National Library of Energy BETA

Sample records for nuclear reactor ownership

  1. Reactor Technology | Nuclear Science | ORNL

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Reactor Technology Advanced Reactor Concepts Advanced Instrumentation & Controls Light Water Reactor Sustainability Safety and Regulatory Technology Small Modular Reactors Nuclear...

  2. Nuclear reactor

    DOE Patents [OSTI]

    Thomson, Wallace B. (Severna Park, MD)

    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.

  3. Nuclear reactor apparatus

    DOE Patents [OSTI]

    Wade, Elman E. (Ruffs Dale, PA)

    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. Nuclear reactor reflector

    DOE Patents [OSTI]

    Hopkins, Ronald J. (Pensacola, FL); Land, John T. (Pensacola, FL); Misvel, Michael C. (Pensacola, FL)

    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.

  5. Nuclear reactor reflector

    DOE Patents [OSTI]

    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.

  6. Nuclear reactor control column

    DOE Patents [OSTI]

    Bachovchin, Dennis M. (Plum Borough, PA)

    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.

  7. Nuclear Reactors and Technology

    SciTech Connect (OSTI)

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

    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.

  8. Heat dissipating nuclear reactor

    DOE Patents [OSTI]

    Hunsbedt, Anstein (Los Gatos, CA); Lazarus, Jonathan D. (Sunnyvale, CA)

    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. Heat dissipating nuclear reactor

    DOE Patents [OSTI]

    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.

  10. Nuclear reactor safety device

    DOE Patents [OSTI]

    Hutter, Ernest (Wilmette, IL)

    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.

  11. Nuclear reactor building

    DOE Patents [OSTI]

    Gou, Perng-Fei (Saratoga, CA); Townsend, Harold E. (Campbell, CA); Barbanti, Giancarlo (Sirtori, IT)

    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.

  12. Nuclear reactor building

    DOE Patents [OSTI]

    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.

  13. Nuclear reactor safety device

    DOE Patents [OSTI]

    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.

  14. Nuclear reactor control apparatus

    DOE Patents [OSTI]

    Sridhar, Bettadapur N. (Cupertino, CA)

    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.

  15. Nuclear power reactor instrumentation systems handbook. Volume...

    Office of Scientific and Technical Information (OSTI)

    Nuclear power reactor instrumentation systems handbook. Volume 1 Citation Details In-Document Search Title: Nuclear power reactor instrumentation systems handbook. Volume 1 You...

  16. Nuclear reactor control

    DOE Patents [OSTI]

    Cawley, William E. (Phoenix, AZ); Warnick, Robert F. (Pasco, WA)

    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.

  17. Nuclear reactor control apparatus

    DOE Patents [OSTI]

    Sridhar, Bettadapur N. (Cupertino, CA)

    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.

  18. Incorporation of Hydride Nuclear Fuels in Commercial Light Water Reactors

    E-Print Network [OSTI]

    Terrani, Kurt Amir

    2010-01-01

    Fundamental aspects of nuclear reactor fuel elements.Unlike permanent nuclear reactor core components, nuclearof the first nuclear reactors, commercial nuclear fuel still

  19. Nuclear Reactors and Technology; (USA)

    SciTech Connect (OSTI)

    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.

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

    E-Print Network [OSTI]

    Djurcic, Zelimir

    2009-01-01

    neutrino Production at Nuclear Reactors Z. Djurcic 1 , ?emission rates from nuclear reactors are determined fromlarge commercial nuclear reactors are playing an important

  1. NUCLEAR POWER REACTORS AND ASSOCIATED PLANTS; 05 NUCLEAR FUELS...

    Office of Scientific and Technical Information (OSTI)

    Title list of documents made publicly available, January 1-31, 1998 NONE 21 NUCLEAR POWER REACTORS AND ASSOCIATED PLANTS; 05 NUCLEAR FUELS; BIBLIOGRAPHIES; NUCLEAR POWER PLANTS;...

  2. Reactor- Nuclear Science Center 

    E-Print Network [OSTI]

    Unknown

    2011-08-17

    A neutronic evaluation of two reactor benchmark problems was performed. The benchmark problems describe typical PWR uranium and plutonium (mixed oxide) fueled lattices. WIMSd4m, a neutron transport lattice code, was used to evaluate multigroup...

  3. Nuclear reactor downcomer flow deflector

    DOE Patents [OSTI]

    Gilmore, Charles B. (Greensburg, PA); Altman, David A. (Pittsburgh, PA); Singleton, Norman R. (Murrysville, PA)

    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 Safety Design Criteria

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    1993-01-19

    The order establishes nuclear safety criteria applicable to the design, fabrication, construction, testing, and performance requirements of nuclear reactor facilities and safety class structures, systems, and components (SSCs) within these facilities. Cancels paragraphs 8a and 8b of DOE 5480.6. Cancels DOE O 5480.6 in part. Supersedes DOE 5480.1, dated 1-19-93. Certified 11-18-10.

  5. Reactor and Nuclear Systems Division (RNSD)

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    RNSD Home Research Groups Advanced Reactor Systems & Safety Nuclear Data & Criticality Safety Nuclear Security Modeling Radiation Safety Information Computational Center Radiation...

  6. Research Reactor Conversion | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

    Reactor Conversion | National Nuclear Security Administration Facebook Twitter Youtube Flickr RSS People Mission Managing the Stockpile Preventing Proliferation Powering the...

  7. Aerial of Nuclear Science Reactor 

    E-Print Network [OSTI]

    Unknown

    2011-08-17

    Small graphite-moderated and gas-cooled reactors have been around since the beginning of the atomic age. Though their existence in the past has been associated with nuclear weapons programs, they are capable of being used in civilian power programs...

  8. Computer aided nuclear reactor modeling 

    E-Print Network [OSTI]

    Warraich, Khalid Sarwar

    1995-01-01

    Nuclear reactor modeling is an important activity that lets us analyze existing as well as proposed systems for safety, correct operation, etc. The quality of a analysis is directly proportional to the quality of the model used. In this work we look...

  9. Horizontal baffle for nuclear reactors

    DOE Patents [OSTI]

    Rylatt, John A. (Monroeville, PA)

    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.

  10. Propellant actuated nuclear reactor steam depressurization valve

    DOE Patents [OSTI]

    Ehrke, Alan C. (San Jose, CA); Knepp, John B. (San Jose, CA); Skoda, George I. (Santa Clara, CA)

    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.

  11. Flow duct for nuclear reactors

    DOE Patents [OSTI]

    Straalsund, Jerry L. (Richland, WA)

    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. Energy Secretary to Visit Georgia Nuclear Reactor Site and Tennessee...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    to Visit Georgia Nuclear Reactor Site and Tennessee Laboratory to Highlight Administration Support for Nuclear Energy Energy Secretary to Visit Georgia Nuclear Reactor Site and...

  13. Reactivity control assembly for nuclear reactor

    DOE Patents [OSTI]

    Bollinger, Lawrence R. (Schenectady, NY)

    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.

  14. Effect of nuclear ownership on utility bond ratings and yields

    SciTech Connect (OSTI)

    Nesse, R.J.

    1982-02-01

    The major objective of this study was to test the hypothesis that investors have required an additional interest rate premium before purchasing bonds of utilities with large investments in nuclear facilities. The study required several tasks. First, the literature relating to firm bankruptcy and default was reviewed. Second, the failing financial health of the electric utility industry was assessed in terms of construction problems, the impact of federal and state regulations, and the impact of Three Mile Island. Finally, data were collected on 63 electric utilities. This allowed statistical estimation of the magnitude of the risk premium associated with utility involvement in nuclear power. The effect of this involvement on a utility's bond ratings was also examined. Multiple regression was the statistical tool used for the statistical testing and estimation.

  15. Nuclear reactor multiphysics via bond graph formalism

    E-Print Network [OSTI]

    Sosnovsky, Eugeny

    2014-01-01

    This work proposes a simple and effective approach to modeling nuclear reactor multiphysics problems using bond graphs. Conventional multiphysics simulation paradigms normally use operator splitting, which treats the ...

  16. Simulated nuclear reactor fuel assembly

    DOE Patents [OSTI]

    Berta, Victor T. (Idaho Falls, ID)

    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.

  17. Simulated nuclear reactor fuel assembly

    DOE Patents [OSTI]

    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.

  18. APPLICATION OF DATA ANALYSIS TECHNIQUES TO NUCLEAR REACTOR

    E-Print Network [OSTI]

    Kunz, Robert Francis

    1 APPLICATION OF DATA ANALYSIS TECHNIQUES TO NUCLEAR REACTOR SYSTEMS CODE ACCURACY ASSESSMENT) has been developed by the authors to provide quantitative comparisons between nuclear reactor systems. 1. INTRODUCTION In recent years, the commercial nuclear reactor industry has focused significant

  19. Proliferation Resistant Nuclear Reactor Fuel

    SciTech Connect (OSTI)

    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 we posit that the exploration, development, and implementation of intrinsic mechanisms such as discussed here are part of a balanced approach aimed at preventing the misuse of nuclear material for nuclear-energy applications.

  20. Liquid metal cooled nuclear reactor plant system

    DOE Patents [OSTI]

    Hunsbedt, Anstein (Los Gatos, CA); Boardman, Charles E. (Saratoga, CA)

    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.

  1. Digital computer operation of a nuclear reactor

    DOE Patents [OSTI]

    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.

  2. Digital computer operation of a nuclear reactor

    DOE Patents [OSTI]

    Colley, Robert W. (Richland, WA)

    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.

  3. Physics-based multiscale coupling for full core nuclear reactor...

    Office of Scientific and Technical Information (OSTI)

    multiscale coupling for full core nuclear reactor simulation Numerical simulation of nuclear reactors is a key technology in the quest for improvements in efficiency, safety,...

  4. Fundamental aspects of nuclear reactor fuel elements (Technical...

    Office of Scientific and Technical Information (OSTI)

    Fundamental aspects of nuclear reactor fuel elements Citation Details In-Document Search Title: Fundamental aspects of nuclear reactor fuel elements You are accessing a document...

  5. Nuclear reactor internals alignment configuration

    DOE Patents [OSTI]

    Gilmore, Charles B. (Greensburg, PA); Singleton, Norman R. (Murrysville, PA)

    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.

  6. Nuclear reactor composite fuel assembly

    DOE Patents [OSTI]

    Burgess, Donn M. (Richland, WA); Marr, Duane R. (West Richland, WA); Cappiello, Michael W. (Richland, WA); Omberg, Ronald P. (Richland, WA)

    1980-01-01

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

  7. Shutdown system for a nuclear reactor

    DOE Patents [OSTI]

    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.

  8. Fast-acting nuclear reactor control device

    DOE Patents [OSTI]

    Kotlyar, Oleg M. (Idaho Falls, ID); West, Phillip B. (Idaho Falls, ID)

    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. Shutdown system for a nuclear reactor

    DOE Patents [OSTI]

    Groh, Edward F. (Naperville, IL); Olson, Arne P. (Western Springs, IL); Wade, David C. (Naperville, IL); Robinson, Bryan W. (Oak Lawn, IL)

    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.

  10. Gas-cooled nuclear reactor

    DOE Patents [OSTI]

    Peinado, Charles O. (La Jolla, CA); Koutz, Stanley L. (San Diego, CA)

    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.

  11. Nuclear reactor safeguards and monitoring with antineutrino detectors A. Bernsteina)

    E-Print Network [OSTI]

    Gratta, Giorgio

    Nuclear reactor safeguards and monitoring with antineutrino detectors A. Bernsteina) Sandia of nuclear reactor types, including power reactors, research reactors, and plutonium production reactors-understood principles that govern the core's evolution in time, can be used to determine whether the reactor is being

  12. Today and Future Neutrino Experiments at Krasnoyarsk Nuclear Reactor

    E-Print Network [OSTI]

    Yu. V. Kozlov; S. V. Khalturtsev; I. N. Machulin; A. V. Martemyanov; V. P. Martemyanov; A. A. Sabelnikov; V. G. Tarasenkov; E. V. Turbin; V. N. Vyrodov; L. A. Popeko; A. V. Cherny; G. A. Shishkina

    1999-12-22

    The results of undergoing experiments and new experiment propositions at Krasnoyarsk underground nuclear reactor are presented

  13. Nuclear reactor shield including magnesium oxide

    DOE Patents [OSTI]

    Rouse, Carl A. (Del Mar, CA); Simnad, Massoud T. (La Jolla, CA)

    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.

  14. Large Scale Weather Control Using Nuclear Reactors

    E-Print Network [OSTI]

    Moninder Singh Modgil

    2002-10-02

    It is pointed out that controlled release of thermal energy from fission type nuclear reactors can be used to alter weather patterns over significantly large geographical regions. (1) Nuclear heat creates a low pressure region, which can be used to draw moist air from oceans, onto deserts. (2) Creation of low pressure zones over oceans using Nuclear heat can lead to Controlled Cyclone Creation (CCC).(3) Nuclear heat can also be used to melt glaciers and control water flow in rivers.

  15. Large Scale Weather Control Using Nuclear Reactors

    E-Print Network [OSTI]

    Singh-Modgil, M

    2002-01-01

    It is pointed out that controlled release of thermal energy from fission type nuclear reactors can be used to alter weather patterns over significantly large geographical regions. (1) Nuclear heat creates a low pressure region, which can be used to draw moist air from oceans, onto deserts. (2) Creation of low pressure zones over oceans using Nuclear heat can lead to Controlled Cyclone Creation (CCC).(3) Nuclear heat can also be used to melt glaciers and control water flow in rivers.

  16. Nuclear reactor vessel fuel thermal insulating barrier

    DOE Patents [OSTI]

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

    2013-03-19

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

  17. Nuclear reactor construction with bottom supported reactor vessel

    DOE Patents [OSTI]

    Sharbaugh, John E. (Bullskin Township, Fayette County, PA)

    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 structure base mat so as to insulate the reactor vessel bottom end wall from the containment structure base mat and allow the reactor vessel bottom end wall to freely expand as it heats up while providing continuous support thereof. Further, a deck is supported upon the side wall of the containment structure above the top open end of the reactor vessel, and a plurality of serially connected extendible and retractable annular bellows extend between the deck and the top open end of the reactor vessel and flexibly and sealably interconnect the reactor vessel at its top end to the deck. An annular guide ring is disposed on the containment structure and extends between its side wall and the top open end of the reactor vessel for providing lateral support of the reactor vessel top open end by limiting imposition of lateral loads on the annular bellows by the occurrence of a lateral seismic event.

  18. Nuclear facilities: criteria for the design and operation of ventilation systems for nuclear installations other than nuclear reactors

    E-Print Network [OSTI]

    International Organization for Standardization. Geneva

    2004-01-01

    Nuclear facilities: criteria for the design and operation of ventilation systems for nuclear installations other than nuclear reactors

  19. NUCLEAR POWER AND RESEARCH REACTORS 1939 1942 1943 1944

    E-Print Network [OSTI]

    #12;#12;11 #12;2 NUCLEAR POWER AND RESEARCH REACTORS 1939 1942 1943 1944 Nuclear fission discovered 430 nuclear power reactors are operating in the world, and 103 nuclear power plants produce 20, naval reactors, and nuclear power plants. Oak Ridge experiments byArt Snell in 1944 showed that 10 tons

  20. Heat dissipating nuclear reactor with metal liner

    DOE Patents [OSTI]

    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.

  1. Heat dissipating nuclear reactor with metal liner

    DOE Patents [OSTI]

    Gluekler, Emil L. (San Jose, CA); Hunsbedt, Anstein (Los Gatos, CA); Lazarus, Jonathan D. (Sunnyvale, CA)

    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.

  2. Reactivity control assembly for nuclear reactor. [LMFBR

    DOE Patents [OSTI]

    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.

  3. ME 361E Nuclear Reactor Engineering ABET EC2000 syllabus

    E-Print Network [OSTI]

    Ben-Yakar, Adela

    ME 361E ­ Nuclear Reactor Engineering Page 1 ABET EC2000 syllabus ME 361E ­ Nuclear Reactor; neutron diffusion and moderation; reactor equations; Fermi Age theory; multigroup and multiregional students should be able to: · Compare and contrast numerous nuclear reactor designs · Calculate the effects

  4. Optimally moderated nuclear fission reactor and fuel source therefor

    DOE Patents [OSTI]

    Ougouag, Abderrafi M. (Idaho Falls, ID); Terry, William K. (Shelley, ID); Gougar, Hans D. (Idaho Falls, ID)

    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.

  5. NUCLEAR REACTORS; PIPES; SEISMIC EFFECTS; SUPPORTS; DYNAMIC LOADS...

    Office of Scientific and Technical Information (OSTI)

    Limit analysis of pipe clamps Flanders, H.E. Jr. 22 GENERAL STUDIES OF NUCLEAR REACTORS; PIPES; SEISMIC EFFECTS; SUPPORTS; DYNAMIC LOADS; HEAT TRANSFER; HYDRAULICS; REACTOR SAFETY;...

  6. Design and Transient Analysis of Passive Safety Cooling Systems for Advanced Nuclear Reactors

    E-Print Network [OSTI]

    Galvez, Cristhian

    2011-01-01

    L. J. Hamilton Nuclear Reactor Analysis John Wiley and Sons,R. J. Neuhold, Introductury Nuclear Reactor Dynamics. ANSL. J. Hamilton Nuclear Reactor Analysis John Wiley and Sons,

  7. White paper report on using nuclear reactors to search for a value of theta13

    E-Print Network [OSTI]

    2004-01-01

    PAPER REPORT on Using Nuclear Reactors to Search for a valuetimely new experiment at a nuclear reactor sensitive to theand judicious choice of a nuclear reactor. The dominant

  8. Cooling system for a nuclear reactor

    DOE Patents [OSTI]

    Amtmann, Hans H. (Rancho Santa Fe, CA)

    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.

  9. Nuclear reactor fissile isotopes antineutrino spectra

    E-Print Network [OSTI]

    V. Sinev

    2012-07-30

    Positron spectrum from inverse beta decay reaction on proton was measured in 1988-1990 as a result of neutrino exploration experiment. The measured spectrum has the largest statistics and lowest energy threshold between other neutrino experiments made that time at nuclear reactors. On base of the positron spectrum the standard antineutrino spectrum for typical reactor fuel composition was restored. In presented analysis the partial spectra forming this standard spectrum were extracted using specific method. They could be used for neutrino experiments data analysis made at any fuel composition of reactor core.

  10. Theta 13 Determination with Nuclear Reactors

    E-Print Network [OSTI]

    F. Dalnoki-Veress

    2004-06-24

    Recently there has been a lot of interest around the world in the use of nuclear reactors to measure theta 13, the last undetermined angle in the 3-neutrino mixing scenario. In this paper the motivations for theta 13 measurement using short baseline nuclear reactor experiments are discussed. The features of such an experiment are described in the context of Double Chooz, which is a new project planned to start data-taking in 2008, and to reach a sensitivity of sinsq(2 theta 13) < 0.03.

  11. 22.312 Engineering of Nuclear Reactors, Fall 2004

    E-Print Network [OSTI]

    Buongiorno, Jacopo, 1971-

    Engineering principles of nuclear reactors, emphasizing power reactors. Power plant thermodynamics, reactor heat generation and removal (single-phase as well as two-phase coolant flow and heat transfer), and structural ...

  12. 22.312 Engineering of Nuclear Reactors, Fall 2002

    E-Print Network [OSTI]

    Todreas, Neil E.

    Engineering principles of nuclear reactors, emphasizing power reactors. Power plant thermodynamics, reactor heat generation and removal (single-phase as well as two-phase coolant flow and heat transfer), and structural ...

  13. CRAD, Nuclear Reactor Facility Operations - December 4, 2014...

    Broader source: Energy.gov (indexed) [DOE]

    CRAD, Nuclear Reactor Facility Operations - December 4, 2014 (EA CRAD 31-08, Rev. 0) Nuclear Reactor Faclity Operations Criteria Review and Approach Document (EA CRAD 31-08, Rev....

  14. Nuclear reactor alignment plate configuration

    DOE Patents [OSTI]

    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.

  15. Nuclear reactor shutdown control rod assembly

    DOE Patents [OSTI]

    Bilibin, Konstantin (North Hollywood, CA)

    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.

  16. Current Abstracts Nuclear Reactors and Technology

    SciTech Connect (OSTI)

    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.

  17. Passive heat transfer means for nuclear reactors

    DOE Patents [OSTI]

    Burelbach, James P. (Glen Ellyn, IL)

    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.

  18. MA50177: Scientific Computing Nuclear Reactor Simulation Generalised Eigenvalue Problems

    E-Print Network [OSTI]

    Wirosoetisno, Djoko

    MA50177: Scientific Computing Case Study Nuclear Reactor Simulation ­ Generalised Eigenvalue of a malfunction or of an accident experimentally, the numerical simulation of nuclear reactors is of utmost balance in a nuclear reactor are the two-group neutron diffusion equations -div (K1 u1) + (a,1 + s) u1 = 1

  19. ARTIGO INTERNET Professores visitam o maior reactor de Fuso Nuclear

    E-Print Network [OSTI]

    Instituto de Sistemas e Robotica

    ARTIGO INTERNET Professores visitam o maior reactor de Fusão Nuclear in http reactor de Fusão Nuclear Experiência aproxima investigação das futuras gerações Doze professores do ensino secundário visitaram o maior reactor de fusão nuclear da Terra (JET), no Reino Unido, na semana passada

  20. THE ECONOMICS OF NUCLEAR REACTORS: RENAISSANCE OR RELAPSE?

    E-Print Network [OSTI]

    Laughlin, Robert B.

    THE ECONOMICS OF NUCLEAR REACTORS: RENAISSANCE OR RELAPSE? MARK COOPER SENIOR FELLOW FOR ECONOMIC Findings Approach Hope and Hype vs. Reality in Nuclear Reactor Costs The Economic Cost of Low Carbon. INTRODUCTION 10 A. The Troubling History of Nuclear Reactor Costs B. Purpose and Outline II. THE STRUCTURE

  1. Nuclear Thermal Rockets: The Physics of the Fission Reactor

    E-Print Network [OSTI]

    Ross, Shane

    Nuclear Thermal Rockets: The Physics of the Fission Reactor Shane D. Ross Control and Dynamical heats up when it passes through a nuclear reactor, where controlled fission of some fissionable material, with the nuclear fission reactor as a heat source [Lawrence, Witter, and Humble, 1992]. it works essentially

  2. Nuclear Regulatory Commission Handling of Beyond Design Basis Events for Nuclear Power Reactors

    Broader source: Energy.gov [DOE]

    Presenter: Bill Reckley, Chief, Policy and Support Branch, Japan Lessons-Learned Project Directorate, Office of Nuclear Reactor Regulation, US Nuclear Regulatory Commission US Nuclear Regulatory Commission

  3. Rodded shutdown system for a nuclear reactor

    DOE Patents [OSTI]

    Golden, Martin P. (Penn Township, Allegheny County, PA); Govi, Aldo R. (Greensburg, PA)

    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.

  4. Nuclear reactor control room construction

    DOE Patents [OSTI]

    Lamuro, Robert C. (Pittsburgh, PA); Orr, Richard (Pittsburgh, PA)

    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.

  5. Nuclear reactor control room construction

    DOE Patents [OSTI]

    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.

  6. DOE fundamentals handbook: Nuclear physics and reactor theory

    SciTech Connect (OSTI)

    Not Available

    1993-01-01

    The Nuclear Physics and Reactor Theory Handbook was developed to assist nuclear facility operating contractors in providing operators, maintenance personnel, and the technical staff with the necessary fundamentals training to ensure a basic understanding of nuclear physics and reactor theory. The handbook includes information on atomic and nuclear physics; neutron characteristics; reactor theory and nuclear parameters; and the theory of reactor operation. This information will provide personnel with a foundation for understanding the scientific principles that are associated with various DOE nuclear facility operations and maintenance.

  7. DOE fundamentals handbook: Nuclear physics and reactor theory. Volume 1

    SciTech Connect (OSTI)

    Not Available

    1993-01-01

    The Nuclear Physics and Reactor Theory Handbook was developed to assist nuclear facility operating contractors in providing operators, maintenance personnel, and the technical staff with the necessary fundamentals training to ensure a basic understanding of nuclear physics and reactor theory. The handbook includes information on atomic and nuclear physics; neutron characteristics; reactor theory and nuclear parameters; and the theory of reactor operation. This information will provide personnel with a foundation for understanding the scientific principles that are associated with various DOE nuclear facility operations and maintenance.

  8. DOE fundamentals handbook: Nuclear physics and reactor theory. Volume 2

    SciTech Connect (OSTI)

    Not Available

    1993-01-01

    The Nuclear Physics and Reactor Theory Handbook was developed to assist nuclear facility operating contractors in providing operators, maintenance personnel, and the technical staff with the necessary fundamentals training to ensure a basic understanding of nuclear physics and reactor theory. The handbook includes information on atomic and nuclear physics; neutron characteristics; reactor theory and nuclear parameters; and the theory of reactor operation. This information will provide personnel with a foundation for understanding the scientific principles that are associated with various DOE nuclear facility operations and maintenance.

  9. Method for automatically scramming a nuclear reactor

    DOE Patents [OSTI]

    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.

  10. Oklo reactors and implications for nuclear science

    E-Print Network [OSTI]

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

    2014-04-19

    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 $^{175}$Lu ground state cross section for thermal neutron capture leading to the isomer $^{176\\mathrm{m}}$ Lu, 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 $\\alpha$ and the ratio $X_q=m_q/\\Lambda$ (where $m_q$ is the average of the $u$ and $d$ current quark masses and $\\Lambda$ is the mass scale of quantum chromodynamics). We suggest a formula for the combined sensitivity to $\\alpha$ and $X_q$ that exhibits the dependence on proton number $Z$ and mass number $A$, potentially allowing quantum electrodynamic and quantum chromodynamic effects to be disentangled if a broader range of isotopic abundance data becomes available.

  11. Thermoacoustic Thermometry for Nuclear Reactor Monitoring

    SciTech Connect (OSTI)

    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 prevention as quickly as possible. This is the question which we are attempting to answer: Is it possible to implement a self-powered sensor that could transmit data independently of electronic networks while taking advantage of the harsh operating environment of the nuclear reactor?

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

    E-Print Network [OSTI]

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

    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 electron anti-neutrino detectors is examined. We discuss reactor-to-reactor correlations between the leading uncertainties and their relevance to reactor anti-neutrino experiments.

  13. LIMITED POWER BURSTS IN DISTRIBUTED MODELS OF NUCLEAR REACTORS

    E-Print Network [OSTI]

    Bazhenov, Maxim

    LIMITED POWER BURSTS IN DISTRIBUTED MODELS OF NUCLEAR REACTORS M. V. Bazhenov and E. F. Sabaev UDC employed for analyzing reactor dynamics. Equations of this type are used for analyzing the stability of the reactor power, etc. Among these problems the question of the boundedness of reactor power bursts

  14. ASSESSMENT OF SMALL AND MODULAR REACTOR NUCLEAR FUEL COST 

    E-Print Network [OSTI]

    Pannier, Christopher 1992-

    2012-05-03

    INCAS INtegrated model for the Competitiveness Analysis of Small modular reactors LWR Light Water Reactor NEI Nuclear Energy Institute PWR Pressurized Water Reactor PHWR Pressurized Heavy Water Reactor SEMER Système d’Évaluation et de Modélisation... ...................................................... 27 8 LWR Fuel Cost ..................................................................................................... 28 9 SMR Fuel Cost ..................................................................................................... 29...

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

    E-Print Network [OSTI]

    Djurcic, Zelimir

    2009-01-01

    CALCULATIONS During the power cycle of a nuclear reactor,depleted. At the end of the power cycle, some frac- tion offuel throughout the power cycle is of interest to reactor

  16. Advanced nuclear reactor public opinion project

    SciTech Connect (OSTI)

    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.

  17. A comparison of nuclear reactor control room display panels 

    E-Print Network [OSTI]

    Bowers, Frances Renae

    1988-01-01

    . PAGE 38 79 CHAPTER I INTRODUCTION At approximately 4:00 am on March 28, 1979, several reactor coolant feedwater pumps malfunctioned in the Three Mile Island Unit 2 nuclear power plant. Thus began the worst accident to date in the U. S. nuclear...: Dr. Rodger S. Koppa A study was conducted to investigate the use of computer generated displays to operate nuclear reactor power plants. The AGN-201 reactor at Texas A&M university was the reactor studied. After observing several licensed reactor...

  18. Nuclear reactor flow control method and apparatus

    DOE Patents [OSTI]

    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.

  19. Neutrino Oscillation Experiments at Nuclear Reactors

    E-Print Network [OSTI]

    Giorgio Gratta

    1999-05-06

    In this paper I give an overview of the status of neutrino oscillation experiments performed using nuclear reactors as sources of neutrinos. I review the present generation of experiments (Chooz and Palo Verde) with baselines of about 1 km as well as the next generation that will search for oscillations with a baseline of about 100 km. While the present detectors provide essential input towards the understanding of the atmospheric neutrino anomaly, in the future, the KamLAND reactor experiment represents our best opportunity to study very small mass neutrino mixing in laboratory conditions. In addition KamLAND with its very large fiducial mass and low energy threshold, will also be sensitive to a broad range of different physics.

  20. Nuclear reactor flow control method and apparatus

    DOE Patents [OSTI]

    Church, John P. (1204 Woodbine Rd., Aiken, SC 29803)

    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.

  1. Fuel handling system for a nuclear reactor

    DOE Patents [OSTI]

    Saiveau, James G. (Hickory Hills, IL); Kann, William J. (Park Ridge, IL); Burelbach, James P. (Glen Ellyn, IL)

    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.

  2. Idaho Site Obtains Patent for Nuclear Reactor Sodium Cleanup Treatment

    Broader source: Energy.gov [DOE]

    IDAHO FALLS, Idaho – An innovative idea for cleaning up sodium in a decommissioned nuclear reactor at EM’s Idaho site grew from a carpool discussion.

  3. Foreign Research Reactor Spent Nuclear Fuel Acceptance Program

    National Nuclear Security Administration (NNSA)

    * Complete reactor control rod system. * Note: Does not include the steam turbine generator portion of the power plant. - Sensitive nuclear technology: Any information...

  4. Modeling and Simulation for Nuclear Reactors Hub | Department...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    to help the nuclear industry make reactors more efficient through computer modeling and simulation. The Department's Energy Innovation Hubs are helping to advance promising areas...

  5. Nuclear Reactor Technologies | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious Rank EERE: Alternative Fuelsof EnergyApril 2014Department ofWind CareerEnergy Nuclear FuelsReactor

  6. Fluid sampling system for a nuclear reactor

    DOE Patents [OSTI]

    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

    DOE Patents [OSTI]

    Lau, Louis K. (Monroeville, PA); Alper, Naum I. (Monroeville, PA)

    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. PHYSICS OF NUCLEAR REACTORS Nuclear reactions and cross sections 1-10

    E-Print Network [OSTI]

    Danon, Yaron

    PHYSICS OF NUCLEAR REACTORS Nuclear reactions and cross sections 1-10 10 11 12 13 14 15 16 17 18 19 neutron wavelength, D is given by: cE mM Mm 2 + = h D , (1.22) 1 Bell and Glasstone, Nuclear Reactor Theory, p. 392, 1970. #12;PHYSICS OF NUCLEAR REACTORS Nuclear reactions and cross sections 1-11 Where m

  9. Neutron transport analysis for nuclear reactor design

    DOE Patents [OSTI]

    Vujic, Jasmina L. (Lisle, IL)

    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.

  10. Minimizing or eliminating refueling of nuclear reactor

    DOE Patents [OSTI]

    Doncals, Richard A. (Washington, PA); Paik, Nam-Chin (Pittsburgh, PA); Andre, Sandra V. (Hempfield Township, Westmoreland County, PA); Porter, Charles A. (Rostraver Township, Westmoreland County, PA); Rathbun, Roy W. (Greensburg, PA); Schwallie, Ambrose L. (Greensburg, PA); Petras, Diane S. (Penn Township, Westmoreland County, PA)

    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.

  11. Neutron transport analysis for nuclear reactor design

    DOE Patents [OSTI]

    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.

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

    SciTech Connect (OSTI)

    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.

  13. C Produced by Nuclear Power Reactors Generation and Characterization of

    E-Print Network [OSTI]

    Haviland, David

    14 C Produced by Nuclear Power Reactors ­ Generation and Characterization of Gaseous, Liquid in the terrestrial environment in the vicinity of two European nuclear power plants. Radiocarbon 46(2)863­868. III levels in the vicinity of the Lithuanian nuclear power plant Ignalina. Nuclear Instruments and Methods

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

    SciTech Connect (OSTI)

    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.

  15. Weld monitor and failure detector for nuclear reactor system

    DOE Patents [OSTI]

    Sutton, Jr., Harry G. (Mt. Lebanon, PA)

    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.

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

    SciTech Connect (OSTI)

    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.

  17. P.C. 21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; BWR...

    Office of Scientific and Technical Information (OSTI)

    Erosioncorrosion-induced pipe wall thinning in US Nuclear Power Plants Wu, P.C. 21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; BWR TYPE REACTORS; PIPES; CORROSION; EROSION;...

  18. MOOSE simulating nuclear reactor CRUD buildup

    SciTech Connect (OSTI)

    2014-02-06

    This simulation uses multiple physical models to show how the buildup of boron deposits on reactor fuel can affect performance and the reactor's power profile.

  19. MOOSE simulating nuclear reactor CRUD buildup

    ScienceCinema (OSTI)

    None

    2014-07-21

    This simulation uses multiple physical models to show how the buildup of boron deposits on reactor fuel can affect performance and the reactor's power profile.

  20. Nuclear reactor cooling system decontamination reagent regeneration

    DOE Patents [OSTI]

    Anstine, Larry D. (San Jose, CA); James, Dean B. (Saratoga, CA); Melaika, Edward A. (Berkeley, CA); Peterson, Jr., John P. (Livermore, CA)

    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.

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

    SciTech Connect (OSTI)

    Sekimoto, Hiroshi, E-mail: hsekimot@gmail.com [Emeritus Professor, Tokyo Institute of Technology (Japan)

    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.

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

    SciTech Connect (OSTI)

    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).

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

    SciTech Connect (OSTI)

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

    2012-09-01

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

  4. Nuclear Graphite -Fission Reactor Brief Outline of Experience and

    E-Print Network [OSTI]

    McDonald, Kirk

    Nuclear Graphite - Fission Reactor Brief Outline of Experience and Understanding Professor Barry J Marsden and Dr. Graham N Hall Nuclear Graphite Research Group The University of Manchester 20 March 201313 9PL Tel: +44 (0) 161 275 4399, barry.marsden@manchester.ac.uk #12;Overview · Nuclear Graphite

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

    DOE Patents [OSTI]

    Hunsbedt, Anstein (Los Gatos, CA); Boardman, Charles E. (Saratoga, CA)

    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.

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

    DOE Patents [OSTI]

    Hunsbedt, Anstein (Los Gatos, CA); Boardman, Charles E. (Saratoga, CA); Hui, Marvin M. (Sunnyvale, CA); Berglund, Robert C. (Saratoga, CA)

    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.

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

    SciTech Connect (OSTI)

    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.

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

    SciTech Connect (OSTI)

    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. Standard practice for evaluation of surveillance capsules from light-water moderated nuclear power reactor vessels

    E-Print Network [OSTI]

    American Society for Testing and Materials. Philadelphia

    2010-01-01

    Standard practice for evaluation of surveillance capsules from light-water moderated nuclear power reactor vessels

  10. Solid0Core Heat-Pipe Nuclear Batterly Type Reactor

    SciTech Connect (OSTI)

    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).

  11. Discovery sheds light on nuclear reactor fuel behavior during...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Discovery sheds light on nuclear reactor fuel behavior during a severe event By Angela Hardin * November 20, 2014 Tweet EmailPrint A new discovery about the atomic structure of...

  12. PEBBLE-BED NUCLEAR REACTOR SYSTEM PHYSICS AND FUEL UTILIZATION 

    E-Print Network [OSTI]

    Kelly, Ryan 1989-

    2011-04-20

    The Generation IV Pebble Bed Modular Reactor (PMBR) design may be used for electricity production, co-generation applications (industrial heat, hydrogen production, desalination, etc.), and could potentially eliminate some high level nuclear wastes...

  13. Fuel assembly transfer basket for pool type nuclear reactor vessels

    DOE Patents [OSTI]

    Fanning, Alan W. (San Jose, CA); Ramsour, Nicholas L. (San Jose, CA)

    1991-01-01

    A fuel assembly transfer basket for a pool type, liquid metal cooled nuclear reactor having a side access loading and unloading port for receiving and relinquishing fuel assemblies during transfer.

  14. Observer-based fault detection for nuclear reactors

    E-Print Network [OSTI]

    Li, Qing, 1972-

    2001-01-01

    This is a study of fault detection for nuclear reactor systems. Basic concepts are derived from fundamental theories on system observers. Different types of fault- actuator fault, sensor fault, and system dynamics fault ...

  15. Development of Technical Nuclear Forensics for Spent Research Reactor Fuel 

    E-Print Network [OSTI]

    Sternat, Matthew Ryan 1982-

    2012-11-20

    Pre-detonation technical nuclear forensics techniques for research reactor spent fuel were developed in a collaborative project with Savannah River National Lab ratory. An inverse analysis method was employed to reconstruct ...

  16. Liquid metal cooled nuclear reactors with passive cooling system

    DOE Patents [OSTI]

    Hunsbedt, Anstein (Los Gatos, CA); Fanning, Alan W. (San Jose, CA)

    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.

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

    SciTech Connect (OSTI)

    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).

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

    SciTech Connect (OSTI)

    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).

