10, and IET-11) were conducted in CTRF, which is a 1/6{sup th} scale model of the Surry reactor containment building (RCB). The 1/10{sup th} scale IET test (IET-12) was performed in the Surtsey vessel, which had been configured as a 1/10{sup th} scale Surry RCB. Scale models were constructed in each of the facilities of the Surry structures, including the reactor pressure vessel, reactor support skirt, control rod drive missile shield, biological shield wall, cavity, instrument tunnel, residual heat removal platform and heat exchangers, seal table room and seal table, operating deck, and crane wall. This report describes these experiments and gives the results.

  19. 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.

  20. 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.

  1. Foundational development of an advanced nuclear reactor integrated safety code.

    SciTech Connect

    Clarno, Kevin; Lorber, Alfred Abraham; Pryor, Richard J.; Spotz, William F.; Schmidt, Rodney Cannon; Belcourt, Kenneth; Hooper, Russell Warren; Humphries, Larry LaRon

    2010-02-01

    This report describes the activities and results of a Sandia LDRD project whose objective was to develop and demonstrate foundational aspects of a next-generation nuclear reactor safety code that leverages advanced computational technology. The project scope was directed towards the systems-level modeling and simulation of an advanced, sodium cooled fast reactor, but the approach developed has a more general applicability. The major accomplishments of the LDRD are centered around the following two activities. (1) The development and testing of LIME, a Lightweight Integrating Multi-physics Environment for coupling codes that is designed to enable both 'legacy' and 'new' physics codes to be combined and strongly coupled using advanced nonlinear solution methods. (2) The development and initial demonstration of BRISC, a prototype next-generation nuclear reactor integrated safety code. BRISC leverages LIME to tightly couple the physics models in several different codes (written in a variety of languages) into one integrated package for simulating accident scenarios in a liquid sodium cooled 'burner' nuclear reactor. Other activities and accomplishments of the LDRD include (a) further development, application and demonstration of the 'non-linear elimination' strategy to enable physics codes that do not provide residuals to be incorporated into LIME, (b) significant extensions of the RIO CFD code capabilities, (c) complex 3D solid modeling and meshing of major fast reactor components and regions, and (d) an approach for multi-physics coupling across non-conformal mesh interfaces.

  2. Passive decay heat removal system for water-cooled nuclear reactors

    DOEpatents

    Forsberg, Charles W.

    1991-01-01

    A passive decay-heat removal system for a water-cooled nuclear reactor employs a closed heat transfer loop having heat-exchanging coils inside an open-topped, insulated box located inside the reactor vessel, below its normal water level, in communication with a condenser located outside of containment and exposed to the atmosphere. The heat transfer loop is located such that the evaporator is in a position where, when the water level drops in the reactor, it will become exposed to steam. Vapor produced in the evaporator passes upward to the condenser above the normal water level. In operation, condensation in the condenser removes heat from the system, and the condensed liquid is returned to the evaporator. The system is disposed such that during normal reactor operations where the water level is at its usual position, very little heat will be removed from the system, but during emergency, low water level conditions, substantial amounts of decay heat will be removed.

  3. Neutron Capture and the Antineutrino Yield from Nuclear Reactors.

    PubMed

    Huber, Patrick; Jaffke, Patrick

    2016-03-25

    We identify a new, flux-dependent correction to the antineutrino spectrum as produced in nuclear reactors. The abundance of certain nuclides, whose decay chains produce antineutrinos above the threshold for inverse beta decay, has a nonlinear dependence on the neutron flux, unlike the vast majority of antineutrino producing nuclides, whose decay rate is directly related to the fission rate. We have identified four of these so-called nonlinear nuclides and determined that they result in an antineutrino excess at low energies below 3.2 MeV, dependent on the reactor thermal neutron flux. We develop an analytic model for the size of the correction and compare it to the results of detailed reactor simulations for various real existing reactors, spanning 3 orders of magnitude in neutron flux. In a typical pressurized water reactor the resulting correction can reach ∼0.9% of the low energy flux which is comparable in size to other, known low-energy corrections from spent nuclear fuel and the nonequilibrium correction. For naval reactors the nonlinear correction may reach the 5% level by the end of cycle. PMID:27058075

  4. Neutron Capture and the Antineutrino Yield from Nuclear Reactors

    NASA Astrophysics Data System (ADS)

    Huber, Patrick; Jaffke, Patrick

    2016-03-01

    We identify a new, flux-dependent correction to the antineutrino spectrum as produced in nuclear reactors. The abundance of certain nuclides, whose decay chains produce antineutrinos above the threshold for inverse beta decay, has a nonlinear dependence on the neutron flux, unlike the vast majority of antineutrino producing nuclides, whose decay rate is directly related to the fission rate. We have identified four of these so-called nonlinear nuclides and determined that they result in an antineutrino excess at low energies below 3.2 MeV, dependent on the reactor thermal neutron flux. We develop an analytic model for the size of the correction and compare it to the results of detailed reactor simulations for various real existing reactors, spanning 3 orders of magnitude in neutron flux. In a typical pressurized water reactor the resulting correction can reach ˜0.9 % of the low energy flux which is comparable in size to other, known low-energy corrections from spent nuclear fuel and the nonequilibrium correction. For naval reactors the nonlinear correction may reach the 5% level by the end of cycle.

  5. Chernobyl nuclear reactor accident fallout: Measurement and consequences. (Latest citations from the NTIS bibliographic database). Published Search

    SciTech Connect

    1994-01-01

    The bibliography contains citations concerning the consequences of radioactive fallout from the Chernobyl nuclear reactor accident. Citations discuss radioactive monitoring, health hazards, and radiation dosimetry. Radiation contamination in the air, soil, vegetation, and food is examined. (Contains a minimum of 210 citations and includes a subject term index and title list.)

  6. Chernobyl Nuclear Reactor accident fallout: Measurement and consequences. (Latest citations from the NTIS bibliographic database). Published Search

    SciTech Connect

    1996-11-01

    The bibliography contains citations concerning the consequences of radioactive fallout from the Chernobyl nuclear reactor accident. Citations discuss radioactive monitoring, health hazards, and radiation dosimetry. Radiation contamination in the air, soil, vegetation, and food is examined. (Contains 50-250 citations and includes a subject term index and title list.) (Copyright NERAC, Inc. 1995)

  7. Chernobyl nuclear reactor accident fallout: Measurement and consequences. (Latest citations from the NTIS Bibliographic database). Published Search

    SciTech Connect

    1993-09-01

    The bibliography contains citations concerning the consequences of radioactive fallout from the Chernobyl nuclear reactor accident. Citations discuss radioactive monitoring, health hazards, and radiation dosimetry. Radiation contamination in the air, soil, vegetation, and food is examined. (Contains a minimum of 208 citations and includes a subject term index and title list.)

  8. Chernobyl nuclear reactor accident fallout: Measurement and consequences. (Latest citations from the NTIS bibliographic database). Published Search

    SciTech Connect

    1995-02-01

    The bibliography contains citations concerning the consequences of radioactive fallout from the Chernobyl nuclear reactor accident. Citations discuss radioactive monitoring, health hazards, and radiation dosimetry. Radiation contamination in the air, soil, vegetation, and food is examined. (Contains a minimum of 247 citations and includes a subject term index and title list.)

  9. Chernobyl nuclear reactor accident fallout: Measurement and consequences. (Latest citations from the NTIS bibliographic database). Published Search

    SciTech Connect

    1996-01-01

    The bibliography contains citations concerning the consequences of radioactive fallout from the Chernobyl nuclear reactor accident. Citations discuss radioactive monitoring, health hazards, and radiation dosimetry. Radiation contamination in the air, soil, vegetation, and food is examined. (Contains 50-250 citations and includes a subject term index and title list.) (Copyright NERAC, Inc. 1995)

  10. Piezoelectric material for use in a nuclear reactor core

    NASA Astrophysics Data System (ADS)

    Parks, D. A.; Reinhardt, Brian; Tittmann, B. R.

    2012-05-01

    In radiation environments ultrasonic nondestructive evaluation has great potential for improving reactor safety and furthering the understanding of radiation effects and materials. In both nuclear power plants and materials test reactors, elevated temperatures and high levels of radiation present challenges to ultrasonic NDE methodologies. The challenges are primarily due to the degradation of the ultrasonic sensors utilized. We present results from the operation of a ultrasonic piezoelectric transducer, composed of bulk single crystal AlN, in a nuclear reactor core for over 120 MWHrs. The transducer was coupled to an aluminum cylinder and operated in pulse echo mode throughout the irradiation. In addition to the pulse echo testing impedance data were obtained. Further, the piezoelectric coefficient d33 was measured prior to irradiation and found to be 5.5 pC/N which is unchanged from as-grown samples, and in fact higher than the measured d33 for many as-grown samples.

  11. Modern control technology for improved nuclear reactor performance

    SciTech Connect

    Oakes, L.C.

    1986-12-01

    One of the main complaints leveled at reactor control systems by utility spokesmen is complexity. One only has to look inside a power reactor control room to appreciate this viewpoint. The high reliability and versatility of modern microprocessors makes possible distributed control systems with only performance data and abnormal conditions being relayed to the control room. In a sense, this emulates the human-body control system where routine repetitive actions are handled in an involuntary manner. The significance of expert systems to the nuclear reactor control and safety systems is their ability to capture human and other expertise and make it available, upon demand, and under almost all circumstances. Thus, human problem-solving skills acquired by the learning process over a long period of time can be captured and employed with the reliability inherent in computers. This is especially important in nuclear plants when human operators are burdened by stress and emotional factors that have a dramatic effect on performance level.

  12. Piezoelectric material for use in a nuclear reactor core

    SciTech Connect

    Parks, D. A.; Reinhardt, Brian; Tittmann, B. R.

