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Sample records for power reactor 1400apr

  1. POWER REACTOR

    DOEpatents

    Zinn, W.H.

    1958-07-01

    A fast nuclear reactor system ls described for producing power and radioactive isotopes. The reactor core is of the heterogeneous, fluid sealed type comprised of vertically arranged elongated tubular fuel elements having vertical coolant passages. The active portion is surrounded by a neutron reflector and a shield. The system includes pumps and heat exchangers for the primary and secondary coolant circuits. The core, primary coolant pump and primary heat exchanger are disposed within an irapenforate tank which is filled with the primary coolant, in this case a liquid metal such as Na or NaK, to completely submerge these elements. The tank is completely surrounded by a thick walled concrete shield. This reactor system utilizes enriched uranium or plutonium as the fissionable material, uranium or thorium as a diluent and thorium or uranium containing less than 0 7% of the U/sup 235/ isotope as a fertile material.

  2. NEUTRONIC REACTOR POWER PLANT

    DOEpatents

    Metcalf, H.E.

    1962-12-25

    This patent relates to a nuclear reactor power plant incorporating an air-cooled, beryllium oxide-moderated, pebble bed reactor. According to the invention means are provided for circulating a flow of air through tubes in the reactor to a turbine and for directing a sidestream of the circu1ating air through the pebble bed to remove fission products therefrom as well as assist in cooling the reactor. (AEC)

  3. Operating US power reactors

    SciTech Connect

    Silver, E.G.

    1982-07-01

    The operation of US power reactors during March and April 1982 is summarized. Events of special note are discussed in the text, and the operational performance of all licensed power reactors is presented. These data are taken from the monthly Operating Units Status Report prepared by the Nuclear Regulatory Commission (NRC).

  4. HOMOGENEOUS NUCLEAR POWER REACTOR

    DOEpatents

    King, L.D.P.

    1959-09-01

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

  5. Compact power reactor

    DOEpatents

    Wetch, Joseph R.; Dieckamp, Herman M.; Wilson, Lewis A.

    1978-01-01

    There is disclosed a small compact nuclear reactor operating in the epithermal neutron energy range for supplying power at remote locations, as for a satellite. The core contains fuel moderator elements of Zr hydride with 7 w/o of 93% enriched uranium alloy. The core has a radial beryllium reflector and is cooled by liquid metal coolant such as NaK. The reactor is controlled and shut down by moving portions of the reflector.

  6. Multimegawatt space power reactors

    SciTech Connect

    Dearien, J.A.; Whitbeck, J.F.

    1989-01-01

    In response to the need of the Strategic Defense Initiative (SDI) and long range space exploration and extra-terrestrial basing by the National Air and Space Administration (NASA), concepts for nuclear power systems in the multi-megawatt levels are being designed and evaluated. The requirements for these power systems are being driven primarily by the need to minimize weight and maximize safety and reliability. This paper will discuss the present requirements for space based advanced power systems, technological issues associated with the development of these advanced nuclear power systems, and some of the concepts proposed for generating large amounts of power in space. 31 figs.

  7. POWER BREEDER REACTOR

    DOEpatents

    Monson, H.O.

    1960-11-22

    An arrangement is offered for preventing or minimizing the contraction due to temperature rise, of a reactor core comprising vertical fuel rods in sodium. Temperature rise of the fuel rods would normally make them move closer together by inward bowing, with a resultant undesired increase in reactivity. According to the present invention, assemblies of the fuel rods are laterally restrained at the lower ends of their lower blanket sections and just above the middle of the fuel sections proper of the rods, and thus the fuel sections move apart, rather than together, with increase in temperature.

  8. PUSH-PULL POWER REACTOR

    DOEpatents

    Froman, D.K.

    1959-02-24

    Power generating nuclear reactors of the homogeneous liquid fuel type are discussed. The apparatus utilizes two identical reactors interconnected by conduits through heat exchanging apparatus. Each reactor contains a critical geometry region and a vapor region separated from the critical region by a baffle. When the liquid in the first critical region becomes critical, the vapor pressure above the fuel is increased due to the rise in the temperature until it forces the liquid fuel out of the first critical region through the heat exchanger and into the second critical region, which is at a lower temperature and consequently a lower vapor pressure. The above reaction is repeated in the second critical region and the liquid fuel is forced back into the first critical region. In this manner criticality is achieved alternately in each critical region and power is extracted by the heat exchanger from the liquid fuel passing therethrough. The vapor region and the heat exchanger have a non-critical geometry and reactivity control is effected by conventional control rods in the critical regions.

  9. Transmutation of actinides in power reactors.

    PubMed

    Bergelson, B R; Gerasimov, A S; Tikhomirov, G V

    2005-01-01

    Power reactors can be used for partial short-term transmutation of radwaste. This transmutation is beneficial in terms of subsequent storage conditions for spent fuel in long-term storage facilities. CANDU-type reactors can transmute the main minor actinides from two or three reactors of the VVER-1000 type. A VVER-1000-type reactor can operate in a self-service mode with transmutation of its own actinides.

  10. Transmutation of actinides in power reactors.

    PubMed

    Bergelson, B R; Gerasimov, A S; Tikhomirov, G V

    2005-01-01

    Power reactors can be used for partial short-term transmutation of radwaste. This transmutation is beneficial in terms of subsequent storage conditions for spent fuel in long-term storage facilities. CANDU-type reactors can transmute the main minor actinides from two or three reactors of the VVER-1000 type. A VVER-1000-type reactor can operate in a self-service mode with transmutation of its own actinides. PMID:16604724

  11. Zirconium Hydride Space Power Reactor design.

    NASA Technical Reports Server (NTRS)

    Asquith, J. G.; Mason, D. G.; Stamp, S.

    1972-01-01

    The Zirconium Hydride Space Power Reactor being designed and fabricated at Atomics International is intended for a wide range of potential applications. Throughout the program a series of reactor designs have been evaluated to establish the unique requirements imposed by coupling with various power conversion systems and for specific applications. Current design and development emphasis is upon a 100 kilowatt thermal reactor for application in a 5 kwe thermoelectric space power generating system, which is scheduled to be fabricated and ground tested in the mid 70s. The reactor design considerations reviewed in this paper will be discussed in the context of this 100 kwt reactor and a 300 kwt reactor previously designed for larger power demand applications.

  12. Thermionic reactors for space nuclear power

    NASA Technical Reports Server (NTRS)

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

    1985-01-01

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

  13. MODERATOR ELEMENTS FOR UNIFORM POWER NUCLEAR REACTOR

    DOEpatents

    Balent, R.

    1963-03-12

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

  14. Thermionic reactors for space nuclear power

    NASA Astrophysics Data System (ADS)

    Griaznov, Georgii M.; Zhabotinskii, Evgenii E.; Serbin, Victor I.; Zrodnikov, Anatolii V.; Pupko, Victor Ia.; Ponomarev-Stepnoi, Nikolai N.; Usov, V. A.; Nikolaev, Iu. V.

    Compact thermionic nuclear reactor systems with satisfactory mass performance are competitive with space nuclear power systems based on the organic Rankine and closed Brayton cycles. The mass characteristics of the thermionic space nuclear power system are better than that of the solar power system for power levels beyond about 10 kWe. Longlife thermionic fuel element requirements, including their optimal dimensions, and common requirements for the in-core thermionic reactor design are formulated. Thermal and fast in-core thermionic reactors are considered and the ranges of their sensible use are discussed. Some design features of the fast in-core thermionic reactors cores (power range to 1 MWe) including a choice of coolants are discussed. Mass and dimensional performance for thermionic nuclear power reactor system are assessed. It is concluded that thermionic space nuclear power systems are promising power supplies for spacecrafts and that a single basic type of thermionic fuel element may be used for power requirements ranging to several hundred kWe.

  15. Low power reactor for remote applications

    NASA Astrophysics Data System (ADS)

    Meier, K. L.; Palmer, R. G.; Kirchner, W. L.