  19. Neutron spectrometer for fast nuclear reactors

    E-Print Network [OSTI]

    M. Osipenko; M. Ripani; G. Ricco; B. Caiffi; F. Pompili; M. Pillon; M. Angelone; G. Verona-Rinati; R. Cardarelli; G. Mila; S. Argiro

    2015-05-25

    In this paper we describe the development and first tests of a neutron spectrometer designed for high flux environments, such as the ones found in fast nuclear reactors. The spectrometer is based on the conversion of neutrons impinging on $^6$Li into $\\alpha$ and $t$ whose total energy comprises the initial neutron energy and the reaction $Q$-value. The $^6$LiF layer is sandwiched between two CVD diamond detectors, which measure the two reaction products in coincidence. The spectrometer was calibrated at two neutron energies in well known thermal and 3 MeV neutron fluxes. The measured neutron detection efficiency varies from 4.2$\\times 10^{-4}$ to 3.5$\\times 10^{-8}$ for thermal and 3 MeV neutrons, respectively. These values are in agreement with Geant4 simulations and close to simple estimates based on the knowledge of the $^6$Li(n,$\\alpha$)$t$ cross section. The energy resolution of the spectrometer was found to be better than 100 keV when using 5 m cables between the detector and the preamplifiers.

  20. Neutron spectrometer for fast nuclear reactors

    E-Print Network [OSTI]

    Osipenko, M; Ricco, G; Caiffi, B; Pompili, F; Pillon, M; Angelone, M; Verona-Rinati, G; Cardarelli, R; Mila, G; Argiro, S

    2015-01-01

    In this paper we describe the development and first tests of a neutron spectrometer designed for high flux environments, such as the ones found in fast nuclear reactors. The spectrometer is based on the conversion of neutrons impinging on $^6$Li into $\\alpha$ and $t$ whose total energy comprises the initial neutron energy and the reaction $Q$-value. The $^6$LiF layer is sandwiched between two CVD diamond detectors, which measure the two reaction products in coincidence. The spectrometer was calibrated at two neutron energies in well known thermal and 3 MeV neutron fluxes. The measured neutron detection efficiency varies from 4.2$\\times 10^{-4}$ to 3.5$\\times 10^{-8}$ for thermal and 3 MeV neutrons, respectively. These values are in agreement with Geant4 simulations and close to simple estimates based on the knowledge of the $^6$Li(n,$\\alpha$)$t$ cross section. The energy resolution of the spectrometer was found to be better than 100 keV when using 5 m cables between the detector and the preamplifiers.

  1. Improved Design of Nuclear Reactor Control System | U.S. DOE...

    Office of Science (SC) Website

    Improved Design of Nuclear Reactor Control System Nuclear Physics (NP) NP Home About Research Facilities Science Highlights Benefits of NP Applications of Nuclear Science...

  2. Nuclear reactor characteristics and operational history

    Gasoline and Diesel Fuel Update (EIA)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of Natural GasAdjustments (Billion Cubic Feet)Decade Year-0 Year-1 Year-2 Year-3 Year-4Barrels)(Dollars2. Ownership Data,

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

    SciTech Connect (OSTI)

    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.

  4. Generating unstructured nuclear reactor core meshes in parallel

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    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

  5. Generating unstructured nuclear reactor core meshes in parallel

    SciTech Connect (OSTI)

    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.

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

    SciTech Connect (OSTI)

    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 cost of decommissioning a PWR is lowest for ENTOMB and highest for SAFSTOR • the estimated cost of decommissioning a BWR is lowest for OECON and highest for SAFSTOR. In all cases, SAFSTOR has the lowest occupational radiation dose and the highest cost.

  7. Nuclear Reactor Safeguards and Monitoring with Antineutrino Detectors

    E-Print Network [OSTI]

    Adam Bernstein; Yifang Wang; Giorgio Gratta; Todd West

    2001-08-01

    Cubic-meter-sized antineutrino detectors can be used to non-intrusively, robustly and automatically monitor and safeguard a wide variety of nuclear reactor types, including power reactors, research reactors, and plutonium production reactors. Since the antineutrino spectra and relative yields of fissioning isotopes depend on the isotopic composition of the core, changes in composition can be observed without ever directly accessing the core itself. Information from a modest-sized antineutrino detector, coupled with the well-understood principles that govern the core's evolution in time, can be used to determine whether the reactor is being operated in an illegitimate way. A group at Sandia is currently constructing a one cubic meter antineutrino detector at the San Onofre reactor site in California to demonstrate these principles.

  8. Spent nuclear fuel discharges from US reactors 1992

    SciTech Connect (OSTI)

    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.

  9. SNIF: A Futuristic Neutrino Probe for Undeclared Nuclear Fission Reactors

    E-Print Network [OSTI]

    Lasserre, Thierry; Mention, Guillaume; Reboulleau, Romain; Cribier, Michel; Letourneau, Alain; Lhuillier, David

    2010-01-01

    Today reactor neutrino experiments are at the cutting edge of fundamental research in particle physics. Understanding the neutrino is far from complete, but thanks to the impressive progress in this field over the last 15 years, a few research groups are seriously considering that neutrinos could be useful for society. The International Atomic Energy Agency (IAEA) works with its Member States to promote safe, secure and peaceful nuclear technologies. In a context of international tension and nuclear renaissance, neutrino detectors could help IAEA to enforce the Treaty on the Non-Proliferation of Nuclear Weapons (NPT). In this article we discuss a futuristic neutrino application to detect and localize an undeclared nuclear reactor from across borders. The SNIF (Secret Neutrino Interactions Finder) concept proposes to use a few hundred thousand tons neutrino detectors to unveil clandestine fission reactors. Beyond previous studies we provide estimates of all known background sources as a function of the detecto...

  10. Spent nuclear fuel discharges from US reactors 1993

    SciTech Connect (OSTI)

    Not Available

    1995-02-01

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

  11. http://arXiv.org/physics/0507088 Teaching About Nature's Nuclear Reactors

    E-Print Network [OSTI]

    Learned, John

    http://arXiv.org/physics/0507088 Teaching About Nature's Nuclear Reactors J. Marvin Herndon reactors existed in uranium deposits on Earth long before Enrico Fermi built the first man-made nuclear reactors. The subject of planetocentric nuclear fission reactors can be a jumping off point for stimulating

  12. Nuclear data requirements for fission reactor neutronics calculations.

    SciTech Connect (OSTI)

    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.

  13. June 28, 2005 France to Be Site of World's First Nuclear Fusion Reactor

    E-Print Network [OSTI]

    June 28, 2005 France to Be Site of World's First Nuclear Fusion Reactor By CRAIG S. SMITH PARIS the reactor in the southern French city of Cadarache. Nuclear fusion is the process by which the atomic nuclei than burning fossil fuels or even nuclear fission, which is used in nuclear reactors today but produces

  14. Decision-support tool for assessing future nuclear reactor generation portfolios.

    E-Print Network [OSTI]

    Oosterlee, Cornelis W. "Kees"

    Decision-support tool for assessing future nuclear reactor generation portfolios. Shashi Jain, where especially capital costs are known to be highly uncertain. Differ- ent nuclear reactor types uncertainties in the cost elements of a nuclear power plant, to provide an optimal portfolio of nuclear reactors

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

    DOE Patents [OSTI]

    Jacox, Michael G. (Idaho Falls, ID); Drexler, Robert L. (Idaho Falls, ID); Hunt, Robert N. M. (Idaho Falls, ID); Lake, James A. (Idaho Falls, ID)

    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.

  16. A Compact Nuclear Fusion Reactor for Space Flights

    SciTech Connect (OSTI)

    Nastoyashchiy, Anatoly F. [SRC Troitsk Institute for Innovation and Fusion Research, TRINITI 142190 Troitsk Moscow Reg. (Russian Federation)

    2006-05-02

    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.

  17. Spectral Structure of Electron Antineutrinos from Nuclear Reactors

    E-Print Network [OSTI]

    D. A. Dwyer; T. J. Langford

    2014-07-04

    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-principle 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 substructure in the antineutrino spectrum due to Coulomb effects in beta decay. Precise measurement of this substructure can constrain nuclear reactor physics. The substructure can be a systematic uncertainty for measurements utilizing the detailed spectral shape.

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

    SciTech Connect (OSTI)

    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 Generation Safeguards Initiative (NGSI).

  19. 288 Int. J. Nuclear Energy Science and Technology, Vol. 7, No. 4, 2013 Multi-physics modelling of nuclear reactors

    E-Print Network [OSTI]

    Demazière, Christophe

    of nuclear reactors: current practices in a nutshell Christophe Demazière Department of Applied Physics of nuclear reactors are based on the use of different solvers for resolving the different physical fields and the corresponding approximations. Keywords: nuclear reactors; multi-physics; multi-scale; modelling; deterministic

  20. Low exchange element for nuclear reactor

    DOE Patents [OSTI]

    Brogli, Rudolf H. (Aarau, CH); Shamasunder, Bangalore I. (Encinitas, CA); Seth, Shivaji S. (Encinitas, CA)

    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.

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

    SciTech Connect (OSTI)

    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.

  2. Exploratory Nuclear Reactor Safety Analysis and Visualization...

    Office of Scientific and Technical Information (OSTI)

    via Integrated Topological and Geometric Techniques A recent trend in the nuclear power engineering field is the implementation of heavily computational and time consuming...

  3. Systems Issues in Nuclear Reactor Safety

    E-Print Network [OSTI]

    de Weck, Olivier L.

    postulated Loss-of-Coolant Accident (LOCA): 9 (LOCA): a double-ended break of the largest reactor coolant line, the concurrent loss of offsite power, and a single failure of an active ECCS component Loss Of Offsite Power Initiating Event 51,940 Steam Generator Tube Rupture Initiating Event 41,200 12

  4. Production capabilities in US nuclear reactors for medical radioisotopes

    SciTech Connect (OSTI)

    Mirzadeh, S.; Callahan, A.P.; Knapp, F.F. Jr. [Oak Ridge National Lab., TN (United States); Schenter, R.E. [Westinghouse Hanford Co., Richland, WA (United States)

    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.

  5. Reactor control system upgrade for the McClellan Nuclear Radiation Center

    E-Print Network [OSTI]

    Power, Michael A.

    1999-01-01

    a new reactor control system for the McClellan NuclearI REACTOR CONTROL SYSTEM UPGRADE FOR THE McCLELLAN NUCLEARReactor Control System Upgrade for the McClellan Nuclear

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

    SciTech Connect (OSTI)

    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.

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

    SciTech Connect (OSTI)

    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.

  8. Safety Culture in the US Nuclear Regulatory Commission's Reactor Oversight Process

    Broader source: Energy.gov [DOE]

    Presenter: Undine Shoop, Chief, Health Physics and Human Performance Branch, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission

  9. Neutron capture and the antineutrino yield from nuclear reactors

    E-Print Network [OSTI]

    Patrick Huber; Patrick Jaffke

    2015-10-30

    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 3.2MeV, 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 non-equilibrium correction. For naval reactors the nonlinear correction may reach the 10% level.

  10. Passive cooling system for nuclear reactor containment structure

    DOE Patents [OSTI]

    Gou, Perng-Fei (Saratoga, CA); Wade, Gentry E. (Saratoga, CA)

    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.

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

    DOE Patents [OSTI]

    Gou, Perng-Fei (Saratoga, CA); Wade, Gentry E. (Saratoga, CA)

    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.

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

    DOE Patents [OSTI]

    Lloyd, Milton H. (Oak Ridge, TN); Collins, Jack L. (Knoxville, TN); Shell, Sam E. (Oak Ridge, TN)

    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.

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

    DOE Patents [OSTI]

    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.

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

    DOE Patents [OSTI]

    Hutter, Ernest (Wilmette, IL)

    1986-01-01

    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.

  15. Operating strategy generators for nuclear reactors

    SciTech Connect (OSTI)

    Solovyev, D. A., E-mail: and@est.mephi.ru; Semenov, A. A.; Shchukin, N. V. [National Research Nuclear University MEPhI (Russian Federation)

    2011-12-15

    Operating strategy generators, i.e., the software intended for increasing the efficiency of work of nuclear power plant operators, are discussed. The possibilities provided by the domestic and foreign operating-strategy generators are analyzed.

  16. Identification and localization of absorbers of variable strength in nuclear reactors

    E-Print Network [OSTI]

    Demazière, Christophe

    Identification and localization of absorbers of variable strength in nuclear reactors C. Demazie evenly distrib- uted throughout the core of a commercial nuclear reactor. The novelty and ergodic in time, can be used for many diagnostic purposes in nuclear reactors. Many examples can be found

  17. Walking and Climbing Service Robots for Safety Inspection of Nuclear Reactor Pressure Vessels

    E-Print Network [OSTI]

    Chen, Sheng

    Walking and Climbing Service Robots for Safety Inspection of Nuclear Reactor Pressure Vessels B of Electronics and Computer Science, University of Southampton, Southampton, UK Abstract: Nuclear reactor and the usefulness of these robots for improving safety inspection of nuclear reactors in general are discussed

  18. FRP Retrofit of the Ring-Beam of a Nuclear Reactor Containment Structure

    E-Print Network [OSTI]

    SP·215-18 FRP Retrofit of the Ring-Beam of a Nuclear Reactor Containment Structure by M. Demers. A for the storage of the moderately contaminated nuclear reactor. The enforcement of more rigorous environmental. 1. HISTORY 1.1 Decommissioning of the Reactor The Gentilly-I nuclear power plant, located

  19. Analysis of granular flow in a pebble-bed nuclear reactor Chris H. Rycroft,1

    E-Print Network [OSTI]

    Bazant, Martin Z.

    Analysis of granular flow in a pebble-bed nuclear reactor Chris H. Rycroft,1 Gary S. Grest,2 James February 2006; published 24 August 2006 Pebble-bed nuclear reactor technology, which is currently being States, the Modular Pebble Bed Reactor MPBR 4,8 is a candidate for the next generation nuclear plant

  20. Japanese set to direct `sun-power' nuclear reactor in France September 16, 2005

    E-Print Network [OSTI]

    Japanese set to direct `sun-power' nuclear reactor in France September 16, 2005 Japan has been develop three generations of nuclear reactors and includes six low-capacity experimental reactors and a 17 asked to nominate the chief of an international project to build a multi- billion-dollar nuclear fusion

  1. Nuclear Reactor Technology Subcommittee of NEAC

    Energy Savers [EERE]

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on DeliciousMathematicsEnergyInterested Parties -DepartmentAvailable forSite |n t e OfficeResearch andFacts:Reactor

  2. Neutron capture and the antineutrino yield from nuclear reactors

    E-Print Network [OSTI]

    Huber, Patrick

    2015-01-01

    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 3.2MeV, 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 non-equilibrium correction...

  3. Nuclear plant-aging research on reactor protection systems

    SciTech Connect (OSTI)

    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.

  4. Heat barrier for use in a nuclear reactor facility

    DOE Patents [OSTI]

    Keegan, Charles P. (South Huntingdon Twp., Westmoreland County, PA)

    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.

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

    SciTech Connect (OSTI)

    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.

  6. SNIF: A Futuristic Neutrino Probe for Undeclared Nuclear Fission Reactors

    E-Print Network [OSTI]

    Thierry Lasserre; Maximilien Fechner; Guillaume Mention; Romain Reboulleau; Michel Cribier; Alain Letourneau; David Lhuillier

    2010-11-16

    Today reactor neutrino experiments are at the cutting edge of fundamental research in particle physics. Understanding the neutrino is far from complete, but thanks to the impressive progress in this field over the last 15 years, a few research groups are seriously considering that neutrinos could be useful for society. The International Atomic Energy Agency (IAEA) works with its Member States to promote safe, secure and peaceful nuclear technologies. In a context of international tension and nuclear renaissance, neutrino detectors could help IAEA to enforce the Treaty on the Non-Proliferation of Nuclear Weapons (NPT). In this article we discuss a futuristic neutrino application to detect and localize an undeclared nuclear reactor from across borders. The SNIF (Secret Neutrino Interactions Finder) concept proposes to use a few hundred thousand tons neutrino detectors to unveil clandestine fission reactors. Beyond previous studies we provide estimates of all known background sources as a function of the detector's longitude, latitude and depth, and we discuss how they impact the detectability.

  7. Nuclear reactor spacer grid and ductless core component

    DOE Patents [OSTI]

    Christiansen, David W. (Kennewick, WA); Karnesky, Richard A. (Richland, WA)

    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.

  8. Removable check valve for use in a nuclear reactor

    DOE Patents [OSTI]

    Dunn, Charlton (Calabasas, CA); Gutzmann, Edward A. (Simi Valley, CA)

    1988-01-01

    A removable check valve for interconnecting the discharge duct of a pump and an inlet coolant duct of a reactor core in a pool-type nuclear reactor. A manifold assembly is provided having an outer periphery affixed to and in fluid communication with the discharge duct of the pump and has an inner periphery having at least one opening therethrough. A housing containing a check valve is located within the inner periphery of the manifold. The upper end of the housing has an opening in alignment with the opening in the manifold assembly, and seals are provided above and below the openings. The lower end of the housing is adapted for fluid communication with the inlet duct of the reactor core.

  9. Collecting and recirculating condensate in a nuclear reactor containment

    DOE Patents [OSTI]

    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.

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

    DOE Patents [OSTI]

    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.

  11. Support arrangement for core modules of nuclear reactors

    DOE Patents [OSTI]

    Bollinger, Lawrence R. (Schenectady, NY)

    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.

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

    DOE Patents [OSTI]

    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.

  13. Collecting and recirculating condensate in a nuclear reactor containment

    DOE Patents [OSTI]

    Schultz, Terry L. (Murrysville Boro, PA)

    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.

  14. Nuclear reactor characteristics and operational history

    Gasoline and Diesel Fuel Update (EIA)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of Natural GasAdjustments (Billion Cubic Feet)DecadeYear Jan Feb MarthroughFeet)Feet) YearThousand81Nuclear > U.S.

  15. Multiphysics Simulation of Nuclear Reactors F

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity ofkandz-cm11 Outreach Home Room NewsInformationJessework usesof Energy Moving Forward to Address Nuclear3-000MSGPSolarNo.

  16. Temperature measuring analysis of the nuclear reactor fuel assembly

    SciTech Connect (OSTI)

    Urban, F. E-mail: zdenko.zavodny@stuba.sk; Ku?ák, 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.

  17. Systems and methods for dismantling a nuclear reactor

    SciTech Connect (OSTI)

    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.

  18. Detachable connection for a nuclear reactor fuel assembly

    DOE Patents [OSTI]

    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.

  19. Search for Neutrino Oscillations at the Palo Verde Nuclear Reactors

    E-Print Network [OSTI]

    F. Boehm; J. Busenitz; B. Cook; G. Gratta; H. Henrikson; J. Kornis; D. Lawrence; K. B. Lee; K. McKinny; L. Miller; V. Novikov; A. Piepke; B. Ritchie; D. Tracy; P. Vogel; Y-F. Wang; J. Wolf

    1999-12-22

    We report on the initial results from a measurement of the anti-neutrino flux and spectrum at a distance of about 800 m from the three reactors of the Palo Verde Nuclear Generating Station using a segmented gadolinium-loaded scintillation detector. We find that the anti-neutrino flux agrees with that predicted in the absence of oscillations to better than 5%, excluding at 90% CL $\\rm\\bar\

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

    DOE Patents [OSTI]

    Gross, Kenny C. (Bolingbrook, IL); Singer, Ralph M. (Naperville, IL); Humenik, Keith E. (Columbia, MD)

    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.

  1. INVESTIGATIONS ON NUCLEAR SPECTROSCOPY AT THE REACTOR AND THEIR APPLICATIONS1

    E-Print Network [OSTI]

    Titov, Anatoly

    1 INVESTIGATIONS ON NUCLEAR SPECTROSCOPY AT THE REACTOR AND THEIR APPLICATIONS1 I.A. Kondurov , E. However the first work on nuclear spectroscopy was carried out before the reactor was launched; namely.M. Korotkikh, Yu.E. Loginov, V.V. Martynov Introduction Physical launch of the WWR-M reactor in the branch

  2. Method for passive cooling liquid metal cooled nuclear reactors, and system thereof

    DOE Patents [OSTI]

    Hunsbedt, Anstein (Los Gatos, CA); Busboom, Herbert J. (San Jose, CA)

    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.

  3. Woo, H.H.; Chou, C.K. 21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED...

    Office of Scientific and Technical Information (OSTI)

    Piping-reliability analysis for pressurized-water-reactor feedwater lines Woo, H.H.; Chou, C.K. 21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; PIPES; CRACKS; RELIABILITY; PWR...

  4. Reactor Vessel and Reactor Vessel Internals Segmentation at Zion Nuclear Power Station - 13230

    SciTech Connect (OSTI)

    Cooke, Conrad; Spann, Holger

    2013-07-01

    Zion Nuclear Power Station (ZNPS) is a dual-unit Pressurized Water Reactor (PWR) nuclear power plant located on the Lake Michigan shoreline, in the city of Zion, Illinois approximately 64 km (40 miles) north of Chicago, Illinois and 67 km (42 miles) south of Milwaukee, Wisconsin. Each PWR is of the Westinghouse design and had a generation capacity of 1040 MW. Exelon Corporation operated both reactors with the first unit starting production of power in 1973 and the second unit coming on line in 1974. The operation of both reactors ceased in 1996/1997. In 2010 the Nuclear Regulatory Commission approved the transfer of Exelon Corporation's license to ZionSolutions, the Long Term Stewardship subsidiary of EnergySolutions responsible for the decommissioning of ZNPS. In October 2010, ZionSolutions awarded Siempelkamp Nuclear Services, Inc. (SNS) the contract to plan, segment, remove, and package both reactor vessels and their respective internals. This presentation discusses the tools employed by SNS to remove and segment the Reactor Vessel Internals (RVI) and Reactor Vessels (RV) and conveys the recent progress. SNS's mechanical segmentation tooling includes the C-HORCE (Circumferential Hydraulically Operated Cutting Equipment), BMT (Bolt Milling Tool), FaST (Former Attachment Severing Tool) and the VRS (Volume Reduction Station). Thermal segmentation of the reactor vessels will be accomplished using an Oxygen- Propane cutting system. The tools for internals segmentation were designed by SNS using their experience from other successful reactor and large component decommissioning and demolition (D and D) projects in the US. All of the designs allow for the mechanical segmentation of the internals remotely in the water-filled reactor cavities. The C-HORCE is designed to saw seven circumferential cuts through the Core Barrel and Thermal Shield walls with individual thicknesses up to 100 mm (4 inches). The BMT is designed to remove the bolts that fasten the Baffle Plates to the Baffle Former Plates. The FaST is designed to remove the Baffle Former Plates from the Core Barrel. The VRS further volume reduces segmented components using multiple configurations of the 38i and horizontal reciprocating saws. After the successful removal and volume reduction of the Internals, the RV will be segmented using a 'First in the US' thermal cutting process through a co-operative effort with Siempelkamp NIS Ingenieurgesellschaft mbH using their experience at the Stade NPP and Karlsruhe in Germany. SNS mobilized in the fall of 2011 to commence execution of the project in order to complete the RVI segmentation, removal and packaging activities for the first unit (Unit 2) by end of the 2012/beginning 2013 and then mobilize to the second unit, Unit 1. Parallel to the completion of the segmentation of the reactor vessel internals at Unit 1, SNS will segment the Unit 2 pressure vessel and at completion move to Unit 1. (authors)

  5. Advanced nuclear reactor public opinion project. Interim report

    SciTech Connect (OSTI)

    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.

  6. Designing a Component-Based Architecture for the Modeling and Simulation of Nuclear Fuels and Reactors

    E-Print Network [OSTI]

    Pennycook, Steve

    interest in nuclear energy in the U. S. Applications for 26 new reactors have been sub- mitted to the U. S. The NEAMS program is organized around four technical areas of the nuclear fuel cycle: fuels, reactorsDesigning a Component-Based Architecture for the Modeling and Simulation of Nuclear Fuels

  7. Print this article Close This Window EU OKs India joining ITER nuclear reactor project

    E-Print Network [OSTI]

    Print this article Close This Window EU OKs India joining ITER nuclear reactor project Fri Dec 2-billion-euro project to build an experimental nuclear fusion reactor that in the long-run could provide virtually unlimited, cheap and clean energy. The EU's willingness to work with India on a civil nuclear

  8. Nuclear reactor power for an electrically powered orbital transfer vehicle

    SciTech Connect (OSTI)

    Jaffe, L.; Beatty, R.; Bhandari, P.; Chow, E.; Deininger, W.; Ewell, R.; Fujita, T.; Grossman, M.; Kia, T.; Nesmith, B.

    1987-01-01

    To help determine the systems requirements for a 300-kWe space nuclear reactor power system, a mission and spacecraft have been examined which utilize electric propulsion and this nuclear reactor power for multiple transfers of cargo between low Earth orbit (LEO) and geosynchronous Earth orbit (GEO). A propulsion system employing ion thrusters and xenon propellant was selected. Propellant and thrusters are replaced after each sortie to GEO. The mass of the Orbital Transfer Vehicle (OTV), empty and dry, is 11,000 kg; nominal propellant load is 5000 kg. The OTV operates between a circular orbit at 925 km altitude, 28.5 deg inclination, and GEO. Cargo is brought to the OTV by Shuttle and an Orbital Maneuvering Vehicle (OMV); the OTV then takes it to GEO. The OTV can also bring cargo back from GEO, for transfer by OMV to the Shuttle. OTV propellant is resupplied and the ion thrusters are replaced by the OMV before each trip to GEO. At the end of mission life, the OTV's electric propulsion is used to place it in a heliocentric orbit so that the reactor will not return to Earth. The nominal cargo capability to GEO is 6000 kg with a transit time of 120 days; 1350 kg can be transferred in 90 days, and 14,300 kg in 240 days. These capabilities can be considerably increased by using separate Shuttle launches to bring up propellant and cargo, or by changing to mercury propellant.

  9. Electric Power Produced from Nuclear Reactor | National Nuclear Security

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantityBonneville Power Administration would like submitKansas Nuclear ProfileMultiferroic Electric FieldAdministration

  10. Monitoring system for a liquid-cooled nuclear fission reactor

    DOE Patents [OSTI]

    DeVolpi, Alexander (Bolingbrook, IL)

    1987-01-01

    A monitoring system for detecting changes in the liquid levels in various regions of a water-cooled nuclear power reactor, viz., in the downcomer, in the core, in the inlet and outlet plenums, at the head, and elsewhere; and also for detecting changes in the density of the liquid in these regions. A plurality of gamma radiation detectors are used, arranged vertically along the outside of the reactor vessel, and collimator means for each detector limits the gamma-radiation it receives as emitting from only isolated regions of the vessel. Excess neutrons produced by the fission reaction will be captured by the water coolant, by the steel reactor walls, or by the fuel or control structures in the vessel. Neutron capture by steel generates gamma radiation having an energy level of the order of 5-12 MeV, whereas neutron capture by water provides an energy level of approximately 2.2 MeV, and neutron capture by the fission fuel or its cladding provides an energy level of 1 MeV or less. The intensity of neutron capture thus changes significantly at any water-metal interface. Comparative analysis of adjacent gamma detectors senses changes from the normal condition with liquid coolant present to advise of changes in the presence and/or density of the coolant at these specific regions. The gamma detectors can also sense fission-product gas accumulation at the reactor head to advise of a failure of fuel-pin cladding.

  11. Thermal barrier and support for nuclear reactor fuel core

    DOE Patents [OSTI]

    Betts, Jr., William S. (Del Mar, CA); Pickering, J. Larry (Del Mar, CA); Black, William E. (San Diego, CA)

    1987-01-01

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

  12. Pressurized water nuclear reactor system with hot leg vortex mitigator

    DOE Patents [OSTI]

    Lau, Louis K. S. (Monroeville, PA)

    1990-01-01

    A pressurized water nuclear reactor system includes a vortex mitigator in the form of a cylindrical conduit between the hot leg conduit and a first section of residual heat removal conduit, which conduit leads to a pump and a second section of residual heat removal conduit leading back to the reactor pressure vessel. The cylindrical conduit is of such a size that where the hot leg has an inner diameter D.sub.1, the first section has an inner diameter D.sub.2, and the cylindrical conduit or step nozzle has a length L and an inner diameter of D.sub.3 ; D.sub.3 /D.sub.1 is at least 0.55, D.sub.2 is at least 1.9, and L/D.sub.3 is at least 1.44, whereby cavitation of the pump by a vortex formed in the hot leg is prevented.

  13. Retrievable fuel pin end member for a nuclear reactor

    DOE Patents [OSTI]

    Rosa, Jerry M. (Los Gatos, CA)

    1982-01-01

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

  14. Summary of space nuclear reactor power systems, 1983--1992

    SciTech Connect (OSTI)

    Buden, D.

    1993-08-11

    This report summarizes major developments in the last ten years which have greatly expanded the space nuclear reactor power systems technology base. In the SP-100 program, after a competition between liquid-metal, gas-cooled, thermionic, and heat pipe reactors integrated with various combinations of thermoelectric thermionic, Brayton, Rankine, and Stirling energy conversion systems, three concepts:were selected for further evaluation. In 1985, the high-temperature (1,350 K), lithium-cooled reactor with thermoelectric conversion was selected for full scale development. Since then, significant progress has been achieved including the demonstration of a 7-y-life uranium nitride fuel pin. Progress on the lithium-cooled reactor with thermoelectrics has progressed from a concept, through a generic flight system design, to the design, development, and testing of specific components. Meanwhile, the USSR in 1987--88 orbited a new generation of nuclear power systems beyond the, thermoelectric plants on the RORSAT satellites. The US has continued to advance its own thermionic fuel element development, concentrating on a multicell fuel element configuration. Experimental work has demonstrated a single cell operating time of about 1 1/2-y. Technology advances have also been made in the Stirling engine; an advanced engine that operates at 1,050 K is ready for testing. Additional concepts have been studied and experiments have been performed on a variety of systems to meet changing needs; such as powers of tens-to-hundreds of megawatts and highly survivable systems of tens-of-kilowatts power.

  15. N.R. 20 FOSSIL-FUELED POWER PLANTS; 21 SPECIFIC NUCLEAR REACTORS...

    Office of Scientific and Technical Information (OSTI)

    20 FOSSIL-FUELED POWER PLANTS; 21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; 14 SOLAR ENERGY; 15 GEOTHERMAL ENERGY; GEOTHERMAL POWER PLANTS; COMPUTERIZED SIMULATION; HEAT...

  16. A New Nuclear Reactor Neutrino Experiment to Measure theta 13

    E-Print Network [OSTI]

    K. Anderson

    2004-02-26

    An International Working Group has been meeting to discuss ideas for a new Nuclear Reactor Neutrino Experiment at meetings in May 2003 (Alabama), October 2003 (Munich) and plans for March 2004 (Niigata). This White Paper Report on the Motivation and Feasibility of such an experiment is the result of these meetings. After a discussion of the context and opportunity for such an experiment, there are sections on detector design, calibration, overburden and backgrounds, systematic errors, other physics, tunneling issues, safety and outreach. There are 7 appendices describing specific site opportunities.

  17. Reactor Subsystem Simulation for Nuclear Hybrid Energy Systems

    SciTech Connect (OSTI)

    Shannon Bragg-Sitton; J. Michael Doster; Alan Rominger

    2012-09-01

    Preliminary system models have been developed by Idaho National Laboratory researchers and are currently being enhanced to assess integrated system performance given multiple sources (e.g., nuclear + wind) and multiple applications (i.e., electricity + process heat). Initial efforts to integrate a Fortran-based simulation of a small modular reactor (SMR) with the balance of plant model have been completed in FY12. This initial effort takes advantage of an existing SMR model developed at North Carolina State University to provide initial integrated system simulation for a relatively low cost. The SMR subsystem simulation details are discussed in this report.

  18. A REVIEW OF LIGHT-WATER REACTOR SAFETY STUDIES. VOLUME 3 OF THE FINAL REPORT ON HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL ELECTRIC GENERATION IN CALIFORNIA

    E-Print Network [OSTI]

    Nero, A.V.

    2010-01-01

    Charges Relating to Nuclear Reactor Safety," 1976, availablestudies of light-water nuclear reactor safety, emphasizingstudies of overall nuclear reactor safety have been

  19. Incorporation of Hydride Nuclear Fuels in Commercial Light Water Reactors

    E-Print Network [OSTI]

    Terrani, Kurt Amir

    2010-01-01

    re- actor (PWR) and boiling-water reactor (BWR) designsin integral boiling water super heat reactors. Technical

  20. 309NUCLEAR ENGINEERING AND TECHNOLOGY, VOL.37 NO.4, AUGUST 2005 A NEW BOOK: "LIGHT-WATER REACTOR MATERIALS"

    E-Print Network [OSTI]

    Motta, Arthur T.

    309NUCLEAR ENGINEERING AND TECHNOLOGY, VOL.37 NO.4, AUGUST 2005 A NEW BOOK: "LIGHT-WATER REACTOR review; it is a book preview. Thirty years ago, "Fundamental Aspects of Nuclear Reactor Fuel Elements of nuclear fuels among other topics pertinent to the materials in the ensemble of the nuclear reactor

  1. W.B.; Allison, G.S. 21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED...

    Office of Scientific and Technical Information (OSTI)

    nuclear fuel bundle data for use in fuel bundle handling Weihermiller, W.B.; Allison, G.S. 21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; FUEL ELEMENT CLUSTERS; REMOTE...

  2. Hadlock, R.K.; Abbey, O.B. 22 GENERAL STUDIES OF NUCLEAR REACTORS...

    Office of Scientific and Technical Information (OSTI)

    on ultimate heat sinks--cooling ponds Hadlock, R.K.; Abbey, O.B. 22 GENERAL STUDIES OF NUCLEAR REACTORS; 20 FOSSIL-FUELED POWER PLANTS; COOLING PONDS; PERFORMANCE TESTING; NUCLEAR...

  3. Gas tagging and cover gas combination for nuclear reactor

    DOE Patents [OSTI]

    Gross, Kenny C. (Lemont, IL); Laug, Matthew T. (Idaho Falls, ID)

    1985-01-01

    The invention discloses the use of stable isotopes of neon and argon, that are grouped in preselected different ratios one to the other and are then sealed as tags in different cladded nuclear fuel elements to be used in a liquid metal fast breeder reactor. Failure of the cladding of any fuel element allows fission gases generated in the reaction and these tag isotopes to escape and to combine with the cover gas held in the reactor over the fuel elements. The isotopes specifically are Ne.sup.20, Ne.sup.21 and Ne.sup.22 of neon and Ar.sup.36, Ar.sup.38 and Ar.sup.40 of argon, and the cover gas is helium. Serially connected cryogenically operated charcoal beds are used to clean the cover gas and to separate out the tags. The first or cover gas cleanup bed is held between approximately 0.degree. and -25.degree. C. operable to remove the fission gases from the cover gas and tags and the second or tag recovery system bed is held between approximately -170.degree. and -185.degree. C. operable to isolate the tags from the cover gas. Spectrometric analysis further is used to identify the specific tags that are recovered, and thus the specific leaking fuel element. By cataloging the fuel element tags to the location of the fuel elements in the reactor, the location of the leaking fuel element can then be specifically determined.

  4. Monitoring nuclear reactor systems using neural networks and fuzzy logic

    SciTech Connect (OSTI)

    Ikonomopoulos, A.; Tsoukalas, L.H.; Uhrig, R.E. [Tennessee Univ., Knoxville, TN (United States); Mullens, J.A. [Tennessee Univ., Knoxville, TN (United States)]|[Oak Ridge National Lab., TN (United States)

    1991-12-01

    A new approach is presented that demonstrates the potential of trained artificial neural networks (ANNs) as generators of membership functions for the purpose of monitoring nuclear reactor systems. ANN`s provide a complex-to-simple mapping of reactor parameters in a process analogous to that of measurement. Through such ``virtual measurements`` the value of parameters with operational significance, e.g., control-valve-disk-position, valve-line-up or performance can be determined. In the methodology presented the output of a virtual measuring device is a set of membership functions which independently represent different states of the system. Utilizing a fuzzy logic representation offers the advantage of describing the state of the system in a condensed form, developed through linguistic descriptions and convenient for application in monitoring, diagnostics and generally control algorithms. The developed methodology is applied to the problem of measuring the disk position of the secondary flow control valve of an experimental reactor using data obtained during a start-up. The enhanced noise tolerance of the methodology is clearly demonstrated as well as a method for selecting the actual output. The results suggest that it is possible to construct virtual measuring devices through artificial neural networks mapping dynamic time series to a set of membership functions and thus enhance the capability of monitoring systems. 8 refs., 11 figs., 1 tab.