    2012-05-17

    In radiation environments ultrasonic nondestructive evaluation has great potential for improving reactor safety and furthering the understanding of radiation effects and materials. In both nuclear power plants and materials test reactors, elevated temperatures and high levels of radiation present challenges to ultrasonic NDE methodologies. The challenges are primarily due to the degradation of the ultrasonic sensors utilized. We present results from the operation of a ultrasonic piezoelectric transducer, composed of bulk single crystal AlN, in a nuclear reactor core for over 120 MWHrs. The transducer was coupled to an aluminum cylinder and operated in pulse echo mode throughout the irradiation. In addition to the pulse echo testing impedance data were obtained. Further, the piezoelectric coefficient d{sub 33} was measured prior to irradiation and found to be 5.5 pC/N which is unchanged from as-grown samples, and in fact higher than the measured d{sub 33} for many as-grown samples.

  13. Distribution and detection of positrons from an orbiting nuclear reactor.

    PubMed

    Hones, E W; Higbie, P R

    1989-04-28

    The Solar Maximum Mission (SMM) Gamma-Ray Spectrometer has on many occasions detected nuclear radiation produced by nuclear reactors carried on Soviet satellites. A unique feature of the observations is the measurement of bursts of 511-kiloelectron volt gamma rays that are thought to signal SMM encounters with positrons emanating from the Soviet satellites. A model of positron generation by an orbiting reactor has been developed that describes the resulting time-dependent distribution of positrons temporarily trapped in the geomagnetic field and estimates the response of the SMM spectrometer to passage through such distributions. The model successfully predicts onset times, durations, and intensities of the 511-kiloelectron volt gamma bursts, as we illustrate in a detailed analysis of one event, and thus confirms that these are due to positrons from the Soviet satellites. Reactor-generated positrons are potentially useful in magnetospheric research. PMID:17807611

  14. CONTROL ROD FOR A NUCLEAR REACTOR AND METHOD OF PREPARATION

    DOEpatents

    Hausner, H.H.

    1958-12-30

    BS>An improved control rod is presented for a nuclear reactor. This control rod is comprised of a rare earth metal oxide or rare earth metal carbide such as gadolinium oxide or gadolinium carbide, uniformly distributed in a metal matrix having a low cross sectional area of absorption for thermal neutrons, such as aluminum, beryllium, and zirconium.

  15. Safety program considerations for space nuclear reactor systems

    SciTech Connect

    Cropp, L.O.

    1984-08-01

    This report discusses the necessity for in-depth safety program planning for space nuclear reactor systems. The objectives of the safety program and a proposed task structure is presented for meeting those objectives. A proposed working relationship between the design and independent safety groups is suggested. Examples of safety-related design philosophies are given.

  16. Global risk of radioactive fallout after nuclear reactor accidents

    NASA Astrophysics Data System (ADS)

    Kunkel, D.; Lelieveld, J.; Lawrence, M. G.

    2012-04-01

    Reactor core meltdowns of nuclear power plants are rare, yet the consequences are catastrophic. But what is meant by "rare"? And what can be learned from the Chernobyl and Fukushima incidents? Here we assess the risk of exposure to radioactivity due to atmospheric dispersion of gases and particles following severe nuclear accidents, using particulate 137Cs and gaseous 131I as proxies for the fallout. It appears that previously the occurrence of major accidents and the risks of radioactive contamination have been underestimated. Using a global model of the atmosphere we compute that on average, in the event of a core melt of any nuclear power plant worldwide, more than 90 % of emitted 137Cs would be transported beyond 50 km and about 50 % beyond 1000 km distance. This corroborates that such accidents have large-scale and trans-boundary impacts. Although the emission strengths and atmospheric removal processes of 137Cs and 131I are quite different, the radioactive contamination patterns over land and the human deposition exposure are computed to be similar. High human exposure risks occur around reactors in densely populated regions, notably in southern Asia where a core melt can subject 55 million people to radioactive contamination. The recent decision by Germany to phase out its nuclear reactors will reduce the national risk, though a large risk will still remain from the reactors in neighbouring countries.

  17. Global risk of radioactive fallout after nuclear reactor accidents

    NASA Astrophysics Data System (ADS)

    Lelieveld, J.; Kunkel, D.; Lawrence, M. G.

    2011-11-01

    Reactor core meltdowns of nuclear power plants are rare, yet the consequences are catastrophic. But what is meant by "rare"? And what can be learned from the Chernobyl and Fukushima incidents? Here we assess the risk of exposure to radioactivity due to atmospheric dispersion of gases and particles following severe nuclear accidents, using particulate 137Cs and gaseous 131I as proxies for the fallout. It appears that previously the occurrence of major accidents and the risks of radioactive contamination have been underestimated. Using a global model of the atmosphere we compute that on average, in the event of a core melt of any nuclear power plant worldwide, more than 90% of emitted 137Cs would be transported beyond 50km and about 50% beyond 1000 km distance. This corroborates that such accidents have large-scale and trans-boundary impacts. Although the emission strengths and atmospheric removal processes of 137Cs and 131I are quite different, the radioactive contamination patterns over land and the human deposition exposure are computed to be similar. High human exposure risks occur around reactors in densely populated regions, notably in southern Asia where a core melt can subject 55 million people to radioactive contamination. The recent decision by Germany to phase out its nuclear reactors will reduce the national risk, though a large risk will still remain from the reactors in neighbouring countries.

  18. Packed rod neutron shield for fast nuclear reactors

    DOEpatents

    Eck, John E.; Kasberg, Alvin H.

    1978-01-01

    A fast neutron nuclear reactor including a core and a plurality of vertically oriented neutron shield assemblies surrounding the core. Each assembly includes closely packed cylindrical rods within a polygonal metallic duct. The shield assemblies are less susceptible to thermal stresses and are less massive than solid shield assemblies, and are cooled by liquid coolant flow through interstices among the rods and duct.

  19. 10. Photocopy of drawing, February 1958, NUCLEAR REACTOR FACILITY, STRUCTURAL ...

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

    10. Photocopy of drawing, February 1958, NUCLEAR REACTOR FACILITY, STRUCTURAL CROSS SECTION. Giffals & Vallet, Inc., L. Rosetti, Associated Architects and Engineers, Detroit, Michigan; and U.S. Army Engineer Division, New England Corps of Engineers, Boston, Massachusetts. Drawing Number 35-84-04. (Original: AMTL Engineering Division, Watertown). - Watertown Arsenal, Building No. 100, Wooley Avenue, Watertown, Middlesex County, MA

  20. 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)

  1. 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.

  2. 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.

  3. 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)

  4. Automatic coolant flow control device for a nuclear reactor assembly

    DOEpatents

    Hutter, E.

    1984-01-27

    A device which controls coolant flow through a nuclear reactor assembly comprises a baffle means at the exit end of said assembly having a plurality of orifices, and a bimetallic member in operative relation to the baffle means such that at increased temperatures said bimetallic member deforms to unblock some of said orifices and allow increased coolant flow therethrough.

  5. Pin-Type Gas Cooled Reactor for Nuclear Electric Propulsion

    NASA Astrophysics Data System (ADS)

    Wright, Steven A.; Lipinski, Ronald J.

    2003-01-01

    This paper describes a point design for a pin-type Gas-Cooled Reactor concept that uses a fuel pin design similar to the SP100 fuel pin. The Gas-Cooled Reactor is designed to operate at 100 kWe for 7 years plus have a reduced power mode of 20% power for a duration of 5 years. The power system uses a gas-cooled, UN-fueled, pin-type reactor to heat He/Xe gas that flows directly into a recuperated Brayton system to produce electricity. Heat is rejected to space via a thermal radiator that unfolds in space. The reactor contains approximately 154 kg of 93.15 % enriched UN in 313 fuel pins. The fuel is clad with rhenium-lined Nb-1Zr. The pressures vessel and ducting are cooled by the 900 K He/Xe gas inlet flow or by thermal radiation. This permits all pressure boundaries to be made of superalloy metals rather than refractory metals, which greatly reduces the cost and development schedule required by the project. The reactor contains sufficient rhenium (a neutron poison) to make the reactor subcritical under water immersion accidents without the use of internal shutdown rods. The mass of the reactor and reflectors is about 750 kg.

  6. Nuclear Data Measurements for 21st Century Reactor Physics Applications

    SciTech Connect

    Rahmat Aryaeinejad; Jerald D. Cole; Mark W. Drigert; James K. Jewell; Christopher A. McGrath; David W. Nigg; Edward L. Reber

    2003-03-01

    The United States Department of Energy (DOE), Office of Nuclear Energy (NE) has embarked on a long-term program to significantly advance the science and technology of nuclear energy. This is in response to the overall national plan for accelerated development of domestic energy resources on several fronts, punctuated by recent dramatic events that have emphasized the need for the US to reduce its dependence on foreign petroleum supplies. Key aspects of the DOE-NE agenda are embodied in the Generation-IV (Gen-IV) advanced nuclear energy systems development program and in the Advanced Fuel Cycle (AFC) program. The planned efforts involve near-term and intermediate-term improvements in fuel utilization and recycling in current nuclear power reactor systems as well as the longer-term development of new nuclear energy systems that offer much improved fuel utilization and proliferation resistance, along with continued advances in operational safety. The success of the overall NE effort will depend not only on sophisticated system development and engineering, but also on the advances in the supporting sciences and technologies. Of these, one of the most important is the improvement of the relevant fundamental nuclear science data bases, especially the evaluated neutron interaction cross section files that serve as the foundation of all reactor system designs, operating strategies, and fuel cycle engineering activities. The new concepts for reactors and fuel cycles involve the use of transuranic nuclides that were previously of little interest, and where experimentally measured information is lacking. The current state of the cross section database for some of these nuclides is such that design computations for advanced fast-spectrum reactor systems and fuel cycles that incorporate such materials in significant quantities are meaningful only for approximate conceptual applications. No actual system could reliably be designed according to currently accepted standards, nor

  7. Improved Nuclear Reactor and Shield Mass Model for Space Applications

    NASA Technical Reports Server (NTRS)