    1985-05-01

    A compact, low power reactor is being designed to provide electric power for remote, unattended applications. Because of the high fuel and maintenance costs for conventional power sources such as diesel generators, a reactor power supply appears especially attractive for remote and inaccessible locations. Operating at a thermal power level of 135 kWt, the power supply achieves a gross electrical output of 25 kWe from an organic Rankine cycle (ORC) engine. By intentional selection of design features stressing inherent safety, operation in an unattended mode is possible with minimal risk to the environment. Reliability is achieved through the use of components representing existing, proven technology. Low enrichment uranium particle fuel, in graphite core blocks, cooled by heat pipes coupled to an ORC converter insures long term, virtually maintenance free, operation of this reactor for remote applications.

  16. Low power reactor for remote applications

    SciTech Connect

    Meier, K.L.; Palmer, R.G.; Kirchner, W.L.

    1985-01-01

    A compact, low power reactor is being designed to provide electric power for remote, unattended applications. Because of the high fuel and maintenance costs for conventional power sources such as diesel generators, a reactor power supply appears especially attractive for remote and inaccessible locations. Operating at a thermal power level of 135 kWt, the power supply achieves a gross electrical output of 25 kWe from an organic Rankine cycle (ORC) engine. By intentional selection of design features stressing inherent safety, operation in an unattended mode is possible with minimal risk to the environment. Reliability is achieved through the use of components representing existing, proven technology. Low enrichment uranium particle fuel, in graphite core blocks, cooled by heat pipes coupled to an ORC converter insures long-term, virtually maintenance free, operation of this reactor for remote applications. 10 refs., 7 figs., 3 tabs.

  17. Compact reactor/ORC power source

    SciTech Connect

    Meier, K.L.; Kirchner, W.L.; Willcutt, G.J.

    1986-01-01

    A compact power source that combines an organic Rankine Cycle (ORC) electric generator with a nuclear reactor heat source is being designed and fabricated. Incorporating existing ORC technology with proven reactor technology, the compact reactor/ORC power source offers high reliability while minimizing the need for component development. Thermal power at 125 kWt is removed from the coated particle fueled, graphite moderated reactor by heat pipes operating at 500/sup 0/C. Outside the reactor vessel and connected to the heat pipes are vaporizers in which the toluene ORC working fluid is heated to 370/sup 0/C. In the turbine-alternator-pump (TAP) combined-rotating unit, the thermal energy of the toluene is converted to 25 kWe of electric power. Lumped parameter systems analyses combined with a finite element thermal analysis have aided in the power source design. The analyses have provided assurance of reliable multiyear normal operation as well as full power operation with upset conditions, such as failed heat pipes and inoperative ORC vaporizers. Because of inherent high reliability, long life, and insensitivity to upset conditions, this power source is especially suited for use in remote, inaccessible locations where fuel delivery and maintenance costs are high. 10 refs.

  18. Parliament votes against building fifth power reactor

    SciTech Connect

    Not Available

    1993-11-01

    After a heated three-day debate, Finland's parliament voted on September 24 to reject the proposal to build the country's fifth nuclear power reactor. As predicted, the vote was close: 107 voted against more nuclear power, 90 were in favor, two members of the 200-seat parliament were not present, and the speaker did not vote.

  19. Heat pipe reactors for space power applications

    NASA Technical Reports Server (NTRS)

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

    1977-01-01

    A family of heat pipe reactors design concepts has been developed to provide heat to a variety of electrical conversion systems. Three power plants are described that span the power range 1-500 kWe and operate in the temperature range 1200-1700 K. The reactors are fast, compact, heat-pipe cooled, high-temperature nuclear reactors fueled with fully enriched refractory fuels, UC-ZrC or UO2. Each fuel element is cooled by an axially located molybdenum heat pipe containing either sodium or lithium vapor. Virtues of the reactor designs are the avoidance of single-point failure mechanisms, the relatively high operating temperature, and the expected long lifetimes of the fuel element components.

  20. Gaseous fuel reactors for power systems

    NASA Technical Reports Server (NTRS)

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

    1977-01-01

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

  1. POWER GENERATING NEUTRONIC REACTOR SYSTEM

    DOEpatents

    Vernon, H.C.

    1958-03-01

    This patent relates to reactor systems of the type wherein the cooiing medium is a liquid which is converted by the heat of the reaction to steam which is conveyed directly to a pnime mover such as a steam turbine driving a generatore after which it is condensed and returred to the coolant circuit. In this design, the reactor core is disposed within a tank for containing either a slurry type fuel or an aggregation of solid fuel elements such as elongated rods submerged in a liquid moderator such as heavy water. The top of the tank is provided with a nozzle which extends into an expansion chamber connected with the upper end of the tank, the coolant being maintained in the expansion chamber at a level above the nozzle and the steam being formed in the expansion chamber.

  2. Liquid Metal Cooled Reactor for Space Power

    NASA Astrophysics Data System (ADS)

    Weitzberg, Abraham

    2003-01-01

    The conceptual design is for a liquid metal (LM) cooled nuclear reactor that would provide heat to a closed Brayton cycle (CBC) power conversion subsystem to provide electricity for electric propulsion thrusters and spacecraft power. The baseline power level is 100 kWe to the user. For long term power generation, UN pin fuel with Nb1Zr alloy cladding was selected. As part of the SP-100 Program this fuel demonstrated lifetime with greater than six atom percent burnup, at temperatures in the range of 1400-1500 K. The CBC subsystem was selected because of the performance and lifetime database from commercial and aircraft applications and from prior NASA and DOE space programs. The high efficiency of the CBC also allows the reactor to operate at relatively low power levels over its 15-year life, minimizing the long-term power density and temperature of the fuel. The scope of this paper is limited to only the nuclear components that provide heated helium-xenon gas to the CBC subsystem. The principal challenge for the LM reactor concept was to design the reactor core, shield and primary heat transport subsystems to meet mission requirements in a low mass configuration. The LM concept design approach was to assemble components from prior programs and, with minimum change, determine if the system met the objective of the study. All of the components are based on technologies having substantial data bases. Nuclear, thermalhydraulic, stress, and shielding analyses were performed using available computer codes. Neutronics issues included maintaining adequate operating and shutdown reactivities, even under accident conditions. Thermalhydraulic and stress analyses calculated fuel and material temperatures, coolant flows and temperatures, and thermal stresses in the fuel pins, components and structures. Using conservative design assumptions and practices, consistent with the detailed design work performed during the SP-100 Program, the mass of the reactor, shield, primary heat

  3. Reactor power system deployment and startup

    NASA Technical Reports Server (NTRS)

    Wetch, J. R.; Nelin, C. J.; Britt, E. J.; Klein, G.

    1985-01-01

    This paper addresses issues that should receive further examination in the near-term as concept selection for development of a U.S. space reactor power system is approached. The issues include: the economics, practicality and system reliability associated with transfer of nuclear spacecraft from low earth shuttle orbits to operational orbits, via chemical propulsion versus nuclear electric propulsion; possible astronaut supervised reactor and nuclear electric propulsion startup in low altitude Shuttle orbit; potential deployment methods for nuclear powered spacecraft from Shuttle; the general public safety of low altitude startup and nuclear safe and disposal orbits; the question of preferred reactor power level; and the question of frozen versus molten alkali metal coolant during launch and deployment. These issues must be considered now because they impact the SP-100 concept selection, power level selection, weight and size limits, use of deployable radiators, reliability requirements, and economics, as well as the degree of need for and the urgency of developing space reactor power systems.

  4. A small, 1400 K, reactor for Brayton space power systems.

    NASA Technical Reports Server (NTRS)

    Lantz, E.; Mayo, W.

    1972-01-01

    An investigation was conducted to determine minimum dimensions and minimum weight obtainable in a design for a reactor using uranium-233 nitride or plutonium-239 nitride as fuel. Such a reactor had been considered by Krasner et al. (1971). Present space power status is discussed, together with questions of reactor design and power distribution in the reactor. The characteristics of various reactor types are compared, giving attention also to a zirconium hydride reactor.

  5. High Flux Isotope Reactor power upgrade status

    SciTech Connect

    Rothrock, R.B.; Hale, R.E.; Cheverton, R.D.

    1997-03-01

    A return to 100-MW operation is being planned for the High Flux Isotope Reactor (HFIR). Recent improvements in fuel element manufacturing procedures and inspection equipment will be exploited to reduce hot spot and hot streak factors sufficiently to permit the power upgrade without an increase in primary coolant pressure. Fresh fuel elements already fabricated for future use are being evaluated individually for power upgrade potential based on their measured coolant channel dimensions.