  5. Closed Brayton cycle power conversion systems for nuclear reactors :

    SciTech Connect (OSTI)

    Wright, Steven A.; Lipinski, Ronald J.; Vernon, Milton E.; Sanchez, Travis

    2006-04-01

    This report describes the results of a Sandia National Laboratories internally funded research program to study the coupling of nuclear reactors to gas dynamic Brayton power conversion systems. The research focused on developing integrated dynamic system models, fabricating a 10-30 kWe closed loop Brayton cycle, and validating these models by operating the Brayton test-loop. The work tasks were performed in three major areas. First, the system equations and dynamic models for reactors and Closed Brayton Cycle (CBC) systems were developed and implemented in SIMULINKTM. Within this effort, both steady state and dynamic system models for all the components (turbines, compressors, reactors, ducting, alternators, heat exchangers, and space based radiators) were developed and assembled into complete systems for gas cooled reactors, liquid metal reactors, and electrically heated simulators. Various control modules that use proportional-integral-differential (PID) feedback loops for the reactor and the power-conversion shaft speed were also developed and implemented. The simulation code is called RPCSIM (Reactor Power and Control Simulator). In the second task an open cycle commercially available Capstone C30 micro-turbine power generator was modified to provide a small inexpensive closed Brayton cycle test loop called the Sandia Brayton test-Loop (SBL-30). The Capstone gas-turbine unit housing was modified to permit the attachment of an electrical heater and a water cooled chiller to form a closed loop. The Capstone turbine, compressor, and alternator were used without modification. The Capstone systems nominal operating point is 1150 K turbine inlet temperature at 96,000 rpm. The annular recuperator and portions of the Capstone control system (inverter) and starter system also were reused. The rotational speed of the turbo-machinery is controlled by adjusting the alternator load by using the electrical grid as the load bank. The SBL-30 test loop was operated at the manufacturers site (Barber-Nichols Inc.) and installed and operated at Sandia. A sufficiently detailed description of the loop is provided in this report along with the design characteristics of the turbo-alternator-compressor set to allow other researchers to compare their results with those measured in the Sandia test-loop. The third task consisted of a validation effort. In this task the test loop was operated and compared with the modeled results to develop a more complete understanding of this electrically heated closed power generation system and to validate the model. The measured and predicted system temperatures and pressures are in good agreement, indicating that the model is a reasonable representation of the test loop. Typical deviations between the model and the hardware results are less than 10%. Additional tests were performed to assess the capability of the Brayton engine to continue to remove decay heat after the reactor/heater is shutdown, to develop safe and effective control strategies, and to access the effectiveness of gas inventory control as an alternative means to provide load following. In one test the heater power was turned off to simulate a rapid reactor shutdown, and the turbomachinery was driven solely by the sensible heat stored in the heater for over 71 minutes without external power input. This is an important safety feature for CBC systems as it means that the closed Brayton loop will keep cooling the reactor without the need for auxiliary power (other than that needed to circulate the waste heat rejection coolant) provided the heat sink is available.

  6. Selecting a radiation tolerant piezoelectric material for nuclear reactor applications

    SciTech Connect (OSTI)

    Parks, D. A.; Reinhardt, B. T.; Tittmann, B. R. [Department of Engineering Science and Mechanics, Penn State, University Park, PA 16803 (United States)

    2013-01-25

    Bringing systems for online monitoring of nuclear reactors to fruition has been delayed by the lack of suitable ultrasonic sensors. Recent work has demonstrated the capability of an AlN sensor to perform ultrasonic evaluation in an actual nuclear reactor. Although the AlN demonstrated sustainability, no loss in signal amplitude and d{sub 33} up to a fast and thermal neutron fluence of 1.85 Multiplication-Sign 1018 n/cm{sup 2} and 5.8 Multiplication-Sign 1018 n/cm{sup 2} respectively, no formal process to selecting a suitable sensor material was made. It would be ideal to use first principles approaches to somehow reduce each candidate piezoelectric material to a simple ranking showing directly which materials one should expect to be most radiation tolerant. However, the complexity of the problem makes such a ranking impractical and one must appeal to experimental observations. This should not be of any surprise to one whom is familiar with material science as most material properties are obtained in this manner. Therefore, this work adopts a similar approach, the mechanisms affecting radiation tolerance are discussed and a good engineering sense is used for material qualification of the candidate piezoelectric materials.

  7. Critical technologies for reactors used in nuclear electric propulsion

    SciTech Connect (OSTI)

    Bhattacharyya, S.K. (Argonne National Lab., IL (United States))

    1993-01-01

    Nuclear electric Propulsion (NEP) systems are expected to play a significant role in the exploration and exploitation of space. Unlike nuclear thermal propulsion (NTP) systems in which the hot reactor coolant is directly discharged from nozzles to provide the required thrust, NEP systems include electric power generation and conditioning units that in turn are used to drive electric thrusters. These thrusters accelerate sub atomic particles to produce thrust. The major advantage of NEP systems is the ability to provide very high specific impulses ([approximately]5000 s) that minimize the requirement for propellants. In addition, the power systems used in NEP could pro vide the dual purpose of also providing power for the missions at the destination. This synergism can be exploited in shared development costs. The NEP systems produce significantly lower thrust that NTP systems and are generally more massive. Both systems have their appropriate roles in a balanced space program. The technology development needs of NEP systems differ in many important ways from the development needs for NTP systems because of the significant differences in the operating conditions of the systems. The NEP systems require long-life reactor power systems operating at power levels that are considerably lower than those for NTP systems. In contrast, the operational lifetime of an NEP system (years) is orders of magnitude longer than the operational lifetime of NTP systems (thousands of second). Thus, the critical issue of NEP is survivability and reliable operability for very long times at temperatures that are considerably more modest than the temperatures required for effective NTP operations but generally much higher than those experienced in terrestrial reactors.

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

    SciTech Connect (OSTI)

    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)

  9. Liquid level, void fraction, and superheated steam sensor for nuclear-reactor cores. [PWR; BWR

    DOE Patents [OSTI]

    Tokarz, R.D.

    1981-10-27

    This disclosure relates to an apparatus for monitoring the presence of coolant in liquid or mixed liquid and vapor, and superheated gaseous phases at one or more locations within an operating nuclear reactor core, such as pressurized water reactor or a boiling water reactor.

  10. Physics of Nuclear Reactors, March,21 2011 What do we know ?

    E-Print Network [OSTI]

    Danon, Yaron

    Dr. Danon Physics of Nuclear Reactors, March,21 2011 #12;What do we know ? All the information we have is from the media. More reliable; nuclear related information: www.nei.org www.iaea.org THE REST IS INTERPRETATION OF THIS DATA #12;BWR Reactor (Mark I containment) #12;BWR containment in more

  11. Vital area identification for U.S. Nuclear Regulatory Commission nuclear power reactor licensees and new reactor applicants.

    SciTech Connect (OSTI)

    Whitehead, Donnie Wayne; Varnado, G. Bruce

    2008-09-01

    U.S. Nuclear Regulatory Commission nuclear power plant licensees and new reactor applicants are required to provide protection of their plants against radiological sabotage, including the placement of vital equipment in vital areas. This document describes a systematic process for the identification of the minimum set of areas that must be designated as vital areas in order to ensure that all radiological sabotage scenarios are prevented. Vital area identification involves the use of logic models to systematically identify all of the malicious acts or combinations of malicious acts that could lead to radiological sabotage. The models available in the plant probabilistic risk assessment and other safety analyses provide a great deal of the information and basic model structure needed for the sabotage logic model. Once the sabotage logic model is developed, the events (or malicious acts) in the model are replaced with the areas in which the events can be accomplished. This sabotage area logic model is then analyzed to identify the target sets (combinations of areas the adversary must visit to cause radiological sabotage) and the candidate vital area sets (combinations of areas that must be protected against adversary access to prevent radiological sabotage). Any one of the candidate vital area sets can be selected for protection. Appropriate selection criteria will allow the licensee or new reactor applicant to minimize the impacts of vital area protection measures on plant safety, cost, operations, or other factors of concern.

  12. Measuring of fissile isotopes partial antineutrino spectra in direct experiment at nuclear reactor

    E-Print Network [OSTI]

    V. V. Sinev

    2009-02-22

    The direct measuring method is considered to get nuclear reactor antineutrino spectrum. We suppose to isolate partial spectra of the fissile isotopes by using the method of antineutrino spectrum extraction from the inverse beta decay positron spectrum applied at Rovno experiment. This admits to increase the accuracy of partial antineutrino spectra forming the total nuclear reactor spectrum. It is important for the analysis of the reactor core fuel composition and could be applied for non-proliferation purposes.

  13. Porous nuclear fuel element with internal skeleton for high-temperature gas-cooled nuclear reactors

    DOE Patents [OSTI]

    Youchison, Dennis L.; Williams, Brian E.; Benander, Robert E.

    2013-09-03

    Porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's), and to processes for fabricating them. Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, the nuclear fuel may be deposited inside of a highly porous skeletal structure made of, for example, reticulated vitreous carbon foam.

  14. Porous nuclear fuel element for high-temperature gas-cooled nuclear reactors

    DOE Patents [OSTI]

    Youchison, Dennis L. (Albuquerque, NM); Williams, Brian E. (Pacoima, CA); Benander, Robert E. (Pacoima, CA)

    2011-03-01

    Porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's), and to processes for fabricating them. Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, the nuclear fuel may be deposited inside of a highly porous skeletal structure made of, for example, reticulated vitreous carbon foam.

  15. Methods for manufacturing porous nuclear fuel elements for high-temperature gas-cooled nuclear reactors

    DOE Patents [OSTI]

    Youchison, Dennis L. (Albuquerque, NM); Williams, Brian E. (Pocoima, CA); Benander, Robert E. (Pacoima, CA)

    2010-02-23

    Methods for manufacturing porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's). Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, a thin coating of nuclear fuel may be deposited inside of a highly porous skeletal structure made, for example, of reticulated vitreous carbon foam.

  16. Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR): Programmer's guide

    SciTech Connect (OSTI)

    Call, O. J.; Jacobson, J. A.

    1988-09-01

    The Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR) is an automated data base management system for processing and storing human error probability and hardware component failure data. The NUCLARR system software resides on an IBM (or compatible) personal micro-computer and can be used to furnish data inputs for both human and hardware reliability analysis in support of a variety of risk assessment activities. The NUCLARR system is documented in a five-volume series of reports. Volume 2 of this series is the Programmer's Guide for maintaining the NUCLARR system software. This Programmer's Guide provides, for the software engineer, an orientation to the software elements involved, discusses maintenance methods, and presents useful aids and examples. 4 refs., 75 figs., 1 tab.

  17. Reactor and shielding design implications of clustering nuclear thermal rockets

    SciTech Connect (OSTI)

    Buksa, J.J.; Houts, M.G. (Los Alamos National Laboratory, NM (United States))

    1992-07-01

    This paper examines design considerations in the context of engine-out accidents in clustered nuclear-thermal rocket stages, and an accident-management protocol is devised. Safety and performance issues are considered in the light of designs for the reactor and shielding elements of ROVER/NERVA-type engines. The engine-out management process involves: phase one, in which sufficient propulsive power is guaranteed for mission completion; and phase two, in which engine failure is isolated and not allowed to propagate to other engines or to the spacecraft. Phase-one designs can employ spare engines, throttled engines, and/or long-burning engines. Phase-two safety concepts can include techniques for cooling or jettisoning the failed engines. Engine-out management philosophies are shown to be shaped by a combination of safety and mission-trajectory requirements. 6 refs.

  18. Thermal insulating barrier and neutron shield providing integrated protection for a nuclear reactor vessel

    DOE Patents [OSTI]

    Schreiber, Roger B. (Penn Twp., PA); Fero, Arnold H. (New Kensington, PA); Sejvar, James (Murrysville, PA)

    1997-01-01

    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 to form 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 valving also includes bistable vents at the upper end of the thermal insulating barrier for releasing steam. A removable, modular neutron shield extending around the upper end of the reactor cavity below the nozzles forms with the upwardly and outwardly tapered transition on the outer surface of the reactor vessel, a labyrinthine channel which reduces neutron streaming while providing a passage for the escape of steam during a severe accident, and for the cooling air which is circulated along the reactor cavity walls outside the thermal insulating barrier during normal operation of the reactor.

  19. Thermal insulating barrier and neutron shield providing integrated protection for a nuclear reactor vessel

    DOE Patents [OSTI]

    Schreiber, R.B.; Fero, A.H.; Sejvar, J.

    1997-12-16

    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 to form 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 valving also includes bistable vents at the upper end of the thermal insulating barrier for releasing steam. A removable, modular neutron shield extending around the upper end of the reactor cavity below the nozzles forms with the upwardly and outwardly tapered transition on the outer surface of the reactor vessel, a labyrinthine channel which reduces neutron streaming while providing a passage for the escape of steam during a severe accident, and for the cooling air which is circulated along the reactor cavity walls outside the thermal insulating barrier during normal operation of the reactor. 8 figs.

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

    DOE Patents [OSTI]

    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.

  1. Above-ground Antineutrino Detection for Nuclear Reactor Monitoring

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Sweany, Melinda; Brennan, James S.; Cabrera-Palmer, Belkis; Kiff, Scott D.; Reyna, David; Throckmorton, Daniel J.

    2014-08-01

    Antineutrino monitoring of nuclear reactors has been demonstrated many times, however the technique has not as of yet been developed into a useful capability for treaty verification purposes. The most notable drawback is the current requirement that detectors be deployed underground, with at least several meters-water-equivalent of shielding from cosmic radiation. In addition, the deployment of liquid-based detector media presents a challenge in reactor facilities. We are currently developing a detector system that has the potential to operate above ground and circumvent deployment problems associated with a liquid detection media: the system is composed of segments of plastic scintillator surroundedmore »by 6LiF/ZnS:Ag. ZnS:Ag is a radio-luminescent phosphor used to detect the neutron capture products of lithium-6. Because of its long decay time compared to standard plastic scintillators, pulse-shape discrimination can be used to distinguish positron and neutron interactions resulting from the inverse beta decay (IBD) of antineutrinos within the detector volume, reducing both accidental and correlated backgrounds. Segmentation further reduces backgrounds by identifying the positron’s annihilation gammas, which are absent for most correlated and uncorrelated backgrounds. This work explores different configurations in order to maximize the size of the detector segments without reducing the intrinsic neutron detection efficiency. We believe this technology will ultimately be applicable to potential safeguards scenarios such as those recently described.« less

  2. Self-actuating and locking control for nuclear reactor

    DOE Patents [OSTI]

    Chung, Dong K. (Chatsworth, CA)

    1982-01-01

    A self-actuating, self-locking flow cutoff valve particularly suited for use in a nuclear reactor of the type which utilizes a plurality of fluid support neutron absorber elements to provide for the safe shutdown of the reactor. The valve comprises a substantially vertical elongated housing and an aperture plate located in the housing for the flow of fluid therethrough, a substantially vertical elongated nozzle member located in the housing and affixed to the housing with an opening in the bottom for receiving fluid and apertures adjacent a top end for discharging fluid. The nozzle further includes two sealing means, one located above and the other below the apertures. Also located in the housing and having walls surrounding the nozzle is a flow cutoff sleeve having a fluid opening adjacent an upper end of the sleeve, the sleeve being moveable between an upper open position wherein the nozzle apertures are substantially unobstructed and a closed position wherein the sleeve and nozzle sealing surfaces are mated such that the flow of fluid through the apertures is obstructed. It is a particular feature of the present invention that the valve further includes a means for utilizing any increase in fluid pressure to maintain the cutoff sleeve in a closed position. It is another feature of the invention that there is provided a means for automatically closing the valve whenever the flow of fluid drops below a predetermined level.

  3. Improved gas tagging and cover gas combination for nuclear reactor

    DOE Patents [OSTI]

    Gross, K.C.; Laug, M.T.

    1983-09-26

    The invention discloses the use of stable isotopes of neon and argon, sealed as tags in different cladding nuclear fuel elements to be used in a liquid metal fast breeder reactor. Cladding failure allows fission gases and these tag isotopes to escape and to combine with the cover gas. The isotopes are Ne/sup 20/, Ne/sup 21/ and Ne/sup 22/ and Ar/sup 36/, Ar/sup 38/ and Ar/sup 40/, and the cover gas is He. Serially connected cryogenically operated charcoal beds are used to clean the cover gas and to separate out the tags. The first or cover gas cleanup bed is held between 0 and -25/sup 0/C to remove the fission gases from the cover gas and tags, and the second or tag recovery system bed between -170 and -185/sup 0/C to isolate the tags from the cover gas. Spectrometric analysis is used to identify the specific tags that are recovered, and thus the specific leaking fuel element. By cataloging the fuel element tags to the location of the fuel elements in the reactor, the location of the leaking fuel element can then be determined.

  4. Record of Cycling Operation of the Natural Nuclear Reactor in the Oklo/Okelobondo Area in Gabon

    E-Print Network [OSTI]

    Record of Cycling Operation of the Natural Nuclear Reactor in the Oklo/Okelobondo Area in Gabon A billion yr old Oklo natural nuclear reactor. In addition to elevated abundances of fission-produced Zr, Ce nuclear chain reaction was predicted by Kuroda [1] 20 years before the remnants of the natural reactor

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

    SciTech Connect (OSTI)

    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.

  6. System aspects of a Space Nuclear Reactor Power System

    SciTech Connect (OSTI)

    Jaffe, L.; Fujita, T.; Beatty, R.; Bhandari, P.; Chow, E.; Deininger, W.; Ewell, R.; Grossman, M.; Kia, T.; Nesmith, B.

    1988-01-01

    Selected systems aspects of a 300 kW nuclear reactor power system for spacecraft have been studied. The approach included examination of two candidate missions and their associated spacecraft, and a number of special topics dealing with the power system design and operation. The missions considered were a reusable orbital transfer vehicle and a space-based radar. The special topics included: power system configuration and scaling, launch vehicle integration, operating altitude, orbital storage, start-up, thawing, control, load following, procedures in case of malfunction, restart, thermal and nuclear radiation to other portions of the spacecraft, thermal stresses between subsystems, boom and cable designs, vibration modes, altitude control, reliability, and survivability. Among the findings are 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. The power system design evolved during the study and has continued to evolve; the current design differs somewhat from that examined in this paper.

  7. Underground collocation of nuclear power plant reactors and repository to facilitate the post-renaissance expansion of nuclear power

    SciTech Connect (OSTI)

    Myers, Carl W [Los Alamos National Laboratory; Elkins, Ned Z [Los Alamos National Laboratory

    2008-01-01

    Underground collocation of nuclear power reactors and the nuclear waste management facilities supporting those reactors, termed an underground nuclear park (UNP), appears to have several advantages compared to the conventional approach to siting reactors and waste management facilities. These advantages include the potential to lower reactor capital and operating cost, lower nuclear waste management cost, and increase margins of physical security and safety. Envirorunental impacts related to worker health, facility accidents, waste transportation, and sabotage and terrorism appear to be lower for UNPs compared to the current approach. In-place decommissioning ofUNP reactors appears to have cost, safety, envirorunental and waste disposal advantages. The UNP approach has the potential to lead to greater public acceptance for the deployment of new power reactors. Use of the UNP during the post-nuclear renaissance time frame has the potential to enable a greater expansion of U.S. nuclear power generation than might otherwise result. Technical and economic aspects of the UNP concept need more study to determine the viability of the concept.

  8. A Preliminary Report on Static Analysis of C Code for Nuclear Reactor Protection System

    E-Print Network [OSTI]

    A Preliminary Report on Static Analysis of C Code for Nuclear Reactor Protection System Jong: Cybersecurity regulations require new I&C (Instrumentation & Control) systems in nuclear power plants to develop if the C code is produced mechanically. Keywords: Nuclear Plant Protection System , I&C , PLC software

  9. Underground nuclear power station using self-regulating heat-pipe controlled reactors

    DOE Patents [OSTI]

    Hampel, Viktor E. (Pleasanton, CA)

    1989-01-01

    A nuclear reactor for generating electricity is disposed underground at the bottom of a vertical hole that can be drilled using conventional drilling technology. The primary coolant of the reactor core is the working fluid in a plurality of thermodynamically coupled heat pipes emplaced in the hole between the heat source at the bottom of the hole and heat exchange means near the surface of the earth. Additionally, the primary coolant (consisting of the working flud in the heat pipes in the reactor core) moderates neutrons and regulates their reactivity, thus keeping the power of the reactor substantially constant. At the end of its useful life, the reactor core may be abandoned in place. Isolation from the atmosphere in case of accident or for abandonment is provided by the operation of explosive closures and mechanical valves emplaced along the hole. This invention combines technology developed and tested for small, highly efficient, space-based nuclear electric power plants with the technology of fast-acting closure mechanisms developed and used for underground testing of nuclear weapons. This invention provides a nuclear power installation which is safe from the worst conceivable reactor accident, namely, the explosion of a nuclear weapon near the ground surface of a nuclear power reactor.

  10. An underground nuclear power station using self-regulating heat-pipe controlled reactors

    DOE Patents [OSTI]

    Hampel, V.E.

    1988-05-17

    A nuclear reactor for generating electricity is disposed underground at the bottom of a vertical hole that can be drilled using conventional drilling technology. The primary coolant of the reactor core is the working fluid in a plurality of thermodynamically coupled heat pipes emplaced in the hole between the heat source at the bottom of the hole and heat exchange means near the surface of the earth. Additionally, the primary coolant (consisting of the working fluid in the heat pipes in the reactor core) moderates neutrons and regulates their reactivity, thus keeping the power of the reactor substantially constant. At the end of its useful life, the reactor core may be abandoned in place. Isolation from the atmosphere in case of accident or for abandonment is provided by the operation of explosive closures and mechanical valves emplaced along the hole. This invention combines technology developed and tested for small, highly efficient, space-based nuclear electric power plants with the technology of fast- acting closure mechanisms developed and used for underground testing of nuclear weapons. This invention provides a nuclear power installation which is safe from the worst conceivable reactor accident, namely, the explosion of a nuclear weapon near the ground surface of a nuclear power reactor. 5 figs.

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

    SciTech Connect (OSTI)

    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.

  12. Passive cooling system for liquid metal cooled nuclear reactors with backup coolant flow path

    DOE Patents [OSTI]

    Hunsbedt, Anstein (Los Gatos, CA); Boardman, Charles E. (Saratoga, CA)

    1993-01-01

    A liquid metal cooled nuclear fission reactor plant having a passive auxiliary safety cooling system for removing residual heat resulting from fuel decay during reactor shutdown, or heat produced during a mishap. This reactor plant is enhanced by a backup or secondary passive safety cooling system which augments the primary passive auxiliary cooling system when in operation, and replaces the primary system when rendered inoperable.

  13. Development of Nuclear Reactor remote Monitoring software (NRM) for the Star project 

    E-Print Network [OSTI]

    Gautier, Vincent Charles

    2002-01-01

    As a response to the needs of developing countries to meet their rapidly growing energy requirements, the Safe, Transportable, Autonomous Reactor (STAR) program originated. This concept relies on small, passively safe, and highly autonomous nuclear...

  14. The development of a remote monitoring system for the Nuclear Science Center reactor 

    E-Print Network [OSTI]

    Jiltchenkov, Dmitri Victorovich

    2002-01-01

    With funding provided by Nuclear Energy Research Initiative (NERI), design of Secure, Transportable, Autonomous Reactors (STAR) to aid countries with insufficient energy supplies is underway. The development of a new monitoring system that allows...

  15. Examination of offsite radiological emergency protective measures for nuclear reactor accidents involving core melt

    E-Print Network [OSTI]

    Aldrich, David C.

    1979-01-01

    Evacuation, sheltering followed by population relocation, and iodine prophylaxis are evaluated as offsite public protective measures in response to nuclear reactor accidents involving core-melt. Evaluations were conducted ...

  16. Space nuclear-power reactor design based on combined neutronic and thermal-fluid analyses

    SciTech Connect (OSTI)

    Koenig, D.R.; Gido, R.G.; Brandon, D.I.

    1985-01-01

    The design and performance analysis of a space nuclear-power system requires sophisticated analytical capabilities such as those developed during the nuclear rocket propulsion (Rover) program. In particular, optimizing the size of a space nuclear reactor for a given power level requires satisfying the conflicting requirements of nuclear criticality and heat removal. The optimization involves the determination of the coolant void (volume) fraction for which the reactor diameter is a minimum and temperature and structural limits are satisfied. A minimum exists because the critical diameter increases with increasing void fraction, whereas the reactor diameter needed to remove a specified power decreases with void fraction. The purpose of this presentation is to describe and demonstrate our analytical capability for the determination of minimum reactor size. The analysis is based on combining neutronic criticality calculations with OPTION-code thermal-fluid calculations.

  17. EU in push for support on nuclear fusion reactor September 26, 2004

    E-Print Network [OSTI]

    to obtain power through nuclear fusion, a clean energy source. But views are split on where the ITER reactor the green light to the project at the end of November, with or without Washington. In a bid to end

  18. Power conversion system design for supercritical carbon dioxide cooled indirect cycle nuclear reactors

    E-Print Network [OSTI]

    Gibbs, Jonathan Paul

    2008-01-01

    The supercritical carbon dioxide (S-CO?) cycle is a promising advanced power conversion cycle which couples nicely to many Generation IV nuclear reactors. This work investigates the power conversion system design and ...

  19. Proceedings of the 1984 DOE nuclear reactor and facility safety conference. Volume II

    SciTech Connect (OSTI)

    Not Available

    1984-01-01

    This report is a collection of papers on reactor safety. The report takes the form of proceedings from the 1984 DOE Nuclear Reactor and Facility Safety Conference, Volume II of two. These proceedings cover Safety, Accidents, Training, Task/Job Analysis, Robotics and the Engineering Aspects of Man/Safety interfaces.

  20. Novel Approach to Plasma Facing Materials in Nuclear Fusion Reactors

    SciTech Connect (OSTI)

    Livramento, V.; Correia, J. B.; Shohoji, N.; Osawa, E.; Nunes, D.

    2008-04-07

    A novel material design in nuclear fusion reactors is proposed based on W-nDiamond nanostructured composites. Generally, a microstructure refined to the nanometer scale improves the mechanical strength due to modification of plasticity mechanisms. Moreover, highly specific grain-boundary area raises the number of sites for annihilation of radiation induced defects. However, the low thermal stability of fine-grained and nanostructured materials demands the presence of particles at the grain boundaries that can delay coarsening by a pinning effect. As a result, the concept of a composite is promising in the field of nanostructured materials. The hardness of diamond renders nanodiamond dispersions excellent reinforcing and stabilization candidates and, in addition, diamond has extremely high thermal conductivity. Consequently, W-nDiamond nanocomposites are promising candidates for thermally stable first-wall materials. The proposed design involves the production of W/W-nDiamond/W-Cu/Cu layered castellations. The W, W-nDiamond and W-Cu layers are produced by mechanical alloying followed by a consolidation route that combines hot rolling with spark plasma sintering (SPS). Layer welding is achieved by spark plasma sintering. The present work describes the mechanical alloying processsing and consolidation route used to produce W-nDiamond composites, as well as microstructural features and mechanical properties of the material produced Long term plasma exposure experiments are planned at ISTTOK and at FTU (Frascati)

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

    DOE Patents [OSTI]

    Miller, John V. (Munhall, PA); Carlson, William R. (Scott Township, Allegheny County, PA); Yarbrough, Michael B. (Hempfield Township, Westmoreland County, PA)

    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.

  2. Eastern Europe Research Reactor Initiative nuclear education and training courses - Current activities and future challenges

    SciTech Connect (OSTI)

    Snoj, L.; Sklenka, L.; Rataj, J.; Boeck, H.

    2012-07-01

    The Eastern Europe Research Reactor Initiative was established in January 2008 to enhance cooperation between the Research Reactors in Eastern Europe. It covers three areas of research reactor utilisation: irradiation of materials and fuel, radioisotope production, neutron beam experiments, education and training. In the field of education and training an EERRI training course was developed. The training programme has been elaborated with the purpose to assist IAEA Member States, which consider building a research reactor (RR) as a first step to develop nuclear competence and infrastructure in the Country. The major strength of the reactor is utilisation of three different research reactors and a lot of practical exercises. Due to high level of adaptability, the course can be tailored to specific needs of institutions with limited or no access to research reactors. (authors)

  3. The role of actinide burning and the Integral Fast Reactor in the future of nuclear power

    SciTech Connect (OSTI)

    Hollaway, W.R.; Lidsky, L.M.; Miller, M.M.

    1990-12-01

    A preliminary assessment is made of the potential role of actinide burning and the Integral Fast Reactor (IFR) in the future of nuclear power. The development of a usable actinide burning strategy could be an important factor in the acceptance and implementation of a next generation of nuclear power. First, the need for nuclear generating capacity is established through the analysis of energy and electricity demand forecasting models which cover the spectrum of bias from anti-nuclear to pro-nuclear. The analyses take into account the issues of global warming and the potential for technological advances in energy efficiency. We conclude, as do many others, that there will almost certainly be a need for substantial nuclear power capacity in the 2000--2030 time frame. We point out also that any reprocessing scheme will open up proliferation-related questions which can only be assessed in very specific contexts. The focus of this report is on the fuel cycle impacts of actinide burning. Scenarios are developed for the deployment of future nuclear generating capacity which exploit the advantages of actinide partitioning and actinide burning. Three alternative reactor designs are utilized in these future scenarios: The Light Water Reactor (LWR); the Modular Gas-Cooled Reactor (MGR); and the Integral Fast Reactor (FR). Each of these alternative reactor designs is described in some detail, with specific emphasis on their spent fuel streams and the back-end of the nuclear fuel cycle. Four separation and partitioning processes are utilized in building the future nuclear power scenarios: Thermal reactor spent fuel preprocessing to reduce the ceramic oxide spent fuel to metallic form, the conventional PUREX process, the TRUEX process, and pyrometallurgical reprocessing.

  4. NRC (Nuclear Regulatory Commission) staff evaluation of the General Electric Company Nuclear Reactor Study (''Reed Report'')

    SciTech Connect (OSTI)

    1987-07-01

    In 1975, the General Electric Company (GE) published a Nuclear Reactor Study, also referred to as ''the Reed Report,'' an internal product-improvement study. GE considered the document ''proprietary'' and thus, under the regulations of the Nuclear Regulatory Commission (NRC), exempt from mandatory public disclosure. Nonetheless, members of the NRC staff reviewed the document in 1976 and determined that it did not raise any significant new safety issues. The staff also reached the same conclusion in subsequent reviews. However, in response to recent inquiries about the report, the staff reevaluated the Reed Report from a 1987 perspective. This re-evaluation, documented in this staff report, concluded that: (1) there are no issues raised in the Reed Report that support a need to curtail the operation of any GE boiling water reactor (BWR); (2) there are no new safety issues raised in the Reed Report of which the staff was unaware; and (3) although certain issues addressed by the Reed Report are still being studied by the NRC and the industry, there is no basis for suspending licensing and operation of GE BWR plants while these issues are being resolved.

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

    SciTech Connect (OSTI)

    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.

  6. A REVIEW OF LIGHT-WATER REACTOR SAFETY STUDIES. VOLUME 3 OF THE FINAL REPORT ON HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL ELECTRIC GENERATION IN CALIFORNIA

    E-Print Network [OSTI]

    Nero, A.V.

    2010-01-01

    discussed, between the Nuclear safety assurance and riskCharges Relating to Nuclear Reactor Safety," 1976, availableof light-water nuclear reactor safety, emphasizing the

  7. CONSTRUCTION OF WEB-ACCESSIBLE MATERIALS HANDBOOK FORGENERATION IV NUCLEAR REACTORS

    SciTech Connect (OSTI)

    Ren, Weiju

    2005-01-01

    The development of a web-accessible materials handbook in support of the materials selection and structural design for the Generation IV nuclear reactors is being planned. Background of the reactor program is briefly introduced. Evolution of materials handbooks for nuclear reactors over years is reviewed in light of the trends brought forth by the rapid advancement in information technologies. The framework, major features, contents, and construction considerations of the web-accessible Gen IV Materials Handbook are discussed. Potential further developments and applications of the handbook are also elucidated.

  8. SL-1 Accident Briefing Report - 1961 Nuclear Reactor Meltdown Educational Documentary

    ScienceCinema (OSTI)

    None

    2014-03-11

    U.S. Atomic Energy Commission (Idaho Operations Office) briefing about the SL-1 Nuclear Reactor Meltdown. The SL-1, or Stationary Low-Power Reactor Number One, was a United States Army experimental nuclear power reactor which underwent a steam explosion and meltdown on January 3, 1961, killing its three operators. The direct cause was the improper withdrawal of the central control rod, responsible for absorbing neutrons in the reactor core. The event is the only known fatal reactor accident in the United States. The accident released about 80 curies (3.0 TBq) of Iodine-131, which was not considered significant due to its location in a remote desert of Idaho. About 1,100 curies (41 TBq) of fission products were released into the atmosphere. The facility, located at the National Reactor Testing Station approximately 40 miles (64 km) west of Idaho Falls, Idaho, was part of the Army Nuclear Power Program and was known as the Argonne Low Power Reactor (ALPR) during its design and build phase. It was intended to provide electrical power and heat for small, remote military facilities, such as radar sites near the Arctic Circle, and those in the DEW Line. The design power was 3 MW (thermal). Operating power was 200 kW electrical and 400 kW thermal for space heating. In the accident, the core power level reached nearly 20 GW in just four milliseconds, precipitating the reactor accident and steam explosion.

  9. SL-1 Accident Briefing Report - 1961 Nuclear Reactor Meltdown Educational Documentary

    SciTech Connect (OSTI)

    2013-09-25

    U.S. Atomic Energy Commission (Idaho Operations Office) briefing about the SL-1 Nuclear Reactor Meltdown. The SL-1, or Stationary Low-Power Reactor Number One, was a United States Army experimental nuclear power reactor which underwent a steam explosion and meltdown on January 3, 1961, killing its three operators. The direct cause was the improper withdrawal of the central control rod, responsible for absorbing neutrons in the reactor core. The event is the only known fatal reactor accident in the United States. The accident released about 80 curies (3.0 TBq) of Iodine-131, which was not considered significant due to its location in a remote desert of Idaho. About 1,100 curies (41 TBq) of fission products were released into the atmosphere. The facility, located at the National Reactor Testing Station approximately 40 miles (64 km) west of Idaho Falls, Idaho, was part of the Army Nuclear Power Program and was known as the Argonne Low Power Reactor (ALPR) during its design and build phase. It was intended to provide electrical power and heat for small, remote military facilities, such as radar sites near the Arctic Circle, and those in the DEW Line. The design power was 3 MW (thermal). Operating power was 200 kW electrical and 400 kW thermal for space heating. In the accident, the core power level reached nearly 20 GW in just four milliseconds, precipitating the reactor accident and steam explosion.

  10. Nuclear reactor having a polyhedral primary shield and removable vessel insulation

    DOE Patents [OSTI]

    Ekeroth, D.E.; Orr, R.

    1993-12-07

    A nuclear reactor is provided having a generally cylindrical reactor vessel disposed within an opening in a primary shield. The opening in the primary shield is defined by a plurality of generally planar side walls forming a generally polyhedral-shaped opening. The reactor vessel is supported within the opening in the primary shield by reactor vessel supports which are in communication and aligned with central portions of some of the side walls. The reactor vessel is connected to the central portions of the reactor vessel supports. A thermal insulation polyhedron formed from a plurality of slidably insertable and removable generally planar insulation panels substantially surrounds at least a portion of the reactor vessel and is disposed between the reactor vessel and the side walls of the primary shield. The shape of the insulation polyhedron generally corresponds to the shape of the opening in the primary shield. Reactor monitoring instrumentation may be mounted in the corners of the opening in the primary shield between the side walls and the reactor vessel such that insulation is not disposed between the instrumentation and the reactor vessel. 5 figures.

  11. Nuclear reactor having a polyhedral primary shield and removable vessel insulation

    DOE Patents [OSTI]

    Ekeroth, Douglas E. (Delmont, PA); Orr, Richard (Pittsburgh, PA)

    1993-01-01

    A nuclear reactor is provided having a generally cylindrical reactor vessel disposed within an opening in a primary shield. The opening in the primary shield is defined by a plurality of generally planar side walls forming a generally polyhedral-shaped opening. The reactor vessel is supported within the opening in the primary shield by reactor vessel supports which are in communication and aligned with central portions of some of the side walls. The reactor vessel is connected to the central portions of the reactor vessel supports. A thermal insulation polyhedron formed from a plurality of slidably insertable and removable generally planar insulation panels substantially surrounds at least a portion of the reactor vessel and is disposed between the reactor vessel and the side walls of the primary shield. The shape of the insulation polyhedron generally corresponds to the shape of the opening in the primary shield. Reactor monitoring instrumentation may be mounted in the corners of the opening in the primary shield between the side walls and the reactor vessel such that insulation is not disposed between the instrumentation and the reactor vessel.

  12. Safety of Department of Energy-Owned Nuclear Reactors

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    1986-09-23

    To establish reactor safety program requirements assure that the safety of each Department of Energy-owned (DOE-owned) reactor is properly analyzed, evaluated, documented, and approved by DOE; and reactors are sited, designed, constructed, modified, operated, maintained, and decommissioned in a manner that gives adequate protection for health and safety and will be in accordance with uniform standards, guides, and codes which are consistent with those applied to comparable licensed reactors. Cancels Chap. 6 of DOE O 5480.1A. Paragraphs 7b(3), 7e(3) & 8c canceled by DOE O 5480.23 & canceled by DOE N 251.4 of 9-29-95.

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

    E-Print Network [OSTI]

    Djurcic, Zelimir

    2009-01-01

    1996. [41] US Nuclear Regulatory Commission, “NRC Sta? Posi-by the US Nuclear Regulatory Commission in a discus- sion on

  14. CRAD, Nuclear Safety- Oak Ridge National Laboratory High Flux Isotope Reactor

    Broader source: Energy.gov [DOE]

    A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a February 2007 assessment of the Nuclear Safety Program in preparation for restart of the Oak Ridge National Laboratory High Flux Isotope Reactor.

  15. Hanging core support system for a nuclear reactor. [LMFBR

    DOE Patents [OSTI]

    Burelbach, J.P.; Kann, W.J.; Pan, Y.C.; Saiveau, J.G.; Seidensticker, R.W.

    1984-04-26

    For holding the reactor core in the confining reactor vessel, a support is disclosed that is structurally independent of the vessel, that is dimensionally accurate and stable, and that comprises tandem tension linkages that act redundantly of one another to maintain stabilized core support even in the unlikely event of the complete failure of one of the linkages. The core support has a mounting platform for the reactor core, and unitary structure including a flange overlying the top edge of the reactor vessels, and a skirt and box beams between the flange and platform for establishing one of the linkages. A plurality of tension rods connect between the deck closing the reactor vessel and the platform for establishing the redundant linkage. Loaded Belleville springs flexibly hold the tension rods at the deck and separable bayonet-type connections hold the tension rods at the platform.