    Robb, Kevin

    2004-01-01

    New technologies are being developed to explore the distant reaches of the solar system. Beyond Mars, solar energy is inadequate to power advanced scientific instruments. One technology that can meet the energy requirements is the space nuclear reactor. The nuclear reactor is used as a heat source for which a heat-to-electricity conversion system is needed. Examples of such conversion systems are the Brayton, Rankine, and Stirling cycles. Since launch cost is proportional to the amount of mass to lift, mass is always a concern in designing spacecraft. Estimations of system masses are an important part in determining the feasibility of a design. I worked under Michael Barrett in the Thermal Energy Conversion Branch of the Power & Electric Propulsion Division. An in-house Closed Cycle Engine Program (CCEP) is used for the design and performance analysis of closed-Brayton-cycle energy conversion systems for space applications. This program also calculates the system mass including the heat source. CCEP uses the subroutine RSMASS, which has been updated to RSMASS-D, to estimate the mass of the reactor. RSMASS was developed in 1986 at Sandia National Laboratories to quickly estimate the mass of multi-megawatt nuclear reactors for space applications. In response to an emphasis for lower power reactors, RSMASS-D was developed in 1997 and is based off of the SP-100 liquid metal cooled reactor. The subroutine calculates the mass of reactor components such as the safety systems, instrumentation and control, radiation shield, structure, reflector, and core. The major improvements in RSMASS-D are that it uses higher fidelity calculations, is easier to use, and automatically optimizes the systems mass. RSMASS-D is accurate within 15% of actual data while RSMASS is only accurate within 50%. My goal this summer was to learn FORTRAN 77 programming language and update the CCEP program with the RSMASS-D model.

  8. Simulated antineutrino signatures of nuclear reactors for nonproliferation applications

    NASA Astrophysics Data System (ADS)

    Misner, Alex C.

    2008-10-01

    Antineutrino detectors could provide a valuable addition to current safeguards regimes. Antineutrinos are an attractive emission to monitor due to their low interaction cross-section that prevents them from being shielded and the dependence of their spectrum on the power level and isotopic content of a reactor core. While there are antineutrino detectors currently deployed at an operational reactor, such observations cannot predict the effect of the diversion of nuclear material on the antineutrino emissions. Utilizing simulation tools, one can predict the antineutrino signatures of such abnormal operations and other reactor types that have not been experimentally measured. This study simulates reactor cores with assembly-level resolution for both baseline and diversion cases in order to predict the properties of a detector for measuring the differences in the antineutrino signatures.

  9. The case for extended nuclear reactor operation

    NASA Astrophysics Data System (ADS)

    Szilard, R.; Planchon, P.; Busby, J.

    2009-07-01

    Greenhouse gas emission is a mounting problem that threatens the future production of electricity from both coal and natural gas. In 2006, 70% of domestic electricity generation relied upon fossil fuels. Projections by the Energy Information Agency1 show U.S. demand for electricity increasing 30% to 40% by 2030. Today nuclear power is the largest source of non-greenhouse-gas emitting energy generation and will be an important source of energy production in the future. This paper considers potential construction of new plants as well as the long-term prospects of existing nuclear power plants in the United States.

  10. Hybrid nuclear reactor grey rod to obtain required reactivity worth

    DOEpatents

    Miller, John V.; Carlson, William R.; Yarbrough, Michael B.

    1991-01-01

    Hybrid nuclear reactor grey rods are described, wherein geometric combinations of relatively weak neutron absorber materials such as stainless steel, zirconium or INCONEL, and relatively strong neutron absorber materials, such as hafnium, silver-indium cadmium and boron carbide, are used to obtain the reactivity worths required to reach zero boron change load follow. One embodiment includes a grey rod which has combinations of weak and strong neutron absorber pellets in a stainless steel cladding. The respective pellets can be of differing heights. A second embodiment includes a grey rod with a relatively thick stainless steel cladding receiving relatively strong neutron absorber pellets only. A third embodiment includes annular relatively weak netron absorber pellets with a smaller diameter pellet of relatively strong absorber material contained within the aperture of each relatively weak absorber pellet. The fourth embodiment includes pellets made of a homogeneous alloy of hafnium and a relatively weak absorber material, with the percentage of hafnium chosen to obtain the desired reactivity worth.

  11. Antineutrino analysis for continuous monitoring of nuclear reactors: Sensitivity study

    SciTech Connect

    Stewart, Christopher; Erickson, Anna

    2015-10-28

    This paper explores the various contributors to uncertainty on predictions of the antineutrino source term which is used for reactor antineutrino experiments and is proposed as a safeguard mechanism for future reactor installations. The errors introduced during simulation of the reactor burnup cycle from variation in nuclear reaction cross sections, operating power, and other factors are combined with those from experimental and predicted antineutrino yields, resulting from fissions, evaluated, and compared. The most significant contributor to uncertainty on the reactor antineutrino source term when the reactor was modeled in 3D fidelity with assembly-level heterogeneity was found to be the uncertainty on the antineutrino yields. Using the reactor simulation uncertainty data, the dedicated observation of a rigorously modeled small, fast reactor by a few-ton near-field detector was estimated to offer reduction of uncertainty on antineutrino yields in the 3.0–6.5 MeV range to a few percent for the primary power-producing fuel isotopes, even with zero prior knowledge of the yields.

  12. Antineutrino analysis for continuous monitoring of nuclear reactors: Sensitivity study

    NASA Astrophysics Data System (ADS)

    Stewart, Christopher; Erickson, Anna

    2015-10-01

    This paper explores the various contributors to uncertainty on predictions of the antineutrino source term which is used for reactor antineutrino experiments and is proposed as a safeguard mechanism for future reactor installations. The errors introduced during simulation of the reactor burnup cycle from variation in nuclear reaction cross sections, operating power, and other factors are combined with those from experimental and predicted antineutrino yields, resulting from fissions, evaluated, and compared. The most significant contributor to uncertainty on the reactor antineutrino source term when the reactor was modeled in 3D fidelity with assembly-level heterogeneity was found to be the uncertainty on the antineutrino yields. Using the reactor simulation uncertainty data, the dedicated observation of a rigorously modeled small, fast reactor by a few-ton near-field detector was estimated to offer reduction of uncertainty on antineutrino yields in the 3.0-6.5 MeV range to a few percent for the primary power-producing fuel isotopes, even with zero prior knowledge of the yields.

  13. 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.

  14. Five Requirements for Nuclear Energy and CANDLE Fast Reactor

    NASA Astrophysics Data System (ADS)

    Sekimoto, Hiroshi

    2010-06-01

    The Center for Research into Innovative Nuclear Energy Systems (CRINES) was established in order to succeed the COE-INES mission after finishing this program in Tokyo Tech. CRINES considers nuclear energy should satisfy 5 requirements; sustainability as basic energy, solving 3 problems inherent to accidents, radioactive waste and nuclear bomb, and economical acceptance. Characteristics of CANDLE fast reactor are discussed for these requirements. It satisfies 4 requirements; sustainability and solving 3 inherent problems. For the remaining requirement for economy, a high potential to satisfy this requirement is also shown.

  15. Five Requirements for Nuclear Energy and CANDLE Fast Reactor

    SciTech Connect

    Sekimoto, Hiroshi

    2010-06-22

    The Center for Research into Innovative Nuclear Energy Systems (CRINES) was established in order to succeed the COE-INES mission after finishing this program in Tokyo Tech. CRINES considers nuclear energy should satisfy 5 requirements; sustainability as basic energy, solving 3 problems inherent to accidents, radioactive waste and nuclear bomb, and economical acceptance. Characteristics of CANDLE fast reactor are discussed for these requirements. It satisfies 4 requirements; sustainability and solving 3 inherent problems. For the remaining requirement for economy, a high potential to satisfy this requirement is also shown.

  16. 77 FR 8902 - Draft Regulatory Guide: Issuance, Availability Decommissioning of Nuclear Power Reactors

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-02-15

    ... COMMISSION Draft Regulatory Guide: Issuance, Availability Decommissioning of Nuclear Power Reactors AGENCY... ``Decommissioning of Nuclear Power Reactors.'' This guide describes a method NRC considers acceptable for use in... Revision 1 of Regulatory Guide 1.184, ``Decommissioning of Nuclear Power Reactors,'' dated July 2000....

  17. Characterization of Oversized Crates containing Nuclear Waste

    SciTech Connect

    Berg, Randal K.; Haggard, Daniel L.; Hilliard, Jim; Mozhayev, Andrey V.

    2007-11-01

    The 212-N Building at the Hanford Site held fifteen large crates containing glove boxes and process equipment associated with the development and fabrication of mixed oxide (MOX) fuel. The gloveboxes and associated equipment originated from the 308 Building of the Hanford Site and had been placed in the crates after a process upset in the 1960s. The crates were transported to the 212-N Building and had been in storage since 1972. In an effort to reduce the hazard categorization of 212-N the crates were removed from the building and Nondestructive Assay (NDA) was performed to characterize the crate contents meeting both Safeguards and Waste Management interests. A measurement system consisting of four configurable neutron slab detectors and high purity germanium (HPGe) detectors was deployed. Since no viable information regarding the waste matrix and configuration was available it was essential to correct for attenuation with a series of transmission measurements using californium and europium sources for both neutron and gamma applications. The gamma and neutron results obtained during this measurement campaign are compared and discussed in the paper.

  18. 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.

  19. How much nuclear fuel is present in the lavalike fuel-containing mass in the fourth power-generating unit of the Chernobyl nuclear power plant?

    SciTech Connect

    Kiselev, A.N.