  6. Static conversion systems. [for space power reactors

    NASA Technical Reports Server (NTRS)

    Ewell, R.; Mondt, J.

    1985-01-01

    Historically, all space power systems that have actually flown in space have relied on static energy conversion technology. Thus, static conversion is being considered for space nuclear power systems as well. There are four potential static conversion technologies which should be considered. These include: the alkali metal thermoelectric converter (AMTEC), the thermionic converter, the thermoelectric converter, and the thermophotovoltaic converter (TPV). These four conversion technologies will be described in brief detail along with their current status and development needs. In addition, the systems implications of using each of these conversion technologies with a space nuclear reactor power system will be evaluated and some comparisons made.

  7. Gas-core reactor power transient analysis

    NASA Technical Reports Server (NTRS)

    Kascak, A. F.

    1972-01-01

    The gas core reactor is a proposed device which features high temperatures. It has applications in high specific impulse space missions, and possibly in low thermal pollution MHD power plants. The nuclear fuel is a ball of uranium plasma radiating thermal photons as opposed to gamma rays. This thermal energy is picked up before it reaches the solid cavity liner by an inflowing seeded propellant stream and convected out through a rocket nozzle. A wall-burnout condition will exist if there is not enough flow of propellant to convect the energy back into the cavity. A reactor must therefore operate with a certain amount of excess propellant flow. Due to the thermal inertia of the flowing propellant, the reactor can undergo power transients in excess of the steady-state wall burnout power for short periods of time. The objective of this study was to determine how long the wall burnout power could be exceeded without burning out the cavity liner. The model used in the heat-transfer calculation was one-dimensional, and thermal radiation was assumed to be a diffusion process.

  8. NACA Zero Power Reactor Facility Hazards Summary

    NASA Technical Reports Server (NTRS)

    1957-01-01

    The Lewis Flight Propulsion Laboratory of the National Advisory Committee for Aeronautics proposes to build a zero power research reactor facility which will be located in the laboratory grounds near Clevelaurd, Ohio. The purpose of this report is to inform the Advisory Commit tee on Reactor Safeguards of the U. S. Atomic Energy Commission in re gard to the design of the reactor facility, the cha,acteristics of th e site, and the hazards of operation at this location, The purpose o f this reactor is to perform critical experiments, to measure reactiv ity effects, to serve as a neutron source, and to serve as a training tool. The reactor facility is described. This is followed by a discu ssion of the nuclear characteristics and the control system. Site cha racteristics are then discussed followed by a discussion of the exper iments which may be conducted in the facility. The potential hazards of the facility are then considered, particularly, the maximum credib le accident. Finally, the administrative procedure is discussed.

  9. 77 FR 38742 - Non-Power Reactor License Renewal

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-06-29

    ...; ] NUCLEAR REGULATORY COMMISSION 10 CFR Part 50 RIN 3150-AI96 Non-Power Reactor License Renewal AGENCY... renewal requirements for non-power reactors. This contemplated rulemaking would also make conforming changes to address technical issues in existing non-power reactor regulations. The NRC is seeking...

  10. Modular stellarator reactor: a fusion power plant

    SciTech Connect

    Miller, R.L.; Bathke, C.G.; Krakowski, R.A.; Heck, F.M.; Green, L.; Karbowski, J.S.; Murphy, J.H.; Tupper, R.B.; DeLuca, R.A.; Moazed, A.

    1983-07-01

    A comparative analysis of the modular stellarator and the torsatron concepts is made based upon a steady-state ignited, DT-fueled, reactor embodiment of each concept for use as a central electric-power station. Parametric tradeoff calculations lead to the selection of four design points for an approx. 4-GWt plant based upon Alcator transport scaling in l = 2 systems of moderate aspect ratio. The four design points represent high-aspect ratio. The four design points represent high-(0.08) and low-(0.04) beta versions of the modular stellarator and torsatron concepts. The physics basis of each design point is described together with supporting engineering and economic analyses. The primary intent of this study is the elucidation of key physics and engineering tradeoffs, constraints, and uncertainties with respect to the ultimate power reactor embodiment.

  11. Power conditioning for space nuclear reactor systems

    NASA Technical Reports Server (NTRS)

    Berman, Baruch

    1987-01-01

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

  12. Direct conversion nuclear reactor space power systems

    SciTech Connect

    Britt, E.J.; Fitzpatrick, G.O.

    1982-08-01

    This paper presents the results of a study of space nuclear reactor power systems using either thermoelectric or thermionic energy converters. An in-core reactor design and two heat pipe cooled out-of-core reactor designs were considered. One of the out-of-core cases utilized, long heat pipes (LHP) directly coupled to the energy converter. The second utilized a larger number of smaller heat pipes (mini-pipe) radiatively coupled to the energy converter. In all cases the entire system, including power conditioning, was constrained to be launched in a single shuttle flight. Assuming presently available performance, both the LHP thermoelectric system and minipipe thermionic system, designed to produce 100 kWe for seven years, would have a specific mass near 22kg/kWe. The specific mass of the thermionic minipipe system designed for a one year mission is 165 kg/kWe due to less fuel swelling. Shuttle imposed growth limits are near 300 kWe and 1.2 MWe for the thermoelectric and thermionic systems, respectively. Converter performance improvements could double this potential, and over 10 MWe may be possible for very short missions.

  13. Analysis of UF6 breeder reactor power plants

    NASA Technical Reports Server (NTRS)

    Clement, J. D.; Rust, J. H.

    1976-01-01

    Gaseous UF6 fueled breeder reactor design and technical applications of such concepts are summarized. Special attention was given to application in nuclear power plants and to reactor efficiency and safety factors.

  14. REACTOR-FLASH BOILER-FLYWHEEL POWER PLANT

    DOEpatents

    Loeb, E.

    1961-01-17

    A power generator in the form of a flywheel with four reactors positioned about its rim is described. The reactors are so positioned that steam, produced in the reactor, exists tangentially to the flywheel, giving it a rotation. The reactors are incompletely moderated without water. The water enters the flywheel at its axis, under sufficient pressure to force it through the reactors, where it is converted to steam. The fuel consists of parallel twisted ribbons assembled to approximate a cylinder.

  15. 78 FR 73898 - Operator Licensing Examination Standards for Power Reactors

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-12-09

    ... COMMISSION Operator Licensing Examination Standards for Power Reactors AGENCY: Nuclear Regulatory Commission... Standards for Power Reactors.'' DATES: Submit comments by February 7, 2014. Comments received after this..., and grading of examinations used for licensing operators at nuclear power plants pursuant to...

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

    NASA Technical Reports Server (NTRS)

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

    1971-01-01

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

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

    NASA Technical Reports Server (NTRS)

    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.

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

    SciTech Connect

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

    1987-05-01

    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.

  19. UF6 breeder reactor power plants for electric power generation

    NASA Technical Reports Server (NTRS)

    Rust, J. H.; Clement, J. D.; Hohl, F.

    1976-01-01

    The reactor concept analyzed is a U-233F6 core surrounded by a molten salt (Li(7)F, BeF2, ThF4) blanket. Nuclear survey calculations were carried out for both spherical and cylindrical geometries. Thermodynamic cycle calculations were performed for a variety of Rankine cycles. A conceptual design is presented along with a system layout for a 1000 MW stationary power plant. Advantages of the gas core breeder reactor (GCBR) are as follows: (1) high efficiency; (2) simplified on-line reprocessing; (3) inherent safety considerations; (4) high breeding ratio; (5) possibility of burning all or most of the long-lived nuclear waste actinides; and (6) possibility of extrapolating the technology to higher temperatures and MHD direct conversion.

  20. Programmable AC power supply for simulating power transient expected in fusion reactor

    SciTech Connect

    Halimi, B.; Suh, K. Y.

    2012-07-01

    This paper focus on control engineering of the programmable AC power source which has capability to simulate power transient expected in fusion reactor. To generate the programmable power source, AC-AC power electronics converter is adopted to control the power of a set of heaters to represent the transient phenomena of heat exchangers or heat sources of a fusion reactor. The International Thermonuclear Experimental Reactor (ITER) plasma operation scenario is used as the basic reference for producing this transient power source. (authors)

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

    NASA Technical Reports Server (NTRS)

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

    1988-01-01

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

  2. 77 FR 60039 - Non-Power Reactor License Renewal

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-10-02

    ... FR 38742), for comment from the public, licensees, certificate holders, and other stakeholders. The... COMMISSION 10 CFR Part 50 RIN 3150-AI96 Non-Power Reactor License Renewal AGENCY: Nuclear Regulatory... streamline non-power reactor license renewal. This final regulatory basis incorporates input from the...