  16. Safeguards Issues at Nuclear Reactors and Enrichment Plants

    SciTech Connect (OSTI)

    Boyer, Brian D

    2012-08-15

    The Agency's safeguards technical objective is the timely detection of diversion of significant quantities of nuclear material from peaceful nuclear activities to the manufacture of nuclear weapons or of other nuclear explosive devices or for purposes unknown, and deterrence of such diversion by the risk of early detection.

  17. Supplying the nuclear arsenal: Production reactor technology, management, and policy, 1942--1992

    SciTech Connect (OSTI)

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

    1994-01-01

    This book focuses on the lineage of America`s production reactors, those three at Hanford and their descendants, the reactors behind America`s nuclear weapons. The work will take only occasional sideways glances at the collateral lines of descent, the reactor cousins designed for experimental purposes, ship propulsion, and electric power generation. Over the decades from 1942 through 1992, fourteen American production reactors made enough plutonium to fuel a formidable arsenal of more than twenty thousand weapons. In the last years of that period, planners, nuclear engineers, and managers struggled over designs for the next generation of production reactors. The story of fourteen individual machines and of the planning effort to replace them might appear relatively narrow. Yet these machines lay at the heart of the nation`s nuclear weapons complex. The story of these machines is the story of arming the winning weapon, supplying the nuclear arms race. This book is intended to capture the history of the first fourteen production reactors, and associated design work, in the face of the end of the Cold War.

  18. Nuclear reactor with low-level core coolant intake

    DOE Patents [OSTI]

    Challberg, Roy C. (Livermore, CA); Townsend, Harold E. (Campbell, CA)

    1993-01-01

    A natural-circulation boiling-water reactor has skirts extending downward from control rod guide tubes to about 10 centimeters from the reactor vessel bottom. The skirts define annular channels about control rod drive housings that extend through the reactor vessel bottom. Recirculating water is forced in through the low-level entrances to these channels, sweeping bottom water into the channels in the process. The sweeping action prevents cooler water from accumulating at the bottom. This in turn minimizes thermal shock to bottom-dwelling components as would occur when accumulated cool water is swept away and suddenly replaced by warmer water.

  19. Neural net controlled tag gas sampling system for nuclear reactors

    DOE Patents [OSTI]

    Gross, K.C.; Laug, M.T.; Lambert, J.B.; Herzog, J.P.

    1997-02-11

    A method and system are disclosed for providing a tag gas identifier to a nuclear fuel rod and analyze escaped tag gas to identify a particular failed nuclear fuel rod. The method and system include disposing a unique tag gas composition into a plenum of a nuclear fuel rod, monitoring gamma ray activity, analyzing gamma ray signals to assess whether a nuclear fuel rod has failed and is emitting tag gas, activating a tag gas sampling and analysis system upon sensing tag gas emission from a failed nuclear rod and evaluating the escaped tag gas to identify the particular failed nuclear fuel rod. 12 figs.

  20. Neural net controlled tag gas sampling system for nuclear reactors

    DOE Patents [OSTI]

    Gross, Kenneth C. (Bolingbrook, IL); Laug, Matthew T. (Idaho Fall, ID); Lambert, John D. B. (Wheaton, IL); Herzog, James P. (Downers Grove, IL)

    1997-01-01

    A method and system for providing a tag gas identifier to a nuclear fuel rod and analyze escaped tag gas to identify a particular failed nuclear fuel rod. The method and system include disposing a unique tag gas composition into a plenum of a nuclear fuel rod, monitoring gamma ray activity, analyzing gamma ray signals to assess whether a nuclear fuel rod has failed and is emitting tag gas, activating a tag gas sampling and analysis system upon sensing tag gas emission from a failed nuclear rod and evaluating the escaped tag gas to identify the particular failed nuclear fuel rod.

  1. The Neutrino Mass Hierarchy from Nuclear Reactor Experiments

    E-Print Network [OSTI]

    Emilio Ciuffoli; Jarah Evslin; Xinmin Zhang

    2013-08-14

    10 years from now reactor neutrino experiments will attempt to determine which neutrino mass eigenstate is the most massive. In this letter we present the results of more than seven million detailed simulations of such experiments, studying the dependence of the probability of successfully determining the mass hierarchy upon the analysis method, the neutrino mass matrix parameters, reactor flux models, geoneutrinos and, in particular, combinations of baselines. We show that a recently reported spurious dependence of the data analysis upon the high energy tail of the reactor spectrum can be removed by using a weighted Fourier transform. We determine the optimal baselines and corresponding detector locations. For most values of the CP-violating, leptonic Dirac phase delta, a degeneracy prevents NOvA and T2K from determining either delta or the hierarchy. We determine the confidence with which a reactor experiment can determine the hierarchy, breaking the degeneracy.

  2. NA 30 - Deputy Administrator for Naval Reactors | National Nuclear...

    National Nuclear Security Administration (NNSA)

    Annual Report FY14 Year End Report Semi Annual Report FY12 Semi Annual Report FY11 Year End Report NX 3 - Naval Reactors Laboratory Field Office FY11 NX3 Year End Report FY10 NX3...

  3. Applicability of reactor code WIMS for nuclear criticality safety studies

    SciTech Connect (OSTI)

    Matausek, M.V.; Marinkovic, N.

    1995-12-31

    The purpose of this paper is to examine applicability of the reactor code WIMS for calculating criticality parameters of nonreactor configurations containing fissile materials. Results are given and discussed for some typical configurations containing {sup 235}U.

  4. Containment building : architecture between the city and advanced nuclear reactors

    E-Print Network [OSTI]

    Pauli, Lisa M

    2011-01-01

    Since the inception of nuclear energy research, the element thorium (Th) has been considered the superior fuel for nuclear reactions because of its potency, safety, abundance and reduced waste. Cold War agendas broke from ...

  5. The Japan Times Printer Friendly Articles France has won the competition to host the International Thermonuclear Experimental Reactor (ITER), the world's first nuclear-

    E-Print Network [OSTI]

    the International Thermonuclear Experimental Reactor (ITER), the world's first nuclear- fusion reactor. Japan fought for nuclear energy in Japan. The controversies surrounding the Monju fast-breeder-reactor project -- nuclear-fusion reactions -- to produce energy. Scientists at the ITER plant will create nuclear

  6. Dual annular rotating [open quotes]windowed[close quotes] nuclear reflector reactor control system

    DOE Patents [OSTI]

    Jacox, M.G.; Drexler, R.L.; Hunt, R.N.M.; Lake, J.A.

    1994-03-29

    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. 4 figures.

  7. Zirconium-based alloys, nuclear fuel rods and nuclear reactors including such alloys, and related methods

    DOE Patents [OSTI]

    Mariani, Robert Dominick

    2014-09-09

    Zirconium-based metal alloy compositions comprise zirconium, a first additive in which the permeability of hydrogen decreases with increasing temperatures at least over a temperature range extending from 350.degree. C. to 750.degree. C., and a second additive having a solubility in zirconium over the temperature range extending from 350.degree. C. to 750.degree. C. At least one of a solubility of the first additive in the second additive over the temperature range extending from 350.degree. C. to 750.degree. C. and a solubility of the second additive in the first additive over the temperature range extending from 350.degree. C. to 750.degree. C. is higher than the solubility of the second additive in zirconium over the temperature range extending from 350.degree. C. to 750.degree. C. Nuclear fuel rods include a cladding material comprising such metal alloy compositions, and nuclear reactors include such fuel rods. Methods are used to fabricate such zirconium-based metal alloy compositions.

  8. Blue Ribbon Commission, Yucca Mountain Closure, Court Actions - Future of Decommissioned Reactors, Operating Reactors and Nuclear Power - 13249

    SciTech Connect (OSTI)

    Devgun, Jas S. [Nuclear Power Technologies, Sargent and Lundy LLC1, Chicago, IL (United States)] [Nuclear Power Technologies, Sargent and Lundy LLC1, Chicago, IL (United States)

    2013-07-01

    Issues related to back-end of the nuclear fuel cycle continue to be difficult for the commercial nuclear power industry and for the decision makers at the national and international level. In the US, the 1982 NWPA required DOE to develop geological repositories for SNF and HLW but in spite of extensive site characterization efforts and over ten billion dollars spent, a repository opening is nowhere in sight. There has been constant litigation against the DOE by the nuclear utilities for breach of the 'standard contract' they signed with the DOE under the NWPA. The SNF inventory continues to rise both in the US and globally and the nuclear industry has turned to dry storage facilities at reactor locations. In US, the Blue Ribbon Commission on America's Nuclear Future issued its report in January 2012 and among other items, it recommends a new, consent-based approach to siting of facilities, prompt efforts to develop one or more geologic disposal facilities, and prompt efforts to develop one or more consolidated storage facilities. In addition, the March 2011 Fukushima Daiichi accident had a severe impact on the future growth of nuclear power. The nuclear industry is focusing on mitigation strategies for beyond design basis events and in the US, the industry is in the process of implementing the recommendations from NRC's Near Term Task Force. (authors)

  9. Passive cooling system for top entry liquid metal cooled nuclear reactors

    DOE Patents [OSTI]

    Boardman, Charles E. (Saratoga, CA); Hunsbedt, Anstein (Los Gatos, CA); Hui, Marvin M. (Cupertino, CA)

    1992-01-01

    A liquid metal cooled nuclear fission reactor plant having a top entry loop joined satellite assembly with a passive auxiliary safety cooling system for removing residual heat resulting from fuel decay during shutdown, or heat produced during a mishap. This satellite type reactor plant is enhanced by a backup or secondary passive safety cooling system which augments the primary passive auxiliary cooling system when in operation, and replaces the primary cooling system when rendered inoperative.

  10. Systems and methods for managing shared-path instrumentation and irradiation targets in a nuclear reactor

    DOE Patents [OSTI]

    Heinold, Mark R.; Berger, John F.; Loper, Milton H.; Runkle, Gary A.

    2015-12-29

    Systems and methods permit discriminate access to nuclear reactors. Systems provide penetration pathways to irradiation target loading and offloading systems, instrumentation systems, and other external systems at desired times, while limiting such access during undesired times. Systems use selection mechanisms that can be strategically positioned for space sharing to connect only desired systems to a reactor. Selection mechanisms include distinct paths, forks, diverters, turntables, and other types of selectors. Management methods with such systems permits use of the nuclear reactor and penetration pathways between different systems and functions, simultaneously and at only distinct desired times. Existing TIP drives and other known instrumentation and plant systems are useable with access management systems and methods, which can be used in any nuclear plant with access restrictions.

  11. Detecting a Nuclear Fission Reactor at the Center of the Earth

    E-Print Network [OSTI]

    R. S. Raghavan

    2002-08-24

    A natural nuclear fission reactor with a power output of 3- 10 terawatt at the center of the earth has been proposed as the energy source of the earth's magnetic field. The proposal can be directly tested by a massive liquid scintillation detector that can detect the signature spectrum of antineutrinos from the geo-reactor as well as the direction of the antineutrino source. Such detectors are now in operation or under construction in Japan/Europe. However, the clarity of both types of measurements may be limited by background from antineutrinos from surface power reactors. Future U. S. detectors, relatively more remote from power reactors, may be more suitable for achieving unambiguous spectral and directional evidence for a 3TW geo-reactor.

  12. Civilian nuclear power on the drawing board: the development of Experimental Breeder Reactor-II.

    SciTech Connect (OSTI)

    Westfall, C.

    2003-02-20

    On September 28, 2001 a symposium was held at Argonne National Laboratory as part of the festivities to mark the 100th birthday of Enrico Fermi. The symposium celebrated Fermi's ''contribution to the development of nuclear power'' and focused on one particular ''line of development'' resulting from Fermi's interest in power reactors: Argonne's fast reactor program. Symposium participants made many references to the ways in which the program was linked to Fermi, who led the team which created the world's first self-sustaining nuclear chain reaction. For example, one presentation featured an April, 1944 memo that described a meeting attended by Fermi and others. The memo came from the time when research on plutonium and the nuclear chain reaction at Chicago's WWII Metallurgical Laboratory was nearing its end. Even as other parts of the Manhattan Engineering Project were building on this effort to create the bombs that would end the war, Fermi and his colleagues were taking the first steps to plan the use of nuclear energy in the postwar era. After noting that Fermi ''viewed the use of [nuclear] power for the heating of cities with sympathy,'' the group outlined several power reactor designs. In the course of discussion, Fermi and his colleagues took the first steps in conjuring the vision that would later be brought to life with Experimental Breeder Reactor I (EBR-I) and Experimental Breeder Reactor II (EBR-II), the celebrated achievements of the Argonne fast reactor program. Group members considered various schemes for a breeder reactor in which the relatively abundant U-238 would be placed near a core of fissionable material. The reactor would be a fast reactor; that is, neutrons would not be moderated, as were most wartime reactors. Thus, the large number of neutrons emitted in fast neutron fission would hit the U-238 and create ''extra'' fissionable material, that is, more than ''invested,'' and at the same time produce power. The group identified the problem of removing heat in such a reactor and presaged the eventual solution by suggesting the use of sodium coolant, which has minimal interaction with neutrons.

  13. Hanging core support system for a nuclear reactor

    DOE Patents [OSTI]

    Burelbach, James P. (Glen Ellyn, IL); Kann, William J. (Park Ridge, IL); Pan, Yen-Cheng (Naperville, IL); Saiveau, James G. (Hickory Hills, IL); Seidensticker, Ralph W. (Wheaton, IL)

    1987-01-01

    For holding the reactor core in the confining reactor vessel, a support is disclosed that is structurally independent of the vessel, that is dimensionally accurate and stable, and that comprises tandem tension linkages that act redundantly of one another to maintain stabilized core support even in the unlikely event of the complete failure of one of the linkages. The core support has a mounting platform for the reactor core, and unitary structure including a flange overlying the top edge of the reactor vessels, and a skirt and box beams between the flange and platform for establishing one of the linkages. A plurality of tension rods connect between the deck closing the reactor vessel and the platform for establishing the redundant linkage. Loaded Belleville springs flexibly hold the tension rods at the deck and separable bayonet-type connections hold the tension rods at the platform. Motion or radiation sensing detectors can be provide at the lower ends of the tension rods for obtaining pertinent readings proximate the core.

  14. France gets nuclear fusion plant France will get to host the project to build a 10bn-euro (6.6bn) nuclear fusion reactor, in

    E-Print Network [OSTI]

    ) nuclear fusion reactor, in the face of strong competition from Japan. The International ThermonuclearFrance gets nuclear fusion plant France will get to host the project to build a 10bn-euro (£6.6bn Experimental Reactor (Iter) will be the most expensive joint scientific project after the International Space

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

    SciTech Connect (OSTI)

    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.

  16. Incorporation of Hydride Nuclear Fuels in Commercial Light Water Reactors

    E-Print Network [OSTI]

    Terrani, Kurt Amir

    2010-01-01

    1.2.1 PWRs . . . . . . . . . . . . . . . . . . . . 1.2.2Actinides Multi-Recycling in PWR Using Hydride Fuels. InRecycling in Hydride Fueled PWR Cores. Nuclear Engineering

  17. Nuclear breeder reactor fuel element with silicon carbide getter

    DOE Patents [OSTI]

    Christiansen, David W. (Kennewick, WA); Karnesky, Richard A. (Richland, WA)

    1987-01-01

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

  18. Advanced Nuclear Technology: Advanced Light Water Reactors Utility Requirements Document Small Modular Reactors Inclusion Summary

    SciTech Connect (OSTI)

    Loflin, Leonard; McRimmon, Beth

    2014-12-18

    This report summarizes a project by EPRI to include requirements for small modular light water reactors (smLWR) into the EPRI Utility Requirements Document (URD) for Advanced Light Water Reactors. The project was jointly funded by EPRI and the U.S. Department of Energy (DOE). The report covers the scope and content of the URD, the process used to revise the URD to include smLWR requirements, a summary of the major changes to the URD to include smLWR, and how to use the URD as revised to achieve value on new plant projects.

  19. Terracentric Nuclear Fission Reactor: Background, Basis, Feasibility, Structure, Evidence, and Geophysical Implications

    E-Print Network [OSTI]

    J. Marvin Herndon

    2013-12-31

    The background, basis, feasibility, structure, evidence, and geophysical implications of a naturally occurring Terracentric nuclear fission georeactor are reviewed. For a nuclear fission reactor to exist at the center of the Earth, all of the following conditions must be met: (1) There must originally have been a substantial quantity of uranium within Earth's core; (2) There must be a natural mechanism for concentrating the uranium; (3) The isotopic composition of the uranium at the onset of fission must be appropriate to sustain a nuclear fission chain reaction; (4) The reactor must be able to breed a sufficient quantity of fissile nuclides to permit operation over the lifetime of Earth to the present; (5) There must be a natural mechanism for the removal of fission products; (6) There must be a natural mechanism for removing heat from the reactor; (7) There must be a natural mechanism to regulate reactor power level, and; (8) The location of the reactor or must be such as to provide containment and prevent meltdown. Herndon's georeactor alone is shown to meet those conditions. Georeactor existence evidence based upon helium measurements and upon antineutrino measurements is described. Geophysical implications discussed include georeactor origin of the geomagnetic field, geomagnetic reversals from intense solar outbursts and severe Earth trauma, as well as georeactor heat contributions to global dynamics.

  20. Nuclear processes in magnetic fusion reactors with polarized fuel

    E-Print Network [OSTI]

    Michail P. Rekalo; Egle Tomasi-Gustafsson

    2000-10-16

    We consider the processes $d +d \\to n +{^3He}$, $d +{^3He} \\to p +{^4He}$, $d +{^3H} \\to n +{^4He}$, ${^3He} +{^3He}\\to p+p +{^4He}$, ${^3H} +{^3He}\\to d +{^4He}$, with particular attention for applications in fusion reactors. After a model independent parametrization of the spin structure of the matrix elements for these processes at thermal colliding energies, in terms of partial amplitudes, we study polarization phenomena in the framework of a formalism of helicity amplitudes. The strong angular dependence of the final nuclei and of the polarization observables on the polarizations of the fuel components can be helpful in the design of the reactor shielding, blanket arrangement etc..We analyze also the angular dependence of the neutron polarization for the processes $\\vec d +\\vec d \\to n +{^3He}$ and $\\vec d +\\vec {^3H} \\to n +{^4He}$.

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

    SciTech Connect (OSTI)

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

    2012-10-30

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

  2. Passive and inherent safety technologies for light-water nuclear reactors

    SciTech Connect (OSTI)

    Forsberg, C.W.

    1990-07-01

    Passive/inherent safety implies a technical revolution in our approach to nuclear power safety. This direction is discussed herein for light-water reactors (LWRs) -- the predominant type of power reactor used in the world today. At Oak Ridge National Laboratory (ORNL) the approach to the development of passive/inherent safety for LWRs consists of four steps: identify and quantify safety requirements and goals; identify and quantify the technical functional requirements needed for safety; identify, invent, develop, and quantify technical options that meet both of the above requirements; and integrate safety systems into designs of economic and reliable nuclear power plants. Significant progress has been achieved in the first three steps of this program. The last step involves primarily the reactor vendors. These activities, as well as related activities worldwide, are described here. 27 refs., 7 tabs.

  3. Fuel-Cycle and Nuclear Material Disposition Issues Associated with High-Temperature Gas Reactors

    SciTech Connect (OSTI)

    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 storage, packaging, transportation, waste forms, waste treatment, decontamination and decommissioning issues; and low-level waste (LLW) and high-level waste (HLW) disposal.

  4. Engineering analysis of a power upgrade for the Texas A&M Nuclear Science Center Reactor 

    E-Print Network [OSTI]

    Rearden, Bradley Thomas

    1995-01-01

    Heat transfer, neutronics, and accident analyses are presented to support an increase of the power level of the Texas A&M University Nuclear Science Center Reactor. The upgraded steady state power level is to be 1.49 MW, from the current level of 1...

  5. Investigation of Neutrino Properties in Experiments at Nuclear Reactors: Present Status and Prospects

    E-Print Network [OSTI]

    L. A. Mikaelyan

    2002-10-07

    This paper was submitted in Russian edition of Journal Physics of Atomic Nuclei in 2001. The present status of experiments that are being performed at nuclear reactors in order to seek the neutrino masses, mixing, and magnetic moments, whose discovery would be a signal of the existence of physics beyond the Standard Model, is considered, along with their future prospects.

  6. Spring design for use in the core of a nuclear reactor

    DOE Patents [OSTI]

    Willard, Jr., H. James (Bethel Park, PA)

    1993-01-01

    A spring design particularly suitable for use in the core of a nuclear reactor includes one surface having a first material oriented in a longitudinal direction, and another surface having a second material oriented in a transverse direction. The respective surfaces exhibit different amounts of irraditation induced strain.

  7. Assessement of Codes and Standards Applicable to a Hydrogen Production Plant Coupled to a Nuclear Reactor

    SciTech Connect (OSTI)

    M. J. Russell

    2006-06-01

    This is an assessment of codes and standards applicable to a hydrogen production plant to be coupled to a nuclear reactor. The result of the assessment is a list of codes and standards that are expected to be applicable to the plant during its design and construction.

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

    DOE Patents [OSTI]

    Forsberg, Charles W. (Oak Ridge, TN)

    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.

  9. Fast neutron fluence of yonggwang nuclear unit 1 reactor pressure vessel

    SciTech Connect (OSTI)

    Yoo, C.; Km, B.; Chang, K.; Leeand, S.; Park, J.

    2006-07-01

    The Code of Federal Regulations, Title 10, Part 50, Appendix H, requires that the neutron dosimetry be present to monitor the reactor vessel throughout plant life. The Ex-Vessel Neutron Dosimetry System has been installed for Yonggwang Nuclear Unit 1 after complete withdrawal of all six in-vessel surveillance capsules. This system has been installed in the reactor cavity annulus in order to measure the fast neutron spectrum coming out through the reactor pressure vessel. Cycle specific neutron transport calculations were performed to obtain the energy dependent neutron flux throughout the reactor geometry including dosimetry positions. Comparisons between calculations and measurements were performed for the reaction rates of each dosimetry sensors and results show good agreements. (authors)

  10. Japan hopes for rapid accord with EU on nuclear reactor Japanese Prime Minister Junichiro Koizumi said Monday that he hopes to resolve a dispute with the

    E-Print Network [OSTI]

    Japan hopes for rapid accord with EU on nuclear reactor 02/05/2005 Japanese Prime Minister nuclear reactor as soon as possible. Koizumi discussed the row over plans for the International Terms and Conditions 5/2/05 8:59 AMEUbusiness - Japan hopes for rapid accord with EU on nuclear reactor

  11. For more information, contact Michele Boyd at mboyd@psr.org. Updated July 13, 2009. Existing Subsidies and Incentives for New Nuclear Reactors

    E-Print Network [OSTI]

    Laughlin, Robert B.

    Subsidies and Incentives for New Nuclear Reactors Research and Development · Generation IV program in construction and operation licensing for 6 new reactors, including delays due to the Nuclear Regulatory-hour paid by ratepayers receiving electricity from nuclear reactors to pay for a geologic repository

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

    SciTech Connect (OSTI)

    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.

  13. Reactivity worth measurements at the IPEN/MB-01 nuclear reactor

    SciTech Connect (OSTI)

    Pinto, Leticia Negrao; Santos, Adimir dos [Nuclear Engineering Center, Nuclear and Energy Research Institute- IPEN/CNEN-SP (Brazil)

    2013-05-06

    Researches that aim to improve the performance of neutron transport codes and quality of nuclear cross section databases are very important to increase the accuracy of simulations and the quality of the analysis and prediction of phenomena in the nuclear field. In this context, relevant experimental data such as reactivity worth measurements are needed. The objective of this work was to perform a series of experiments of reactivity worth measurements, using a digital reactivity meter developed at IPEN. The experiments employed small metallic and ceramic samples inserted in the central region of the core of the experimental IPEN/MB-01 reactor. The theoretical analysis was performed by the MCNP-5 reactor physics code, developed and maintained by Los Alamos National Laboratory, and the ENDF/B-VII.0 nuclear data library.

  14. Method for removing cesium from a nuclear reactor coolant

    DOE Patents [OSTI]

    Colburn, Richard P. (Pasco, WA)

    1986-01-01

    A method of and system for removing cesium from a liquid metal reactor coolant including a carbon packing trap in the primary coolant system for absorbing a major portion of the radioactive cesium from the coolant flowing therethrough at a reduced temperature. A regeneration subloop system having a secondary carbon packing trap is selectively connected to the primary system for isolating the main trap therefrom and connecting it to the regeneration system. Increasing the temperature of the sodium flowing through the primary trap diffuses a portion of the cesium

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

    DOE Patents [OSTI]

    Gabor, John D. (Western Springs, IL); Cassulo, John C. (Stickney, IL); Pedersen, Dean R. (Naperville, IL); Baker, Jr., Louis (Downers Grove, IL)

    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.

  16. Standard Practice for Design of Surveillance Programs for Light-Water Moderated Nuclear Power Reactor Vessels

    E-Print Network [OSTI]

    American Society for Testing and Materials. Philadelphia

    2010-01-01

    1.1 This practice covers procedures for designing a surveillance program for monitoring the radiation-induced changes in the mechanical properties of ferritic materials in light-water moderated nuclear power reactor vessels. This practice includes the minimum requirements for the design of a surveillance program, selection of vessel material to be included, and the initial schedule for evaluation of materials. 1.2 This practice was developed for all light-water moderated nuclear power reactor vessels for which the predicted maximum fast neutron fluence (E > 1 MeV) at the end of license (EOL) exceeds 1 × 1021 neutrons/m2 (1 × 1017 n/cm2) at the inside surface of the reactor vessel. 1.3 This practice applies only to the planning and design of surveillance programs for reactor vessels designed and built after the effective date of this practice. Previous versions of Practice E185 apply to earlier reactor vessels. 1.4 This practice does not provide specific procedures for monitoring the radiation induced cha...

  17. Multi-unit Operations in Non-Nuclear Systems: Lessons Learned for Small Modular Reactors

    SciTech Connect (OSTI)

    OHara J. M.; Higgins, J.; DAgostino, A.

    2012-01-17

    The nuclear-power community has reached the stage of proposing advanced reactor designs to support power generation for decades to come. Small modular reactors (SMRs) are one approach to meet these energy needs. While the power output of individual reactor modules is relatively small, they can be grouped to produce reactor sites with different outputs. Also, they can be designed to generate hydrogen, or to process heat. Many characteristics of SMRs are quite different from those of current plants and may be operated quite differently. One difference is that multiple units may be operated by a single crew (or a single operator) from one control room. The U.S. Nuclear Regulatory Commission (NRC) is examining the human factors engineering (HFE) aspects of SMRs to support licensing reviews. While we reviewed information on SMR designs to obtain information, the designs are not completed and all of the design and operational information is not yet available. Nor is there information on multi-unit operations as envisioned for SMRs available in operating experience. Thus, to gain a better understanding of multi-unit operations we sought the lesson learned from non-nuclear systems that have experience in multi-unit operations, specifically refineries, unmanned aerial vehicles and tele-intensive care units. In this paper we report the lessons learned from these systems and the implications for SMRs.

  18. Environmental Assessment of Urgent-Relief Acceptance of Foreign Research Reactor Spent Nuclear Fuel

    SciTech Connect (OSTI)

    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.

  19. Evaluation of a Business Case for Safeguards by Design in Nuclear Power Reactors

    SciTech Connect (OSTI)

    Wood, Thomas W.; Seward, Amy M.; Lewis, Valerie A.; Gitau, Ernest TN; Zentner, Michael D.

    2012-12-01

    Safeguards by Design (SbD) is a well-known paradigm for consideration and incorporation of safeguards approaches and associated design features early in the nuclear facility development process. This paradigm has been developed as part of the Next Generation Safeguards Initiative (NGSI), and has been accepted as beneficial in many discussions and papers on NGSI or specific technologies under development within NGSI. The Office of Nuclear Safeguards and Security funded the Pacific Northwest National Laboratory to examine the business case justification of SbD for nuclear power reactors. Ultimately, the implementation of SbD will rely on the designers of nuclear facilities. Therefore, it is important to assess the incentives which will lead designers to adopt SbD as a standard practice for nuclear facility design. This report details the extent to which designers will have compelling economic incentives to adopt SbD.

  20. Monitoring system for a liquid-cooled nuclear fission reactor. [PWR

    DOE Patents [OSTI]

    DeVolpi, A.

    1984-07-20

    The invention provides improved means for detecting the water levels in various regions of a water-cooled nuclear power reactor, viz., in the downcomer, in the core, in the inlet and outlet plenums, at the head, and elsewhere; and also for detecting the density of the water in these regions. The invention utilizes a plurality of exterior gamma radiation detectors and a collimator technique operable to sense separate regions of the reactor vessel to give respectively, unique signals for these regions, whereby comparative analysis of these signals can be used to advise of the presence and density of cooling water in the vessel.

  1. The Next Generation Nuclear Plant Graphite Creep Experiment Irradiation in the Advanced Test Reactor

    SciTech Connect (OSTI)

    Blaine Grover

    2010-10-01

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Program will be irradiating six gas reactor graphite creep experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the United States Department of Energy’s lead laboratory for nuclear energy development. The ATR is one of the world’s premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These graphite irradiations are being accomplished to support development of the next generation reactors in the United States. The graphite experiments will be irradiated over the next six to eight years to support development of a graphite irradiation performance data base on the new nuclear grade graphites now available for use in high temperature gas reactors. The goals of the irradiation experiments are to obtain irradiation performance data, including irradiation creep, at different temperatures and loading conditions to support design of the Next Generation Nuclear Plant (NGNP) Very High Temperature Gas Reactor, as well as other future gas reactors. The experiments will each consist of a single capsule that will contain six stacks of graphite specimens, with half of the graphite specimens in each stack under a compressive load, while the other half of the specimens will not be subjected to a compressive load during irradiation. The six stacks will have differing compressive loads applied to the top half of each pair of specimen stacks, while a seventh stack will not have a compressive load. The specimens will be irradiated in an inert sweep gas atmosphere with on-line temperature and compressive load monitoring and control. There will also be the capability of sampling the sweep gas effluent to determine if any oxidation or off-gassing of the specimens occurs during initial start-up of the experiment. The first experiment was inserted in the ATR in August 2009 and started its irradiation in September 2009. It is anticipated to complete its irradiation in early calendar 2011. This paper will discuss the design of the experiment including the test train and the temperature and compressive load monitoring, control, and the irradiation experience to date.

  2. Acoustic emission monitoring of hot functional testing: Watts Bar Unit 1 Nuclear Reactor

    SciTech Connect (OSTI)

    Hutton, P.H.; Dawson, J.F.; Friesel, M.A.; Harris, J.C.; Pappas, R.A.

    1984-06-01

    Acoustic emission (AE) monitoring of selected pressure boundary areas at TVA's Watts Bar, Unit 1 Nuclear Power Plant during hot functional preservice testing is described in this report. The report deals with background, methodology, and results. The work discussed here is a major milestone in a program supported by NRC to develop and demonstrate application of AE monitoring for continuous surveillance of reactor pressure boundaries to detect and evaluate growing flaws. The subject work demonstrated that anticipated problem areas can be overcome. Work is continuing toward AE monitoring during reactor operation.

  3. TRIGA Mark II nuclear reactor facility. Final report, 1 July 1980--30 June 1995

    SciTech Connect (OSTI)

    Ryan, B.C.

    1997-05-01

    This report is a final culmination of activities funded through the Department of Energy`s (DOE) University Reactor Sharing Program, Grant DE-FG02-80ER10273, during the period 1 July 1980 through 30 June 1995. Progress reports have been periodically issued to the DOE, namely the Reactor Facility Annual Reports C00-2082/2219-7 through C00-2082/10723-21, which are contained as an appendix to this report. Due to the extent of time covered by this grant, summary tables are presented. Table 1 lists the fiscal year financial obligations of the grant. As listed in the original grant proposals, the DOE grant financed 70% of project costs, namely the total amount spent of these projects minus materials costs and technical support. Thus the bulk of funds was spent directly on reactor operations. With the exception of a few years, spending was in excess of the grant amount. As shown in Tables 2 and 3, the Reactor Sharing grant funded a immense number of research projects in nuclear engineering, geology, animal science, chemistry, anthropology, veterinary medicine, and many other fields. A list of these users is provided. Out of the average 3000 visitors per year, some groups participated in classes involving the reactor such as Boy Scout Merit Badge classes, teacher`s workshops, and summer internships. A large number of these projects met the requirements for the Reactor Sharing grant, but were funded by the University instead.

  4. Considerations Associated with Reactor Technology Selection for the Next Generation Nuclear Plant Project

    SciTech Connect (OSTI)

    L.E. Demick

    2010-09-01

    At the inception of the Next Generation Nuclear Plant Project and during predecessor activities, alternative reactor technologies have been evaluated to determine the technology that best fulfills the functional and performance requirements of the targeted energy applications and market. Unlike the case of electric power generation where the reactor performance is primarily expressed in terms of economics, the targeted energy applications involve industrial applications that have specific needs in terms of acceptable heat transport fluids and the associated thermodynamic conditions. Hence, to be of interest to these industrial energy applications, the alternative reactor technologies are weighed in terms of the reactor coolant/heat transport fluid, achievable reactor outlet temperature, and practicality of operations to achieve the very high reliability demands associated with the petrochemical, petroleum, metals and related industries. These evaluations have concluded that the high temperature gas-cooled reactor (HTGR) can uniquely provide the required ranges of energy needs for these target applications, do so with promising economics, and can be commercialized with reasonable development risk in the time frames of current industry interest – i.e., within the next 10-15 years.

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

    DOE Patents [OSTI]

    Tutu, Narinder K. (Manorville, NY); Ginsberg, Theodore (East Setauket, NY); Klages, John R. (Mattituck, NY)

    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.

  6. Method for removing cesium from a nuclear reactor coolant

    DOE Patents [OSTI]

    Colburn, R.P.

    1983-08-10

    A method of and system for removing cesium from a liquid metal reactor coolant including a carbon packing trap in the primary coolant system for absorbing a major portion of the radioactive cesium from the coolant flowing therethrough at a reduced temperature. A regeneration subloop system having a secondary carbon packing trap is selectively connected to the primary system for isolating the main trap therefrom and connecting it to the regeneration system. Increasing the temperature of the sodium flowing through the primary trap diffuses a portion of the cesium inventory thereof further into the carbon matrix while simultaneously redispersing a portion into the regeneration system for absorption at a reduced temperature by the secondary trap.

  7. Nuclear reactor heat transport system component low friction support system

    DOE Patents [OSTI]

    Wade, Elman E. (Ruffs Dale, PA)

    1980-01-01

    A support column for a heavy component of a liquid metal fast breeder reactor heat transport system which will deflect when the pipes leading coolant to and from the heavy component expand or contract due to temperature changes includes a vertically disposed pipe, the pipe being connected to the heavy component by two longitudinally spaced cycloidal dovetail joints wherein the distal end of each of the dovetails constitutes a part of the surface of a large diameter cylinder and the centerlines of these large diameter cylinders intersect at right angles and the pipe being supported through two longitudinally spaced cycloidal dovetail joints wherein the distal end of each of the dovetails constitutes a part of the surface of a large diameter cylinder and the centerlines of these large diameter cylinders intersect at right angles, each of the cylindrical surfaces bearing on a flat and horizontal surface.

  8. Can Next-Generation Reactors Power a Safe Nuclear Futur By Clay Dillow Posted 03.17.2011 at 12:18 pm

    E-Print Network [OSTI]

    Danon, Yaron

    Can Next-Generation Reactors Power a Safe Nuclear Futur By Clay Dillow Posted 03.17.2011 at 12 of nuclear reactors are designed to prevent exactly what we old Fukushima Daiichi plant. Which is good the world rush to reconsider their nuclear plans, nuclear experts look toward a future of smaller, safer

  9. Universal Ownership: Why Environmental Externalities Matter to...

    Open Energy Info (EERE)

    Universal Ownership: Why Environmental Externalities Matter to Institutional Investors Jump to: navigation, search Tool Summary LAUNCH TOOL Name: Universal Ownership: Why...

  10. Multi-reactor power system configurations for multimegawatt nuclear electric propulsion

    SciTech Connect (OSTI)

    George, J.A.

    1991-09-01

    A modular, multi-reactor power system and vehicle configuration for piloted nuclear electric propulsion (NEP) missions to Mars is presented. Such a design could provide enhanced system and mission reliability, allowing a comfortable safety margin for early manned flights, and would allow a range of piloted and cargo missions to be performed with a single power system design. Early use of common power modules for cargo missions would also provide progressive flight experience and validation of standardized systems for use in later piloted applications. System and mission analysis are presented to compare single and multi-reactor configurations for piloted Mars missions. A conceptual design for the Hydra modular multi-reactor NEP vehicle is presented.

  11. Standard Guide for Conducting Supplemental Surveillance Tests for Nuclear Power Reactor Vessels, E 706 (IH)

    E-Print Network [OSTI]

    American Society for Testing and Materials. Philadelphia

    2010-01-01

    1.1 This guide discusses test procedures that can be used in conjunction with, but not as alternatives to, those required by Practices E185 and E2215 for the surveillance of nuclear reactor vessels. The supplemental mechanical property tests outlined permit the acquisition of additional information on radiation-induced changes in fracture toughness, notch ductility, and yield strength properties of the reactor vessel steels. 1.2 This guide provides recommendations for the preparation of test specimens for irradiation, and identifies special precautions and requirements for reactor surveillance operations and postirradiation test planning. Guidance on data reduction and computational procedures is also given. Reference is made to other ASTM test methods for the physical conduct of specimen tests and for raw data acquisition.