    1995-10-01

    At the time of the accident in the reactor of the fourth power-generating unit of the Chernobyl nuclear power plant on April 26, 1986, the core contained 1659 fuel assemblies. Each assembly contained 114.7 kg of uranium, and therefore the reactor contained a total of 114.7 x 1659 = 190,287.3 kg of uranium. If the amount of fuel is calculated according to the uranium dioxide, i.e. in the form in which the fuel was loaded into the fuel elements, then its mass in each fuel element was 3.6 kg. A fuel assembly consists of 36 fuel elements, and therefore the reactor contained 3.6 x 36 x 1659 = 215,006.4 kg of uranium dioxide. The investigations performed in the destroyed buildings showed that the nuclear fuel after the accident is found in three main modifications: in the form of uranium dioxide tablets, in the form in which it was loaded into the reactor; in a dispersed form as dust and aerosol; and in a remelted state, in the form of a lavalike fuel-containing mass. This paper discusses the amount of nuclear fuel in the lavalike mass at the Chernobyl Nuclear Power Plant.

  20. SEPARATION OF METAL VALUES FROM NUCLEAR REACTOR

    DOEpatents

    Campbell, D.O.; Cathers, G.I.

    1962-06-19

    A method is given for separating beryllium fluoride and an alkali metal fluoride from a mixture containing same and rare earth fluorides. The method comprises contacting said mixture with a liquid hydrogen fluoride solvent containing no more than about 30 per cent water by weight and saturated with a fluoride salt characterized by its solubility in anhydrous hydrogen fluoride for a period of time sufficient to dissolve said beryllium fluoride in said solvent. (AEC)

  1. Guidelines for preparing criticality safety evaluations at Department of Energy non-reactor nuclear facilities

    SciTech Connect

    1998-09-01

    This Department of Energy (DOE) is approved for use by all components of DOE. It contains guidelines that should be followed when preparing Criticality Safety Evaluations that will be used to demonstrate the safety of operations performed at DOE Non-Reactor Nuclear Facilities. Adherence with these guidelines will provide consistency and uniformity in Criticality Safety Evaluations (CSEs) across the complex and will document compliance with DOE Order 5480.24 requirements as they pertain to CSEs.

  2. Technology, safety, and costs of decommissioning reference nuclear research and test reactors. Main report

    SciTech Connect

    Konzek, G.J.; Ludwick, J.D.; Kennedy, W.E. Jr.; Smith, R.I.

    1982-03-01

    Safety and Cost Information is developed for the conceptual decommissioning of two representative licensed nuclear research and test reactors. Three decommissioning alternatives are studied to obtain comparisons between costs (in 1981 dollars), occupational radiation doses, potential radiation dose to the public, and other safety impacts. The alternatives considered are: DECON (immediate decontamination), SAFSTOR (safe storage followed by deferred decontamination), and ENTOMB (entombment). The study results are presented in two volumes. Volume 1 (Main Report) contains the results in summary form.

  3. Mössbauer study of ilmenite-limonite concrete for nuclear reactor shield

    NASA Astrophysics Data System (ADS)

    Eissa, N. A.; Sallam, H. A.; Gomaa, S. S.; Kama, G. M.; Deriu, A.; Albanese, G.

    1986-02-01

    A heavy type concrete for nuclear reactor shield prepared by adding to the portland cement Egyptian iron ores has been studied. This concrete contains heavy elements and a large amount of water of crystallization that make it more effective than ordinary concrete for neutron and gamma attenuation. The measurement of the Mössbauer parameters allows a deeper understanding of the relations between attenuation and physical properties.

  4. Sodium leak detection system for liquid metal cooled nuclear reactors

    DOEpatents

    Modarres, Dariush

    1991-01-01

    A light source is projected across the gap between the containment vessel and the reactor vessel. The reflected light is then analyzed with an absorption spectrometer. The presence of any sodium vapor along the optical path results in a change of the optical transmissivity of the media. Since the absorption spectrum of sodium is well known, the light source is chosen such that the sensor is responsive only to the presence of sodium molecules. The optical sensor is designed to be small and require a minimum of amount of change to the reactor containment vessel.

  5. Corrosion problems in light water nuclear reactors

    SciTech Connect

    Berry, W.E.

    1984-06-01

    The corrosion problems encountered during the author's career are reviewed. Attention is given to the development of Zircaloys and attendant factors that affect corrosion; the caustic and chloride stress corrosion cracking (SCC) of austenitic stainless steel steam generator tubing; the qualification of Inconel Alloy 600 for steam generator tubing and the subsequent corrosion problem of secondary side wastage, caustic SCC, pitting, intergranular attack, denting, and primary side SCC; and SCC in weld and furnace sensitized stainless steel piping and internals in boiling water reactor primary coolants. Also mentioned are corrosion of metallic uranium alloy fuels; corrosion of aluminum and niobium candidate fuel element claddings; crevice corrosion and seizing of stainless steel journal-sleeve combinations; SCC of precipitation hardened and martensitic stainless steels; low temperature SCC of welded austenitic stainless steels by chloride, fluoride, and sulfur oxy-anions; and corrosion problems experienced by condensers.

  6. Nuclear reactor melt arrest and coolability device

    DOEpatents

    Theofanous, Theo G.; Dinh, Nam Truc; Wachowiak, Richard M.

    2016-06-14

    Example embodiments provide a Basemat-Internal Melt Arrest and Coolability device (BiMAC) that offers improved spatial and mechanical characteristics for use in damage prevention and risk mitigation in accident scenarios. Example embodiments may include a BiMAC having an inclination of less than 10-degrees from the basemat floor and/or coolant channels of less than 4 inches in diameter, while maintaining minimum safety margins required by the Nuclear Regulatory Commission.

  7. Nuclear plant-aging research on reactor protection systems

    SciTech Connect

    Meyer, L.C.

    1988-01-01

    This report presents the rsults of a review of the Reactor Trip System (RTS) and the Engineered Safety Feature Actuating System (ESFAS) operating experiences reported in Licensee Event Reports (LER)s, the Nuclear Power Experience data base, Nuclear Plant Reliability Data System, and plant maintenance records. Our purpose is to evaluate the potential significance of aging, including cycling, trips, and testing as contributors to degradation of the RTS and ESFAS. Tables are presented that show the percentage of events for RTS and ESFAS classified by cause, components, and subcomponents for each of the Nuclear Steam Supply System vendors. A representative Babcock and Wilcox plant was selected for detailed study. The US Nuclear Regulatory Commission's Nuclear Plant Aging Research guidelines were followed in performing the detailed study that identified materials susceptible to aging, stressors, environmental factors, and failure modes for the RTS and ESFAS as generic instrumentation and control systems. Functional indicators of degradation are listed, testing requirements evaluated, and regulatory issues discussed.

  8. Aging of steel containments and liners in nuclear power plants

    SciTech Connect

    Naus, D.J.; Oland, C.B.; Ellingwood, B.; Norris, W.E.

    1998-01-01

    Aging of the containment pressure boundary in light water reactor plants is being addressed to understand the significant factors relating occurrence of corrosion efficacy of inspection and structural capacity reduction of steel containments and liners of concrete containments. and to make recommendations on use of risk models in regulatory decisions. Current regulatory in-service inspection requirements are reviewed and a summary of containment related degradation experience is presented. Current and emerging nondestructive examination techniques and a degradation assessment methodology for characterizing and quantifying the amount of damage present are described. Quantitative tools for condition assessment of aging structures using time dependent structural reliability analysis methods are summarized. Such methods provide a framework for addressing the uncertainties attendant to aging in the decision process. Results of this research provide a means for establishing current and estimating future structural capacity margins of containments, and to address the significance of incidences of reported containment degradation.

  9. 10 CFR Appendix J to Part 50 - Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... and feedwater piping and other systems which penetrate containment of direct-cycle boiling water power... CFR 50.12, are still applicable to Option B of this appendix if necessary, unless specifically revoked... 10 Energy 1 2011-01-01 2011-01-01 false Primary Reactor Containment Leakage Testing for...

  10. 10 CFR Appendix J to Part 50 - Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... and feedwater piping and other systems which penetrate containment of direct-cycle boiling water power... CFR 50.12, are still applicable to Option B of this appendix if necessary, unless specifically revoked... 10 Energy 1 2013-01-01 2013-01-01 false Primary Reactor Containment Leakage Testing for...

  11. 10 CFR Appendix J to Part 50 - Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... and feedwater piping and other systems which penetrate containment of direct-cycle boiling water power... CFR 50.12, are still applicable to Option B of this appendix if necessary, unless specifically revoked... 10 Energy 1 2010-01-01 2010-01-01 false Primary Reactor Containment Leakage Testing for...

  12. 10 CFR Appendix J to Part 50 - Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... and feedwater piping and other systems which penetrate containment of direct-cycle boiling water power... CFR 50.12, are still applicable to Option B of this appendix if necessary, unless specifically revoked... 10 Energy 1 2014-01-01 2014-01-01 false Primary Reactor Containment Leakage Testing for...

  13. 10 CFR Appendix J to Part 50 - Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... and feedwater piping and other systems which penetrate containment of direct-cycle boiling water power... CFR 50.12, are still applicable to Option B of this appendix if necessary, unless specifically revoked... 10 Energy 1 2012-01-01 2012-01-01 false Primary Reactor Containment Leakage Testing for...

  14. Parallelization and automatic data distribution for nuclear reactor simulations

    SciTech Connect

    Liebrock, L.M.

    1997-07-01

    Detailed attempts at realistic nuclear reactor simulations currently take many times real time to execute on high performance workstations. Even the fastest sequential machine can not run these simulations fast enough to ensure that the best corrective measure is used during a nuclear accident to prevent a minor malfunction from becoming a major catastrophe. Since sequential computers have nearly reached the speed of light barrier, these simulations will have to be run in parallel to make significant improvements in speed. In physical reactor plants, parallelism abounds. Fluids flow, controls change, and reactions occur in parallel with only adjacent components directly affecting each other. These do not occur in the sequentialized manner, with global instantaneous effects, that is often used in simulators. Development of parallel algorithms that more closely approximate the real-world operation of a reactor may, in addition to speeding up the simulations, actually improve the accuracy and reliability of the predictions generated. Three types of parallel architecture (shared memory machines, distributed memory multicomputers, and distributed networks) are briefly reviewed as targets for parallelization of nuclear reactor simulation. Various parallelization models (loop-based model, shared memory model, functional model, data parallel model, and a combined functional and data parallel model) are discussed along with their advantages and disadvantages for nuclear reactor simulation. A variety of tools are introduced for each of the models. Emphasis is placed on the data parallel model as the primary focus for two-phase flow simulation. Tools to support data parallel programming for multiple component applications and special parallelization considerations are also discussed.