  3. 78 FR 64028 - Decommissioning of Nuclear Power Reactors

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-10-25

    ... Register on February 14, 2012 (77 FR 8902), for a 60-day public comment period. The public comment period... COMMISSION Decommissioning of Nuclear Power Reactors AGENCY: Nuclear Regulatory Commission. ACTION... regulatory guide (RG) 1.184 ``Decommissioning of Nuclear Power Reactors.'' This guide describes a method...

  4. Power ascension strategy following a reactor trip during EOC coastdown

    SciTech Connect

    Beard, C.L.; Heibel, M.D. ); Lesnick, D.C. )

    1992-01-01

    The difficulties associated with returning a reactor to the pretrip power level following a reactor trip during an end-of-cycle (EOC) power coastdown maneuver, and maintaining it once achieved, have caused utilities to abandon the restart and enter their refueling outages ahead of schedule. The Commonwealth Edison Company (CECo) Braidwood and Byron units have experienced reactor trips during EOC power coastdown maneuvers and have successfully performed restarts. The installation of the BEACON core monitoring system, which provides core monitoring, measurement reduction, core analysis and follow, and core prediction capability utilizing a very fast and accurate three-dimensional nodal code, at the CECo Byron, Braidwood, and Zion stations allows the reactor engineers at these units to accurately determine reactor response. The capabilities of the BEACON system allow an optimal return to power strategy to be developed and continuously updated. This paper presents a method for establishing the optimal return to power strategy utilizing the BEACON system.

  5. Liquid metal cooled reactors for space power applications

    NASA Technical Reports Server (NTRS)

    Bailey, S.; Vaidyanathan, S.; Van Hoomissen, J.

    1985-01-01

    The technology basis for evaluation of liquid metal cooled space reactors is summarized. Requirements for space nuclear power which are relevant to selection of the reactor subsystem are then reviewed. The attributes of liquid metal cooled reactors are considered in relation to these requirements in the areas of liquid metal properties, neutron spectrum characteristics, and fuel form. Key features of typical reactor designs are illustrated. It is concluded that liquid metal cooled fast spectrum reactors provide a high confidence, flexible option for meeting requirements for SP-100 and beyond.

  6. Diversion assumptions for high-powered research reactors

    SciTech Connect

    Binford, F.T.

    1984-01-01

    This study deals with diversion assumptions for high-powered research reactors -- specifically, MTR fuel; pool- or tank-type research reactors with light-water moderator; and water, beryllium, or graphite reflectors, and which have a power level of 25 MW(t) or more. The objective is to provide assistance to the IAEA in documentation of criteria and inspection observables related to undeclared plutonium production in the reactors described above, including: criteria for undeclared plutonium production, necessary design information for implementation of these criteria, verification guidelines including neutron physics and heat transfer, and safeguards measures to facilitate the detection of undeclared plutonium production at large research reactors.

  7. Gas-cooled reactor power systems for space

    NASA Astrophysics Data System (ADS)

    Walter, Carl E.

    Large amounts of electric power are required for some of the systems envisioned in support of SDI. Since various applications are being considered, and an overall power architecture study has not been completed, the required power levels and corresponding operating times for specific systems are not known. The characteristics of six designs for power levels of 2, 10 and 20 MWe for operating time of 1 and 7 yrs are described. The operating conditions for these arbitrary designs were chosen to minimize system specific mass. Both gas and liquid cooled reactors are considered. The designs discussed draw heavily on the Pluto project experience. Gas cooled thermal reactors coupled with Brayton cycle power conversion appear to provide reasonable multimegawatt space power systems. An advanced radiation design must be developed which can meet the mass limit assumed. The inherent high temperature capability of the reactors considered removes the reactor as a limiting condition on system performance.

  8. Consumption of the electric power inside silent discharge reactors

    SciTech Connect

    Yehia, Ashraf

    2015-01-15

    An experimental study was made in this paper to investigate the relation between the places of the dielectric barriers, which cover the surfaces of the electrodes in the coaxial cylindrical reactors, and the rate of change of the electric power that is consumed in forming silent discharges. Therefore, silent discharges have been formed inside three coaxial cylindrical reactors. The dielectric barriers in these reactors were pasted on both the internal surface of the outer electrode in the first reactor and the external surface of the inner electrode in the second reactor as well as the surfaces of the two electrodes in the third reactor. The reactor under study has been fed by atmospheric air that flowed inside it with a constant rate at normal temperature and pressure, in parallel with the application of a sinusoidal ac voltage between the electrodes of the reactor. The electric power consumed in forming the silent discharges inside the three reactors was measured as a function of the ac peak voltage. The validity of the experimental results was investigated by applying Manley's equation on the same discharge conditions. The results have shown that the rate of consumption of the electric power relative to the ac peak voltage per unit width of the discharge gap improves by a ratio of either 26.8% or 80% or 128% depending on the places of the dielectric barriers that cover the surfaces of the electrodes inside the three reactors.

  9. EBT reactor characteristics consistent with stability and power balance requirements

    SciTech Connect

    Uckan, N.A.; Santoro, R.T.

    1983-01-01

    This paper summarizes the results of a recent EBT reactor study that includes both ring and core plasma properties and consistent treatment of coupled ring-core stability criteria and power balance requirements. The principal finding is that constraints imposed by these coupling and other physics and technology considerations permit a broad operating window for reactor design optimization. A number of concept improvements are also proposed that are found to offer the potential for further improvement of the reactor size and parameters.

  10. Ultrasonic level and temperature sensor for power reactor applications

    SciTech Connect

    Dress, W.B.: Miller, G.N.

    1983-01-01

    An ultrasonic waveguide employing torsional and extensional acoustic waves has been developed for use as a level and temperature sensor in pressurized and boiling water nuclear power reactors. Features of the device include continuous measurement of level, density, and temperature producing a real-time profile of these parameters along a chosen path through the reactor vessel.

  11. Status of power-reactor projects undergoing licensing review

    SciTech Connect

    Silver, E.G.

    1982-07-01

    Recent regulatory and other actions relating to power reactors undergoing licensing review are summarized in Table 1 as of May 1, 1982. Except as otherwise noted, all the information presented in this article is taken from NRC press releases or from the reactor docket file, both of which are available at the NRC Public Document Room, 1717 H Street NW, Washington, DC 20555.

  12. High power density reactors based on direct cooled particle beds

    NASA Astrophysics Data System (ADS)

    Powell, J. R.; Horn, F. L.

    Reactors based on direct cooled High Temperature Gas Cooled Reactor (HTGR) type particle fuel are described. The small diameter particle fuel is packed between concentric porous cylinders to make annular fuel elements, with the inlet coolant gas flowing inwards. Hot exit gas flows out along the central channel of each element. Because of the very large heat transfer area in the packed beds, power densities in particle bed reactors (PBRs) are extremely high resulting in compact, lightweight systems. Coolant exit temperatures are high, because of the ceramic fuel temperature capabilities, and the reactors can be ramped to full power and temperature very rapidly. PBR systems can generate very high burst power levels using open cycle hydrogen coolant, or high continuous powers using closed cycle helium coolant. PBR technology is described and development requirements assessed.

  13. Computer optimization of reactor-thermoelectric space power systems

    NASA Technical Reports Server (NTRS)

    Maag, W. L.; Finnegan, P. M.; Fishbach, L. H.

    1973-01-01

    A computer simulation and optimization code that has been developed for nuclear space power systems is described. The results of using this code to analyze two reactor-thermoelectric systems are presented.

  14. Heat pipe nuclear reactor for space power

    NASA Technical Reports Server (NTRS)

    Koening, D. R.

    1976-01-01

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

  15. COUPLED FAST-THERMAL POWER BREEDER REACTOR

    DOEpatents

    Avery, R.