  12. Standard Guide for In-Service Annealing of Light-Water Moderated Nuclear Reactor Vessels

    E-Print Network [OSTI]

    American Society for Testing and Materials. Philadelphia

    2003-01-01

    1.1 This guide covers the general procedures to be considered for conducting an in-service thermal anneal of a light-water moderated nuclear reactor vessel and demonstrating the effectiveness of the procedure. The purpose of this in-service annealing (heat treatment) is to improve the mechanical properties, especially fracture toughness, of the reactor vessel materials previously degraded by neutron embrittlement. The improvement in mechanical properties generally is assessed using Charpy V-notch impact test results, or alternatively, fracture toughness test results or inferred toughness property changes from tensile, hardness, indentation, or other miniature specimen testing (1). 1.2 This guide is designed to accommodate the variable response of reactor-vessel materials in post-irradiation annealing at various temperatures and different time periods. Certain inherent limiting factors must be considered in developing an annealing procedure. These factors include system-design limitations; physical constrain...

  13. NA 30 - Deputy Administrator for Naval Reactors | National Nuclear Security

    National Nuclear Security Administration (NNSA)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of NaturalDukeWakefield Municipal GasAdministration Medal01 Sandia4) August 2012 Guidance for HighNational Nuclear

  14. University Research Reactor Task Force to the Nuclear Energy Research

    Energy Savers [EERE]

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on QA:QA J-E-1 SECTION J APPENDIX E LIST OF APPLICABLEStatutoryin theNuclear Energy Research andofRod BeeverDataAdvisory

  15. NEAC Nuclear Reactor Technology (NRT) Subcommittee Advanced Test and/or Demonstration Reactor Planning Study

    Energy Savers [EERE]

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on DeliciousMathematicsEnergyInterested Parties -Department of EnergyNEW YORKFuel Cycle TechnologiesNuclear

  16. Dynamic Complexity Study of Nuclear Reactor and Process Heat Application Integration

    SciTech Connect (OSTI)

    J'Tia Patrice Taylor; David E. Shropshire

    2009-09-01

    Abstract This paper describes the key obstacles and challenges facing the integration of nuclear reactors with process heat applications as they relate to dynamic issues. The paper also presents capabilities of current modeling and analysis tools available to investigate these issues. A pragmatic approach to an analysis is developed with the ultimate objective of improving the viability of nuclear energy as a heat source for process industries. The extension of nuclear energy to process heat industries would improve energy security and aid in reduction of carbon emissions by reducing demands for foreign derived fossil fuels. The paper begins with an overview of nuclear reactors and process application for potential use in an integrated system. Reactors are evaluated against specific characteristics that determine their compatibility with process applications such as heat outlet temperature. The reactor system categories include light water, heavy water, small to medium, near term high-temperature, and far term high temperature reactors. Low temperature process systems include desalination, district heating, and tar sands and shale oil recovery. High temperature processes that support hydrogen production include steam reforming, steam cracking, hydrogen production by electrolysis, and far-term applications such as the sulfur iodine chemical process and high-temperature electrolysis. A simple static matching between complementary systems is performed; however, to gain a true appreciation for system integration complexity, time dependent dynamic analysis is required. The paper identifies critical issues arising from dynamic complexity associated with integration of systems. Operational issues include scheduling conflicts and resource allocation for heat and electricity. Additionally, economic and safety considerations that could impact the successful integration of these systems are considered. Economic issues include the cost differential arising due to an integrated system and the economic allocation of electricity and heat resources. Safety issues include changes in regulatory constraints imposed on the facilities. Modeling and analysis tools, such as System Dynamics for time dependent operational and economic issues and RELAP5 3D for chemical transient affects, are evaluated. The results of this study advance the body of knowledge toward integration of nuclear reactors and process heat applications.

  17. Progress in the Development of Compressible, Multiphase Flow Modeling Capability for Nuclear Reactor Flow Applications

    SciTech Connect (OSTI)

    R. A. Berry; R. Saurel; F. Petitpas; E. Daniel; O. Le Metayer; S. Gavrilyuk; N. Dovetta

    2008-10-01

    In nuclear reactor safety and optimization there are key issues that rely on in-depth understanding of basic two-phase flow phenomena with heat and mass transfer. Within the context of multiphase flows, two bubble-dynamic phenomena – boiling (heterogeneous) and flashing or cavitation (homogeneous boiling), with bubble collapse, are technologically very important to nuclear reactor systems. The main difference between boiling and flashing is that bubble growth (and collapse) in boiling is inhibited by limitations on the heat transfer at the interface, whereas bubble growth (and collapse) in flashing is limited primarily by inertial effects in the surrounding liquid. The flashing process tends to be far more explosive (and implosive), and is more violent and damaging (at least in the near term) than the bubble dynamics of boiling. However, other problematic phenomena, such as crud deposition, appear to be intimately connecting with the boiling process. In reality, these two processes share many details.

  18. Magnetic Field Generation in Planets and Satellites by Natural Nuclear Fission Reactors

    E-Print Network [OSTI]

    J. Marvin Herndon

    2007-09-27

    One of the most fundamental problems in physics has been to understand the nature of the mechanism that generates the geomagnetic field and the magnetic fields of other planets and satellites. For decades, the dynamo mechanism, thought to be responsible for generating the geomagnetic field and other planetary magnetic fields, has been ascribed to convection in each planet's iron-alloy core. Recently, I described the problems inherent in Earth-core convection and proposed instead that the geomagnetic field is produced by a dynamo mechanism involving convection, not in the fluid core, but in the electrically conductive, fluid, fission-product sub-shell of a natural nuclear fission reactor at the center of the Earth, called the georeactor. Here I set forth in detail the commonality in the Solar System of the matter like that of the inside of the Earth, which is my basis for generalizing the concept of planetary magnetic field generation by natural planetocentric nuclear fission reactors.

  19. Advanced dry head-end reprocessing of light water reactor spent nuclear fuel

    SciTech Connect (OSTI)

    Collins, Emory D.; Delcul, Guillermo D.; Hunt, Rodney D.; Johnson, Jared A.; Spencer, Barry B.

    2014-06-10

    A method for reprocessing spent nuclear fuel from a light water reactor includes the step of reacting spent nuclear fuel in a voloxidation vessel with an oxidizing gas having nitrogen dioxide and oxygen for a period sufficient to generate a solid oxidation product of the spent nuclear fuel. The reacting step includes the step of reacting, in a first zone of the voloxidation vessel, spent nuclear fuel with the oxidizing gas at a temperature ranging from 200-450.degree. C. to form an oxidized reaction product, and regenerating nitrogen dioxide, in a second zone of the voloxidation vessel, by reacting oxidizing gas comprising nitrogen monoxide and oxygen at a temperature ranging from 0-80.degree. C. The first zone and the second zone can be separate. A voloxidation system is also disclosed.

  20. Advanced dry head-end reprocessing of light water reactor spent nuclear fuel

    DOE Patents [OSTI]

    Collins, Emory D; Delcul, Guillermo D; Hunt, Rodney D; Johnson, Jared A; Spencer, Barry B

    2013-11-05

    A method for reprocessing spent nuclear fuel from a light water reactor includes the step of reacting spent nuclear fuel in a voloxidation vessel with an oxidizing gas having nitrogen dioxide and oxygen for a period sufficient to generate a solid oxidation product of the spent nuclear fuel. The reacting step includes the step of reacting, in a first zone of the voloxidation vessel, spent nuclear fuel with the oxidizing gas at a temperature ranging from 200-450.degree. C. to form an oxidized reaction product, and regenerating nitrogen dioxide, in a second zone of the voloxidation vessel, by reacting oxidizing gas comprising nitrogen monoxide and oxygen at a temperature ranging from 0-80.degree. C. The first zone and the second zone can be separate. A voloxidation system is also disclosed.

  1. Hydrogen loops in existing reactors for testing fuel elements for nuclear propulsion

    SciTech Connect (OSTI)

    Olsen, C.S.; Welland, H.; Abraschoff, J. (Idaho National Engineering Laboratory, EG G Idaho Inc., P.O. Box 1625, Idaho Falls, Idaho 83415 (United States)); Thoms, K. (Oak Ridge National Laboratory, P.O. Box, Oak Ridge, Tennessee 37831-8087 (United States))

    1993-01-15

    The Space Exploration Initiative (SEI) has revitalized interest in adapting nuclear energy for power and propulsion. Prior to the selection of a nuclear thermal propulsion (NTP) system, extensive testing of the various proposed concepts will be required. In today's environmental, safety and health culture, full size rocket engine tests as were done under the Rover/NERVA program will be extremely difficult and expensive to perform and meet NASA's schedules. A different test strategy uses a hydrogen loop in an existing reactor to test a wide variety of single elements or clusters of elements for fuel qualification. This approach is expected to reduce operating and capital costs and expedite the testing schedule. This paper examines the potential of performing subscale tests in a hydrogen loop in an existing reactor such as the Advanced Test Reactor (ATR) at the Idaho National Engineering Laboratory or the High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory. The HFIR is expected to achieve power densities comparable to those achieved in ATR because of the 85 MWt power level and the high thermal and fast flux levels. The available length and diameter of the test region of FHIR are 60 cm and 10 cm whereas the available length and diameter of the test region of ATR are 120 cm and 12 cm respectively.

  2. TITAN : an advanced three dimensional coupled neutronicthermal-hydraulics code for light water nuclear reactor core analysis

    E-Print Network [OSTI]

    Griggs, D. P.

    1984-01-01

    The accurate analysis of nuclear reactor transients frequently requires that neutronics, thermal-hydraulics and feedback be included. A number of coupled neutronics/thermal-hydraulics codes have been developed for this ...

  3. The Use of experiments on a single fuel element to determine the nuclear parameters of reactor lattices

    E-Print Network [OSTI]

    Pilat, E. E., 1937-

    1967-01-01

    The nuclear parameters of a reactor lattice may be determined by critical experiments on that lattice, by theoretical calculations in which only cross sections are used as input, or by methods which combine theory and ...

  4. Fission neutron/gamma irradiation of Bacillus thuringiensis bacteria at the Texas A&M University Nuclear Science Center Reactor 

    E-Print Network [OSTI]

    Hearnsberger, David Wayne

    2001-01-01

    The objective of this research is to fully characterize the effectiveness of the Texas A&M University Nuclear Science Center Reactor (TAMU NSCR) neutrons for bacterial sterilization, and to assess the secondary gamma flux produced when neutrons...

  5. Thermal-hydraulic analysis of advanced reactor concepts: The Gas Core Nuclear Rocket

    SciTech Connect (OSTI)

    Banjac, V.; Heger, A.S.

    1995-12-31

    The Gas Core Nuclear Rocket (GCNR), a design first proposed in the 1960s for fast round-trip missions to Mars and the outer planets, is generally considered to be the most advanced, and therefore the most complex, iteration of the fission reactor concept. The GCNR technology involves the extraction of fission energy, by means of thermal radiation, from a high-temperature plasma core to a working fluid. A specific derivative of GCNR technology is the nuclear fight bulb (NLB) rocket engine, first proposed by the then United Aircraft Research Laboratories (UARL) in the early 1960s. The potential operating parameters provided the motivation for a detailed thermal hydraulics analysis.

  6. Integral Validation of Minor Actinide Nuclear Data by using Samples Irradiated at Dounreay Prototype Fast Reactor

    SciTech Connect (OSTI)

    Tsujimoto, Kazufumi; Oigawa, Hiroyuki; Shinohara, Nobuo [Japan Atomic Energy Research Institute, Shirakata Shirane 2-4, Tokai, Ibaraki 319-1195 (Japan)

    2005-05-24

    The reliability of nuclear data for minor actinides was evaluated by using the results of the post-irradiation experiment for actinide samples irradiated at the Dounreay Prototype Fast Reactor. The burnup calculations with JENDL-3.3, ENDF/B-VI.8, and JEFF-3.0 were performed. From the comparison between the experimental data and the calculational results, in general, the reliability of nuclear data for the minor actinides are at an adequate level for the conceptual design study of transmutation systems. It is, however, found that improvement of the accuracy is necessary for some nuclides, such as 238Pu, 242Pu, and 241Am.

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

    SciTech Connect (OSTI)

    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.

  8. Fuel element design for the enhanced destruction of plutonium in a nuclear reactor

    DOE Patents [OSTI]

    Crawford, Douglas C. (Idaho Falls, ID); Porter, Douglas L. (Idaho Falls, ID); Hayes, Steven L. (Idaho Falls, ID); Hill, Robert N. (Bolingbrook, IL)

    1999-01-01

    A uranium-free fuel for a fast nuclear reactor comprising an alloy of Pu, Zr and Hf, wherein Hf is present in an amount less than about 10% by weight of the alloy. The fuel may be in the form of a Pu alloy surrounded by a Zr--Hf alloy or an alloy of Pu--Zr--Hf or a combination of both.

  9. Descriptions of selected accidents that have occurred at nuclear reactor facilities

    SciTech Connect (OSTI)

    Bertini, H.W.

    1980-04-01

    This report was prepared at the request of the President's Commission on the Accident at Three Mile Island to provide the members of the Commission with some insight into the nature and significance of accidents that have occurred at nuclear reactor facilities in the past. Toward that end, this report presents a brief description of 44 accidents which have occurred throughout the world and which meet at least one of the severity criteria that were established.

  10. Fuel element design for the enhanced destruction of plutonium in a nuclear reactor

    DOE Patents [OSTI]

    Crawford, D.C.; Porter, D.L.; Hayes, S.L.; Hill, R.N.

    1999-03-23

    A uranium-free fuel for a fast nuclear reactor comprising an alloy of Pu, Zr and Hf, wherein Hf is present in an amount less than about 10% by weight of the alloy. The fuel may be in the form of a Pu alloy surrounded by a Zr--Hf alloy or an alloy of Pu--Zr--Hf or a combination of both. 7 figs.

  11. Temperature dependent scattering cross section effects on nuclear reactor control 

    E-Print Network [OSTI]

    Biggs, Charles Leon

    1968-01-01

    Tglil . ?', 'i'l. 'i:i D;;!:i;":O &lT SC'. I", EIi!g CROSS S'CTIG. 'i hgl g( I. . G". I llgisE. ', R EE. '. G" OR CO;ITI. 'OL A Thoe1e CriIRIES I, "Gli BIGSS, JR, SubmItted to the Craduate Co11ege Of ths Ts:caa Ag? UnIversfty 1n partIal fulf...IIlment oi' tho requIrenonts for ths degree of NASTER OF SCIENCE January 1958 Ma)or Subgectm Nuclear EngInssrIng A Tn ops by Cr'PEG. S& LEO'! LEGGS, JB. APl. p'Col'col c 8 tG 8t yl P ClgG cQ"It'. ". t bye (- (Ci (. i:. .E". . eu oC Ci'. nfttso...

  12. Small Reactor Designs Suitable for Direct Nuclear Thermal Propulsion: Interim Report

    SciTech Connect (OSTI)

    Bruce G. Schnitzler

    2012-01-01

    Advancement of U.S. scientific, security, and economic interests requires high performance propulsion systems to support missions beyond low Earth orbit. A robust space exploration program will include robotic outer planet and crewed missions to a variety of destinations including the moon, near Earth objects, and eventually Mars. Past studies, in particular those in support of both the Strategic Defense Initiative (SDI) and the Space Exploration Initiative (SEI), have shown nuclear thermal propulsion systems provide superior performance for high mass high propulsive delta-V missions. In NASA's recent Mars Design Reference Architecture (DRA) 5.0 study, nuclear thermal propulsion (NTP) was again selected over chemical propulsion as the preferred in-space transportation system option for the human exploration of Mars because of its high thrust and high specific impulse ({approx}900 s) capability, increased tolerance to payload mass growth and architecture changes, and lower total initial mass in low Earth orbit. The recently announced national space policy2 supports the development and use of space nuclear power systems where such systems safely enable or significantly enhance space exploration or operational capabilities. An extensive nuclear thermal rocket technology development effort was conducted under the Rover/NERVA, GE-710 and ANL nuclear rocket programs (1955-1973). Both graphite and refractory metal alloy fuel types were pursued. The primary and significantly larger Rover/NERVA program focused on graphite type fuels. Research, development, and testing of high temperature graphite fuels was conducted. Reactors and engines employing these fuels were designed, built, and ground tested. The GE-710 and ANL programs focused on an alternative ceramic-metallic 'cermet' fuel type consisting of UO2 (or UN) fuel embedded in a refractory metal matrix such as tungsten. The General Electric program examined closed loop concepts for space or terrestrial applications as well as open loop systems for direct nuclear thermal propulsion. Although a number of fast spectrum reactor and engine designs suitable for direct nuclear thermal propulsion were proposed and designed, none were built. This report summarizes status results of evaluations of small nuclear reactor designs suitable for direct nuclear thermal propulsion.

  13. A REVIEW OF LIGHT-WATER REACTOR SAFETY STUDIES. VOLUME 3 OF THE FINAL REPORT ON HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL ELECTRIC GENERATION IN CALIFORNIA

    E-Print Network [OSTI]

    Nero, A.V.

    2010-01-01

    Testimony from the Nuclear Regulatory Commission indivi­ ofPlants", U.S. Nuclear Regulatory Commission Report WASH-Yellin, "The Nuclear Regulatory Commission's Reactor Safety

  14. Knowledge and abilities catalog for nuclear power plant operators: boiling water reactors

    SciTech Connect (OSTI)

    Not Available

    1986-09-01

    The Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Boiling-Water Reactors (BWR) (NUREG-1123) provides the basis for the development of content-valid licensing examinations for reactor operators (ROs) and senior reactor operators (SROs). The examinations developed using the BWR Catalog and Examiners' Handbook for Developing Operator Licensing Examinations (NUREG-1121) will cover those topics listed under Title 10, Code of Federal Regulations, Part 55. The BWR Catalog contains approximately 7000 knowledge and ability (K/A) statements for ROs and SROs at boiling water reactors. Each K/A statement has been rated for its importance to the safe operation of the plant in a manner ensuring personnel and public health and safety. The BWR K/A Catalog is organized into five major sections: Plant-wide Generic Knowledge and Ability Statements, Plant Systems grouped by Safety Function, Emergency and Abnormal Plant Evolutions, Components, and Theory. The BWR Catalog represents a modification of the form and content of the K/A Catalog for Nuclear Power Plant Operators: Pressurized Water Reactors (NUREG-1122). First, categories of knowledge and ability statements have been redefined. Second, the scope of the definition of emergency and abnormal plant evolutions has been revised in line with a symptom-based approach. Third, K/As related to the operational applications of theory have been incorporated into the delineations for both plant systems and emergency and abnormal plant evolutions, while K/As pertaining to theory fundamental to plant operation have been delineated in a separate theory section. Finally, the components section has been revised.

  15. Apparatus and method for classifying fuel pellets for nuclear reactor

    DOE Patents [OSTI]

    Wilks, Robert S. (Plum Borough, PA); Sternheim, Eliezer (Pittsburgh, PA); Breakey, Gerald A. (Penn Township, Allegheny County, PA); Sturges, Jr., Robert H. (Plum Borough, PA); Taleff, Alexander (Churchill Borough, PA); Castner, Raymond P. (Monroeville, PA)

    1984-01-01

    Control for the operation of a mechanical handling and gauging system for nuclear fuel pellets. The pellets are inspected for diameters, lengths, surface flaws and weights in successive stations. The control includes, a computer for commanding the operation of the system and its electronics and for storing and processing the complex data derived at the required high rate. In measuring the diameter, the computer enables the measurement of a calibration pellet, stores that calibration data and computes and stores diameter-correction factors and their addresses along a pellet. To each diameter measurement a correction factor is applied at the appropriate address. The computer commands verification that all critical parts of the system and control are set for inspection and that each pellet is positioned for inspection. During each cycle of inspection, the measurement operation proceeds normally irrespective of whether or not a pellet is present in each station. If a pellet is not positioned in a station, a measurement is recorded, but the recorded measurement indicates maloperation. In measuring diameter and length a light pattern including successive shadows of slices transverse for diameter or longitudinal for length are projected on a photodiode array. The light pattern is scanned electronically by a train of pulses. The pulses are counted during the scan of the lighted diodes. For evaluation of diameter the maximum diameter count and the number of slices for which the diameter exceeds a predetermined minimum is determined. For acceptance, the maximum must be less than a maximum level and the minimum must exceed a set number. For evaluation of length, the maximum length is determined. For acceptance, the length must be within maximum and minimum limits.

  16. Next Generation Nuclear Plant Reactor Pressure Vessel Materials Research and Development Plan (PLN-2803)

    SciTech Connect (OSTI)

    J. K. Wright; R. N. Wright

    2010-07-01

    The U.S. Department of Energy (DOE) has selected the High-Temperature Gas-cooled Reactor (HTGR) design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production, with an outlet gas temperature in the range of 750°C, and a design service life of 60 years. The reactor design will be a graphite-moderated, helium-cooled, prismatic, or pebble bed reactor and use low-enriched uranium, Tri-Isotopic (TRISO)-coated fuel. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. Selection of the technology and design configuration for the NGNP must consider both the cost and risk profiles to ensure that the demonstration plant establishes a sound foundation for future commercial deployments. The NGNP challenge is to achieve a significant advancement in nuclear technology while setting the stage for an economically viable deployment of the new technology in the commercial sector soon after 2020. This technology development plan details the additional research and development (R&D) required to design and license the NGNP RPV, assuming that A 508/A 533 is the material of construction. The majority of additional information that is required is related to long-term aging behavior at NGNP vessel temperatures, which are somewhat above those commonly encountered in the existing database from LWR experience. Additional data are also required for the anticipated NGNP environment. An assessment of required R&D for a Grade 91 vessel has been retained from the first revision of the R&D plan in Appendix B in somewhat less detail. Considerably more development is required for this steel compared to A 508/A 533 including additional irradiation testing for expected NGNP operating temperatures, high-temperature mechanical properties, and extensive studies of long-term microstructural stability.

  17. Apparatus for suppressing formation of vortices in the coolant fluid of a nuclear reactor and associated method

    DOE Patents [OSTI]

    Ekeroth, Douglas E. (Delmont, PA); Garner, Daniel C. (Murrysville, PA); Hopkins, Ronald J. (Pensacola, FL); Land, John T. (Pensacola, FL)

    1993-01-01

    An apparatus and method are provided for suppressing the formation of vortices in circulating coolant fluid of a nuclear reactor. A vortex-suppressing plate having a plurality of openings therein is suspended within the lower plenum of a reactor vessel below and generally parallel to the main core support of the reactor. The plate is positioned so as to intersect vortices which may form in the circulating reactor coolant fluid. The intersection of the plate with such vortices disrupts the rotational flow pattern of the vortices, thereby disrupting the formation thereof.

  18. Apparatus for suppressing formation of vortices in the coolant fluid of a nuclear reactor and associated method

    DOE Patents [OSTI]

    Ekeroth, D.E.; Garner, D.C.; Hopkins, R.J.; Land, J.T.

    1993-11-30

    An apparatus and method are provided for suppressing the formation of vortices in circulating coolant fluid of a nuclear reactor. A vortex-suppressing plate having a plurality of openings therein is suspended within the lower plenum of a reactor vessel below and generally parallel to the main core support of the reactor. The plate is positioned so as to intersect vortices which may form in the circulating reactor coolant fluid. The intersection of the plate with such vortices disrupts the rotational flow pattern of the vortices, thereby disrupting the formation thereof. 3 figures.

  19. Next Generation Nuclear Plant Reactor Pressure Vessel Materials Research and Development Plan (PLN-2803)

    SciTech Connect (OSTI)

    J. K. Wright; R. N. Wright

    2008-04-01

    The U.S. Department of Energy has selected the High Temperature Gas-cooled Reactor design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production. It will have an outlet gas temperature in the range of 900°C and a plant design service life of 60 years. The reactor design will be a graphite moderated, helium-cooled, prismatic, or pebble-bed reactor and use low-enriched uranium, Tri-Isotopic-coated fuel. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. The NGNP Materials Research and Development Program is responsible for performing research and development on likely NGNP materials in support of the NGNP design, licensing, and construction activities. Selection of the technology and design configuration for the NGNP must consider both the cost and risk profiles to ensure that the demonstration plant establishes a sound foundation for future commercial deployments. The NGNP challenge is to achieve a significant advancement in nuclear technology while setting the stage for an economically viable deployment of the new technology in the commercial sector soon after 2020. Studies of potential Reactor Pressure Vessel (RPV) steels have been carried out as part of the pre-conceptual design studies. These design studies generally focus on American Society of Mechanical Engineers (ASME) Code status of the steels, temperature limits, and allowable stresses. Three realistic candidate materials have been identified by this process: conventional light water reactor RPV steels A508/533, 2¼Cr-1Mo in the annealed condition, and modified 9Cr 1Mo ferritic martenistic steel. Based on superior strength and higher temperature limits, the modified 9Cr-1Mo steel has been identified by the majority of design engineers as the preferred choice for the RPV. All of the vendors have concluded, however, that with adequate engineered cooling of the vessel, the A508/533 steels are also acceptable.

  20. EU, U.S., Russia, Asian States Sign Nuclear-Fusion Reactor May 24 (Bloomberg) --The European Union, the U.S., Russia and Asian nations

    E-Print Network [OSTI]

    -gas emissions from fossil fuels, rising oil prices and worries about the safety of existing nuclear plantsEurope EU, U.S., Russia, Asian States Sign Nuclear-Fusion Reactor Pact May 24 (Bloomberg nuclear-fusion reactor, forging ahead with a 4.6 billion- euro ($5.9 billion) project to cut oil demand

  1. BFS, a Legacy to the International Reactor Physics, Criticality Safety, and Nuclear Data Communities

    SciTech Connect (OSTI)

    J. Blair Briggs; Anatoly Tsibulya; Yevgeniy Rozhikhin

    2012-03-01

    Interest in high-quality integral benchmark data is increasing as efforts to quantify and reduce calculational uncertainties accelerate to meet the demands of next generation reactor and advanced fuel cycle concepts. Two Organization for Economic Cooperation and Development (OECD) Nuclear Energy Agency (NEA) activities, the International Criticality Safety Benchmark Evaluation Project (ICSBEP), initiated in 1992, and the International Reactor Physics Experiment Evaluation Project (IRPhEP), initiated in 2003, have been identifying existing integral experiment data, evaluating those data, and providing integral benchmark specifications for methods and data validation for nearly two decades. Thus far, 14 countries have contributed to the IRPhEP, and 20 have contributed to the ICSBEP. Data provided by these two projects will be of use to the international reactor physics, criticality safety, and nuclear data communities for future decades The Russian Federation has been a major contributor to both projects with the Institute of Physics and Power Engineering (IPPE) as the major contributor from the Russian Federation. Included in the benchmark specifications from the BFS facilities are 34 critical configurations from BFS-49, 61, 62, 73, 79, 81, 97, 99, and 101; spectral characteristics measurements from BFS-31, 42, 57, 59, 61, 62, 73, 97, 99, and 101; reactivity effects measurements from BFS-62-3A; reactivity coefficients and kinetics measurements from BFS-73; and reaction rate measurements from BFS-42, 61, 62, 73, 97, 99, and 101.

  2. Nuclear reactor with internal thimble-type delayed neutron detection system

    DOE Patents [OSTI]

    Gross, Kenny C. (Lemont, IL); Poloncsik, John (Downers Grove, IL); Lambert, John D. B. (Wheaton, IL)

    1990-01-01

    This invention teaches improved apparatus for the method of detecting a breach in cladded fuel used in a nuclear reactor. The detector apparatus is located in the primary heat exchanger which conveys part of the reactor coolant past at least three separate delayed-neutron detectors mounted in this heat exchanger. The detectors are spaced apart such that the coolant flow time from the core to each detector is different, and these differences are known. The delayed-neutron activity at the detectors is a function of the delay time after the reaction in the fuel until the coolant carrying the delayed-neutron emitter passes the respective detector. This time delay is broken down into separate components including an isotopic holdup time required for the emitter to move through the fuel from the reaction to the coolant at the breach, and two transit times required for the emitter now in the coolant to flow from the breach to the detector loop and then via the loop to the detector. At least two of these time components are determined during calibrated operation of the reactor. Thereafter during normal reactor operation, repeated comparisons are made by the method of regression approximation of the third time component for the best-fit line correlating measured delayed-neutron activity against activity that is approximated according to specific equations. The equations use these time-delay components and known parameter values of the fuel and of the part and emitting daughter isotopes.

  3. Analysis of Granular Flow in a Pebble-Bed Nuclear Reactor

    E-Print Network [OSTI]

    Chris H. Rycroft; Gary S. Grest; James W. Landry; Martin Z. Bazant

    2006-02-16

    Pebble-bed nuclear reactor technology, which is currently being revived around the world, raises fundamental questions about dense granular flow in silos. A typical reactor core is composed of graphite fuel pebbles, which drain very slowly in a continuous refueling process. Pebble flow is poorly understood and not easily accessible to experiments, and yet it has a major impact on reactor physics. To address this problem, we perform full-scale, discrete-element simulations in realistic geometries, with up to 440,000 frictional, viscoelastic 6cm-diameter spheres draining in a cylindrical vessel of diameter 3.5m and height 10m with bottom funnels angled at 30 degrees or 60 degrees. We also simulate a bidisperse core with a dynamic central column of smaller graphite moderator pebbles and show that little mixing occurs down to a 1:2 diameter ratio. We analyze the mean velocity, diffusion and mixing, local ordering and porosity (from Voronoi volumes), the residence-time distribution, and the effects of wall friction and discuss implications for reactor design and the basic physics of granular flow.

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

    DOE Patents [OSTI]

    Corletti, Michael M. (New Kensington, PA); Schulz, Terry L. (Murrysville, PA)

    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.

  5. Physics-based multiscale coupling for full core nuclear reactor simulation

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Gaston, Derek R.; Permann, Cody J.; Peterson, John W.; Slaughter, Andrew E.; Andrš, David; Wang, Yaqi; Short, Michael P.; Perez, Danielle M.; Tonks, Michael R.; Ortensi, Javier; et al

    2015-10-01

    Numerical simulation of nuclear reactors is a key technology in the quest for improvements in efficiency, safety, and reliability of both existing and future reactor designs. Historically, simulation of an entire reactor was accomplished by linking together multiple existing codes that each simulated a subset of the relevant multiphysics phenomena. Recent advances in the MOOSE (Multiphysics Object Oriented Simulation Environment) framework have enabled a new approach: multiple domain-specific applications, all built on the same software framework, are efficiently linked to create a cohesive application. This is accomplished with a flexible coupling capability that allows for a variety of different datamore »exchanges to occur simultaneously on high performance parallel computational hardware. Examples based on the KAIST-3A benchmark core, as well as a simplified Westinghouse AP-1000 configuration, demonstrate the power of this new framework for tackling—in a coupled, multiscale manner—crucial reactor phenomena such as CRUD-induced power shift and fuel shuffle. 2014 The Authors. Published by Elsevier Ltd. This is an open access article under the CC BY-NC-SA license« less

  6. Jacking mechanism for upper internals structure of a liquid metal nuclear reactor

    DOE Patents [OSTI]

    Gillett, James E. (Greensburg, PA); Wineman, Arthur L. (Greensburg, PA)

    1984-01-01

    A jacking mechanism for raising the upper internals structure of a liquid metal nuclear reactor which jacking mechanism uses a system of gears and drive shafts to transmit force from a single motor to four mechanically synchronized ball jacks to raise and lower support columns which support the upper internals structure. The support columns have a pin structure which rides up and down in a slot in a housing fixed to the reactor head. The pin has two locking plates which can be rotated around the pin to bring bolt holes through the locking plates into alignment with a set of bolt holes in the housing, there being a set of such housing bolt holes corresponding to both a raised and a lowered position of the support column. When the locking plate is so aligned, a surface of the locking plate mates with a surface in the housing such that the support column is then supported by the locking plate and not by the ball jacks. Since the locking plates are to be installed and bolted to the housing during periods of reactor operation, the ball jacks need not be sized to react the large forces which occur or potentially could occur on the upper internals structure of the reactor during operation. The locking plates react these loads. The ball jacks, used only during refueling, can be smaller, which enable conventionally available equipment to fulfill the precision requirements for the task within available space.

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

    DOE Patents [OSTI]

    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.

  8. Reflected kinetics model for nuclear space reactor kinetics and control scoping calculations

    SciTech Connect (OSTI)

    Washington, K.E.

    1986-05-01

    The objective of this research is to develop a model that offers an alternative to the point kinetics (PK) modelling approach in the analysis of space reactor kinetics and control studies. Modelling effort will focus on the explicit treatment of control drums as reactivity input devices so that the transition to automatic control can be smoothly done. The proposed model is developed for the specific integration of automatic control and the solution of the servo mechanism problem. The integration of the kinetics model with an automatic controller will provide a useful tool for performing space reactor scoping studies for different designs and configurations. Such a tool should prove to be invaluable in the design phase of a space nuclear system from the point of view of kinetics and control limitations.

  9. Exploration of High-Dimensional Scalar Function for Nuclear Reactor Safety Analysis and Visualization

    SciTech Connect (OSTI)

    Dan Maljovec; Bei Wang; Valerio Pascucci; Peer-Timo Bremer; Michael Pernice; Robert Nourgaliev

    2013-05-01

    The next generation of methodologies for nuclear reactor Probabilistic Risk Assessment (PRA) explicitly accounts for the time element in modeling the probabilistic system evolution and uses numerical simulation tools to account for possible dependencies between failure events. The Monte-Carlo (MC) and the Dynamic Event Tree (DET) approaches belong to this new class of dynamic PRA methodologies. A challenge of dynamic PRA algorithms is the large amount of data they produce which may be difficult to visualize and analyze in order to extract useful information. We present a software tool that is designed to address these goals. We model a large-scale nuclear simulation dataset as a high-dimensional scalar function defined over a discrete sample of the domain. First, we provide structural analysis of such a function at multiple scales and provide insight into the relationship between the input parameters and the output. Second, we enable exploratory analysis for users, where we help the users to differentiate features from noise through multi-scale analysis on an interactive platform, based on domain knowledge and data characterization. Our analysis is performed by exploiting the topological and geometric properties of the domain, building statistical models based on its topological segmentations and providing interactive visual interfaces to facilitate such explorations. We provide a user’s guide to our software tool by highlighting its analysis and visualization capabilities, along with a use case involving dataset from a nuclear reactor safety simulation.

  10. Nuclear Systems Enhanced Performance Program, Maintenance Cycle Extension in Advanced Light Water Reactor Design

    SciTech Connect (OSTI)

    Professor Neill Todreas

    2001-10-01

    A renewed interest in new nuclear power generation in the US has spurred interest in developing advanced reactors with features which will address the public's concerns regarding nuclear generation. However, it is economic performance which will dictate whether any new orders for these plants will materialize. Economic performance is, to a great extent, improved by maximizing the time that the plant is on-line generating electricity relative to the time spent off-line conducting maintenance and refueling. Indeed, the strategy for the advanced light water reactor plant IRIS (International Reactor, Innovative and Secure) is to utilize an eight year operating cycle. This report describes a formalized strategy to address, during the design phase, the maintenance-related barriers to an extended operating cycle. The top-level objective of this investigation was to develop a methodology for injecting component and system maintainability issues into the reactor plant design process to overcome these barriers. A primary goal was to demonstrate the applicability and utility of the methodology in the context of the IRIS design. The first step in meeting the top-level objective was to determine the types of operating cycle length barriers that the IRIS design team is likely to face. Evaluation of previously identified regulatory and investment protection surveillance program barriers preventing a candidate operating PWR from achieving an extended (48 month) cycle was conducted in the context of the IRIS design. From this analysis, 54 known IRIS operating cycle length barriers were identified. The resolution methodology was applied to each of these barriers to generate design solution alternatives for consideration in the IRIS design. The methodology developed has been demonstrated to narrow the design space to feasible design solutions which enable a desired operating cycle length, yet is general enough to have broad applicability. Feedback from the IRIS design team indicates that the proposed solutions to the investigated operating cycle length barriers are both feasible and consistent with sound design practice.

  11. An American company thinks it can have a commercial reactor ready and working within a decade ONE of the clichs of nuclear-power research is that a commercial fusion reactor is only 30 years away, and always1

    E-Print Network [OSTI]

    South Bohemia, University of

    of the clichés of nuclear-power research is that a commercial fusion reactor is only 30 years away, and always1An American company thinks it can have a commercial reactor ready and working within a decade ONE that Lockheed Martin,2 a big American engineering and defence company, has a new design for a fusion reactor

  12. Assessment of fission product yields data needs in nuclear reactor applications

    SciTech Connect (OSTI)

    Kern, K.; Becker, M.; Broeders, C.