  15. iDREAM: an industrial detector for nuclear reactor monitoring

    NASA Astrophysics Data System (ADS)

    Gribov, I. V.; Gromov, M. B.; Lukjanchenko, G. A.; Novikova, G. J.; Obinyakov, B. A.; Oralbaev, A. Y.; Skorokhvatov, M. D.; Sukhotin, S. V.; Chepurnov, A. S.; Etenko, A. V.

    2016-02-01

    Prototype of industrial reactor antineutrino detector iDREAM is dedicated for an experiment to demonstrate the possibility of remote monitoring of PWR reactor operational modes by neutrino method in real-time in order to avoid undeclared exposure modes for nuclear fuel and unauthorized removal of isotopes. The prototype detector was started up in 2014. To test the detector elements and components of electronics distilled water has been used as a target, which enables the use of Cerenkov radiation from cosmic muons as a physical signal. Also parallel measuring of the long-term stability has been doing for samples of liquid organic scintillator doped with gadolinium and synthesized by different methods

  16. Bottom shield for a gas cooled high temperature nuclear reactor

    SciTech Connect

    Schoening, J.; Elter, C.; Kolodzey, H. J.; Schwiers, H. G.; Stracke, W.

    1984-12-25

    A gas cooled, high temperature nuclear reactor is provided with a base plate arranged under the reactor core and over the bottom of the prestressed concrete pressure vessel serving as the bottom shield. The bottom shield comprises at least two plates arranged coaxially with respect to each other, one above the other. Each plate comprises several partially interconnected parts with the lower plate being placed at an axial and vertical distance from the bottom liner of the prestressed concrete pressure vessel and also from the upper plate.

  17. Chemistry of fission product iodine under nuclear reactor accident conditions

    SciTech Connect

    Malinauskas, A.P.; Bell, J.T.

    1986-01-01

    The radioisotopes of iodine are generally acknowledged to be the species whose release into the biosphere as a result of a nuclear reactor accident is of the greatest concern. In the course of its release, the fission product is subjected to differing chemical environments; these can alter the physicochemical form of the fission product and thus modify the manner and extent to which release occurs. Both the chemical environments which are characteristic of reactor accidents and their effect in determining physical and chemical form of fission product iodine have been studied extensively, and are reviewed in this report. 76 refs.

  18. Thermohydraulic and nuclear modeling of natural fission reactors

    NASA Astrophysics Data System (ADS)

    Viggato, Jason Charles

    Experimental verification of proposed nuclear waste storage schemes in geologic repositories is not possible, however, a natural analog exists in the form of ancient natural reactors that existed in uranium-rich ores. Two billion years ago, the enrichment of natural uranium was high enough to allow a sustained chain reaction in the presence of water as a moderator. Several natural reactors occurred in Gabon, Africa and were discovered in the early 1970's. These reactors operated at low power levels for hundreds of thousands of years. Heated water generated from the reactors also leached uranium from the surrounding rock strata and deposited it in the reactor cores. This increased the concentration of uranium in the core over time and served to "refuel" the reactor. This has strong implications in the design of modern geologic repositories for spent nuclear fuel. The possibility of accidental fission events in man-made repositories exists and the geologic evidence from Oklo suggests how those events may progress and enhance local concentrations of uranium. Based on a review of the literature, a comprehensive code was developed to model the thermohydraulic behavior and criticality conditions that may have existed in the Oklo reactor core. A two-dimensional numerical model that incorporates modeling of fluid flow, temperatures, and nuclear fission and subsequent heat generation was developed for the Oklo natural reactors. The operating temperatures ranged from about 456 K to about 721 K. Critical reactions were observed for a wide range of concentrations and porosity values (9 to 30 percent UO2 and 10 to 20 percent porosity). Periodic operation occurred in the computer model prediction with UO2 concentrations of 30 percent in the core and 5 percent in the surrounding material. For saturated conditions and 30 percent porosity, the model predicted temperature transients with a period of about 5 hours. Kuroda predicted 3 to 4 hour durations for temperature transients

  19. Heat barrier for use in a nuclear reactor facility

    DOEpatents

    Keegan, Charles P.

    1988-01-01

    A thermal barrier for use in a nuclear reactor facility is disclosed herein. Generally, the thermal barrier comprises a flexible, heat-resistant web mounted over the annular space between the reactor vessel and the guard vessel in order to prevent convection currents generated in the nitrogen atmosphere in this space from entering the relatively cooler atmosphere of the reactor cavity which surrounds these vessels. Preferably, the flexible web includes a blanket of heat-insulating material formed from fibers of a refractory material, such as alumina and silica, sandwiched between a heat-resistant, metallic cloth made from stainless steel wire. In use, the web is mounted between the upper edges of the guard vessel and the flange of a sealing ring which surrounds the reactor vessel with a sufficient enough slack to avoid being pulled taut as a result of thermal differential expansion between the two vessels. The flexible web replaces the rigid and relatively complicated structures employed in the prior art for insulating the reactor cavity from the convection currents generated between the reactor vessel and the guard vessel.

  20. Nuclear reactor spacer grid and ductless core component

    DOEpatents

    Christiansen, David W.; Karnesky, Richard A.

    1989-01-01

    The invention relates to a nuclear reactor spacer grid member for use in a liquid cooled nuclear reactor and to a ductless core component employing a plurality of these spacer grid members. The spacer grid member is of the egg-shell type and is constructed so that the walls of the cell members of the grid member are formed of a single thickness of metal to avoid tolerance problems. Within each cell member is a hydraulic spring which laterally constrains the nuclear material bearing rod which passes through each cell member against a hardstop in response to coolant flow through the cell member. This hydraulic spring is also suitable for use in a water cooled nuclear reactor. A core component constructed of, among other components, a plurality of these spacer grid members, avoids the use of a full length duct by providing spacer sleeves about the sodium tubes passing through the spacer grid members at locations between the grid members, thereby maintaining a predetermined space between adjacent grid members.

  1. Synergistic smart fuel for in-pile nuclear reactor measurements

    SciTech Connect

    Smith, J.A.; Kotter, D.K.; Ali, R.A.; Garrett, S.L.

    2013-07-01

    The thermo-acoustic fuel rod sensor developed in this research has demonstrated a novel technique for monitoring the temperature within the core of a nuclear reactor or the temperature of the surrounding heat-transfer fluid. It uses the heat from the nuclear fuel to generate sustained acoustic oscillations whose frequency will be indicative of the temperature. Converting a nuclear fuel rod into this type of thermo-acoustic sensor simply requires the insertion of a porous material (stack). This sensor has demonstrated a synergy with the elevated temperatures that exist within the nuclear reactor using materials that have only minimal susceptibility to high-energy particle fluxes. When the sensor is in operation, the sound waves radiated from the fuel rod resonator will propagate through the surrounding cooling fluid. The frequency of these oscillations is directly correlated with an effective temperature within the fuel rod resonator. This device is self-powered and is operational even in case of total loss of power of the reactor.

  2. Requirements for advanced simulation of nuclear reactor and chemicalseparation plants.

    SciTech Connect

    Palmiotti, G.; Cahalan, J.; Pfeiffer, P.; Sofu, T.; Taiwo, T.; Wei,T.; Yacout, A.; Yang, W.; Siegel, A.; Insepov, Z.; Anitescu, M.; Hovland,P.; Pereira, C.; Regalbuto, M.; Copple, J.; Willamson, M.

    2006-12-11

    This report presents requirements for advanced simulation of nuclear reactor and chemical processing plants that are of interest to the Global Nuclear Energy Partnership (GNEP) initiative. Justification for advanced simulation and some examples of grand challenges that will benefit from it are provided. An integrated software tool that has its main components, whenever possible based on first principles, is proposed as possible future approach for dealing with the complex problems linked to the simulation of nuclear reactor and chemical processing plants. The main benefits that are associated with a better integrated simulation have been identified as: a reduction of design margins, a decrease of the number of experiments in support of the design process, a shortening of the developmental design cycle, and a better understanding of the physical phenomena and the related underlying fundamental processes. For each component of the proposed integrated software tool, background information, functional requirements, current tools and approach, and proposed future approaches have been provided. Whenever possible, current uncertainties have been quoted and existing limitations have been presented. Desired target accuracies with associated benefits to the different aspects of the nuclear reactor and chemical processing plants were also given. In many cases the possible gains associated with a better simulation have been identified, quantified, and translated into economical benefits.

  3. Nuclear reactor spent fuel storage rack

    SciTech Connect

    Machado, O.J.; Flynn, W.M.; Flanders, H.E. Jr.; Booker, L.W.

    1989-04-11

    A fuel rack is described for use in storing nuclear fuel assemblies in a nuclear fuel storage pool having a floor on which an upwardly projecting stud is mounted; the fuel rack comprising: a base structure at the lower end of the fuel rack including base-plate means having flow openings therein, the base-plate means supporting a first network of interlaced beams which form a multiplicity of polygonal openings; a second network of interlaced beams forming polygonal openings positioned in spaced vertical alignment with corresponding polygonal openings in the first network of beams; a plurality of cells, each cell having sides bounded by inner and outer surfaces and being of a size and configuration designed to hold therein a fuel assembly, each cell positioned in a corresponding pair of the aligned polygonal openings, each cell being open at both ends with a guiding funnel at the upper end, and the cells being positioned over the flow openings in the base-plate to permit flow of coolant through the cells; spaced, outwardly directed, projections on the outer surfaces of the sides of the cells near the tops and bottoms of the sides thereof, each cell being sized to be received within a corresponding of the pair of aligned polygonal openings in which the cells are respectively positioned; and means fixedly securing the projections to the beams in the first and second networks of beams thereby to provide a substantially rigid fuel rack of modular design.