    1961-07-18

    A nuclear reactor having a region operating predominantly on fast neutrons and another region operating predominantly on slow neutrons is described. The fast region is a plutonium core and the slow region is a natural uranium blanket around the core. Both of these regions are free of moderator. A moderating reflector surrounds the uranium blanket. The moderating material and thickness of the reflector are selected so that fissions in the uranium blanket make a substantial contribution to the reactivity of the reactor.

  16. Reference Reactor Module for the Affordable Fission Surface Power System

    NASA Astrophysics Data System (ADS)

    Poston, David I.; Kapernick, Richard J.; Dixon, David D.; Amiri, Benjamin W.; Marcille, Thomas F.

    2008-01-01

    Surface fission power systems on the Moon and Mars may provide the first US application of fission reactor technology in space since 1965. The requirements of many surface power applications allow the consideration of systems with much less development risk than most other space reactor applications, because of modest power (10s of kWe) and no driving need for minimal mass (allowing temperatures <1000 K). The Affordable Fission Surface Power System (AFSPS) study was completed by NASA/DOE to determine the cost of a modest performance, low-technical risk surface power system. This paper describes the reference AFSPS reactor module concept, which is designed to provide a net power of 40 kWe for 8 years on the lunar surface; note, the system has been designed with technologies that are fully compatible with a Martian surface application. The reactor concept uses stainless-steel based, UO2-fueled, liquid metal-cooled fission reactor coupled to free-piston Stirling converters. The reactor shielding approach utilizes both in-situ and launched shielding to keep the dose to astronauts much lower than the natural background radiation on the lunar surface. One of the important ``affordability'' attributes is that the concept has been designed to minimize both the technical and programmatic safety risk.

  17. Reference Reactor Module for the Affordable Fission Surface Power System

    SciTech Connect

    Poston, David I.; Kapernick, Richard J.; Dixon, David D.; Amiri, Benjamin W.; Marcille, Thomas F.

    2008-01-21

    Surface fission power systems on the Moon and Mars may provide the first US application of fission reactor technology in space since 1965. The requirements of many surface power applications allow the consideration of systems with much less development risk than most other space reactor applications, because of modest power (10s of kWe) and no driving need for minimal mass (allowing temperatures <1000 K). The Affordable Fission Surface Power System (AFSPS) study was completed by NASA/DOE to determine the cost of a modest performance, low-technical risk surface power system. This paper describes the reference AFSPS reactor module concept, which is designed to provide a net power of 40 kWe for 8 years on the lunar surface; note, the system has been designed with technologies that are fully compatible with a Martian surface application. The reactor concept uses stainless-steel based, UO{sub 2}-fueled, liquid metal-cooled fission reactor coupled to free-piston Stirling converters. The reactor shielding approach utilizes both in-situ and launched shielding to keep the dose to astronauts much lower than the natural background radiation on the lunar surface. One of the important 'affordability' attributes is that the concept has been designed to minimize both the technical and programmatic safety risk.

  18. Design considerations for an interial confinement fusion reactor power plant

    NASA Astrophysics Data System (ADS)

    Massey, J. V.; Simpson, J. E.

    1981-08-01

    A conceptual design study to further define the engineering and economic concerns for inertial confinement fusion reactors is presented. Alternatives to the Livermore HYLIFE concept were examined and information from liquid metal cooled fast breeder reactor power plant studies was incorporated into the design. Laser and target physics models were employed in a reactor design with a low coolant flowrate and a high driver repetition rate. An example of such a design is the JADE concept. In addition to a power plant design developed using the JADE example, the applicability of the energy absorbing gas lithium ejector concept was investigated.

  19. Gas-cooled reactor power systems for space

    SciTech Connect

    Walter, C.E.

    1987-01-01

    Efficiency and mass characteristics for four gas-cooled reactor power system configurations in the 2- to 20-MWe power range are modeled. The configurations use direct and indirect Brayton cycles with and without regeneration in the power conversion loop. The prismatic ceramic core of the reactor consists of several thousand pencil-shaped tubes made from a homogeneous mixture of moderator and fuel. The heat rejection system is found to be the major contributor to system mass, particularly at high power levels. A direct, regenerated Brayton cycle with helium working fluid permits high efficiency and low specific mass for a 10-MWe system.

  20. Investigation of materials for fusion power reactors

    NASA Astrophysics Data System (ADS)

    Bouhaddane, A.; Slugeň, V.; Sojak, S.; Veterníková, J.; Petriska, M.; Bartošová, I.

    2014-06-01

    The possibility of application of nuclear-physical methods to observe radiation damage to structural materials of nuclear facilities is nowadays a very actual topic. The radiation damage to materials of advanced nuclear facilities, caused by extreme radiation stress, is a process, which significantly limits their operational life as well as their safety. In the centre of our interest is the study of the radiation degradation and activation of the metals and alloys for the new nuclear facilities (Generation IV fission reactors, fusion reactors ITER and DEMO). The observation of the microstructure changes in the reactor steels is based on experimental investigation using the method of positron annihilation spectroscopy (PAS). The experimental part of the work contains measurements focused on model reactor alloys and ODS steels. There were 12 model reactor steels and 3 ODS steels. We were investigating the influence of chemical composition on the production of defects in crystal lattice. With application of the LT 9 program, the spectra of specimen have been evaluated and the most convenient samples have been determined.

  1. Design of megawatt power level heat pipe reactors

    SciTech Connect

    Mcclure, Patrick Ray; Poston, David Irvin; Dasari, Venkateswara Rao; Reid, Robert Stowers

    2015-11-12

    An important niche for nuclear energy is the need for power at remote locations removed from a reliable electrical grid. Nuclear energy has potential applications at strategic defense locations, theaters of battle, remote communities, and emergency locations. With proper safeguards, a 1 to 10-MWe (megawatt electric) mobile reactor system could provide robust, self-contained, and long-term power in any environment. Heat pipe-cooled fast-spectrum nuclear reactors have been identified as a candidate for these applications. Heat pipe reactors, using alkali metal heat pipes, are perfectly suited for mobile applications because their nature is inherently simpler, smaller, and more reliable than “traditional” reactors. The goal of this project was to develop a scalable conceptual design for a compact reactor and to identify scaling issues for compact heat pipe cooled reactors in general. Toward this goal two detailed concepts were developed, the first concept with more conventional materials and a power of about 2 MWe and a the second concept with less conventional materials and a power level of about 5 MWe. A series of more qualitative advanced designs were developed (with less detail) that show power levels can be pushed to approximately 30 MWe.

  2. Small reactor power systems for manned planetary surface bases

    SciTech Connect

    Bloomfield, H.S.

    1987-12-01

    A preliminary feasibility study of the potential application of small nuclear reactor space power systems to manned planetary surface base missions was conducted. The purpose of the study was to identify and assess the technology, performance, and safety issues associated with integration of reactor power systems with an evolutionary manned planetary surface exploration scenario. The requirements and characteristics of a variety of human-rated modular reactor power system configurations selected for a range of power levels from 25 kWe to hundreds of kilowatts is described. Trade-off analyses for reactor power systems utilizing both man-made and indigenous shielding materials are provided to examine performance, installation and operational safety feasibility issues. The results of this study have confirmed the preliminary feasibility of a wide variety of small reactor power plant configurations for growth oriented manned planetary surface exploration missions. The capability for power level growth with increasing manned presence, while maintaining safe radiation levels, was favorably assessed for nominal 25 to 100 kWe modular configurations. No feasibility limitations or technical barriers were identified and the use of both distance and indigenous planetary soil material for human rated radiation shielding were shown to be viable and attractive options.

  3. Reference reactor module for NASA's lunar surface fission power system

    SciTech Connect

    Poston, David I; Kapernick, Richard J; Dixon, David D; Werner, James; Qualls, Louis; Radel, Ross

    2009-01-01

    Surface fission power systems on the Moon and Mars may provide the first US application of fission reactor technology in space since 1965. The Affordable Fission Surface Power System (AFSPS) study was completed by NASA/DOE to determine the cost of a modest performance, low-technical risk surface power system. The AFSPS concept is now being further developed within the Fission Surface Power (FSP) Project, which is a near-term technology program to demonstrate system-level TRL-6 by 2013. This paper describes the reference FSP reactor module concept, which is designed to provide a net power of 40 kWe for 8 years on the lunar surface; note, the system has been designed with technologies that are fully compatible with a Martian surface application. The reactor concept uses stainless-steel based. UO{sub 2}-fueled, pumped-NaK fission reactor coupled to free-piston Stirling converters. The reactor shielding approach utilizes both in-situ and launched shielding to keep the dose to astronauts much lower than the natural background radiation on the lunar surface. The ultimate goal of this work is to provide a 'workhorse' power system that NASA can utilize in near-term and future Lunar and Martian mission architectures, with the eventual capability to evolve to very high power, low mass systems, for either surface, deep space, and/or orbital missions.