    2012-07-01

    Studies on the build-up of fission products in fast reactors have been performed, with particular emphasis on the effects related to the physics of the nuclear fission process. Fission product yields, which are required for burn-up calculations, depend on the proton and neutron number of the target nucleus as well as on the incident neutron energy. Evaluated nuclear data on fission product yields are available for all relevant target nuclides in reactor applications. However, the description of their energy dependence in evaluated data is still rather rudimentary, which is due to the lack of experimental fast fission data and reliable physical models. Additionally, physics studies of evaluated JEFF-3.1.1 fission yields data have shown potential improvements, especially for various fast fission data sets of this evaluation. In recent years, important progress in the understanding of the fission process has been made, and advanced model codes are currently being developed. This paper deals with the semi-empirical approach to the description of the fission process, which is used in the GEF code being developed by K.-H. Schmidt and B. Jurado on behalf of the OECD Nuclear Energy Agency, and with results from the corresponding author's diploma thesis. An extended version of the GEF code, supporting the calculation of spectrum weighted fission product yields, has been developed. It has been applied to the calculation of fission product yields in the fission rate spectra of a MOX fuelled sodium-cooled fast reactor. Important results are compared to JEFF-3.1.1 data and discussed in this paper. (authors)

  13. Reactor-Based Management of Used Nuclear Fuel: Assessment of Major Options

    SciTech Connect (OSTI)

    Phillip J. Finck; Roald A. Wigeland; Robert N. Hill

    2011-01-01

    This paper discusses the current status of the ongoing Advanced Fuel Cycle Initiative (AFCI) program in the U.S. Department of Energy that is investigating the potential for using the processing and recycling of used nuclear fuel to improve radioactive waste management, including used fuel. A key element of the strategies is to use nuclear reactors for further irradiation of recovered chemical elements to transmute certain long-lived highly-radioactive isotopes into less hazardous isotopes. Both thermal and fast neutron spectrum reactors are being studied as part of integrated nuclear energy systems where separations, transmutation, and disposal are considered. Radiotoxicity is being used as one of the metrics for estimating the hazard of used fuel and the processing of wastes resulting from separations and recycle-fuel fabrication. Decay heat from the used fuel and/or wastes destined for disposal is used as a metric for use of a geologic repository. Results to date indicate that the most promising options appear to be those using fast reactors in a repeated recycle mode to limit buildup of higher actinides, since the transuranic elements are a key contributor to the radiotoxicity and decay heat. Using such an approach, there could be much lower environmental impact from the high-level waste as compared to direct disposal of the used fuel, but there would likely be greater generation of low-level wastes that will also require disposal. An additional potential waste management benefit is having the ability to tailor waste forms and contents to one or more targeted disposal environments (i.e., to be able to put waste in environments best-suited for the waste contents and forms).

  14. A REVIEW OF LIGHT-WATER REACTOR SAFETY STUDIES. VOLUME 3 OF THE FINAL REPORT ON HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL ELECTRIC GENERATION IN CALIFORNIA

    E-Print Network [OSTI]

    Nero, A.V.

    2010-01-01

    Nuclear Power Reactors PROTECTION AGAINST SABOTAGE Protection Against Industrial Sabotage I1C-4 Decominarion and Decommissioning of Reactors a Design Features to Control

  15. Features of temperature control of fuel element cladding for pressurized water nuclear reactor “WWER-1000” while simulating reactor accidents

    SciTech Connect (OSTI)

    Zaytsev, P. A.; Priymak, S. V.; Usachev, V. B.; Oleynikov, P. P.; Soldatkin, D. M. [Scientific Research Institute, Scientific Industrial Association LUCH, Podolsk (Russian Federation)] [Scientific Research Institute, Scientific Industrial Association LUCH, Podolsk (Russian Federation)

    2013-09-11

    During the experiments simulating NPR (nuclear power reactor) accidents with a coolant loss fuel elements behavior in a steam-hydrogen medium was studied at the temperature changed with the rate from 1 to 100K/s within the range of 300÷1500 °C. Indications of the thermocouples fixed on the cladding notably differ from real values of the cladding temperatures in the area of measuring junction due to thermal resistance influence of the transition zones “cladding-junction” and “junction-coolant”. The estimating method of a measurement error was considered which can provide adequate accounting of the influence factors. The method is based on thermal probing of a thermocouple by electric current flashing through thermoelements under the coolant presence or absence, a response time registration and processing, calculation of thermal inertia value for a thermocouple junction. A formula was derived for calculation of methodical error under stationary mode and within the stage of linear increase in temperature, which will determine the conditions for the cladding depressurization. Some variants of the formula application were considered, and the values of methodical errors were established which reached ?5% of maximum value by the final moment of the stage of linear increase in the temperature.

  16. BOLD VENTURE COMPUTATION SYSTEM for nuclear reactor core analysis, Version III

    SciTech Connect (OSTI)

    Vondy, D.R.; Fowler, T.B.; Cunningham, G.W. III.

    1981-06-01

    This report is a condensed documentation for VERSION III of the BOLD VENTURE COMPUTATION SYSTEM for nuclear reactor core analysis. An experienced analyst should be able to use this system routinely for solving problems by referring to this document. Individual reports must be referenced for details. This report covers basic input instructions and describes recent extensions to the modules as well as to the interface data file specifications. Some application considerations are discussed and an elaborate sample problem is used as an instruction aid. Instructions for creating the system on IBM computers are also given.

  17. Sensors Synergistic With Nature For In-pile Nuclear Reactor Measurements

    SciTech Connect (OSTI)

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

    2012-10-01

    To be able to evolve fuel and structural microstructure within a nuclear power reactor in an engineered manner, an effective extreme environment sensor must exist. The development of sensor technology for nondestructive and nonintrusive measurements in harsh environments is a very active field. However most of the effort has been in adapting existing sensing technology to meet the harsh environmental requirements. A different approach is being presented. The fundamental question that we are trying to answer is how do we take advantage of the harsh environment and maintain synergy between the sensor and the environment. This paper will discuss the synergistic senor being developed that takes advantage of the harsh environments.

  18. Technology, safety, and costs of decommissioning reference nuclear research and test reactors. Appendices

    SciTech Connect (OSTI)

    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 EMTOMB (entombment). The study results are presented in two volumes. Volume 2 (Appendices) contains the detailed data that support the results given in Volume 1, including unit-component data.

  19. A study of the point reactor dynamics equations as applied to large nuclear excursions 

    E-Print Network [OSTI]

    Perry, Robert Terrell

    1967-01-01

    of delayed neutrons are neglected, (() 3 0, C, (t) = O) the following equation may then be obtained by using equations 1 and 7: t n(v) dT n(t) (8) v, here K, is the reactivity in dollars above prompt critical. The solution of this equation..., J. , The De endence of Reactor Kinetics on Tem er- ature, Brookhaven National Laboratory Report BNL 173 (T-30) 1951 9. King, L, D, P, , "Description of the KIWI-TNT Excursion", Transactions of American Nuclear Societ 8: 126 (1965) 10. Nyer, W. B...

  20. Nuclear propulsion systems for orbit transfer based on the particle bed reactor

    SciTech Connect (OSTI)

    Powell, J.R.; Ludewig, H.; Horn, F.L.; Araj, K.; Benenati, R.; Lazareth, O.; Slovik, G.; Solon, M.; Tappe, W.; Belisle, J.

    1987-01-01

    The technology of nuclear direct propulsion orbit transfer systems based on the Particle Bed Reactor (PBR) is described. A 200 megawatt illustrative design is presented for LEO to GEO and other high ..delta..V missions. The PBR-NOTV can be used in a one-way mode with the shuttle or an expendable launch vehicle, e.g., the Titan 34D7, or as a two-way reusable space tug. In the one-way mode, payload capacity is almost three times greater than that of chemical OTV's. PBR technology status is described and development needs outlined.

  1. KEY DESIGN REQUIREMENTS FOR THE HIGH TEMPERATURE GAS-COOLED REACTOR NUCLEAR HEAT SUPPLY SYSTEM

    SciTech Connect (OSTI)

    L.E. Demick

    2010-09-01

    Key requirements that affect the design of the high temperature gas-cooled reactor nuclear heat supply system (HTGR-NHSS) as the NGNP Project progresses through the design, licensing, construction and testing of the first of a kind HTGR based plant are summarized. These requirements derive from pre-conceptual design development completed to-date by HTGR Suppliers, collaboration with potential end users of the HTGR technology to identify energy needs, evaluation of integration of the HTGR technology with industrial processes and recommendations of the NGNP Project Senior Advisory Group.

  2. Occupational Radiation Exposure at Commercial Nuclear Power Reactors and Other Facilities 2010, Prepared for the Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, May 2012

    SciTech Connect (OSTI)

    D. E. Lewis D. A. Hagemeyer Y. U. McCormick

    2012-07-07

    This report summarizes the occupational exposure data that are maintained in the U.S. Nuclear Regulatory Commission’s (NRC) Radiation Exposure Information and Reporting System (REIRS). The bulk of the information contained in the report was compiled from the 2010 annual reports submitted by five of the seven categories of NRC licensees subject to the reporting requirements of 10 CFR 20.2206. Because there are no geologic repositories for high-level waste currently licensed and no NRC-licensed low-level waste disposal facilities currently in operation, only five categories will be considered in this report. The annual reports submitted by these licensees consist of radiation exposure records for each monitored individual. These records are analyzed for trends and presented in this report in terms of collective dose and the distribution of dose among the monitored individuals. Annual reports for 2010 were received from a total of 190 NRC licensees. The summation of reports submitted by the 190 licensees indicated that 192,424 individuals were monitored, 81,961 of whom received a measurable dose. When adjusted for transient workers who worked at more than one licensee during the year, there were actually 142,471 monitored individuals and 62,782 who received a measurable dose. The collective dose incurred by these individuals was 10,617 person-rem, which represents a 12% decrease from the 2009 value. This decrease was primarily due to the decrease in collective dose at commercial nuclear power reactors, as well as a decrease in the collective dose for most of the other categories of NRC licensees. The number of individuals receiving a measurable dose also decreased, resulting in an average measurable dose of 0.13 rem for 2010. The average measurable dose is defined as the total effective dose equivalent (TEDE) divided by the number of individuals receiving a measurable dose. In calendar year 2010, the average annual collective dose per reactor for light water reactor (LWR) licensees was 83 person-rem. This represents a 14% decrease from the value reported for 2009 (96 person-rem). The decrease in collective dose for commercial nuclear power reactors was due to an 11% decrease in total outage hours in 2010. During outages, activities involving increased radiation exposure such as refueling and maintenance are performed while the reactor is not in operation. The average annual collective dose per reactor for boiling water reactors (BWRs) was 137 personrem for 35 BWRs, and 55 person-rem for 69 pressurized water reactors (PWRs). Analyses of transient individual data indicate that 29,333 individuals completed work assignments at two or more licensees during the monitoring year. The dose distributions are adjusted each year to account for the duplicate reporting of transient individuals by multiple licensees. The adjustment to account for transient individuals has been specifically noted in footnotes in the figures and tables for commercial nuclear power reactors. In 2010, the average measurable dose per individual for all licensees calculated from reported data was 0.13 rem. Although the average measurable dose per individual from data submitted by licensees was 0.13 rem, a corrected dose distribution resulted in an average measurable dose per individual of 0.17 rem.

  3. Mission analysis for hybrid thermionic nuclear reactor LEO-to-GEO transfer applications

    SciTech Connect (OSTI)

    Widman, F.W. Jr.; North, D.M. (Rockwell International/Rocketdyne Division, 6633 Canoga Avenue, Canoga Park, California 91303 (United States)); Choong, P.T.; Teofilo, V.L. (Lockheed Missiles and Space Company, Inc., 1111 Lockheed Way, Synnyvale, California 94088 (United States))

    1993-01-10

    This paper details the results of mission analyses concerning a hybrid STAR-C based system, which is based on a safe solid fuel form for high-temperature reactor core operation and a rugged planar thermionic energy converter for long-life steady-state electric power production. Hybrid power/propulsion system concepts are shown to offer superior performance capabilities for Low-Earth-Orbit (LEO) to Geosynchronous-Earth-Orbit (GEO) orbital transfer applications over chemical propulsion systems. A key feature of the hybrid power/propulsion system is that the propulsion system uses the on-board payload power system. Mission results for hybrid concepts using Nuclear Thermal Propulsion (NTP), Nuclear Electric Propulsion (NEP), and combination of NTP and NEP are discussed.

  4. Systems and methods for retaining and removing irradiation targets in a nuclear reactor

    DOE Patents [OSTI]

    Runkle, Gary A.; Matsumoto, Jack T.; Dayal, Yogeshwar; Heinold, Mark R.

    2015-12-08

    A retainer is placed on a conduit to control movement of objects within the conduit in access-restricted areas. Retainers can prevent or allow movement in the conduit in a discriminatory fashion. A fork with variable-spacing between prongs can be a retainer and be extended or collapsed with respect to the conduit to change the size of the conduit. Different objects of different sizes may thus react to the fork differently, some passing and some being blocked. Retainers can be installed in inaccessible areas and allow selective movement in remote portions of conduit where users cannot directly interface, including below nuclear reactors. Position detectors can monitor the movement of objects through the conduit remotely as well, permitting engagement of a desired level of restriction and object movement. Retainers are useable in a variety of nuclear power plants and with irradiation target delivery, harvesting, driving, and other remote handling or robotic systems.

  5. Direct Test of the Time-Independence of Fundamental Nuclear Constants Using the Oklo Natural Reactor

    E-Print Network [OSTI]

    Alexander I. Shlyakhter

    2003-08-06

    [NOTE: This 1983 preprint is being uploaded to arXiv.org after the death of its author, who supported online distribution of his work. Contact info of the submitter is at http://ilya.cc .] The positions of neutron resonances have been shown to be highly sensitive to the variation of fundamental nuclear constants. The analysis of the measured isotopic shifts in the natural fossil reactor at Oklo gives the following restrictions on the possible rates of the interaction constants variation: strong ~2x10^-19 yr^-1, electromagnetic ~5x10^-18 yr^-1, weak ~10^-12 yr^-1. These limits permit to exclude all the versions of nuclear constants contemporary variation discussed in the literature. URL: http://alexonline.info >. For more recent analyses see hep-ph/9606486, hep-ph/0205206 and astro-ph/0204069 .

  6. Nuclear Data for Criticality Safety and Reactor Applications at the Gaerttner LINAC Center Y. Danon, R.M. Bahran, E.J. Blain, A.M. Daskalakis, B.J. McDermott, D.G. Williams

    E-Print Network [OSTI]

    Danon, Yaron

    Nuclear Data for Criticality Safety and Reactor Applications at the Gaerttner LINAC Center Y. Danon used in reactor and nuclear criticality safety applications. The goal of this program is to provide to nuclear criticality, neutron shielding applications, nuclear reactor design, and to better understand

  7. OVERVIEW OF CRITERIA FOR INTERIM WET & DRY STORAGE OF RESEARCH REACTOR SPENT NUCLEAR FUEL

    SciTech Connect (OSTI)

    Sindelar, R.; Vinson, D.; Iyer, N.; Fisher, D.

    2010-11-03

    Following discharge from research reactors, spent nuclear fuel may be stored 'wet' in water pools or basins, or it may be stored 'dry' in various configurations including non-sealed or sealed containers until retrieved for ultimate disposition. Interim safe storage practices are based on avoiding degradation to the fuel that would impact functions related to safety. Recommended practices including environmental controls with technical bases, are outlined for wet storage and dry storage of aluminum-clad, aluminum-based research reactor fuel. For wet storage, water quality must be maintained to minimize corrosion degradation of aluminum fuel. For dry storage, vented canister storage of aluminum fuel readily provides a safe storage configuration. For sealed dry storage, drying must be performed so as to minimize water that would cause additional corrosion and hydrogen generation. Consideration must also be given to the potential for radiolytically-generated hydrogen from the bound water in the attendant oxyhydroxides on aluminum fuel from reactor operation for dry storage systems.

  8. Knowledge and abilities catalog for nuclear power plant operators: Boiling water reactors, Revision 1

    SciTech Connect (OSTI)

    1995-08-01

    The Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Boiling-Water Reactors (BWRs) (NUREG-1123, Revision 1) provides the basis for the development of content-valid licensing examinations for reactor operators (ROs) and senior reactor operators (SROs). The examinations developed using the BWR Catalog along with the Operator Licensing Examiner Standards (NUREG-1021) and the Examiner`s Handbook for Developing Operator Licensing Written Examinations (NUREG/BR-0122), will cover the topics listed under Title 10, Code of Federal Regulations, Part 55 (10 CFR 55). The BWR Catalog contains approximately 7,000 knowledge and ability (K/A) statements for ROs and SROs at BWRs. The catalog is organized into six major sections: Organization of the Catalog, Generic Knowledge and Ability Statements, Plant Systems grouped by Safety Functions, Emergency and Abnormal Plant Evolutions, Components, and Theory. Revision 1 to the BWR Catalog represents a modification in form and content of the original catalog. The K/As were linked to their applicable 10 CFR 55 item numbers. SRO level K/As were identified by 10 CFR 55.43 item numbers. The plant-wide generic and system generic K/As were combined in one section with approximately one hundred new K/As. Component Cooling Water and Instrument Air Systems were added to the Systems Section. Finally, High Containment Hydrogen Concentration and Plant Fire On Site evolutions added to the Emergency and Abnormal Plant Evolutions section.

  9. Development and application of the dynamic system doctor to nuclear reactor probabilistic risk assessments.

    SciTech Connect (OSTI)

    Kunsman, David Marvin; Aldemir, Tunc (Ohio State University); Rutt, Benjamin (Ohio State University); Metzroth, Kyle (Ohio State University); Catalyurek, Umit (Ohio State University); Denning, Richard (Ohio State University); Hakobyan, Aram (Ohio State University); Dunagan, Sean C.

    2008-05-01

    This LDRD project has produced a tool that makes probabilistic risk assessments (PRAs) of nuclear reactors - analyses which are very resource intensive - more efficient. PRAs of nuclear reactors are being increasingly relied on by the United States Nuclear Regulatory Commission (U.S.N.R.C.) for licensing decisions for current and advanced reactors. Yet, PRAs are produced much as they were 20 years ago. The work here applied a modern systems analysis technique to the accident progression analysis portion of the PRA; the technique was a system-independent multi-task computer driver routine. Initially, the objective of the work was to fuse the accident progression event tree (APET) portion of a PRA to the dynamic system doctor (DSD) created by Ohio State University. Instead, during the initial efforts, it was found that the DSD could be linked directly to a detailed accident progression phenomenological simulation code - the type on which APET construction and analysis relies, albeit indirectly - and thereby directly create and analyze the APET. The expanded DSD computational architecture and infrastructure that was created during this effort is called ADAPT (Analysis of Dynamic Accident Progression Trees). ADAPT is a system software infrastructure that supports execution and analysis of multiple dynamic event-tree simulations on distributed environments. A simulator abstraction layer was developed, and a generic driver was implemented for executing simulators on a distributed environment. As a demonstration of the use of the methodological tool, ADAPT was applied to quantify the likelihood of competing accident progression pathways occurring for a particular accident scenario in a particular reactor type using MELCOR, an integrated severe accident analysis code developed at Sandia. (ADAPT was intentionally created with flexibility, however, and is not limited to interacting with only one code. With minor coding changes to input files, ADAPT can be linked to other such codes.) The results of this demonstration indicate that the approach can significantly reduce the resources required for Level 2 PRAs. From the phenomenological viewpoint, ADAPT can also treat the associated epistemic and aleatory uncertainties. This methodology can also be used for analyses of other complex systems. Any complex system can be analyzed using ADAPT if the workings of that system can be displayed as an event tree, there is a computer code that simulates how those events could progress, and that simulator code has switches to turn on and off system events, phenomena, etc. Using and applying ADAPT to particular problems is not human independent. While the human resources for the creation and analysis of the accident progression are significantly decreased, knowledgeable analysts are still necessary for a given project to apply ADAPT successfully. This research and development effort has met its original goals and then exceeded them.

  10. Search of Axions from a Nuclear Power Reactor with a High-Purity Germanium Detector

    E-Print Network [OSTI]

    H. M. Chang; TEXONO Collaboration

    2007-01-21

    A search of axions produced in nuclear transitions was performed at the Kuo-Sheng Nuclear Power Station with a high-purity germanium detector of mass 1.06 kg at a distance of 28 m from the 2.9 GW reactor core. The expected experimental signatures were mono-energetic lines produced by their Primakoff or Compton conversions at the detector. Based on 459.0/96.3 days of Reactor ON/OFF data, no evidence of axion emissions were observed and constraints on the couplings $\\gagg$ and $\\gaee$ versus axion mass $m_a$ within the framework of invisible axion models were placed. The KSVZ and DFSZ models can be excluded for 10^4 eV < m_a < 10^6 ~eV. Model-independent constraints on \\gagg \\gv1 < 7.7 X 10^{-9} GeV^{-2} for m_{a} < 10^5 eV and \\gaee \\gv1 < 1.3 X 10^{-10} for m_{a} < 10^6 eV at 90% confidence level were derived. This experimental approach provides a unique probe for axion mass at the keV--MeV range not accessible to the other techniques.

  11. Apparatus and method for quantitative assay of generic transuranic wastes from nuclear reactors

    DOE Patents [OSTI]

    Caldwell, J.T.; Kunz, W.E.; Atencio, J.D.

    1982-03-31

    A combination of passive and active neutron measurements which yields quantitative information about the isotopic composition of transuranic wastes from nuclear power or weapons material manufacture reactors is described. From the measurement of prompt and delayed neutron emission and the incidence of two coincidentally emitted neutrons from induced fission of fissile material in the sample, one can quantify /sup 233/U, /sup 235/U and /sup 239/Pu isotopes in waste samples. Passive coincidence counting, including neutron multiplicity measurement and determination of the overall passive neutron flux additionally enables the separate quantitative evaluation of spontaneous fission isotopes such as /sup 240/Pu, /sup 244/Cm and /sup 252/Cf, and the spontaneous alpha particle emitter /sup 241/Am. These seven isotopes are the most important constituents of wastes from nuclear power reactors and once the mass of each isotope present is determined by the apparatus and method of the instant invention, the overall alpha particle activity can be determined to better than 1 nCi/g from known radioactivity data. Therefore, in addition to the quantitative analysis of the waste sample useful for later reclamation purposes, the alpha particle activity can be determined to decide whether permanent low-level burial is appropriate for the waste sample.

  12. Apparatus and method for quantitative assay of generic transuranic wastes from nuclear reactors

    DOE Patents [OSTI]

    Caldwell, John T. (Los Alamos, NM); Kunz, Walter E. (Santa Fe, NM); Atencio, James D. (Los Alamos, NM)

    1984-01-01

    A combination of passive and active neutron measurements which yields quantitative information about the isotopic composition of transuranic wastes from nuclear power or weapons material manufacture reactors is described. From the measurement of prompt and delayed neutron emission and the incidence of two coincidentally emitted neutrons from induced fission of fissile material in the sample, one can quantify .sup.233 U, .sup.235 U and .sup.239 Pu isotopes in waste samples. Passive coincidence counting, including neutron multiplicity measurement and determination of the overall passive neutron flux additionally enables the separate quantitative evaluation of spontaneous fission isotopes such as .sup.240 Pu, .sup.244 Cm and .sup.252 Cf, and the spontaneous alpha particle emitter .sup.241 Am. These seven isotopes are the most important constituents of wastes from nuclear power reactors and once the mass of each isotope present is determined by the apparatus and method of the instant invention, the overall alpha particle activity can be determined to better than 1 nCi/g from known radioactivity data. Therefore, in addition to the quantitative analysis of the waste sample useful for later reclamation purposes, the alpha particle activity can be determined to decide whether "permanent" low-level burial is appropriate for the waste sample.

  13. Topics in nuclear power (Journal Article) | SciTech Connect

    Office of Scientific and Technical Information (OSTI)

    NUCLEAR POWER STATION; GAIN; JAPAN; NATURAL DISASTERS; NUCLEAR INDUSTRY; NUCLEAR POWER; NUCLEAR POWER PLANTS; PROBABILISTIC ESTIMATION; REACTOR ACCIDENTS; REACTOR MAINTENANCE;...

  14. A new type of Neutrino Detector for Sterile Neutrino Search at Nuclear Reactors and Nuclear Nonproliferation Applications

    E-Print Network [OSTI]

    C. Lane; S. M. Usman; J. Blackmon; C. Rasco; H. P. Mumm; D. Markoff; G. R. Jocher; R. Dorrill; M. Duvall; J. G. Learned; V. Li; J. Maricic; S. Matsuno; R. Milincic; S. Negrashov; M. Sakai; M. Rosen; G. Varner; P. Huber; M. L. Pitt; S. D. Rountree; R. B. Vogelaar; T. Wright; Z. Yokley

    2015-01-27

    We describe a new detector, called NuLat, to study electron anti-neutrinos a few meters from a nuclear reactor, and search for anomalous neutrino oscillations. Such oscillations could be caused by sterile neutrinos, and might explain the "Reactor Antineutrino Anomaly". NuLat, is made possible by a natural synergy between the miniTimeCube and mini-LENS programs described in this paper. It features a "Raghavan Optical Lattice" (ROL) consisting of 3375 boron or $^6$Li loaded plastic scintillator cubical cells 6.3\\,cm (2.500") on a side. Cell boundaries have a 0.127\\,mm (0.005") air gap, resulting in total internal reflection guiding most of the light down the 3 cardinal directions. The ROL detector technology for NuLat gives excellent spatial and energy resolution and allows for in-depth event topology studies. These features allow us to discern inverse beta decay (IBD) signals and the putative oscillation pattern, even in the presence of other backgrounds. We discuss here test venues, efficiency, sensitivity and project status.

  15. A new type of Neutrino Detector for Sterile Neutrino Search at Nuclear Reactors and Nuclear Nonproliferation Applications

    E-Print Network [OSTI]

    Lane, C; Blackmon, J; Rasco, C; Mumm, H P; Markoff, D; Jocher, G R; Dorrill, R; Duvall, M; Learned, J G; Li, V; Maricic, J; Matsuno, S; Milincic, R; Negrashov, S; Sakai, M; Rosen, M; Varner, G; Huber, P; Pitt, M L; Rountree, S D; Vogelaar, R B; Wright, T; Yokley, Z

    2015-01-01

    We describe a new detector, called NuLat, to study electron anti-neutrinos a few meters from a nuclear reactor, and search for anomalous neutrino oscillations. Such oscillations could be caused by sterile neutrinos, and might explain the "Reactor Antineutrino Anomaly". NuLat, is made possible by a natural synergy between the miniTimeCube and mini-LENS programs described in this paper. It features a "Raghavan Optical Lattice" (ROL) consisting of 3375 boron or $^6$Li loaded plastic scintillator cubical cells 6.3\\,cm (2.500") on a side. Cell boundaries have a 0.127\\,mm (0.005") air gap, resulting in total internal reflection guiding most of the light down the 3 cardinal directions. The ROL detector technology for NuLat gives excellent spatial and energy resolution and allows for in-depth event topology studies. These features allow us to discern inverse beta decay (IBD) signals and the putative oscillation pattern, even in the presence of other backgrounds. We discuss here test venues, efficiency, sensitivity an...

  16. United States Department of Energy`s reactor core protection evaluation methodology for fires at RBMK and VVER nuclear power plants. Revision 1

    SciTech Connect (OSTI)

    1997-06-01

    This document provides operators of Soviet-designed RBMK (graphite moderated light water boiling water reactor) and VVER (pressurized light water reactor) nuclear power plants with a systematic Methodology to qualitatively evaluate plant response to fires and to identify remedies to protect the reactor core from fire-initiated damage.

  17. Car Ownership and Welfare-to-Work

    E-Print Network [OSTI]

    Ong, Paul M.

    2001-01-01

    Problems Related to Child Car, Transportation, and IllnessCar Ownership and Welfare-to-Work Paul M Ong Reprint UCTC Nofor conte~ts thereof oruse Car Ownership and Welfare-to-Work

  18. Nuclear Transmutations in HFIR's Beryllium Reflector and Their Impact on Reactor Operation and Reflector Disposal

    SciTech Connect (OSTI)

    Chandler, David [ORNL; Maldonado, G Ivan [ORNL; Primm, Trent [ORNL; Proctor, Larry Duane [ORNL

    2012-01-01

    The High Flux Isotope Reactor located at the Oak Ridge National Laboratory utilizes a large cylindrical beryllium reflector that is subdivided into three concentric regions and encompasses the compact reactor core. Nuclear transmutations caused by neutron activation occur in the beryllium reflector regions, which leads to unwanted neutron absorbing and radiation emitting isotopes. During the past year, two topics related to the HFIR beryllium reflector were reviewed. The first topic included studying the neutron poison (helium-3 and lithium-6) buildup in the reflector regions and its affect on beginning-of-cycle reactivity. A new methodology was developed to predict the reactivity impact and estimated symmetrical critical control element positions as a function of outage time between cycles due to helium-3 buildup and was shown to be in better agreement with actual symmetrical critical control element position data than the current methodology. The second topic included studying the composition of the beryllium reflector regions at discharge as well as during decay to assess the viability of transporting, storing, and ultimately disposing the reflector regions currently stored in the spent fuel pool. The post-irradiation curie inventories were used to determine whether the reflector regions are discharged as transuranic waste or become transuranic waste during the decay period for disposal purposes and to determine the nuclear hazard category, which may affect the controls invoked for transportation and temporary storage. Two of the reflector regions were determined to be transuranic waste at discharge and the other region was determined to become transuranic waste in less than 2 years after being discharged due to the initial uranium content (0.0044 weight percent uranium). It was also concluded that all three of the reflector regions could be classified as nuclear hazard category 3 (potential for localized consequences only).

  19. System Upgrades at the Advanced Test Reactor Help Ensure that Nuclear Energy Research Continues at the Idaho National Laboratory

    SciTech Connect (OSTI)

    Craig Wise

    2011-12-01

    Fully operational in 1967, the Advanced Test Reactor (ATR) is a first-of-its-kind materials test reactor. Located on the Idaho National Laboratory’s desert site, this reactor remains at the forefront of nuclear science, producing extremely high neutron irradiation in a relatively short time span. The Advanced Test Reactor is also the only U.S. reactor that can replicate multiple reactor environments concurrently. The Idaho National Laboratory and the Department of Energy recently invested over 13 million dollars to replace three of ATR’s instrumentation and control systems. The new systems offer the latest software and technology advancements, ensuring the availability of the reactor for future energy research. Engineers and project managers successfully completed the four year project in March while the ATR was in a scheduled maintenance outage. “These new systems represent state-of-the-art monitoring and annunciation capabilities,” said Don Feldman, ATR Station Manager. “They are comparable to systems currently used for advanced reactor designs planned for construction in the U.S. and in operation in some foreign countries.”

  20. Nuclear Energy Governance and the Politics of Social Justice: Technology, Public Goods, and Redistribution in Russia and France

    E-Print Network [OSTI]

    Grigoriadis, Theocharis N

    2009-01-01

    waste disposal (Low) and nuclear reactor construction (Construction of Nuclear Reactors Global Sustainability (High) Nuclear Waste

  1. A REVIEW OF LIGHT-WATER REACTOR SAFETY STUDIES. VOLUME 3 OF THE FINAL REPORT ON HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL ELECTRIC GENERATION IN CALIFORNIA

    E-Print Network [OSTI]

    Nero, A.V.

    2010-01-01

    Risks In U.S. Commercial Nuclear Power Plants", U.S. NuclearCommission, "The, Safety of Nuclear Power Reactors (Light-October 1, 1976. "Nuclear Power and the Environment," a

  2. Processes and Procedures for Application of CFD to Nuclear Reactor Safety Analysis

    SciTech Connect (OSTI)

    Richard W. Johnson; Richard R. Schultz; Patrick J. Roache; Ismail B. Celik; William D. Pointer; Yassin A. Hassan

    2006-09-01

    Traditionally, nuclear reactor safety analysis has been performed using systems analysis codes such as RELAP5, which was developed at the INL. However, goals established by the Generation IV program, especially the desire to increase efficiency, has lead to an increase in operating temperatures for the reactors. This increase pushes reactor materials to operate towards their upper temperature limits relative to structural integrity. Because there will be some finite variation of the power density in the reactor core, there will be a potential for local hot spots to occur in the reactor vessel. Hence, it has become apparent that detailed analysis will be required to ensure that local ‘hot spots’ do not exceed safety limits. It is generally accepted that computational fluid dynamics (CFD) codes are intrinsically capable of simulating fluid dynamics and heat transport locally because they are based on ‘first principles.’ Indeed, CFD analysis has reached a fairly mature level of development, including the commercial level. However, CFD experts are aware that even though commercial codes are capable of simulating local fluid and thermal physics, great care must be taken in their application to avoid errors caused by such things as inappropriate grid meshing, low-order discretization schemes, lack of iterative convergence and inaccurate time-stepping. Just as important is the choice of a turbulence model for turbulent flow simulation. Turbulence models model the effects of turbulent transport of mass, momentum and energy, but are not necessarily applicable for wide ranges of flow types. Therefore, there is a well-recognized need to establish practices and procedures for the proper application of CFD to simulate flow physics accurately and establish the level of uncertainty of such computations. The present document represents contributions of CFD experts on what the basic practices, procedures and guidelines should be to aid CFD analysts to obtain accurate estimates of the flow and energy transport as applied to nuclear reactor safety. However, it is expected that these practices and procedures will require updating from time to time as research and development affect them or replace them with better procedures. The practices and procedures are categorized into five groups. These are: 1.Code Verification 2.Code and Calculation Documentation 3.Reduction of Numerical Error 4.Quantification of Numerical Uncertainty (Calculation Verification) 5.Calculation Validation. These five categories have been identified from procedures currently required of CFD simulations such as those required for publication of a paper in the ASME Journal of Fluids Engineering and from the literature such as Roache [1998]. Code verification refers to the demonstration that the equations of fluid and energy transport have been correctly coded in the CFD code. Code and calculation documentation simply means that the equations and their discretizations, etc., and boundary and initial conditions used to pose the fluid flow problem are fully described in available documentation. Reduction of numerical error refers to practices and procedures to lower numerical errors to negligible or very low levels as is reasonably possible (such as avoiding use of first-order discretizations). The quantification of numerical uncertainty is also known as calculation verification. This means that estimates are made of numerical error to allow the characterization of the numerical

  3. Heat Transfer Salts for Nuclear Reactor Systems - Chemistry Control, Corrosion Mitigation, and Modeling

    SciTech Connect (OSTI)

    Anderson, Mark; Sridharan, Kumar; Morgan, Dane; Peterson, Per; Calderoni, Pattrick; Scheele, Randall; Casekka, Andrew; McNamara, Bruce

    2015-01-22

    The concept of a molten salt reactor has existed for nearly sixty years. Previously all work was done during a large collaborative effort at Oak Ridge National Laboratory, culminating in a research reactor which operated for 15,000 hours without major error. This technical success has garnished interest in modern, high temperature, reactor schemes. Research using molten fluoride salts for nuclear applications requires a steady supply of high grade molten salts. There is no bulk supplier of research grade fluoride salts in the world, so a facility which could provide all the salt needed for testing at the University of Wisconsin had to be produced. Two salt purification devices were made for this purpose, a large scale purifier, and a small scale purifier, each designed to clean the salts from impurities and reduce their corrosion potential. As of now, the small scale has performed with flibe salt, hydrogen, and hydrogen fluoride, yielding clean salt. This salt is currently being used in corrosion testing facilities at the Massachusetts Institute of Technology and the University of Wisconsin. Working with the beryllium based salts requires extensive safety measures and health monitoring to prevent the development of acute or chronic beryllium disease, two pulmonary diseases created by an allergic reaction to beryllium in the lungs. Extensive health monitoring, engineering controls, and environment monitoring had to be set up with the University of Wisconsin department of Environment, Health and Safety. The hydrogen fluoride required for purification was also an extreme health hazard requiring thoughtful planning and execution. These dangers have made research a slow and tedious process. Simple processes, such as chemical handling and clean-up, can take large amounts of ingenuity and time. Other work has complemented the experimental research at Wisconsin to advance high temperature reactor goals. Modeling work has been performed in house to re-evaluate thermophysical properties of flibe and flinak. Pacific Northwest National Laboratories has focused on evaluating the fluorinating gas nitrogen trifluoride as a potential salt purification agent. Work there was performed on removing hydroxides and oxides from flinak salt under controlled conditions. Lastly, the University of California Berkeley has spent considerable time designing and simulating reactor components with fluoride salts at high temperatures. Despite the hurdles presented by the innate chemical hazards, considerable progress has been made. The stage has been set to perform new research on salt chemical control which could advance the fluoride salt cooled reactor concept towards commercialization. What were previously thought of as chemical undesirable, but nuclear certified, alloys have been shown to be theoretically compatible with fluoride salts at high temperatures. This preliminary report has been prepared to communicate the construction of the basic infrastructure required for flibe, as well as suggest original research to performed at the University of Wisconsin. Simultaneously, the contents of this report can serve as a detailed, but introductory guide to allow anyone to learn the fundamentals of chemistry, engineering, and safety required to work with flibe salt.

  4. A 48-month extended fuel cycle for the B and W mPower{sup TM} small modular nuclear reactor

    SciTech Connect (OSTI)

    Erighin, M. A. [Babcock and Wilcox Company, 109 Ramsey Place, Lynchburg, VA 24502 (United States)

    2012-07-01

    The B and W mPower{sup TM} reactor is a small, rail-shippable pressurized water reactor (PWR) with an integral once-through steam generator and an electric power output of 150 MW, which is intended to replace aging fossil power plants of similar output. The core is composed of 69 reduced-height, but otherwise standard, PWR assemblies with the familiar 17 x 17 fuel rod array on a 21.5 cm inter-assembly pitch. The B and W mPower core design and cycle management plan, which were performed using the Studsvik core design code suite, follow the pattern of a typical nuclear reactor fuel cycle design and analysis performed by most nuclear fuel management organizations, such as fuel vendors and utilities. However, B and W is offering a core loading and cycle management plan for four years of continuous power operations without refueling and without the hurdles of chemical shim. (authors)

  5. Strengthening the fission reactor nuclear science and engineering program at UCLA. Final technical report

    SciTech Connect (OSTI)

    Okrent, D.