  4. Combined gripper for a nuclear reactor

    SciTech Connect

    Kuhn, R.

    1981-07-21

    A device for use in handling nuclear elements, such as control rods, fuel elements and blind rods in a nuclear energy plant, comprises, a hoist having a gripper rod assembly, which includes a tubular column structure into which is engaged a top movable part which is adapted to be connected to a lifting cable. The top movable part carries a crossbar which is engaged by engaging members which are pivotally mounted within the tubular column structure of the gripper rod assembly and which have pin elements thereon which, when the engaging members are positioned inwardly, engage in recesses between teeth defined on the crossbar. The engaging elements are moved by an actuating rod which is reciprocatable in the tubular column structure and includes a piston portion which is movable in a fluid cylinder portion so as to shift the engaging members between an outwardly oriented postion in which they engage into teeth elements of an annular rim of a gripper casing or an axially inwardly oriented position in which they are moved inwardly sufficiently so that the gripper rod assembly will clear a top opening of the casing and permit disengagement thereof.

  5. Comparison of methodologies for assessing the risks from nuclear weapons and from nuclear reactors

    SciTech Connect

    Benjamin, A.S.

    1996-08-01

    There are important differences between the safety principles for nuclear weapons and for nuclear reactors. For example, a principal concern for nuclear weapons is to prevent electrical energy from reaching the nuclear package during accidents produced by crashes, fires, and other hazards, whereas the foremost concern for nuclear reactors is to maintain coolant around the core in the event of certain system failures. Not surprisingly, new methods have had to be developed to assess the risk from nuclear weapons. These include fault tree transformations that accommodate time dependencies, thermal and structural analysis techniques that are fast and unconditionally stable, and parameter sampling methods that incorporate intelligent searching. This paper provides an overview of the new methods for nuclear weapons and compares them with existing methods for nuclear reactors. It also presents a new intelligent searching process for identifying potential nuclear detonation vulnerabilities. The new searching technique runs very rapidly on a workstation and shows promise for providing an accurate assessment of potential vulnerabilities with far fewer physical response calculations than would be required using a standard Monte Carlo sampling procedure.

  6. CATALYTIC RECOMBINER FOR A NUCLEAR REACTOR

    DOEpatents

    King, L.D.P.

    1960-07-01

    A hydrogen-oxygen recombiner is described for use with water-boiler type reactors. The catalyst used is the wellknown platinized alumina, and the novelty lies in the structural arrangement used to prevent flashback through the gas input system. The recombiner is cylindrical, the gases at the input end being deflected by a baffle plate through a first flashback shield of steel shot into an annular passage adjacent to and extending the full length of the housing. Below the baffle plate the gases flow first through an outer annular array of alumina pellets which serve as a second flashback shield, a means of distributing the flowing gases evenly and as a means of reducing radiation losses to the walls. Thereafter the gases flow inio the centrally disposed catalyst bed where recombination is effected. The steam and uncombined gases flow into a centrally disposed cylindrical passage inside the catalyst bod and thereafter out through the exit port. A high rate of recombination is effected.

  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. Projects of High-Temperature Nuclear Reactors

    NASA Astrophysics Data System (ADS)

    Ekmanis, J.; Tomsons, E.; Zeltiņš, N.

    2013-04-01

    Part 2 of the overview gives emphasis to the projects of high-temperature NRs, whose development is an area of active engagement for the specialists from the USA, France, Japan, Russia, China, the Netherlands, and Germany. Projects of several powerful NRs of the HTGR type for commercial use had been worked out in the USA and Germany already by 1970 but not yet implemented. Augstas temperatūras ar gāzes dzesēšanu HTGR (High Temperature Gas cooled Reactor) tipa kodolreaktoru (KR) izstrādes koncepcija bija piedāvāta 1956. gadā Lielbritānijā. Apmēram tanī pašā laikā minētā tipa KR izstrādi uzsāka Vācijā un ASV. HTGR tipa KR kodoldegviela un kodoldegvielas atražošanas materiāla sīkās daļiņas ar diametru apmēram 0.5 mm pārklātas ar vairākām aizsargkārtām un atrodas grafīta neitronu palēninātājā, kas aizsargā daļiņas no neitronu palēninātāja un dzesētāja iedarbes. Augstas temperatūras KR bez hēlija gāzes siltumnesēja var izmantot šķidrus metālus (nātriju, svinu vai svina-bismuta sakausējumu) un izkausētu sāli. Pašlaik darbojās divi augstas temperatūras ar hēlija gāzi dzesēti eksperimentālie HTGR tipa KR. Viens Japānā "HTTR" no 1998. gada oktobra (sākts būvēt 1991. gada 15. martā) ar 30 MWth siltuma jaudu. Otrs Ķīnā "HTR-10" no 2000.gada decembra (sākts būvēt 1995. gada14. jūnijā) ar 10 MWth siltuma jaudu. Ķīnā Shandong provincē 2011.gada aprīlī uzsāka augstas temperatūras "HTR-PM" (High Temperature Gas-cooled Reactor - Pebble bed Module) tipa kodolreaktora celtniecību ar 250 MWth siltuma jaudu. Augstas temperatūras kodolreaktoru izstrādē pašlaik aktīvi iesaistīti ASV, Francijas, Japānas, Krievijas, Ķīnas, Nīderlandes un Vācijas speciālisti.

  9. COMSOL-based Nuclear Reactor Kinetics Studies at the HFIR

    SciTech Connect

    Chandler, David; Freels, James D; Maldonado, G Ivan; Primm, Trent

    2011-01-01

    The computational ability to accurately predict the dynamic behavior of a nuclear reactor core in response to reactivity-induced perturbations is an important subject in reactor physics. Space-time and point kinetics methodologies were developed for the purpose of studying the transient-induced behavior of the High Flux Isotope Reactor s (HFIR) compact core. The space-time simulations employed the three-energy-group neutron diffusion equations, and transients initiated by control cylinder and hydraulic tube rabbit ejections were studied. The work presented here is the first step towards creating a comprehensive multiphysics methodology for studying the dynamic behavior of the HFIR core during reactivity perturbations. The results of these studies show that point kinetics is adequate for small perturbations in which the power distribution is assumed to be time-independent, but space-time methods must be utilized to determine localized effects.

  10. Support arrangements for core modules of nuclear reactors. [PWR

    DOEpatents

    Bollinger, L.R.

    1983-11-03

    A support arrangement is provided for the core modules of a nuclear reactor which provides support access through the control drive mechanisms of the reactor. This arrangement provides axial support of individual reactor core modules from the pressure vessel head in a manner which permits attachment and detachment of the modules from the head to be accomplished through the control drive mechanisms after their leadscrews have been removed. The arrangement includes a module support nut which is suspended from the pressure vessel head and screw threaded to the shroud housing for the module. A spline lock prevents loosening of the screw connection. An installation tool assembly, including a cell lifting and preloading tool and a torquing tool, fits through the control drive mechanism and provides lifting of the shroud housing while disconnecting the spline lock, as well as application of torque to the module support nut.

  11. Software reliability and safety in nuclear reactor protection systems

    SciTech Connect

    Lawrence, J.D.

    1993-11-01

    Planning the development, use and regulation of computer systems in nuclear reactor protection systems in such a way as to enhance reliability and safety is a complex issue. This report is one of a series of reports from the Computer Safety and Reliability Group, Lawrence Livermore that investigates different aspects of computer software in reactor National Laboratory, that investigates different aspects of computer software in reactor protection systems. There are two central themes in the report, First, software considerations cannot be fully understood in isolation from computer hardware and application considerations. Second, the process of engineering reliability and safety into a computer system requires activities to be carried out throughout the software life cycle. The report discusses the many activities that can be carried out during the software life cycle to improve the safety and reliability of the resulting product. The viewpoint is primarily that of the assessor, or auditor.

  12. Particle bed reactor central to SDI nuclear rocket project

    SciTech Connect

    Asker, J.R.

    1991-04-01

    A classified SDI project designated 'Timberwind' and funded with an estimated $7-8 billion over the project's life is charged with the development and flight testing of nuclear reactor-powered rockets. Timberwind's novel 'particle-bed reactor' technology will employ small pellets of reactor fuel to heat a low molecular weight working fluid, such as hydrogen. The fuel pellets would be 0.5 mm in diameter and may be composed of a kernel of fissionable U together with a carbon alloy, coated by layers of carbon and a sealant. A covering of zirconium carbide would prevent chemical degradation of the pellets by the hydrogen working fluid. Performace projection comparisons are conducted for Timberwind, an advanced Atlas-Centaur, and an advanced Titan launch vehicle.

  13. Support arrangement for core modules of nuclear reactors

    DOEpatents

    Bollinger, Lawrence R.

    1987-01-01

    A support arrangement is provided for the core modules of a nuclear reactor which provides support access through the control drive mechanisms of the reactor. This arrangement provides axial support of individual reactor core modules from the pressure vessel head in a manner which permits attachment and detachment of the modules from the head to be accomplished through the control drive mechanisms after their leadscrews have been removed. The arrangement includes a module support nut which is suspended from the pressure vessel head and screw threaded to the shroud housing for the module. A spline lock prevents loosening of the screw connection. An installation tool assembly, including a cell lifting and preloading tool and a torquing tool, fits through the control drive mechanism and provides lifting of the shroud housing while disconnecting the spline lock, as well as application of torque to the module support nut.