  4. Small reactor power systems for manned planetary surface bases

    NASA Technical Reports Server (NTRS)

    Bloomfield, Harvey S.

    1987-01-01

    A preliminary feasibility study of the potential application of small nuclear reactor space power systems to manned planetary surface base missions was conducted. The purpose of the study was to identify and assess the technology, performance, and safety issues associated with integration of reactor power systems with an evolutionary manned planetary surface exploration scenario. The requirements and characteristics of a variety of human-rated modular reactor power system configurations selected for a range of power levels from 25 kWe to hundreds of kilowatts is described. Trade-off analyses for reactor power systems utilizing both man-made and indigenous shielding materials are provided to examine performance, installation and operational safety feasibility issues. The results of this study have confirmed the preliminary feasibility of a wide variety of small reactor power plant configurations for growth oriented manned planetary surface exploration missions. The capability for power level growth with increasing manned presence, while maintaining safe radiation levels, was favorably assessed for nominal 25 to 100 kWe modular configurations. No feasibility limitations or technical barriers were identified and the use of both distance and indigenous planetary soil material for human rated radiation shielding were shown to be viable and attractive options.

  5. Deployment history and design considerations for space reactor power systems

    NASA Astrophysics Data System (ADS)

    El-Genk, Mohamed S.

    2009-05-01

    The history of the deployment of nuclear reactors in Earth orbits is reviewed with emphases on lessons learned and the operation and safety experiences. The former Soviet Union's "BUK" power systems, with SiGe thermoelectric conversion and fast neutron energy spectrum reactors, powered a total of 31 Radar Ocean Reconnaissance Satellites (RORSATs) from 1970 to 1988 in 260 km orbit. Two of the former Soviet Union's TOPAZ reactors, with in-core thermionic conversion and epithermal neutron energy spectrum, powered two Cosmos missions launched in 1987 in ˜800 km orbit. The US' SNAP-10A system, with SiGe energy conversion and a thermal neutron energy spectrum reactor, was launched in 1965 in 1300 km orbit. The three reactor systems used liquid NaK-78 coolant, stainless steel structure and highly enriched uranium fuel (90-96 wt%) and operated at a reactor exit temperature of 833-973 K. The BUK reactors used U-Mo fuel rods, TOPAZ used UO 2 fuel rods and four ZrH moderator disks, and the SNAP-10A used moderated U-ZrH fuel rods. These low power space reactor systems were designed for short missions (˜0.5 kW e and ˜1 year for SNAP-10A, <3.0 kW e and <6 months for BUK, and ˜5.5 kW e and up to 1 year for TOPAZ). The deactivated BUK reactors at the end of mission, which varied in duration from a few hours to ˜4.5 months, were boosted into ˜800 km storage orbit with a decay life of more than 600 year. The ejection of the last 16 BUK reactor fuel cores caused significant contamination of Earth orbits with NaK droplets that varied in sizes from a few microns to 5 cm. Power systems to enhance or enable future interplanetary exploration, in-situ resources utilization on Mars and the Moon, and civilian missions in 1000-3000 km orbits would generate significantly more power of 10's to 100's kW e for 5-10 years, or even longer. A number of design options to enhance the operation reliability and safety of these high power space reactor power systems are presented and discussed.

  6. Gas-cooled reactor for space power systems

    SciTech Connect

    Walter, C.E.; Pearson, J.S.

    1987-05-01

    Reactor characteristics based on extensive development work on the 500-MWt reactor for the Pluto nuclear ramjet are described for space power systems useful in the range of 2 to 20 MWe for operating times of 1 y. The modest pressure drop through the prismatic ceramic core is supported at the outlet end by a ceramic dome which also serves as a neutron reflector. Three core materials are considered which are useful at temperatures up to about 2000 K. Most of the calculations are based on a beryllium oxide with uranium dioxide core. Reactor control is accomplished by use of a burnable poison, a variable-leakage reflector, and internal control rods. Reactivity swings of 20% are obtained with a dozen internal boron-10 rods for the size cores studied. Criticality calculations were performed using the ALICE Monte Carlo code. The inherent high-temperature capability of the reactor design removes the reactor as a limiting condition on system performance. The low fuel inventories required, particularly for beryllium oxide reactors, make space power systems based on gas-cooled near-thermal reactors a lesser safeguard risk than those based on fast reactors.

  7. Application of reactor-pumped lasers to power beaming

    SciTech Connect

    Repetti, T.E.

    1991-10-01

    Power beaming is the concept of centralized power generation and distribution to remote users via energy beams such as microwaves or laser beams. The power beaming community is presently performing technical evaluations of available lasers as part of the design process for developing terrestrial and space-based power beaming systems. This report describes the suitability of employing a nuclear reactor-pumped laser in a power beaming system. Although there are several technical issues to be resolved, the power beaming community currently believes that the AlGaAs solid-state laser is the primary candidate for power beaming because that laser meets the many design criteria for such a system and integrates well with the GaAs photodiode receiver array. After reviewing the history and physics of reactor-pumped lasers, the advantages of these lasers for power beaming are discussed, along with several technical issues which are currently facing reactor-pumped laser research. The overriding conclusion is that reactor-pumped laser technology is not presently developed to the point of being technially or economically competitive with more mature solid-state technologies for application to power beaming. 58 refs.

  8. Nuclear Power from Fission Reactors. An Introduction.

    ERIC Educational Resources Information Center

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

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

  9. Gas-cooled reactor power systems for space

    SciTech Connect

    Walter, C.E.

    1987-01-01

    In this paper the characteristics of six designs for power levels of 2, 10, and 20 MWe for operating times of 1 and 7 y are described. The operating conditions for these arbitrary designs were chosen to minimize system specific mass. The designs are based on recent work which benefits from earlier analyses of nuclear space power systems conducted at our Laboratory. Both gas- and liquid-cooled reactors had been considered. Pitts and Walter (1970) reported on the results of a detailed study of a 10-MWe lithium-cooled reactor in a potassium Rankine system. Unpublished results (1966) of a computer analysis provide details of an argon-cooled reactor in an argon Brayton system. The gas-cooled reactor design was based on extensive development work on the 500-MWth reactor for the nuclear ramjet (Pluto) as described by Walter (1964). The designs discussed here draw heavily on the Pluto project experience, which culminated in a successful full-power ground test as reported by Reynolds (1964). At higher power levels gas-cooled reactors coupled with Brayton systems with advanced radiator designs become attractive.

  10. Measurements of the reactor neutron power in absolute units

    SciTech Connect

    Lebedev, G. V.

    2015-12-15

    The neutron power of the reactor of the Yenisei space nuclear power plant is measured in absolute units using the modernized method of correlation analysis during the ground-based tests of the Yenisei prototypes. Results of the experiments are given. The desired result is obtained in a series of experiments carried out at the stage of the plant preparation for tests. The acceptability of experimental data is confirmed by the results of measuring the reactor neutron power in absolute units at the nominal level by the thermal balance during the life cycle tests of the ground prototypes.

  11. Protective actions as a factor in power reactor siting

    SciTech Connect

    Gant, K.S.; Schweitzer, M.

    1984-06-01

    This report examines the relationship between a power reactor site and the ease of implementing protective actions (emergency measures a serious accident). Limiting populating density around a reactor lowers the number of people at risk but cannot assure that all protective actions are possible for those who reside near the reactor. While some protective measures can always be taken (i.e., expedient respiratory protection, sheltering) the ability to evacuate the area or find adequate shelter may depend on the characteristics of the area near the reactor site. Generic siting restrictions designed to identify and eliminate these site-specific constraints would be difficult to formulate. The authors suggest identifying possible impediments to protective actions at a proposed reactor site and addressing these problems in the emergency plans. 66 references, 6 figures, 8 tables.