    1997-06-23

    This is the final report on DOE Award No. DE-FG03-92ER75838 A000, a three year matching grant program with Pacific Gas and Electric Company (PG and E) to support strengthening of the fission reactor nuclear science and engineering program at UCLA. The program began on September 30, 1992. The program has enabled UCLA to use its strong existing background to train students in technological problems which simultaneously are of interest to the industry and of specific interest to PG and E. The program included undergraduate scholarships, graduate traineeships and distinguished lecturers. Four topics were selected for research the first year, with the benefit of active collaboration with personnel from PG and E. These topics remained the same during the second year of this program. During the third year, two topics ended with the departure o the students involved (reflux cooling in a PWR during a shutdown and erosion/corrosion of carbon steel piping). Two new topics (long-term risk and fuel relocation within the reactor vessel) were added; hence, the topics during the third year award were the following: reflux condensation and the effect of non-condensable gases; erosion/corrosion of carbon steel piping; use of artificial intelligence in severe accident diagnosis for PWRs (diagnosis of plant status during a PWR station blackout scenario); the influence on risk of organization and management quality; considerations of long term risk from the disposal of hazardous wastes; and a probabilistic treatment of fuel motion and fuel relocation within the reactor vessel during a severe core damage accident.

  6. Application of adaptive hierarchical sparse grid collocation to the uncertainty quantification of nuclear reactor simulators

    SciTech Connect (OSTI)

    Yankov, A.; Downar, T. [University of Michigan, 2355 Bonisteel Blvd, Ann Arbor, MI 48109 (United States)

    2013-07-01

    Recent efforts in the application of uncertainty quantification to nuclear systems have utilized methods based on generalized perturbation theory and stochastic sampling. While these methods have proven to be effective they both have major drawbacks that may impede further progress. A relatively new approach based on spectral elements for uncertainty quantification is applied in this paper to several problems in reactor simulation. Spectral methods based on collocation attempt to couple the approximation free nature of stochastic sampling methods with the determinism of generalized perturbation theory. The specific spectral method used in this paper employs both the Smolyak algorithm and adaptivity by using Newton-Cotes collocation points along with linear hat basis functions. Using this approach, a surrogate model for the outputs of a computer code is constructed hierarchically by adaptively refining the collocation grid until the interpolant is converged to a user-defined threshold. The method inherently fits into the framework of parallel computing and allows for the extraction of meaningful statistics and data that are not within reach of stochastic sampling and generalized perturbation theory. This paper aims to demonstrate the advantages of spectral methods-especially when compared to current methods used in reactor physics for uncertainty quantification-and to illustrate their full potential. (authors)

  7. Large-eddy simulations of turbulent flow for grid-to-rod fretting in nuclear reactors

    E-Print Network [OSTI]

    Bakosi, J; Lowrie, R B; Pritchett-Sheats, L A; Nourgaliev, R R

    2013-01-01

    The grid-to-rod fretting (GTRF) problem in pressurized water reactors is a flow-induced vibration problem that results in wear and failure of the fuel rods in nuclear assemblies. In order to understand the fluid dynamics of GTRF and to build an archival database of turbulence statistics for various configurations, implicit large-eddy simulations of time-dependent single-phase turbulent flow have been performed in 3x3 and 5x5 rod bundles with a single grid spacer. To assess the computational mesh and resolution requirements, a method for quantitative assessment of unstructured meshes with no-slip walls is described. The calculations have been carried out using Hydra-TH, a thermal-hydraulics code developed at Los Alamos for the Consortium for Advanced Simulation of Light water reactors, a United States Department of Energy Innovation Hub. Hydra-TH uses a second-order implicit incremental projection method to solve the single-phase incompressible Navier-Stokes equations. The simulations explicitly resolve the la...

  8. THR-TH: a high-temperature gas-cooled nuclear reactor core thermal hydraulics code

    SciTech Connect (OSTI)

    Vondy, D.R.

    1984-07-01

    The ORNL version of PEBBLE, the (RZ) pebble bed thermal hydraulics code, has been extended for application to a prismatic gas cooled reactor core. The supplemental treatment is of one-dimensional coolant flow in up to a three-dimensional core description. Power density data from a neutronics and exposure calculation are used as the basic information for the thermal hydraulics calculation of heat removal. Two-dimensional neutronics results may be expanded for a three-dimensional hydraulics calculation. The geometric description for the hydraulics problem is the same as used by the neutronics code. A two-dimensional thermal cell model is used to predict temperatures in the fuel channel. The capability is available in the local BOLD VENTURE computation system for reactor core analysis with capability to account for the effect of temperature feedback by nuclear cross section correlation. Some enhancements have also been added to the original code to add pebble bed modeling flexibility and to generate useful auxiliary results. For example, an estimate is made of the distribution of fuel temperatures based on average and extreme conditions regularly calculated at a number of locations.

  9. White paper report on using nuclear reactors to search for a value of theta13

    SciTech Connect (OSTI)

    Anderson, K.; Anjos, J.C.; Ayres, D.; Beacom, J.; Bediaga, I.; de Bellefon, A.; Berger, B.E.; Bilenky, S.; Blucher, E.; Bolton, T.; Buck, C.; Bugg, W.; Busenitz, J.; Choubey, S.; Conrad, J.; Cribier, M.; Dadoun, O.; Dalnoki-Veress, F.; Decowski, M.; de Gouvea, Andre; Demutrh, D.; Dessages-Ardellier, F.; Efremenko, Y.; von Feilitzsch, F.; Finley, D.; Formaggio, J.A.; Freedman, S.J.; Fujikawa, B.K.; Garbini, M.; Giusti, P.; Goger-Neff, M.; Goodman, M.; Gray, F.; Grieb, C.; Grudzinski, J.J.; Guarino, V.J.; Hartmann, F.; Hagner, C.; Heeger, K.M.; Hofmann, W.; Horton-Smith, G.; Huber, P.; Inzhechik, L.; Jochum, J.; Jostlein, H.; Kadel, R.; Kamyshkov, Y.; Kaplan, D.; Kasper, P.; de Kerret, H.; Kersten, J.; Klein, J.; Knopfle, K.T.; Kopeikin, V.; Kozlov, Yu.; Kryn, D.; Kuchler, V.; Kuze, M.; Lachenmaier, T.; Lasserre, T.; Laughton, C.; Lendvai, C.; Li, J.; Lindner, M.; Link, J.; Longo, M.; Lu, Y.S.; Luk, K.B.; Ma, Y.Q.; Martemyanov, V.P.; Mauger, C.; Manghetti, H.; McKeown, R.; Mention, G.; Meyer, J.P.; Mikaelyan, L.; Minakata, H.; Naples, D.; Nunokawa, H.; Oberauer, L.; Obolensky, M.; Parke, S.; Petcov, S.T.; Peres, O.L.G.; Potzel, W.; Pilcher, J.; Plunkett, R.; Raffelt, G.; Rapidis, P.; Reyna, D.; Roe, B.; Rolinec, M.; Sakamoto, Y.; Sartorelli, G.; Schonert, S.; Schwertz, T.; Selvi, M.; Shaevitz, M.; Shellard, R.; Shrock, R.; Sidwell, R.; Sims, J.; Sinev, V.; Stanton, N.; Stancu, I.; Stefanski, R.; Seukane, F.; Sugiyama, H.; Sukhotin, S.; Sumiyoshi, T.; Svoboda, R.; Talaga, R.; Tamura, N.; Tanimoto, M.; Thron, J.; von Toerne, E.; Vignaud, D.; Wagner, C.; Wang, Y.F.; Wang, Z.; Winter, W.; Wong, H.; Yakushev, E.; Yang, C.G.; Yasuda, O.

    2004-02-26

    There has been superb progress in understanding the neutrino sector of elementary particle physics in the past few years. It is now widely recognized that the possibility exists for a rich program of measuring CP violation and matter effects in future accelerator {nu} experiments, which has led to intense efforts to consider new programs at neutrino superbeams, off-axis detectors, neutrino factories and beta beams. However, the possibility of measuring CP violation can be fulfilled only if the value of the neutrino mixing parameter {theta}{sub 13} is such that sin{sup 2} (2{theta}{sub 13}) greater than or equal to on the order of 0.01. The authors of this white paper are an International Working Group of physicists who believe that a timely new experiment at a nuclear reactor sensitive to the neutrino mixing parameter {theta}{sub 13} in this range has a great opportunity for an exciting discovery, a non-zero value to {theta}{sub 13}. This would be a compelling next step of this program. We are studying possible new reactor experiments at a variety of sites around the world, and we have collaborated to prepare this document to advocate this idea and describe some of the issues that are involved.

  10. International Conference on Mathematics, Computational Methods & Reactor Physics (M&C 2009) Saratoga Springs, New York, May 3-7, 2009, on CD-ROM, American Nuclear Society, LaGrange Park, IL (2009)

    E-Print Network [OSTI]

    Vialle, Stéphane

    2009-01-01

    operator such as EDF, the time required to compute nuclear reactor core simulations is rather critical. Introduction As operator of nuclear power plants, EDF needs many nuclear reactor core simulationsInternational Conference on Mathematics, Computational Methods & Reactor Physics (M&C 2009

  11. VOLUME 84, NUMBER 17 P HY S I CA L R E V I E W L E T T E R S 24 APRIL 2000 Search for Neutrino Oscillations at the Palo Verde Nuclear Reactors

    E-Print Network [OSTI]

    Gratta, Giorgio

    Oscillations at the Palo Verde Nuclear Reactors F. Boehm, 3 J. Busenitz, 1 B. Cook, 3 G. Gratta, 4 H. Henrikson at a distance of about 800 m from the three reactors of the Palo Verde Nuclear Generating Station using.15.+g, 14.60.Lm, 25.30.Pt Nuclear reactors have been used as intense sources of â?¢ n e in experiments

  12. VOLUME 84, NUMBER 17 P H Y S I C A L R E V I E W L E T T E R S 24 APRIL 2000 Search for Neutrino Oscillations at the Palo Verde Nuclear Reactors

    E-Print Network [OSTI]

    Gratta, Giorgio

    Oscillations at the Palo Verde Nuclear Reactors F. Boehm,3 J. Busenitz,1 B. Cook,3 G. Gratta,4 H. Henrikson,3 J the three reactors of the Palo Verde Nuclear Generating Station using a segmented gadolinium.30.Pt Nuclear reactors have been used as intense sources of ¯ne in experiments searching for neutrino

  13. 1 hour, 25 minutes ago Japan will ask the European Union to declare it a "joint host" of a revolutionary nuclear energy project even if the reactor is located in

    E-Print Network [OSTI]

    " of a revolutionary nuclear energy project even if the reactor is located in France, a newspaper said. The Nihon to 'joint host' revolutionary nuclear reactor 6/6/05 8:23 AMPrint Story: Japan to ask EU to 'joint host' revolutionary nuclear reactor on Yahoo! News Page 1 of 1http://news.yahoo.com/s/afp/20050606/sc

  14. Print Story: Japan has told EU it is giving up on nuclear reactor bid:... http://news.yahoo.com/s/afp/20050622/wl_asia_afp/japaneufrancenuc... 1 of 1 6/22/05 12:12 AM

    E-Print Network [OSTI]

    Print Story: Japan has told EU it is giving up on nuclear reactor bid:... http it is giving up on nuclear reactor bid: report 1 hour, 51 minutes ago Japan has told the European Union it will give up on hosting a revolutionary nuclear energy reactor after securing large deals in turn

  15. The United States Department of Energy (DOE) has always held the safety and reliability of the nation's nuclear reactor fleet as a top priority. Continual improvements and advancements in nuclear fuels have

    E-Print Network [OSTI]

    of the nation's nuclear reactor fleet as a top priority. Continual improvements and advancements in nuclear fuels have been instrumental in maximizing energy generation from nuclear power plants and minimizing the mechanical properties of uranium dioxide (UO2) for nuclear fuel applications. In an effort to improve

  16. High-Temperature Nuclear Reactors for In-Situ Recovery of Oil from Oil Shale

    SciTech Connect (OSTI)

    Forsberg, Charles W.

    2006-07-01

    The world is exhausting its supply of crude oil for the production of liquid fuels (gasoline, jet fuel, and diesel). However, the United States has sufficient oil shale deposits to meet our current oil demands for {approx}100 years. Shell Oil Corporation is developing a new potentially cost-effective in-situ process for oil recovery that involves drilling wells into oil shale, using electric heaters to raise the bulk temperature of the oil shale deposit to {approx}370 deg C to initiate chemical reactions that produce light crude oil, and then pumping the oil to the surface. The primary production cost is the cost of high-temperature electrical heating. Because of the low thermal conductivity of oil shale, high-temperature heat is required at the heater wells to obtain the required medium temperatures in the bulk oil shale within an economically practical two to three years. It is proposed to use high-temperature nuclear reactors to provide high-temperature heat to replace the electricity and avoid the factor-of-2 loss in converting high-temperature heat to electricity that is then used to heat oil shale. Nuclear heat is potentially viable because many oil shale deposits are thick (200 to 700 m) and can yield up to 2.5 million barrels of oil per acre, or about 125 million dollars/acre of oil at $50/barrel. The concentrated characteristics of oil-shale deposits make it practical to transfer high-temperature heat over limited distances from a reactor to the oil shale deposits. (author)

  17. Large-eddy simulations of turbulent flow for grid-to-rod fretting in nuclear reactors

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Bakosi, J.; Christon, M. A.; Lowrie, R. B.; Pritchett-Sheats, L. A.; Nourgaliev, R. R.

    2013-07-12

    The grid-to-rod fretting (GTRF) problem in pressurized water reactors is a flow-induced vibration problem that results in wear and failure of the fuel rods in nuclear assemblies. In order to understand the fluid dynamics of GTRF and to build an archival database of turbulence statistics for various configurations, implicit large-eddy simulations of time-dependent single-phase turbulent flow have been performed in 3 × 3 and 5 × 5 rod bundles with a single grid spacer. To assess the computational mesh and resolution requirements, a method for quantitative assessment of unstructured meshes with no-slip walls is described. The calculations have been carriedmore »out using Hydra-TH, a thermal-hydraulics code developed at Los Alamos for the Consortium for Advanced Simulation of Light water reactors, a United States Department of Energy Innovation Hub. Hydra-TH uses a second-order implicit incremental projection method to solve the singlephase incompressible Navier-Stokes equations. The simulations explicitly resolve the large scale motions of the turbulent flow field using first principles and rely on a monotonicity-preserving numerical technique to represent the unresolved scales. Each series of simulations for the 3 × 3 and 5 × 5 rod-bundle geometries is an analysis of the flow field statistics combined with a mesh-refinement study and validation with available experimental data. Our primary focus is the time history and statistics of the forces loading the fuel rods. These hydrodynamic forces are believed to be the key player resulting in rod vibration and GTRF wear, one of the leading causes for leaking nuclear fuel which costs power utilities millions of dollars in preventive measures. As a result, we demonstrate that implicit large-eddy simulation of rod-bundle flows is a viable way to calculate the excitation forces for the GTRF problem.« less

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

    DOE Patents [OSTI]

    Travelli, Armando (Hinsdale, IL)

    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.

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

    DOE Patents [OSTI]

    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.

  20. Threshold self-powered gamma detector for use as a monitor of power in a nuclear reactor

    DOE Patents [OSTI]

    LeVert, Francis E. (Chicago, IL); Cox, Samson A. (Darien, IL)

    1978-01-01

    A self-powered gamma monitor for placement near the core of a nuclear reactor comprises a lead prism surrounded by a coaxial thin nickel sheet, the combination forming a collector. A coaxial polyethylene electron barrier encloses the collector and is separated from the nickel sheet by a vacuum region. The electron barrier is enclosed by a coaxial stainless steel emitter which, in turn, is enclosed within a lead casing. When the detector is placed in a flux of gamma rays, a measure of the current flow in an external circuit between emitter and collector provides a measure of the power level of the reactor.

  1. Combined cooling and purification system for nuclear reactor spent fuel pit, refueling cavity, and refueling water storage tank

    DOE Patents [OSTI]

    Corletti, Michael M. (New Kensington, PA); Lau, Louis K. (Monroeville, PA); Schulz, Terry L. (Murrysville Boro, PA)

    1993-01-01

    The spent fuel pit of a pressured water reactor (PWR) nuclear power plant has sufficient coolant capacity that a safety rated cooling system is not required. A non-safety rated combined cooling and purification system with redundant branches selectively provides simultaneously cooling and purification for the spent fuel pit, the refueling cavity, and the refueling water storage tank, and transfers coolant from the refueling water storage tank to the refueling cavity without it passing through the reactor core. Skimmers on the suction piping of the combined cooling and purification system eliminate the need for separate skimmer circuits with dedicated pumps.

  2. Computational Nuclear Forensics Analysis of Weapons-grade Plutonium Separated from Fuel Irradiated in a Thermal Reactor 

    E-Print Network [OSTI]

    Coles, Taylor Marie

    2014-04-27

    of bundle 16.4 kg Weight of U per bundle 13.4 kg Active Core Length 500 cm II.C. Monte Carlo N-Particle Code For the thesis work presented here, a code called Monte Carlo N-Particle (MCNP) was used. The manual describes MCNP as "a general... Working Group LWR- Light Water Reactor MCNP- Monte Carlo Neutral Particle MOX- Mixed Oxide NPT- Non-Proliferation Treaty NSG- Nuclear Suppliers Group ORNL- Oak Ridge National Laboratory PHWR- Pressurized Heavy Water Reactor PUREX- Plutonium Uranium...

  3. Combined cooling and purification system for nuclear reactor spent fuel pit, refueling cavity, and refueling water storage tank

    DOE Patents [OSTI]

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

    1993-12-14

    The spent fuel pit of a pressured water reactor (PWR) nuclear power plant has sufficient coolant capacity that a safety rated cooling system is not required. A non-safety rated combined cooling and purification system with redundant branches selectively provides simultaneously cooling and purification for the spent fuel pit, the refueling cavity, and the refueling water storage tank, and transfers coolant from the refueling water storage tank to the refueling cavity without it passing through the reactor core. Skimmers on the suction piping of the combined cooling and purification system eliminate the need for separate skimmer circuits with dedicated pumps. 1 figures.

  4. Long-term storage facility for reactor compartments in Sayda Bay - German support for utilization of nuclear submarines in Russia

    SciTech Connect (OSTI)

    Wolff, Dietmar; Voelzke, Holger; Weber, Wolfgang; Noack, Volker; Baeuerle, Guenther

    2007-07-01

    The German-Russian project that is part of the G8 initiative on Global Partnership Against the Spread of Weapons and Materials of Mass Destruction focuses on the speedy construction of a land-based interim storage facility for nuclear submarine reactor compartments at Sayda Bay near Murmansk. This project includes the required infrastructure facilities for long-term storage of about 150 reactor compartments for a period of about 70 years. The interim storage facility is a precondition for effective activities of decommissioning and dismantlement of almost all nuclear-powered submarines of the Russian Northern Fleet. The project also includes the establishment of a computer-assisted waste monitoring system. In addition, the project involves clearing Sayda Bay of other shipwrecks of the Russian navy. On the German side the project is carried out by the Energiewerke Nord GmbH (EWN) on behalf of the Federal Ministry of Economics and Labour (BMWi). On the Russian side the Kurchatov Institute holds the project management of the long-term interim storage facility in Sayda Bay, whilst the Nerpa Shipyard, which is about 25 km away from the storage facility, is dismantling the submarines and preparing the reactor compartments for long-term interim storage. The technical monitoring of the German part of this project, being implemented by BMWi, is the responsibility of the Federal Institute for Materials Research and Testing (BAM). This paper gives an overview of the German-Russian project and a brief description of solutions for nuclear submarine disposal in other countries. At Nerpa shipyard, being refurbished with logistic and technical support from Germany, the reactor compartments are sealed by welding, provided with biological shielding, subjected to surface treatment and conservation measures. Using floating docks, a tugboat tows the reactor compartments from Nerpa shipyard to the interim storage facility at Sayda Bay where they will be left on the on-shore concrete storage space to allow the radioactivity to decay. For transport of reactor compartments at the shipyard, at the dock and at the storage facility, hydraulic keel blocks, developed and supplied by German subcontractors, are used. In July 2006 the first stage of the reactor compartment storage facility was commissioned and the first seven reactor compartments have been delivered from Nerpa shipyard. Following transports of reactor compartments to the storage facility are expected in 2007. (authors)

  5. Small Fast Spectrum Reactor Designs Suitable for Direct Nuclear Thermal Propulsion

    SciTech Connect (OSTI)

    Bruce G. Schnitzler; Stanley K. Borowski

    2012-07-01

    Advancement of U.S. scientific, security, and economic interests through a robust space exploration program requires high performance propulsion systems to support a variety of robotic and crewed missions beyond low Earth orbit. Past studies, in particular those in support of both the Strategic Defense Initiative (SDI) and Space Exploration Initiative (SEI), have shown nuclear thermal propulsion systems provide superior performance for high mass high propulsive delta-V missions. The recent NASA Design Reference Architecture (DRA) 5.0 Study re-examined mission, payload, and transportation system requirements for a human Mars landing mission in the post-2030 timeframe. Nuclear thermal propulsion was again identified as the preferred in-space transportation system. A common nuclear thermal propulsion stage with three 25,000-lbf thrust engines was used for all primary mission maneuvers. Moderately lower thrust engines may also have important roles. In particular, lower thrust engine designs demonstrating the critical technologies that are directly extensible to other thrust levels are attractive from a ground testing perspective. An extensive nuclear thermal rocket technology development effort was conducted from 1955-1973 under the Rover/NERVA Program. Both graphite and refractory metal alloy fuel types were pursued. Reactors and engines employing graphite based fuels were designed, built and ground tested. A number of fast spectrum reactor and engine designs employing refractory metal alloy fuel types were proposed and designed, but none were built. The Small Nuclear Rocket Engine (SNRE) was the last engine design studied by the Los Alamos National Laboratory during the program. At the time, this engine was a state-of-the-art graphite based fuel design incorporating lessons learned from the very successful technology development program. The SNRE was a nominal 16,000-lbf thrust engine originally intended for unmanned applications with relatively short engine operations and the engine and stage design were constrained to fit within the payload volume of the then planned space shuttle. The SNRE core design utilized hexagonal fuel elements and hexagonal structural support elements. The total number of elements can be varied to achieve engine designs of higher or lower thrust levels. Some variation in the ratio of fuel elements to structural elements is also possible. Options for SNRE-based engine designs in the 25,000-lbf thrust range were described in a recent (2010) Joint Propulsion Conference paper. The reported designs met or exceeded the performance characteristics baselined in the DRA 5.0 Study. Lower thrust SNRE-based designs were also described in a recent (2011) Joint Propulsion Conference paper. Recent activities have included parallel evaluation and design efforts on fast spectrum engines employing refractory metal alloy fuels. These efforts include evaluation of both heritage designs from the Argonne National Laboratory (ANL) and General Electric Company GE-710 Programs as well as more recent designs. Results are presented for a number of not-yet optimized fast spectrum engine options.

  6. Design and Transient Analysis of Passive Safety Cooling Systems for Advanced Nuclear Reactors

    E-Print Network [OSTI]

    Galvez, Cristhian

    2011-01-01

    heat transfer. Nuclear Technology 163 (2008) 344- [18] V.and Electricity . Nuclear Technology 144 [5] D. F. Williams,Vessel Technology . Nuclear Technology, 78 (1987) 245- [5

  7. The Future of Nuclear Energy: Facts and Fiction Chapter IV: Energy from Breeder Reactors and from Fusion?

    E-Print Network [OSTI]

    Dittmar, Michael

    2009-01-01

    The accumulated knowledge and the prospects for commercial energy production from fission breeder and fusion reactors are analyzed in this report. The publicly available data from past experimental breeder reactors indicate that a large number of unsolved technological problems exist and that the amount of "created" fissile material, either from the U238 --> Pu239 or from the Th232 --> U233 cycle, is still far below the breeder requirements and optimistic theoretical expectations. Thus huge efforts, including many basic research questions with an uncertain outcome, are needed before a large commercial breeder prototype can be designed. Even if such efforts are undertaken by the technologically most advanced countries, it will take several decades before such a prototype can be constructed. We conclude therefore, that ideas about near-future commercial fission breeder reactors are nothing but wishful thinking. We further conclude that, no matter how far into the future we may look, nuclear fusion as an energy ...

  8. Estimation, Analysis and Smoothing of Self-Similar Network Induced Delays in Feedback Control of Nuclear Reactors

    E-Print Network [OSTI]

    Majumder, Basudev; Pan, Indranil; Saha, Sayan; Das, Shantanu; Gupta, Amitava; 10.1109/PACC.2011.5978960

    2012-01-01

    This paper analyzes a nuclear reactor power signal that suffers from network induced random delays in the shared data network while being fed-back to the Reactor Regulating System (RRS). A detailed study is carried out to investigate the self similarity of random delay dynamics due to the network traffic in shared medium. The fractionality or selfsimilarity in the network induced delay that corrupts the measured power signal coming from Self Powered Neutron Detectors (SPND) is estimated and analyzed. As any fractional order randomness is intrinsically different from conventional Gaussian kind of randomness, these delay dynamics need to be handled efficiently, before reaching the controller within the RRS. An attempt has been made to minimize the effect of the randomness in the reactor power transient data with few classes of smoothing filters. The performance measure of the smoothers with fractional order noise consideration is also investigated into.

  9. Pressurized water reactor in-core nuclear fuel management by tabu search

    E-Print Network [OSTI]

    Hill, Natasha J.; Parks, Geoffrey T.

    2014-08-24

    Optimization of the arrangement of fuel assemblies and burnable poisons when reloading pressurized water reactors has, in the past, been performed with many di erent algorithms in an attempt to make reactors more economic and fuel effi cient...

  10. Design and Transient Analysis of Passive Safety Cooling Systems for Advanced Nuclear Reactors

    E-Print Network [OSTI]

    Galvez, Cristhian

    2011-01-01

    high temperature reactors (VHTR), 3) supercritical waterup to 760 ° C. The DOE VHTR program is also sponsoring a

  11. Comparative analysis of compact heat exchangers for application as the intermediate heat exchanger for advanced nuclear reactors

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Bartel, N.; Chen, M.; Utgikar, V. P.; Sun, X.; Kim, I. -H.; Christensen, R.; Sabharwall, P.

    2015-04-04

    A comparative evaluation of alternative compact heat exchanger designs for use as the intermediate heat exchanger in advanced nuclear reactor systems is presented in this article. Candidate heat exchangers investigated included the Printed circuit heat exchanger (PCHE) and offset strip-fin heat exchanger (OSFHE). Both these heat exchangers offer high surface area to volume ratio (a measure of compactness [m2/m3]), high thermal effectiveness, and overall low pressure drop. Helium–helium heat exchanger designs for different heat exchanger types were developed for a 600 MW thermal advanced nuclear reactor. The wavy channel PCHE with a 15° pitch angle was found to offer optimummore »combination of heat transfer coefficient, compactness and pressure drop as compared to other alternatives. The principles of the comparative analysis presented here will be useful for heat exchanger evaluations in other applications as well.« less

  12. Inverse Beta Decay in a Nonequilibrium Antineutrino Flux from a Nuclear Reactor

    E-Print Network [OSTI]

    V. I. Kopeikin; L. A. Mikaelyan; V. V. Sinev

    2001-10-23

    The evolution of the reactor antineutrino spectrum toward equilibrium above the inverse beta-decay threshold during the reactor operating period and the decay of residual antineutrino radiation after reactor shutdown are considered. It is found that, under certain conditions, these processes can play a significant role in experiments seeking neutrino oscillations.

  13. Programme A. Nuclear Power Subprogramme A.4 Technology Development for Advanced Reactor Lines

    E-Print Network [OSTI]

    De Cindio, Fiorella

    , and decommissioning of experimental and power sodium cooled fast reactors. The objectives of the CRP are to: (1 and power sodium cooled fast reactors; (2) enable easy access to the information from this feedback; (3 of experimental and power sodium cooled fast reactors. The output from the CRP will be one or more IAEA Technical

  14. Preliminary results of calculations for heavy-water nuclear-power-plant reactors employing {sup 235}U, {sup 233}U, and {sup 232}Th as a fuel and meeting requirements of a nonproliferation of nuclear weapons

    SciTech Connect (OSTI)

    Ioffe, B. L.; Kochurov, B. P. [Institute of Theoretical and Experimental Physics (Russian Federation)

    2012-02-15

    A physical design is developed for a gas-cooled heavy-water nuclear reactor intended for a project of a nuclear power plant. As a fuel, the reactor would employ thorium with a small admixture of enriched uranium that contains not more than 20% of {sup 235}U. It operates in the open-cycle mode involving {sup 233}U production from thorium and its subsequent burnup. The reactor meets the conditions of a nonproliferation of nuclear weapons: the content of fissionable isotopes in uranium at all stages of the process, including the final one, is below the threshold for constructing an atomic bomb, the amount of product plutonium being extremely small.

  15. The Programmable Logic Controller and its application in nuclear reactor systems

    SciTech Connect (OSTI)

    Palomar, J.; Wyman, R. [Lawrence Livermore National Lab., CA (United States)

    1993-09-01

    This document provides recommendations to guide reviewers in the application of Programmable Logic Controllers (PLCS) to the control, monitoring and protection of nuclear reactors. The first topics addressed are system-level design issues, specifically including safety. The document then discusses concerns about the PLC manufacturing organization and the protection system engineering organization. Supplementing this document are two appendices. Appendix A summarizes PLC characteristics. Specifically addressed are those characteristics that make the PLC more suitable for emergency shutdown systems than other electrical/electronic-based systems, as well as characteristics that improve reliability of a system. Also covered are PLC characteristics that may create an unsafe operating environment. Appendix B provides an overview of the use of programmable logic controllers in emergency shutdown systems. The intent is to familiarize the reader with the design, development, test, and maintenance phases of applying a PLC to an ESD system. Each phase is described in detail and information pertinent to the application of a PLC is pointed out.

  16. Backup Power Cost of Ownership Analysis and Incumbent Technology...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    of ownership analysis identifies the factors impacting the value proposition for fuel cell backup power and presents the estimated annualized cost of ownership for fuel cell...

  17. Applying the Battery Ownership Model in Pursuit of Optimal Battery...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Applying the Battery Ownership Model in Pursuit of Optimal Battery Use Strategies Applying the Battery Ownership Model in Pursuit of Optimal Battery Use Strategies 2012 DOE...

  18. Application of the LBB concept to nuclear power plants with WWER 440 and WWER 1000 reactors

    SciTech Connect (OSTI)

    Zdarek, J.; Pecinka, L. [Nuclear Research Institute Rez (Czech Republic)

    1997-04-01

    Leak-before-break (LBB) analysis of WWER type reactors in the Czech and Sloval Republics is summarized in this paper. Legislative bases, required procedures, and validation and verification of procedures are discussed. A list of significant issues identified during the application of LBB analysis is presented. The results of statistical evaluation of crack length characteristics are presented and compared for the WWER 440 Type 230 and 213 reactors and for the WWER 1000 Type 302, 320 and 338 reactors.

  19. Monitoring the Thermal Power of Nuclear Reactors with a Prototype Cubic Meter Antineutrino Detector

    E-Print Network [OSTI]

    A. Bernstein; N. S. Bowden; A. Misner; T. Palmer

    2008-04-30

    In this paper, we estimate how quickly and how precisely a reactor's operational status and thermal power can be monitored over hour to month time scales, using the antineutrino rate as measured by a cubic meter scale detector. Our results are obtained from a detector we have deployed and operated at 25 meter standoff from a reactor core. This prototype can detect a prompt reactor shutdown within five hours, and monitor relative thermal power to three percent within seven days. Monitoring of short-term power changes in this way may be useful in the context of International Atomic Energy Agency's (IAEA) Reactor Safeguards Regime, or other cooperative monitoring regimes.

  20. Monitoring the Thermal Power of Nuclear Reactors with a Prototype Cubic Meter Antineutrino Detector

    E-Print Network [OSTI]

    Bernstein, A; Misner, A; Palmer, T

    2008-01-01

    In this paper, we estimate how quickly and how precisely a reactor's operational status and thermal power can be monitored over hour to month time scales, using the antineutrino rate as measured by a cubic meter scale detector. Our results are obtained from a detector we have deployed and operated at 25 meter standoff from a reactor core. This prototype can detect a prompt reactor shutdown within five hours, and monitor relative thermal power to three percent within seven days. Monitoring of short-term power changes in this way may be useful in the context of International Atomic Energy Agency's (IAEA) Reactor Safeguards Regime, or other cooperative monitoring regimes.

  1. Knowledge and abilities catalog for nuclear power plant operators: Pressurized water reactors. Revision 1

    SciTech Connect (OSTI)

    1995-08-01

    This document provides the basis for the development of content-valid licensing examinations for reactor operators and senior reactor operators. The examinations developed using the PWR catalog will cover those topics listed under Title 10, (ode of Federal Regulations Part 55. The PWR catalog contains approximately 5100 knowledge and ability (K/A) statements for reactor operators and senior reactor operators. The catalog is organized into six major sections: Catalog Organization; Generic Knowledge and Abilities; Plant Systems; Emergency and Abnormal Plant Evolutions; Components and Theory.

  2. Nuclear Waste: Forever Contaminated?

    E-Print Network [OSTI]

    Wang, Andrew

    2015-01-01

    Went Wrong in Japan’s Nuclear Reactors. Retrieved March 28,went-wrong-in-japans-nuclear-reactors World Statistics. (nuclear disaster since Chernobyl. Chernobyl happened on April 26, 1986, when a reactor

  3. System and method for the analysis of one or more compounds and/or species produced by a solution-based nuclear reactor

    DOE Patents [OSTI]

    Policke, Timothy A; Nygaard, Eric T

    2014-05-06

    The present invention relates generally to both a system and method for determining the composition of an off-gas from a solution nuclear reactor (e.g., an Aqueous Homogeneous Reactor (AHR)) and the composition of the fissioning solution from those measurements. In one embodiment, the present invention utilizes at least one quadrupole mass spectrometer (QMS) in a system and/or method designed to determine at least one or more of: (i) the rate of production of at least one gas and/or gas species from a nuclear reactor; (ii) the effect on pH by one or more nitrogen species; (iii) the rate of production of one or more fission gases; and/or (iv) the effect on pH of at least one gas and/or gas species other than one or more nitrogen species from a nuclear reactor.

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

    SciTech Connect (OSTI)

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

    1994-10-01

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

  5. Use of Sensitivity and Uncertainty Analysis in the Design of Reactor Physics and Criticality Benchmark Experiments for Advanced Nuclear Fuel

    SciTech Connect (OSTI)

    Rearden, B.T. [Oak Ridge National Laboratory (United States); Anderson, W.J. [Framatome ANP, Inc. (France); Harms, G.A. [Sandia National Laboratories (United States)

    2005-08-15

    Framatome ANP, Sandia National Laboratories (SNL), Oak Ridge National Laboratory (ORNL), and the University of Florida are cooperating on the U.S. Department of Energy Nuclear Energy Research Initiative (NERI) project 2001-0124 to design, assemble, execute, analyze, and document a series of critical experiments to validate reactor physics and criticality safety codes for the analysis of commercial power reactor fuels consisting of UO{sub 2} with {sup 235}U enrichments {>=}5 wt%. The experiments will be conducted at the SNL Pulsed Reactor Facility.Framatome ANP and SNL produced two series of conceptual experiment designs based on typical parameters, such as fuel-to-moderator ratios, that meet the programmatic requirements of this project within the given restraints on available materials and facilities. ORNL used the Tools for Sensitivity and Uncertainty Analysis Methodology Implementation (TSUNAMI) to assess, from a detailed physics-based perspective, the similarity of the experiment designs to the commercial systems they are intended to validate. Based on the results of the TSUNAMI analysis, one series of experiments was found to be preferable to the other and will provide significant new data for the validation of reactor physics and criticality safety codes.

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

    DOE Patents [OSTI]

    Gibby, Ronald L. (Richland, WA); Lawrence, Leo A. (Kennewick, WA); Woodley, Robert E. (Richland, WA); Wilson, Charles N. (Richland, WA); Weber, Edward T. (Kennewick, WA); Johnson, Carl E. (Elk Grove, IL)

    1981-01-01

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

  7. TTMC buyout will consolidate ownership

    SciTech Connect (OSTI)

    Morris, G.D.L.

    1997-04-16

    CW has learned that the privatization of the Trinidad and Tobago Methanol Co. (TTMS; Point Lisas, Trinidad) is the first step in a consolidation of ownership and marketing of methanol produced on the island. A consortium made up of chemical distributor Helm (Hamburg), engineering firm Ferrostaal (Essen), and Colonia Life Insurance (Clico; Port of Spain, Trinidad) will buy the government`s 69% of TTMC, displacing a bid by Methanex. TTMC owns and operates two methanol plants, M1 and M3. They produce 460,000 m.t./year and 550,000 m.t./year, respectively, and were designed and built by Ferrostaal. Helm markets half of TTMC`s output, with the remainder locked into contracts that run through 2000.

  8. Practical Static Ownership Inference Ana Milanova and Yin Liu

    E-Print Network [OSTI]

    Bystroff, Chris

    Practical Static Ownership Inference Ana Milanova and Yin Liu Rensselaer Polytechnic Institute. This paper presents novel static analyses for Java that infers ownership according to two known ownership. This paper presents novel static analyses for Java that infer ownership accord- ing to two well

  9. Trace Fission Product Ratios for Nuclear Forensics Attribution of Weapons-Grade Plutonium from Fast Breeder Reactor Blankets 

    E-Print Network [OSTI]

    Osborn, Jeremy

    2014-08-13

    for the 500-MWe Indian PFBR was obtained from Chirayath et al.22 and essential parameters are listed in Table II. There is an active core, one meter in height, which consists of an inner core and outer core of MOX “driver” fuel. The MOX fuel of the active...-Particle MOX – Mixed Oxide MWd/kg – Megawatt-day per kilogram of heavy metal MWe – Megawatt electric NPT – Treaty on the Non-Proliferation of Nuclear Weapons ORNL – Oak Ridge National Laboratory PFBR – Prototype Fast Breeder Reactor PHWR...