  14. Naval Spent Nuclear Fuel disposal Container System Description Document

    SciTech Connect

    N. E. Pettit

    2001-07-13

    The Naval Spent Nuclear Fuel Disposal Container System supports the confinement and isolation of waste within the Engineered Barrier System of the Monitored Geologic Repository (MGR). Disposal containers/waste packages are loaded and sealed in the surface waste handling facilities, transferred underground through the access drifts using a rail mounted transporter, and emplaced in emplacement drifts. The Naval Spent Nuclear Fuel Disposal Container System provides long term confinement of the naval spent nuclear fuel (SNF) placed within the disposal containers, and withstands the loading, transfer, emplacement, and retrieval operations. The Naval Spent Nuclear Fuel Disposal Container System provides containment of waste for a designated period of time and limits radionuclide release thereafter. The waste package maintains the waste in a designated configuration, withstands maximum credible handling and rockfall loads, limits the waste form temperature after emplacement, resists corrosion in the expected handling and repository environments, and provides containment of waste in the event of an accident. Each naval SNF disposal container will hold a single naval SNF canister. There will be approximately 300 naval SNF canisters, composed of long and short canisters. The disposal container will include outer and inner cylinder walls and lids. An exterior label will provide a means by which to identify a disposal container and its contents. Different materials will be selected for the waste package inner and outer cylinders. The two metal cylinders, in combination with the Emplacement Drift System, drip shield, and the natural barrier will support the design philosophy of defense-in-depth. The use of materials with different properties prevents a single mode failure from breaching the waste package. The inner cylinder and inner cylinder lids will be constructed of stainless steel while the outer cylinder and outer cylinder lids will be made of high-nickel alloy.

  15. Use of cermet fueled nuclear reactors for direct nuclear propulsion

    SciTech Connect

    Bhattacharyya, S.K.; Carlson, L.W.; Kuczen, K.D.; Hanan, N.A.; Palmer, R.G.; Von Hoomissen, J.; Chiu, W.; Haaland, R.

    1988-07-01

    There has been a renewal of interest in Direct Nuclear Propulsion (DNP) because of the Air Force Forecast II recommendation for the development of the technology. Several nuclear concepts have been proposed to meet the Direct Nuclear Propulsion challenge. In this paper we will present results of an initial study of the potential of a cermet fueled nuclear system in providing the desired DNP capabilities and featuring a set of unique safety characteristics. The concept of cermet fuel for DNP applications was first developed by ANL and GE working independently more than 20 years ago. The two organizations came to several remarkably consistent conclusions. The present work has consisted of collecting a unified set of design parameters from the set of design results produced in the earlier work. The conclusion of this exercise was that a cermet-fueled DNP design looked extremely promising from performance and safety considerations and that it deserves serious consideration when the decision to develop one or more concepts for DNP is made.

  16. Safety apparatus for nuclear reactor to prevent structural damage from overheating by core debris

    DOEpatents

    Gabor, John D.; Cassulo, John C.; Pedersen, Dean R.; Baker Jr., Louis

    1986-07-01

    The invention teaches safety apparatus that can be included in a nuclear reactor, either when newly fabricated or as a retrofit add-on, that will minimize proliferation of structural damage to the reactor in the event the reactor is experiencing an overheating malfunction whereby radioactive nuclear debris might break away from and be discharged from the reactor core. The invention provides a porous bed or sublayer on the lower surface of the reactor containment vessel so that the debris falls on and piles up on the bed. Vapor release elements upstand from the bed in some laterally spaced array. Thus should the high heat flux of the debris interior vaporize the coolant at that location, the vaporized coolant can be vented downwardly to and laterally through the bed to the vapor release elements and in turn via the release elements upwardly through the debris. This minimizes the pressure buildup in the debris and allows for continuing infiltration of the liquid coolant into the debris interior.

  17. Safety apparatus for nuclear reactor to prevent structural damage from overheating by core debris

    DOEpatents

    Gabor, J.D.; Cassulo, J.C.; Pedersen, D.R.; Baker, L. Jr.

    The invention teaches safety apparatus that can be included in a nuclear reactor, either when newly fabricated or as a retrofit add-on, that will minimize proliferation of structural damage to the reactor in the event the reactor is experiencing an overheating malfunction whereby radioactive nuclear debris might break away from and can be discharged from the reactor core. The invention provides a porous bed of sublayer on the lower surface of the reactor containment vessel so that the debris falls on and piles up on the bed. Vapor release elements upstand from the bed in some laterally spaced array. Thus should the high heat flux of the debris interior vaporize the coolant at that location, the vaporized coolant can be vented downwardly to and laterally through the bed to the vapor release elements and in turn via the release elements upwardly through the debris. This minimizes the pressure buildup in the debris and allows for continuing infiltration of the liquid coolant into the debris interior.

  18. Safety apparatus for nuclear reactor to prevent structural damage from overheating by core debris

    DOEpatents

    Gabor, John D.; Cassulo, John C.; Pedersen, Dean R.; Baker, Jr., Louis

    1986-01-01

    The invention teaches safety apparatus that can be included in a nuclear reactor, either when newly fabricated or as a retrofit add-on, that will minimize proliferation of structural damage to the reactor in the event the reactor is experiencing an overheating malfunction whereby radioactive nuclear debris might break away from and be discharged from the reactor core. The invention provides a porous bed or sublayer on the lower surface of the reactor containment vessel so that the debris falls on and piles up on the bed. Vapor release elements upstand from the bed in some laterally spaced array. Thus should the high heat flux of the debris interior vaporize the coolant at that location, the vaporized coolant can be vented downwardly to and laterally through the bed to the vapor release elements and in turn via the release elements upwardly through the debris. This minimizes the pressure buildup in the debris and allows for continuing infiltration of the liquid coolant into the debris interior.

  19. Nuclear reactor with makeup water assist from residual heat removal system

    DOEpatents

    Corletti, M.M.; Schulz, T.L.

    1993-12-07

    A pressurized water nuclear reactor uses its residual heat removal system to make up water in the reactor coolant circuit from an in-containment refueling water supply during staged depressurization leading up to passive emergency cooling by gravity feed from the refueling water storage tank, and flooding of the containment building. When depressurization commences due to inadvertence or a manageable leak, the residual heat removal system is activated manually and prevents flooding of the containment when such action is not necessary. Operation of the passive cooling system is not impaired. A high pressure makeup water storage tank is coupled to the reactor coolant circuit, holding makeup coolant at the operational pressure of the reactor. The staged depressurization system vents the coolant circuit to the containment, thus reducing the supply of makeup coolant. The level of makeup coolant can be sensed to trigger opening of successive depressurization conduits. The residual heat removal pumps move water from the refueling water storage tank into the coolant circuit as the coolant circuit is depressurized, preventing reaching the final depressurization stage unless the makeup coolant level continues to drop. The residual heat removal system can also be coupled in a loop with the refueling water supply tank, for an auxiliary heat removal path. 2 figures.

  20. Nuclear reactor with makeup water assist from residual heat removal system

    DOEpatents

    Corletti, Michael M.; Schulz, Terry L.

    1993-01-01

    A pressurized water nuclear reactor uses its residual heat removal system to make up water in the reactor coolant circuit from an in-containment refueling water supply during staged depressurization leading up to passive emergency cooling by gravity feed from the refueling water storage tank, and flooding of the containment building. When depressurization commences due to inadvertence or a manageable leak, the residual heat removal system is activated manually and prevents flooding of the containment when such action is not necessary. Operation of the passive cooling system is not impaired. A high pressure makeup water storage tank is coupled to the reactor coolant circuit, holding makeup coolant at the operational pressure of the reactor. The staged depressurization system vents the coolant circuit to the containment, thus reducing the supply of makeup coolant. The level of makeup coolant can be sensed to trigger opening of successive depressurization conduits. The residual heat removal pumps move water from the refueling water storage tank into the coolant circuit as the coolant circuit is depressurized, preventing reaching the final depressurization stage unless the makeup coolant level continues to drop. The residual heat removal system can also be coupled in a loop with the refueling water supply tank, for an auxiliary heat removal path.

  1. Detection of Shielded Nuclear Material in a Cargo Container

    SciTech Connect

    J. L. Jones; D. R. Norman; K. J. Haskell; J. W. Sterbentz; W. Y. Yoon; S. M. Watson; J. T. Johnson; J. M. Zabriskie; B. D. Bennett; R. W. Watson; J. F. Hamon

    2005-06-01

    The Idaho National Laboratory, along with Los Alamos National Laboratory and the Idaho State University’s Idaho Accelerator Center, are developing electron accelerator-based, photonuclear inspection technologies for the detection of shielded nuclear material within air-, rail-, and especially, maritime-cargo transportation containers. This paper describes a developing prototypical cargo container inspection system utilizing the Pulsed Photonuclear Assessment (PPA) technology, incorporates interchangeable, well-defined, contraband shielding structures (i.e., "calibration" pallets) providing realistic detection data for induced radiation signatures from smuggled nuclear material, and provides various shielded nuclear material detection results. Using a 4.8-kg quantity of depleted uranium, neutron and gamma-ray detection responses are presented for well-defined shielded and unshielded configurations evaluated in a selected cargo container inspection configuration. © 2001 Elsevier Science. All rights reserved

  2. Detection of shielded nuclear material in a cargo container

    NASA Astrophysics Data System (ADS)

    Jones, James L.; Norman, Daren R.; Haskell, Kevin J.; Sterbentz, James W.; Yoon, Woo Y.; Watson, Scott M.; Johnson, James T.; Zabriskie, John M.; Bennett, Brion D.; Watson, Richard W.; Moss, Cavin E.; Frank Harmon, J.

    2006-06-01

    The Idaho National Laboratory, along with Los Alamos National Laboratory and the Idaho State University's Idaho Accelerator Center, are developing electron accelerator-based, photonuclear inspection technologies for the detection of shielded nuclear material within air-, rail-, and especially, maritime-cargo transportation containers. This paper describes a developing prototypical cargo container inspection system utilizing the Pulsed Photonuclear Assessment (PPA) technology, incorporates interchangeable, well-defined, contraband shielding structures (i.e., "calibration" pallets) providing realistic detection data for induced radiation signatures from smuggled nuclear material, and provides various shielded nuclear material detection results. Using a 4.8-kg quantity of depleted uranium, neutron and gamma-ray detection responses are presented for well-defined shielded and unshielded configurations evaluated in a selected cargo container inspection configuration.

  3. Advanced Space Nuclear Reactors from Fiction to Reality

    NASA Astrophysics Data System (ADS)

    Popa-Simil, L.