  12. Reactor Lithium Heat Pipes for HP-STMCs Space Reactor Power System

    NASA Astrophysics Data System (ADS)

    Tournier, Jean-Michel; El-Genk, Mohamed S.

    2004-02-01

    Design and performance analysis of the nuclear reactor's lithium heat pipes for a 110-kWe Heat Pipes-Segmented Thermoelectric Module Converters (HP-STMCs) Space Reactor Power system (SRPS) are presented. The evaporator length of the heat pipes is the same as the active core height (0.45 m) and the C-C finned condenser is of the same length as the STMC panels (1.5 m). The C-C finned condenser section is radiatively coupled to the collector shoes of the STMCs placed on both sides. The lengths of the adiabatic section, the values of the power throughput and the evaporator wall temperature depend on the radial location of the heat pipe in the reactor core and the number and dimensions of the potassium heat pipes in the heat rejection radiator. The reactor heat pipes have a total length that varies from 7.57 to 7.73 m, and a 0.2 mm thick Mo-14%Re wick with an average pore radius of 12 μm. The wick is separated from the Mo-14%Re wall by a 0.5 mm annulus filled with liquid lithium, to raise the prevailing capillary limit. The nominal evaporator (or reactor) temperature varies from 1513 to 1591 K and the thermal power of the reactor is 1.6 MW, which averages 12.7 kW for each of the 126 reactor heat pipes. The power throughput per heat pipe increase to a nominal 15.24 kW at the location of the peak power in the core and to 20.31 kW when an adjacent heat pipe fails. The prevailing capillary limit of the reactor heat pipes is 28.3 kW, providing a design margin >= 28%.

  13. Reactor Lithium Heat Pipes for HP-STMCs Space Reactor Power System

    SciTech Connect

    Tournier, Jean-Michel; El-Genk, Mohamed S.

    2004-02-04

    Design and performance analysis of the nuclear reactor's lithium heat pipes for a 110-kWe Heat Pipes-Segmented Thermoelectric Module Converters (HP-STMCs) Space Reactor Power system (SRPS) are presented. The evaporator length of the heat pipes is the same as the active core height (0.45 m) and the C-C finned condenser is of the same length as the STMC panels (1.5 m). The C-C finned condenser section is radiatively coupled to the collector shoes of the STMCs placed on both sides. The lengths of the adiabatic section, the values of the power throughput and the evaporator wall temperature depend on the radial location of the heat pipe in the reactor core and the number and dimensions of the potassium heat pipes in the heat rejection radiator. The reactor heat pipes have a total length that varies from 7.57 to 7.73 m, and a 0.2 mm thick Mo-14%Re wick with an average pore radius of 12 {mu}m. The wick is separated from the Mo-14%Re wall by a 0.5 mm annulus filled with liquid lithium, to raise the prevailing capillary limit. The nominal evaporator (or reactor) temperature varies from 1513 to 1591 K and the thermal power of the reactor is 1.6 MW, which averages 12.7 kW for each of the 126 reactor heat pipes. The power throughput per heat pipe increase to a nominal 15.24 kW at the location of the peak power in the core and to 20.31 kW when an adjacent heat pipe fails. The prevailing capillary limit of the reactor heat pipes is 28.3 kW, providing a design margin {>=} 28%.

  14. Transient thermal analysis of a space reactor power system

    SciTech Connect

    Gaeta, M.J.; Best, F.R. . Dept. of Nuclear Engineering)

    1993-07-01

    Space nuclear power systems utilize materials and processes that are completely different from terrestrial reactor systems. Therefore, the tools used to analyze ground-based systems are inappropriate for space reactor design and analysis. The purpose of this study was to develop a space reactor transient analysis tool and to apply this tool to scenarios of interest. The scope of the simulation includes the thermal and neutronic behavior of a liquid-metal-cooled fast reactor, the electrical and thermal performance of the thermoelectric generators, the thermal dynamics of heat pipe radiators, and the thermal behavior of the coolant piping between major components. The thermal model of the system is explicitly coupled to a momentum model of the primary and secondary coolant loops. A one-dimensional conduction model is employed in all solid component models. The reactor model includes an expression for energy generation due to fission and decay heat. The thermoelectric heat exchanger model accounts for thermal energy conversion to useful electrical output. The two-node radiator heat pipe model includes normal operation as well as limited heat pipe operation under sonic limit conditions. The reactor, thermoelectric heat exchanger, and heat pipe models are coupled explicitly by the coolant piping thermal model. The computer program is used to simulate a variety of transients including reactor power changer, degradation of the radiator, and a temporary open circuit condition on the thermoelectrics.

  15. Thermonuclear inverse magnetic pumping power cycle for stellarator reactors

    NASA Astrophysics Data System (ADS)

    Ho, D. D. M.; Kulsrud, R. M.

    1985-09-01

    A novel power cycle for direct conversion of alpha-particle energy into electricity is proposed for an ignited plasma in a stellerator reactor. The plasma column is alternately compressed and expanded in minor radius by periodic variation of the toroidal magnetic field strength. As a result of the way a stellarator is expected to work, the plasma pressure during expansion is greater than the corresponding pressure during compression. Therefore, negative work is done on the plasma during a complete cycle. This work manifests itself as a back-voltage in the toroidal field coils, and direct electrical energy is obtained from this voltage. For a typical reactor, the average power obtained from this cycle (with a minor radius compression factor on the order of 50%) can be as much as 50% of the electrical power obtained from the thermonuclear neutrons without compressing the plasma. Thus, if it is feasible to vary the toroidal field strength, the power cycle provides an alternative scheme of energy conversion for a deuterium-tritium fueled reactor. The cycle may become an important method of energy conversion for advanced neutron-lean fueled reactors. By operating two or more reactors in tandem, the cycle can be made self-sustaining.

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

    PubMed

    Corradini, M L

    2007-11-01

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

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

    PubMed

    Corradini, M L

    2007-11-01

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

  18. Calculation of the electrical power dissipated in silent discharge reactors

    SciTech Connect

    Yehia, Ashraf; Mizuno, Akira

    2005-08-15

    This paper is aimed to investigating how the electrical power consumed in generating silent discharges is influenced by the width of the annular discharge gap in the reactor. The silent discharges have been generated in conventional tubular reactors forming annular discharge gaps of 1x10{sup -3}-, 2x10{sup -3}-, and 3x10{sup -3}-m width, with one dielectric barrier pasting on the ground electrodes of the reactors. The reactors were fed by dry air flowing with a constant rate at atmospheric pressure and room temperature, and stressed by an ac voltage. The discharge power consumed in generating silent discharges has been measured as a function of the voltage applied to the reactors up to a V{sub p} of 20 000 V, and calculated by Manley's equation under the same discharge conditions. The experimental and calculated results have shown that when the reactor has a discharge gap of 1x10{sup -3}- or 2x10{sup -3}-m width, Manley's equation agrees only with the experimental results in range of the voltage of V{sub 0}{<=}V{sub p}{<=}V{sub c}. In the higher range of the applied voltage of V{sub c}{<=}V{sub p}{<=}20 000 V, Manley's equation has been modified to fit the experimental results. With the reactor that has a discharge gap of 3x10{sup -3}-m width, Manley's equation agrees with the experimental results over the whole range of the voltage applied to the reactor. The results have been explained by recorded oscillograms showing the wave form of the ac voltage applied to the reactors and the corresponding discharge current.