  10. Proceedings of the US Nuclear Regulatory Commission twentieth water reactor safety information meeting; Volume 2, Severe accident research, Thermal hydraulics

    SciTech Connect (OSTI)

    Weiss, A.J. [comp.] [Brookhaven National Lab., Upton, NY (United States)

    1993-03-01

    This three-volume report contains papers presented at the Twentieth Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, during the week of October 21--23, 1992. The papers describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included 10 different papers presented by researchersfrom CEC, China, Finland, France, Germany, Japan, Spain and Taiwan. Selected papers have been processed separately for inclusion in the Energy Science and Technology Database.

  11. EA-0912: Urgent-Relief Acceptance of Foreign Research Reactor Spent Nuclear Fuel

    Office of Energy Efficiency and Renewable Energy (EERE)

    This EA evaluates the environmental impacts of a proposal to accept 409 spent fuel elements from eight foreign research reactors in seven European countries.  The spent fuel would be shipped across...

  12. Advanced Test Reactor Tour

    SciTech Connect (OSTI)

    Miley, Don

    2011-01-01

    The Advanced Test Reactor at Idaho National Laboratory is the foremost nuclear materials test reactor in the world. This virtual tour describes the reactor, how experiments are conducted, and how spent nuclear fuel is handled and stored. For more information about INL research, visit http://www.facebook.com/idahonationallaboratory.

  13. Advanced Test Reactor Tour

    ScienceCinema (OSTI)

    Miley, Don

    2013-05-28

    The Advanced Test Reactor at Idaho National Laboratory is the foremost nuclear materials test reactor in the world. This virtual tour describes the reactor, how experiments are conducted, and how spent nuclear fuel is handled and stored. For more information about INL research, visit http://www.facebook.com/idahonationallaboratory.

  14. Chu Visits Site of America's First New Nuclear Reactor in Three...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    for nuclear energy in 2013, U.S. Secretary of Energy Steven Chu visited the Vogtle nuclear power plant in Waynesboro, Georgia and Oak Ridge National Laboratory to highlight the...

  15. CFD Analysis of Nuclear Fuel Bundles and Spacer Grids for PWR Reactors 

    E-Print Network [OSTI]

    Capone, Luigi

    2012-10-19

    The analysis of the turbulent flows in nuclear fuel bundles is a very interesting task to optimize the efficiency of modern nuclear power plants. The proposed study utilizes Computational Fluid Dynamics (CFD) to characterize the flow pattern...

  16. Nuclear reactor removable radial shielding assembly having a self-bowing feature

    DOE Patents [OSTI]

    Pennell, William E. (Greensburg, PA); Kalinowski, Joseph E. (Smithton, PA); Waldby, Robert N. (New Stanton, PA); Rylatt, John A. (Monroeville, PA); Swenson, Daniel V. (Greensburg, PA)

    1978-01-01

    A removable radial shielding assembly for use in the periphery of the core of a liquid-metal-cooled fast-breeder reactor, for closing interassembly gaps in the reactor core assembly load plane prior to reactor criticality and power operation to prevent positive reactivity insertion. The assembly has a lower nozzle portion for inserting into the core support and a flexible heat-sensitive bimetallic central spine surrounded by blocks of shielding material. At refueling temperature and below the spine is relaxed and in a vertical position so that the tolerances permitted by the interassembly gaps allow removal and replacement of the various reactor core assemblies. During an increase in reactor temperature from refueling to hot standby, the bimetallic spine expands, bowing the assembly toward the core center line, exerting a radially inward gap-closing-force on the above core load plane of the reactor core assembly, closing load plane interassembly gaps throughout the core prior to startup and preventing positive reactivity insertion.

  17. ITER: The International Thermonuclear Experimental Reactor and the nuclear weapons proliferation implications of thermonuclear-fusion energy

    E-Print Network [OSTI]

    Gsponer, A; Gsponer, Andre; Hurni, Jean-Pierre

    2004-01-01

    This paper contains two parts: (I) A list of "points" highlighting the strategic-political and military-technical reasons and implications of the very probable siting of ITER (the International Thermonuclear Experimental Reactor) in Japan, which should be confirmed sometimes in early 2004. (II) A technical analysis of the nuclear weapons proliferation implications of inertial- and magnetic-confinement fusion systems substantiating the technical points highlighted in the first part, and showing that while full access to the physics of thermonuclear weapons is the main implication of ICF, full access to large-scale tritium technology is the main proliferation impact of MCF. The conclusion of the paper is that siting ITER in a country such as Japan, which already has a large separated-plutonium stockpile, and an ambitious laser-driven ICF program (comparable in size and quality to those of the United States or France) will considerably increase its latent (or virtual) nuclear weapons proliferation status, and fo...

  18. Reactor safety method

    DOE Patents [OSTI]

    Vachon, Lawrence J. (Clairton, PA)

    1980-03-11

    This invention relates to safety means for preventing a gas cooled nuclear reactor from attaining criticality prior to start up in the event the reactor core is immersed in hydrogenous liquid. This is accomplished by coating the inside surface of the reactor coolant channels with a neutral absorbing material that will vaporize at the reactor's operating temperature.

  19. Nuclear safety analyses and core design calculations to convert the Texas A & M University Nuclear Science Center reactor to low enrichment uranium fuel. Final report

    SciTech Connect (OSTI)

    Parish, T.A.

    1995-03-02

    This project involved performing the nuclear design and safety analyses needed to modify the license issued by the Nuclear Regulatory Commission to allow operation of the Texas A& M University Nuclear Science Center Reactor (NSCR) with a core containing low enrichment uranium (LEU) fuel. The specific type of LEU fuel to be considered was the TRIGA 20-20 fuel produced by General Atomic. Computer codes for the neutronic analyses were provided by Argonne National Laboratory (ANL) and the assistance of William Woodruff of ANL in helping the NSCR staff to learn the proper use of the codes is gratefully acknowledged. The codes applied in the LEU analyses were WIMSd4/m, DIF3D, NCTRIGA and PARET. These codes allowed full three dimensional, temperature and burnup dependent calculations modelling the NSCR core to be performed for the first time. In addition, temperature coefficients of reactivity and pulsing calculations were carried out in-house, whereas in the past this modelling had been performed at General Atomic. In order to benchmark the newly acquired codes, modelling of the current NSCR core with highly enriched uranium fuel was also carried out. Calculated results were compared to both earlier licensing calculations and experimental data and the new methods were found to achieve excellent agreement with both. Therefore, even if an LEU core is never loaded at the NSCR, this project has resulted in a significant improvement in the nuclear safety analysis capabilities established and maintained at the NSCR.

  20. Japanese suppliers in transition from domestic nuclear reactor vendors to international suppliers

    SciTech Connect (OSTI)

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

    1994-06-27

    Japan is emerging as a major leader and exporter of nuclear power technology. In the 1990s, Japan has the largest and strongest nuclear power supply industry worldwide as a result of the largest domestic nuclear power plant construction program. The Japanese nuclear power supply industry has moved from dependence on foreign technology to developing, design, building, and operating its own power plants. This report describes the Japanese nuclear power supply industry and examines one supplier--the Mitsubishi group--to develop an understanding of the supply industry and its relationship to the utilities, government, and other organizations.

  1. Body Mass Index, Neighborhood Fast Food and Restaurant Concentration, and Car Ownership

    E-Print Network [OSTI]

    Inagami, Sanae; Cohen, Deborah A.; Brown, Arleen F.; Asch, Steven M.

    2009-01-01

    Restaurant Concentration, and Car Ownership Sanae Inagami,body mass index and whether car ownership might moderateRESTAURANT CONCENTRATION, AND CAR OWNERSHIP Quiznos, Little

  2. Identification of two-phase flow patterns in a nuclear reactor by high-frequency contribution fraction

    SciTech Connect (OSTI)

    Wang, Y.W.; Pei, B.S. (National Tsing Hua Univ., Hsinchu (Taiwan). Dept. of Nuclear Engineering); King, C.H.; Lee, S.C. (Institute of Nuclear Energy Research, Lung-Tan (Taiwan))

    1990-02-01

    A method based on noise analysis techniques that can be applied to the identification of two-phase flow patterns in nuclear reactors is proposed. The identifying criterion, the high-frequency contribution fraction (HFCF), offers new potential to the in-core recognition of two-phase flow patterns. By analyzing 76 sets of signals acquired from a research nuclear reactor where two-phase flow patterns are generated in an in-core air/water loop, the typical signal, autocorrelogram, and spectrum of each flow pattern are demonstrated and evaluated. The identification success rate is 87 or 93%, depending on whether churn flow is counted. A method to improve the identification rate is also presented. This study demonstrates that the fluctuation characteristics above 10 Hz are induced by two-phase flow itself and are independent of the driving source; thus, it is adequate to apply the HFCF to the identification of two-phase flow patters. This study shows that it is possible to identify two-phase flow patterns by HFCF values.

  3. Tokamak reactor for treating fertile material or waste nuclear by-products

    DOE Patents [OSTI]

    Kotschenreuther, Michael T.; Mahajan, Swadesh M.; Valanju, Prashant M.

    2012-10-02

    Disclosed is a tokamak reactor. The reactor includes a first toroidal chamber, current carrying conductors, at least one divertor plate within the first toroidal chamber and a second chamber adjacent to the first toroidal chamber surrounded by a section that insulates the reactor from neutrons. The current carrying conductors are configured to confine a core plasma within enclosed walls of the first toroidal chamber such that the core plasma has an elongation of 1.5 to 4 and produce within the first toroidal chamber at least one stagnation point at a perpendicular distance from an equatorial plane through the core plasma that is greater than the plasma minor radius. The at least one divertor plate and current carrying conductors are configured relative to one another such that the current carrying conductors expand the open magnetic field lines at the divertor plate.

  4. Nanostructure of Metallic Particles in Light Water Reactor Used Nuclear Fuel

    SciTech Connect (OSTI)

    Buck, Edgar C.; Mausolf, Edward J.; Mcnamara, Bruce K.; Soderquist, Chuck Z.; Schwantes, Jon M.

    2015-03-11

    The extraordinary nano-structure of metallic particles in light water reactor fuels points to possible high reactivity through increased surface area and a high concentration of high energy defect sites. We have analyzed the metallic epsilon particles from a high burn-up fuel from a boiling water reactor using transmission electron microscopy and have observed a much finer nanostructure in these particles than has been reported previously. The individual round particles that varying in size between ~20 and ~50 nm appear to consist of individual crystallites on the order of 2-3 nm in diameter. It is likely that in-reactor irradiation induce displacement cascades results in the formation of the nano-structure. The composition of these metallic phases is variable yet the structure of the material is consistent with the hexagonal close packed structure of epsilon-ruthenium. These findings suggest that unusual catalytic behavior of these materials might be expected, particularly under accident conditions.

  5. Innovative design of uranium startup fast reactors

    E-Print Network [OSTI]

    Fei, Tingzhou

    2012-01-01

    Sodium Fast Reactors are one of the three candidates of GEN-IV fast reactors. Fast reactors play an important role in saving uranium resources and reducing nuclear wastes. Conventional fast reactors rely on transuranic ...

  6. Analysis of muon radiography of the Toshiba nuclear critical assembly reactor

    SciTech Connect (OSTI)

    Morris, C. L.; Bacon, Jeffery; Borozdin, Konstantin; Fabritius, J. M.; Perry, John; Ramsey, John [Los Alamos National Laboratory, Los Alamos, New Mexico 87545 (United States); Ban, Yuichiro; Izumi, Mikio; Sano, Yuji; Yoshida, Noriyuki [Toshiba Corporation, 8 Shinsugita-cho, Isogo-ku, Yokohama 235-8523 (Japan); Miyadera, Haruo [Los Alamos National Laboratory, Los Alamos, New Mexico 87545 (United States); Toshiba Corporation, 8 Shinsugita-cho, Isogo-ku, Yokohama 235-8523 (Japan); Mizokami, Shinya; Otsuka, Yasuyuki; Yamada, Daichi [Tokyo Electric Power Company, 1-1-3 Uchisaiwai-cho, Chiyoda-ku, Tokyo (Japan); Sugita, Tsukasa; Yoshioka, Kenichi [Toshiba Corporation, 4-1 Ukishima-cho, Kawasaki-ku, Kawasaki 210-0862 (Japan)

    2014-01-13

    A 1.2?×?1.2 m{sup 2} muon tracker was moved from Los Alamos to the Toshiba facility at Kawasaki, Japan, where it was used to take ?4 weeks of data radiographing the Toshiba Critical Assembly Reactor with cosmic ray muons. In this paper, we describe the analysis procedure, show results of this experiment, and compare the results to Monte Carlo predictions. The results validate the concept of using cosmic rays to image the damaged cores of the Fukushima Daiichi reactors.

  7. Measuring Sin^22?_13 with the Daya Bay Nuclear Reactors

    E-Print Network [OSTI]

    Yifang Wang

    2006-10-09

    Angle \\theta_13 is one of the two unknown neutrino mixing parameters to be determined. Its value may determine the future trend of the neutrino physics. We propose to measure sin^22\\theta_13 with a sensitivity better than 0.01 (90% C.L) at the Daya Bay reactor power plant.

  8. Method and apparatus for removing iodine from a nuclear reactor coolant

    DOE Patents [OSTI]

    Cooper, Martin H. (Monroeville, PA)

    1980-01-01

    A method and apparatus for removing iodine-131 and iodine-125 from a liquid sodium reactor coolant. Non-radioactive iodine is dissolved in hot liquid sodium to increase the total iodine concentration. Subsequent precipitation of the iodine in a cold trap removes both the radioactive iodine isotopes as well as the non-radioactive iodine.

  9. Successful Completion of the Largest Shipment of Russian Research Reactor High-Enriched Uranium Spent Nuclear Fuel from Czech Republic to Russian Federation

    SciTech Connect (OSTI)

    Michael Tyacke; Dr. Igor Bolshinsky; Jeff Chamberlin

    2008-07-01

    On December 8, 2007, the largest shipment of high-enriched uranium spent nuclear fuel was successfully made from a Russian-designed nuclear research reactor in the Czech Republic to the Russian Federation. This accomplishment is the culmination of years of planning, negotiations, and hard work. The United States, Russian Federation, and the International Atomic Energy Agency have been working together on the Russian Research Reactor Fuel Return (RRRFR) Program in support of the Global Threat Reduction Initiative. In February 2003, RRRFR Program representatives met with the Nuclear Research Institute in Rež, Czech Republic, and discussed the return of their high-enriched uranium spent nuclear fuel to the Russian Federation for reprocessing. Nearly 5 years later, the shipment was made. This paper discusses the planning, preparations, coordination, and cooperation required to make this important international shipment.

  10. HEALTH EFFECTS OF THE NUCLEAR ACCIDENT AT THREE MILE ISLAND

    E-Print Network [OSTI]

    Fabrikant, J.I.

    2010-01-01

    In) Symposium on Nuclear Reactor Safety: Perspective. Awe learned about nuclear reactor safety and health from the

  11. Argonne's Major Nuclear Energy Milestones | Argonne National...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Sciences Intelligence Analysis Nuclear Engineering Nuclear Milestones Argonne's Major Nuclear Energy Milestones Argonne's reactor tree Argonne's reactor tree December 2, 1942:...

  12. Experimental Validation of Passive Safety System Models: Application to Design and Optimization of Fluoride-Salt-Cooled, High-Temperature Reactors

    E-Print Network [OSTI]

    Zweibaum, Nicolas

    2015-01-01

    Systems for Advanced Nuclear Reactors,” Ph.D. Dissertation,Handbook for Nuclear Reactor Safety,” Luxembourg: Commissiondevelopment of advanced nuclear reactor technology requires

  13. New Revelation of Lightning Ball Observation and Proposal for a Nuclear Reactor Fusion Experiment

    E-Print Network [OSTI]

    Domokos Tar

    2009-10-12

    In this paper, the author brings further details regarding his Lightning Ball observation that were not mentioned in the first one (Ref.1-2). Additionally, he goes more into detail as the three forces that are necessary to allow the residual crescent form the hydrodynamic vortex ring to shrink into a sphere.Further topics are the similarities and analogies between the Lightning Ball formation's theory and the presently undertaken Tokamak-Stellarator-Spheromak fusion reactor experiments. A new theory and its experimental realisation are proposed as to make the shrinking of the hot plasma of reactors into a ball possible by means of the so called long range electromagnetic forces. In this way,the fusion ignition temperature could possibly atteined.

  14. System and method for determining coolant level and flow velocity in a nuclear reactor

    DOE Patents [OSTI]

    Brisson, Bruce William; Morris, William Guy; Zheng, Danian; Monk, David James; Fang, Biao; Surman, Cheryl Margaret; Anderson, David Deloyd

    2013-09-10

    A boiling water reactor includes a reactor pressure vessel having a feedwater inlet for the introduction of recycled steam condensate and/or makeup coolant into the vessel, and a steam outlet for the discharge of produced steam for appropriate work. A fuel core is located within a lower area of the pressure vessel. The fuel core is surrounded by a core shroud spaced inward from the wall of the pressure vessel to provide an annular downcomer forming a coolant flow path between the vessel wall and the core shroud. A probe system that includes a combination of conductivity/resistivity probes and/or one or more time-domain reflectometer (TDR) probes is at least partially located within the downcomer. The probe system measures the coolant level and flow velocity within the downcomer.

  15. Status of materials handbooks for particle accelerator and nuclear reactor applications

    SciTech Connect (OSTI)

    Maloy, Stuart [Los Alamos National Laboratory (LANL); Rogers, Berylene [Los Alamos National Laboratory (LANL); Ren, Weiju [ORNL; Philip, Rittenhouse [Consultant

    2008-01-01

    In support of research and development for accelerator applications, a materials handbook was developed in August of 1998 funded by the Accelerator Production of Tritium Project. This handbook, presently called Advanced Fuel Cycle Initiative (AFCI) Materials Handbook, Materials Data for Particle Accelerator Applications, has just issued Revision 5 and contains detailed information showing the effects of irradiation on many properties for a wide variety of materials. Development of a web-accessible materials database for Generation IV Reactor Programs has been ongoing for about three years. This handbook provides a single authoritative source for qualified materials data applicable to all Generation IV reactor concepts. A beta version of this Gen IV Materials Handbook has been completed and is presently under evaluation.

  16. New Revelation of Lightning Ball Observation and Proposal for a Nuclear Reactor Fusion Experiment

    E-Print Network [OSTI]

    Tar, Domokos

    2009-01-01

    In this paper, the author brings further details regarding his Lightning Ball observation that were not mentioned in the first one (Ref.1-2). Additionally, he goes more into detail as the three forces that are necessary to allow the residual crescent form the hydrodynamic vortex ring to shrink into a sphere.Further topics are the similarities and analogies between the Lightning Ball formation's theory and the presently undertaken Tokamak-Stellarator-Spheromak fusion reactor experiments. A new theory and its experimental realisation are proposed as to make the shrinking of the hot plasma of reactors into a ball possible by means of the so called long range electromagnetic forces. In this way,the fusion ignition temperature could possibly atteined.

  17. International Topical Meeting on Nuclear Reactor Thermal -Hydraulics, NURETH-15 NURETH15-491 Pisa, Italy, May 12-17, 2013

    E-Print Network [OSTI]

    Motta, Arthur T.

    with the idea of providing accurate thermal-hydraulic models for heat transfer by coupling a sub- channel codeThe 15th International Topical Meeting on Nuclear Reactor Thermal - Hydraulics, NURETH-15 NURETH15). CTF's capabilities include modeling two-phase flow in transient or quasi-steady state conditions

  18. A simple setup for neutron tomography at the Portuguese Nuclear Research Reactor

    E-Print Network [OSTI]

    M. A. Stanojev Pereira; J. G. Marques; R. Pugliesi

    2012-05-15

    A simple setup for neutron radiography and tomography was recently installed at the Portuguese Research Reactor. The objective of this work was to determine the operational characteristics of the installed setup, namely the irradiation time to obtain the best dynamic range for individual images and the spatial resolution. The performance of the equipment was demonstrated by imaging a fragment of a 17th century decorative tile.

  19. Mandatory Disclosure Quality, Inside Ownership, and Cost of Capital

    E-Print Network [OSTI]

    Core, John E.

    This paper examines whether and how inside ownership mediates the relation between disclosure quality and the cost of capital. Both ownership and more transparent reporting have the potential to align incentives between ...

  20. Guidelines for Preparing Criticality Safety Evaluations at Department of Energy Non-Reactor Nuclear Facilities

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    2007-02-07

    This standard provides a framework for generating Criticality Safety Evaluations (CSE) supporting fissionable material operations at Department of Energy (DOE) nonreactor nuclear facilities. This standard imposes no new criticality safety analysis requirements.

  1. Used nuclear fuel storage options including implications of small modular reactors

    E-Print Network [OSTI]

    Brinton, Samuel O. (Samuel Otis)

    2014-01-01

    This work addresses two aspects of the nuclear fuel cycle system with significant policy implications. The first is the preferred option for used fuel storage based on economics: local, regional or national storage. The ...

  2. Chu Visits Site of America's First New Nuclear Reactor in Three...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    energy in 2013, U.S. Secretary of Energy Steven Chu visited the Vogtle nuclear power plant in Waynesboro, Georgia and Oak Ridge National Laboratory to highlight the steps the...

  3. Uncertainty analysis on reactivity and discharged inventory for a pressurized water reactor fuel assembly due to {sup 235,238}U nuclear data uncertainties

    SciTech Connect (OSTI)

    Da Cruz, D. F.; Rochman, D.; Koning, A. J. [Nuclear Research and Consultancy Group NRG, Westerduinweg 3, 1755 ZG Petten (Netherlands)

    2012-07-01

    This paper discusses the uncertainty analysis on reactivity and inventory for a typical PWR fuel element as a result of uncertainties in {sup 235,238}U nuclear data. A typical Westinghouse 3-loop fuel assembly fuelled with UO{sub 2} fuel with 4.8% enrichment has been selected. The Total Monte-Carlo method has been applied using the deterministic transport code DRAGON. This code allows the generation of the few-groups nuclear data libraries by directly using data contained in the nuclear data evaluation files. The nuclear data used in this study is from the JEFF3.1 evaluation, and the nuclear data files for {sup 238}U and {sup 235}U (randomized for the generation of the various DRAGON libraries) are taken from the nuclear data library TENDL. The total uncertainty (obtained by randomizing all {sup 238}U and {sup 235}U nuclear data in the ENDF files) on the reactor parameters has been split into different components (different nuclear reaction channels). Results show that the TMC method in combination with a deterministic transport code constitutes a powerful tool for performing uncertainty and sensitivity analysis of reactor physics parameters. (authors)

  4. Elevated-Temperature Ferritic and Martensitic Steels and Their Application to Future Nuclear Reactors

    SciTech Connect (OSTI)

    Klueh, RL

    2005-01-31

    In the 1970s, high-chromium (9-12% Cr) ferritic/martensitic steels became candidates for elevated-temperature applications in the core of fast reactors. Steels developed for conventional power plants, such as Sandvik HT9, a nominally Fe-12Cr-1Mo-0.5W-0.5Ni-0.25V-0.2C steel (composition in wt %), were considered in the United States, Europe, and Japan. Now, a new generation of fission reactors is in the planning stage, and ferritic, bainitic, and martensitic steels are again candidates for in-core and out-of-core applications. Since the 1970s, advances have been made in developing steels with 2-12% Cr for conventional power plants that are significant improvements over steels originally considered. This paper will review the development of the new steels to illustrate the advantages they offer for the new reactor concepts. Elevated-temperature mechanical properties will be emphasized. Effects of alloying additions on long-time thermal exposure with and without stress (creep) will be examined. Information on neutron radiation effects will be discussed as it applies to ferritic and martensitic steels.

  5. Radioactive waste shipments to Hanford retrievable storage from Westinghouse Advanced Reactors and Nuclear Fuels Divisions, Cheswick, Pennsylvania

    SciTech Connect (OSTI)

    Duncan, D.; Pottmeyer, J.A.; Weyns, M.I.; Dicenso, K.D.; DeLorenzo, D.S.

    1994-04-01

    During the next two decades the transuranic (TRU) waste now stored in the burial trenches and storage facilities at the Hanford Sits in southeastern Washington State is to be retrieved, processed at the Waste Receiving and Processing Facility, and shipped to the Waste Isolation Pilot Plant (WIPP), near Carlsbad, New Mexico for final disposal. Approximately 5.7 percent of the TRU waste to be retrieved for shipment to WIPP was generated by the decontamination and decommissioning (D&D) of the Westinghouse Advanced Reactors Division (WARD) and the Westinghouse Nuclear Fuels Division (WNFD) in Cheswick, Pennsylvania and shipped to the Hanford Sits for storage. This report characterizes these radioactive solid wastes using process knowledge, existing records, and oral history interviews.

  6. ASSESSMENT OF THE RADIONUCLIDE COMPOSITION OF "HOT PARTICLES" SAMPLED IN THE CHERNOBYL NUCLEAR POWER PLANT FOURTH REACTOR UNIT

    SciTech Connect (OSTI)

    Farfan, E.; Jannik, T.; Marra, J.

    2011-10-01

    Fuel-containing materials sampled from within the Chernobyl Nuclear Power Plant (ChNPP) 4th Reactor Unit Confinement Shelter were spectroscopically studied for gamma and alpha content. Isotopic ratios for cesium, europium, plutonium, americium, and curium were identified and the fuel burnup in these samples was determined. A systematic deviation in the burnup values based on the cesium isotopes, in comparison with other radionuclides, was observed. The conducted studies were the first ever performed to demonstrate the presence of significant quantities of {sup 242}Cm and {sup 243}Cm. It was determined that there was a systematic underestimation of activities of transuranic radionuclides in fuel samples from inside of the ChNPP Confinement Shelter, starting from {sup 241}Am (and going higher), in comparison with the theoretical calculations.

  7. Proceedings of the international meeting on thermal nuclear reactor safety. Vol. 1

    SciTech Connect (OSTI)

    1983-02-01

    Separate abstracts are included for each of the papers presented concerning current issues in nuclear power plant safety; national programs in nuclear power plant safety; radiological source terms; probabilistic risk assessment methods and techniques; non LOCA and small-break-LOCA transients; safety goals; pressurized thermal shocks; applications of reliability and risk methods to probabilistic risk assessment; human factors and man-machine interface; and data bases and special applications.

  8. Decline of radionuclides in the nearshore environment following nuclear reactor closure: A U.K. case study

    SciTech Connect (OSTI)

    Cundy, A.B.; Croudace, I.W.; Warwick, P.E.; Bains, M.E.D.

    1999-09-01

    Radioactive discharges from nuclear facilities are frequently made into the marine environment and their fate during and after cessation of discharges is a matter of interest and concern. This study examines the decline of the radionuclides {sup 60} and {sup 65}Zn along the southern UK. coast, over the per 1988--1998, following the closure of the steam-generating heavy water (SGHW) reactor at AEA Winfrith, Dorset, UK. {sup 60}Co and {sup 65}Zn (and other activation products such as {sup 63}Ni and {sup 55}Fe) were widely dispersed in the marine environment off the central south coast of England, due to authorized releases from AEA Winfrith. Significant interaction occurred with clay-rich sediments and biota. A general exponential decline in {sup 60}Co activities (and in {sup 65}Zn activity) is found in intertidal mudflat sediments, seaweed and marine fauna in different areas along the south coast following closure of the reactor in 1990. Effective half-lives are determined which vary from 1 to 4 years in surface sediments ({sup 60}Co only), 1--4 years in seaweed and 0.5--2.5 years in crustaceans, bivalves and molluscs. Physical mixing and bioturbation largely control the rate at which {sup 60}Co declines in surface sediments. Both {sup 60}Co and {sup 65}Zn show a relatively slow rate of decline in seaweed and in marine fauna, showing that even after the virtual cessation of discharge from nuclear facilities, contamination of these organisms may persist for a number of years, albeit at reduced activities. Reasons for this persistence are likely to include absorption of radionuclides from sediment, and release and recycling of radionuclides via breakdown of contaminated organic material.

  9. Security Science & Technology | Nuclear Science | ORNL

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Security Treaty Verification Nuclear Systems Modeling, Simulation & Validation Nuclear Systems Technology Reactor Technology Nuclear Science Home | Science & Discovery |...

  10. Analysis of Reference Design for Nuclear-Assisted Hydrogen Production at 750°C Reactor Outlet Temperature

    SciTech Connect (OSTI)

    Michael G. McKellar; Edwin A. Harvego

    2010-05-01

    The use of High Temperature Electrolysis (HTE) for the efficient production of hydrogen without the greenhouse gas emissions associated with conventional fossil-fuel hydrogen production techniques has been under investigation at the Idaho National Engineering Laboratory (INL) for the last several years. The activities at the INL have included the development, testing and analysis of large numbers of solid oxide electrolysis cells, and the analyses of potential plant designs for large scale production of hydrogen using a high-temperature gas-cooled reactor (HTGR) to provide the process heat and electricity to drive the electrolysis process. The results of this research led to the selection in 2009 of HTE as the preferred concept in the U.S. Department of Energy (DOE) hydrogen technology down-selection process. However, the down-selection process, along with continued technical assessments at the INL, has resulted in a number of proposed modifications and refinements to improve the original INL reference HTE design. These modifications include changes in plant configuration, operating conditions and individual component designs. This report describes the resulting new INL reference design coupled to two alternative HTGR power conversion systems, a Steam Rankine Cycle and a Combined Cycle (a Helium Brayton Cycle with a Steam Rankine Bottoming Cycle). Results of system analyses performed to optimize the design and to determine required plant performance and operating conditions when coupled to the two different power cycles are also presented. A 600 MWt high temperature gas reactor coupled with a Rankine steam power cycle at a thermal efficiency of 44.4% can produce 1.85 kg/s of hydrogen and 14.6 kg/s of oxygen. The same capacity reactor coupled with a combined cycle at a thermal efficiency of 42.5% can produce 1.78 kg/s of hydrogen and 14.0 kg/s of oxygen.

  11. Nuclear Fission Reactor Safety Research in FP7 and future perspectives

    E-Print Network [OSTI]

    Garbil, Roger

    2014-01-01

    The European Union (?U) has defined in the Europe 2020 strategy and 2050 Energy Roadmap its long-term vision for establishing a secure, sustainable and competitive energy system and setting up legally binding targets by 2020 for reducing greenhouse emissions, by increasing energy efficiency and the share of renewable energy sources while including a significant share from nuclear fission. Nuclear energy can enable the further reduction in harmful emissions and can contribute to the EU’s competitive energy system, security of supply and independence from fossil fuels. Nuclear fission is a valuable option for those 14 EU countries that promote its use as part of their national energy mix. The European Group on Ethics in Science and New Technologies (EGE) adopted its Opinion No.27 ‘An ethical framework for assessing research, production and use of energy’ and proposed an integrated ethics approach for the research, production and use of energy in the EU, seeking equilibrium among four criteria – access ...

  12. Importance of thermal nonequilibrium considerations for the simulation of nuclear reactor LOCA transients. [PWR

    SciTech Connect (OSTI)

    Fischer, S.R.; Nelson, R.A.; Sullivan, L.H.

    1980-01-01

    The purpose of this paper is to show the importance of considering thermal nonequilibrium effects in computer simulations of the refill and reflood portions of pressurized water reactor (PWR) loss-of-coolnat accident (LOCA) transients. Although RELAP4 assumes thermodynamic equilibrium between phases, models that account for the nonequilibrium phenomena associated with the mixing of subcooled emergency cooling water with steam and the superheating of vapor in the presence of liquid droplets have recently been incorporated into the code. Code calculated results, both with and without these new models, have been compared with experimental test data to assess the importance of including thermal nonequilibrium phenomena in computer code simulations.

  13. Source-term reevaluation for US commercial nuclear power reactors: a status report

    SciTech Connect (OSTI)

    Herzenberg, C.L.; Ball, J.R.; Ramaswami, D.

    1984-12-01

    Only results that had been discussed publicly, had been published in the open literature, or were available in preliminary reports as of September 30, 1984, are included here. More than 20 organizations are participating in source-term programs, which have been undertaken to examine severe accident phenomena in light-water power reactors (including the chemical and physical behavior of fission products under accident conditions), update and reevaluate source terms, and resolve differences between predictions and observations of radiation releases and related phenomena. Results from these source-term activities have been documented in over 100 publications to date.

  14. Liquid level, void fraction, and superheated steam sensor for nuclear reactor cores

    DOE Patents [OSTI]

    Tokarz, Richard D. (West Richland, WA)

    1983-01-01

    An apparatus for detecting nominal phase conditions of coolant in a reactor vessel comprising one or more lengths of tubing each leading from a location being monitored to a closed outer end exterior of the vessel. Temperature is sensed at the open end of each length of tubing. Pressure within the tubing is also sensed. Both measurements are directed to an analyzer which compares the measured temperature to the known saturated temperature of the coolant at the measured pressure. In this manner, the nominal phase conditions of the coolant are constantly monitored.

  15. Compact Reactor

    SciTech Connect (OSTI)

    Williams, Pharis E. [Williams Research, P.O. Box 554, Los Alamos, NM87544 (United States)

    2007-01-30

    Weyl's Gauge Principle of 1929 has been used to establish Weyl's Quantum Principle (WQP) that requires that the Weyl scale factor should be unity. It has been shown that the WQP requires the following: quantum mechanics must be used to determine system states; the electrostatic potential must be non-singular and quantified; interactions between particles with different electric charges (i.e. electron and proton) do not obey Newton's Third Law at sub-nuclear separations, and nuclear particles may be much different than expected using the standard model. The above WQP requirements lead to a potential fusion reactor wherein deuterium nuclei are preferentially fused into helium nuclei. Because the deuterium nuclei are preferentially fused into helium nuclei at temperatures and energies lower than specified by the standard model there is no harmful radiation as a byproduct of this fusion process. Therefore, a reactor using this reaction does not need any shielding to contain such radiation. The energy released from each reaction and the absence of shielding makes the deuterium-plus-deuterium-to-helium (DDH) reactor very compact when compared to other reactors, both fission and fusion types. Moreover, the potential energy output per reactor weight and the absence of harmful radiation makes the DDH reactor an ideal candidate for space power. The logic is summarized by which the WQP requires the above conditions that make the prediction of DDH possible. The details of the DDH reaction will be presented along with the specifics of why the DDH reactor may be made to cause two deuterium nuclei to preferentially fuse to a helium nucleus. The presentation will also indicate the calculations needed to predict the reactor temperature as a function of fuel loading, reactor size, and desired output and will include the progress achieved to date.

  16. High Temperature Gas-Cooled Reactors Lessons Learned Applicable to the Next Generation Nuclear Plant

    SciTech Connect (OSTI)

    J. M. Beck; L. F. Pincock

    2011-04-01

    The purpose of this report is to identify possible issues highlighted by these lessons learned that could apply to the NGNP in reducing technical risks commensurate with the current phase of design. Some of the lessons learned have been applied to the NGNP and documented in the Preconceptual Design Report. These are addressed in the background section of this document and include, for example, the decision to use TRISO fuel rather than BISO fuel used in the Peach Bottom reactor; the use of a reactor pressure vessel rather than prestressed concrete found in Fort St. Vrain; and the use of helium as a primary coolant rather than CO2. Other lessons learned, 68 in total, are documented in Sections 2 through 6 and will be applied, as appropriate, in advancing phases of design. The lessons learned are derived from both negative and positive outcomes from prior HTGR experiences. Lessons learned are grouped according to the plant, areas, systems, subsystems, and components defined in the NGNP Preconceptual Design Report, and subsequent NGNP project documents.

  17. White paper report on using nuclear reactors to search for a value of theta13

    E-Print Network [OSTI]

    2004-01-01

    nuclear power plant in San Luis Obispo County, California,Power Plant (DCPP) on the Central Coast in California.Power Plant is located in San Luis Obispo County north of Santa Barbara and provides electricity to northern and central California.

  18. Proceedings of the US Nuclear Regulatory Commission fifteenth water reactor safety information meeting: Volume 6, Decontamination and decommissioning, accident management, TMI-2

    SciTech Connect (OSTI)

    Weiss, A. J.

    1988-02-01

    This six-volume report contains 140 papers out of the 164 that were presented at the Fifteenth Water Reactor Safety Information Meeting held at the National Bureau of Standards, Gaithersburg, Maryland, during the week of October 26-29, 1987. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. This report, Volume 6, discusses decontamination and decommissioning, accident management, and the Three Mile Island-2 reactor accident. Thirteen reports have been cataloged separately.

  19. Neutronic reactor

    DOE Patents [OSTI]

    Wende, Charles W. J. (Augusta, GA); Babcock, Dale F. (Wilmington, DE); Menegus, Robert L. (Wilmington, DE)

    1983-01-01

    A nuclear reactor includes an active portion with fissionable fuel and neutron moderating material surrounded by neutron reflecting material. A control element in the active portion includes a group of movable rods constructed of neutron-absorbing material. Each rod is movable with respect to the other rods to vary the absorption of neutrons and effect control over neutron flux.

  20. Worldwide assessment of steam-generator problems in pressurized-water-reactor nuclear power plants

    SciTech Connect (OSTI)

    Woo, H.H.; Lu, S.C.

    1981-09-15

    Objective is to assess the reliability of steam generators of pressurized water reactor (PWR) power plants in the United States and abroad. The assessment is based on operation experience of both domestic and foreign PWR plants. The approach taken is to collect and review papers and reports available from the literature as well as information obtained by contacting research institutes both here and abroad. This report presents the results of the assessment. It contains a general background of PWR plant operations, plant types, and materials used in PWR plants. A review of the worldwide distribution of PWR plants is also given. The report describes in detail the degradation problems discovered in PWR steam generators: their causes, their impacts on the performance of steam generators, and the actions to mitigate and avoid them. One chapter is devoted to operating experience of PWR steam generators in foreign countries. Another discusses the improvements in future steam generator design.