    The advanced nuclear power sources are used in a large variety of science fiction movies and novels, but their practical development is, still, in its early conceptual stages, some of the ideas being confirmed by collateral experiments. The novel reactor concept uses the direct conversion of nuclear energy into electricity, has electronic control of reactivity, being surrounded by a transmutation blanket and very thin shielding being small and light that at its very limit may be suitable to power an autonomously flying car. It also provides an improved fuel cycle producing minimal negative impact to environment. The key elements started to lose the fiction attributes, becoming viable actual concepts and goals for the developments to come, and on the possibility to achieve these objectives started to become more real because the theory shows that using the novel nano-technologies this novel reactor might be achievable in less than a century.

  4. Temperature measuring analysis of the nuclear reactor fuel assembly

    SciTech Connect

    Urban, F. E-mail: zdenko.zavodny@stuba.sk; Kučák, L. E-mail: zdenko.zavodny@stuba.sk; Bereznai, J. E-mail: zdenko.zavodny@stuba.sk; Závodný, Z. E-mail: zdenko.zavodny@stuba.sk; Muškát, P. E-mail: zdenko.zavodny@stuba.sk

    2014-08-06

    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.

  5. Inception and evolution of Oklo natural nuclear reactors

    NASA Astrophysics Data System (ADS)

    Bentridi, Salah-Eddine; Gall, Benoît; Gauthier-Lafaye, François; Seghour, Abdeslam; Medjadi, Djamel-Eddine

    2011-11-01

    The occurrence of more than 15 natural nuclear Reactor Zones (RZ) in a geological environment remains a mystery even 40 years after their discovery. The present work gives for the first time an explanation of the chemical and physical processes that caused the start-up of the fission reactions with two opposite processes, uranium enrichments and progressive impoverishment in 235U. Based on Monte-Carlo neutronics simulations, a solution space was defined taking into account realistic combinations of relevant parameters acting on geological conditions and neutron transport physics. This study explains criticality occurrence, operation, expansion and end of life conditions of Oklo natural nuclear reactors, from the smallest to the biggest ones.

  6. A fission fragment reactor concept for nuclear space propulsion

    NASA Astrophysics Data System (ADS)

    Suo-Anttila, A. J.; Parma, E. J.; Wright, S. A.; Vernon, M. E.; Pickard, P. S.

    1991-10-01

    Sandia National Laboratory (SNL) has proposed a new nuclear thermal propulsion concept that uses fission fragments to directly heat the propellant up to 1000 K or higher above the material temperatures. The concept offers significant advantages over traditional solid core nuclear rocket concepts because of higher propellant exit temperatures while at the same time providing for more reliable operation due to lower structure temperatures and lower power densities. The concept can be operated in either steady state or pulsed modes. The engine consists of tubular modules, each with its own pressure boundary and rocket nozzle. The steady state mode requires a large engine with a reflector for criticality, provides high thrust and high ISP. The pulse mode utilizes a driver reactor for criticality and can be considerably smaller with lower but scaleable thrust. The pulse mode does require an external heat radiator for reactor cooling, which limits its duty cycle.

  7. Coolant monitoring apparatus for nuclear reactors

    DOEpatents

    Tokarz, Richard D.

    1983-01-01

    A system for monitoring coolant conditions within a pressurized vessel. A length of tubing extends outward from the vessel from an open end containing a first line restriction at the location to be monitored. The flowing fluid is cooled and condensed before passing through a second line restriction. Measurement of pressure drop at the second line restriction gives an indication of fluid condition at the first line restriction. Multiple lengths of tubing with open ends at incremental elevations can measure coolant level within the vessel.

  8. Variable flow control for a nuclear reactor control rod

    DOEpatents

    Carleton, Richard D.; Bhattacharyya, Ajay

    1978-01-01

    A variable flow control for a control rod assembly of a nuclear reactor that depends on turbulent friction though an annulus. The annulus is formed by a piston attached to the control rod drive shaft and a housing or sleeve fitted to the enclosure housing the control rod. As the nuclear fuel is burned up and the need exists for increased reactivity, the control rods are withdrawn, which increases the length of the annulus and decreases the rate of coolant flow through the control rod assembly.

  9. Float level switch for a nuclear power plant containment vessel

    DOEpatents

    Powell, James G.

    1993-01-01

    This invention is a float level switch used to sense rise or drop in water level in a containment vessel of a nuclear power plant during a loss of coolant accident. The essential components of the device are a guide tube, a reed switch inside the guide tube, a float containing a magnetic portion that activates a reed switch, and metal-sheathed, ceramic-insulated conductors connecting the reed switch to a monitoring system outside the containment vessel. Special materials and special sealing techniques prevent failure of components and allow the float level switch to be connected to a monitoring system outside the containment vessel.

  10. Float level switch for a nuclear power plant containment vessel

    DOEpatents

    Powell, J.G.

    1993-11-16

    This invention is a float level switch used to sense rise or drop in water level in a containment vessel of a nuclear power plant during a loss of coolant accident. The essential components of the device are a guide tube, a reed switch inside the guide tube, a float containing a magnetic portion that activates a reed switch, and metal-sheathed, ceramic-insulated conductors connecting the reed switch to a monitoring system outside the containment vessel. Special materials and special sealing techniques prevent failure of components and allow the float level switch to be connected to a monitoring system outside the containment vessel. 1 figures.

  11. Systems aspects of a space nuclear reactor power system

    NASA Technical Reports Server (NTRS)

    Jaffe, L.; Fujita, T.; Beatty, R.; Bhandari, P.; Chow, E.; Deininger, W.; Ewell, R.; Grossman, M.; Bloomfield, H.; Heller, J.

    1988-01-01

    Various system aspects of a 300-kW nuclear reactor power system for spacecraft have been investigated. Special attention is given to the cases of a reusable OTV and a space-based radar. It is demonstrated that the stowed length of the power system is important to mission design, and that orbital storage for months to years may be needed for missions involving orbital assembly.

  12. 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.

  13. 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.

  14. Evaluation of nuclear-reactor-produced iodine-123

    NASA Technical Reports Server (NTRS)

    Blue, J. W.; Sodd, V. J.

    1976-01-01

    Iodine-123 has such great potential for nuclear medicine that all possible production methods should be considered. In this report, an experimental study related to I-123 production at a high-intensity fast-flux reactor using the reaction Xe-124(n,2n)Xe-123 is considered. The conclusion is that I-123 could be made in small quantities and the cost would be higher than the cyclotron methods presently used.

  15. Expert system for online surveillance of nuclear reactor coolant pumps

    DOEpatents

    Gross, Kenny C.; Singer, Ralph M.; Humenik, Keith E.

    1993-01-01

    An expert system for online surveillance of nuclear reactor coolant pumps. This system provides a means for early detection of pump or sensor degradation. Degradation is determined through the use of a statistical analysis technique, sequential probability ratio test, applied to information from several sensors which are responsive to differing physical parameters. The results of sequential testing of the data provide the operator with an early warning of possible sensor or pump failure.

  16. 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.

  17. Photonuclear-based, nuclear material detection system for cargo containers

    NASA Astrophysics Data System (ADS)

    Jones, J. L.; Yoon, W. Y.; Norman, D. R.; Haskell, K. J.; Zabriskie, J. M.; Watson, S. M.; Sterbentz, J. W.

    2005-12-01

    The Idaho National Laboratory (INL) has been developing electron accelerator-based, photonuclear inspection technologies for over a decade. A current need, having important national implications, has been with the detection of smuggled nuclear material within air- and, especially, sea-cargo transportation containers. This paper describes the latest pulsed, photonuclear inspection system for nuclear material detection and identification in cargo configurations, the numerical responses of 5 kg of a nuclear material placed within selected cargo configurations, and the technology's potential role in addressing future inspection needs.

  18. Determination of antineutrino spectra from nuclear reactors

    SciTech Connect

    Huber, Patrick

    2011-08-15

    In this paper we study the effect of well-known higher-order corrections to the allowed {beta}-decay spectrum on the determination of antineutrino spectra resulting from the decays of fission fragments. In particular, we try to estimate the associated theory errors and find that induced currents like weak magnetism may ultimately limit our ability to improve the current accuracy and under certain circumstance could even greatly increase the theoretical errors. We also perform a critical evaluation of the errors associated with our method to extract the antineutrino spectrum using synthetic {beta} spectra. It turns out that a fit using only virtual {beta} branches with a judicious choice of the effective nuclear charge provides results with a minimal bias. We apply this method to actual data for {sup 235}U, {sup 239}Pu, and {sup 241}Pu and confirm, within errors, recent results, which indicate a net 3% upward shift in energy-averaged antineutrino fluxes. However, we also find significant shape differences which can, in principle, be tested by high-statistics antineutrino data samples.

  19. LOFT. Reactor arrives at containment building (TAN650), now being pushed ...

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

    LOFT. Reactor arrives at containment building (TAN-650), now being pushed by locomotive. Camera facing northerly. Note "Hello Dolly" and "PWR MTA No. 1" (pressurized water reactor mobile test assembly) signs. Date: 1973. INEEL negative no. 73-3710 - Idaho National Engineering Laboratory, Test Area North, Scoville, Butte County, ID

  20. Self locking drive system for rotating plug of a nuclear reactor

    DOEpatents

    Brubaker, James E.

    1979-01-01

    This disclosure describes a self locking drive system for rotating the plugs on the head of a nuclear reactor which is able to restrain plug motion if a seismic event should occur during reactor refueling. A servomotor is engaged via a gear train and a bull gear to the plug. Connected to the gear train is a feedback control system which allows the motor to rotate the plug to predetermined locations for refueling of the reactor. The gear train contains a self locking double enveloping worm gear set. The worm gear set is utilized for its self locking nature to prevent unwanted rotation of the plugs as the result of an earthquake. The double enveloping type is used because its unique contour spreads the load across several teeth providing added strength and allowing the use of a conventional size worm.