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

    NASA Astrophysics Data System (ADS)

    Radulescu, Laura; Pavelescu, Margarit

    2010-01-01

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

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

    SciTech Connect

    Radulescu, Laura; Pavelescu, Margarit

    2010-01-21

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

  1. Hybrid thermionic space reactor for power and propulsion

    SciTech Connect

    Sahin, S. . Teknik Egitim Fakueltesi); Kennel, E.B. )

    1994-08-01

    A thermo-hydrodynamic-neutronic analysis is performed for a fast, uranium carbide (UC) fueled space-craft nuclear in-core thermionic reactor. The thermo-hydrodynamic analysis shows that a hybrid thermionic spacecraft nuclear reactor can be designed for both electricity generation and nuclear thermal propulsion purposes. The neutronic analysis has been conducted in S[sub 8]-P[sub 3] approximation with the help of one- and two-dimensional neutron transport codes ANISN and DORT, respectively. The calculations have shown that a UC fueled electricity generating single mode thermionic nuclear reactor can be designed to be extremely compact because of the high atomic density of the nuclear fuel (by 95% sintering density), namely, with a core radius of 8.7 cm and core height of 25 cm, leading to power levels as low as 5 kW (electric) by an electrical output on an emitter surface of 1.243 W/cm[sup 2]. A reactor control with boronated reflector drums at the outer periphery of the radial reflector of 16-cm thickness would make possible reactivity changes of [Delta]k[sub eff] > 10% -- amply sufficient for a fast reactor -- without a significant distortion of the fission power profile during all phases of the space mission. The hybrid thermionic spacecraft nuclear reactor mode contains cooling channels in the nuclear fuel for the hydrogen propellant. This increase the critical reactor size because of the lower uranium atomic density in this design concept. Calculations have lead to a reactor with a core radius of 22 cm and core height of 35 cm leading to power levels [approximately] 50 kW(electric) under the aforementioned thermionic conversion conditions.

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

    SciTech Connect

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

    1984-01-01

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

  3. Specific power of liquid-metal-cooled reactors

    SciTech Connect

    Dobranich, D.

    1987-10-01

    Calculations of the core specific power for conceptual space-based liquid-metal-cooled reactors, based on heat transfer considerations, are presented for three different fuel types: (1) pin-type fuel; (2) cermet fuel; and (3) thermionic fuel. The calculations are based on simple models and are intended to provide preliminary comparative results. The specific power is of interest because it is a measure of the core mass required to produce a given amount of power. Potential problems concerning zero-g critical heat flux and loss-of-coolant accidents are also discussed because these concerns may limit the core specific power. Insufficient experimental data exists to accurately determine the critical heat flux of liquid-metal-cooled reactors in space; however, preliminary calculations indicate that it may be a concern. Results also indicate that the specific power of the pin-type fuels can be increased significantly if the gap between the fuel and the clad is eliminated. Cermet reactors offer the highest specific power because of the excellent thermal conductivity of the core matrix material. However, it may not be possible to take fuel advantage of this characteristic when loss-of-coolant accidents are considered in the final core design. The specific power of the thermionic fuels is dependent mainly on the emitter temperature. The small diameter thermionic fuels have specific powers comparable to those of pin-type fuels. 11 refs., 12 figs, 2 tabs.

  4. Background radiation measurements at high power research reactors

    SciTech Connect

    Ashenfelter, J.; Balantekin, B.; Baldenegro, C. X.; Band, H. R.; Barclay, G.; Bass, C. D.; Berish, D.; Bowden, N. S.; Bryan, C. D.; Cherwinka, J. J.; Chu, R.; Classen, T.; Davee, D.; Dean, D.; Deichert, G.; Dolinski, M. J.; Dolph, J.; Dwyer, D. A.; Fan, S.; Gaison, J. K.; Galindo-Uribarri, A.; Gilje, K.; Glenn, A.; Green, M.; Han, K.; Hans, S.; Heeger, K. M.; Heffroni, B.; Jaffe, D. E.; Kettell, S.; Langford, T. J.; Littlejohn, B. R.; Martinez, D.; McKeown, R. D.; Morrell, S.; Mueller, P. E.; Mumm, H. P.; Napolitano, J.; Norcini, D.; Pushin, D.; Romero, E.; Rosero, R.; Saldana, L.; Seilhan, B. S.; Sharma, R.; Stemen, N. T.; Surukuchi, P. T.; Thompson, S. J.; Varner, R. L.; Wang, W.; Watson, S. M.; White, B.; White, C.; Wilhelmi, J.; Williams, C.; Wise, T.; Yao, H.; Yeh, M.; Yen, Y. -R.; Zhang, C.; Zhang, X.

    2015-10-23

    Research reactors host a wide range of activities that make use of the intense neutron fluxes generated at these facilities. Recent interest in performing measurements with relatively low event rates, e.g. reactor antineutrino detection, at these facilities necessitates a detailed understanding of background radiation fields. Both reactor-correlated and naturally occurring background sources are potentially important, even at levels well below those of importance for typical activities. Here we describe a comprehensive series of background assessments at three high-power research reactors, including -ray, neutron, and muon measurements. For each facility we describe the characteristics and identify the sources of the background fields encountered. Furthermore, the general understanding gained of background production mechanisms and their relationship to facility features will prove valuable for the planning of any sensitive measurement conducted therein.

  5. Background radiation measurements at high power research reactors

    SciTech Connect

    Ashenfelter, J.; Yeh, M.; Balantekin, B.; Baldenegro, C. X.; Band, H. R.; Barclay, G.; Bass, C. D.; Berish, D.; Bowden, N. S.; Bryan, C. D.; Cherwinka, J. J.; Chu, R.; Classen, T.; Davee, D.; Dean, D.; Deichert, G.; Dolinski, M. J.; Dolph, J.; Dwyer, D. A.; Fan, S.; Gaison, J. K.; Galindo-Uribarri, A.; Gilje, K.; Glenn, A.; Green, M.; Han, K.; Hans, S.; Heeger, K. M.; Heffron, B.; Jaffe, D. E.; Kettell, S.; Langford, T. J.; Littlejohn, B. R.; Martinez, D.; McKeown, R. D.; Morrell, S.; Mueller, P. E.; Mumm, H. P.; Napolitano, J.; Norcini, D.; Pushin, D.; Romero, E.; Rosero, R.; Saldana, L.; Seilhan, B. S.; Sharma, R.; Stemen, N. T.; Surukuchi, P. T.; Thompson, S. J.; Varner, R. L.; Wang, W.; Watson, S. M.; White, B.; White, C.; Wilhelmi, J.; Williams, C.; Wise, T.; Yao, H.; Yen, Y. -R.; Zhang, C.; Zhang, X.

    2015-10-23

    Research reactors host a wide range of activities that make use of the intense neutron fluxes generated at these facilities. Recent interest in performing measurements with relatively low event rates, e.g. reactor antineutrino detection, at these facilities necessitates a detailed understanding of background radiation fields. Both reactor-correlated and naturally occurring background sources are potentially important, even at levels well below those of importance for typical activities. Here we describe a comprehensive series of background assessments at three high-power research reactors, including γ-ray, neutron, and muon measurements. For each facility we describe the characteristics and identify the sources of the background fields encountered. Furthermore, the general understanding gained of background production mechanisms and their relationship to facility features will prove valuable for the planning of any sensitive measurement conducted therein.

  6. Background radiation measurements at high power research reactors

    DOE PAGESBeta

    Ashenfelter, J.; Balantekin, B.; Baldenegro, C. X.; Band, H. R.; Barclay, G.; Bass, C. D.; Berish, D.; Bowden, N. S.; Bryan, C. D.; Cherwinka, J. J.; et al

    2015-10-23

    Research reactors host a wide range of activities that make use of the intense neutron fluxes generated at these facilities. Recent interest in performing measurements with relatively low event rates, e.g. reactor antineutrino detection, at these facilities necessitates a detailed understanding of background radiation fields. Both reactor-correlated and naturally occurring background sources are potentially important, even at levels well below those of importance for typical activities. Here we describe a comprehensive series of background assessments at three high-power research reactors, including -ray, neutron, and muon measurements. For each facility we describe the characteristics and identify the sources of the backgroundmore » fields encountered. Furthermore, the general understanding gained of background production mechanisms and their relationship to facility features will prove valuable for the planning of any sensitive measurement conducted therein.« less

  7. Background radiation measurements at high power research reactors

    DOE PAGESBeta

    Ashenfelter, J.; Yeh, M.; Balantekin, B.; Baldenegro, C. X.; Band, H. R.; Barclay, G.; Bass, C. D.; Berish, D.; Bowden, N. S.; Bryan, C. D.; et al

    2015-10-23

    Research reactors host a wide range of activities that make use of the intense neutron fluxes generated at these facilities. Recent interest in performing measurements with relatively low event rates, e.g. reactor antineutrino detection, at these facilities necessitates a detailed understanding of background radiation fields. Both reactor-correlated and naturally occurring background sources are potentially important, even at levels well below those of importance for typical activities. Here we describe a comprehensive series of background assessments at three high-power research reactors, including γ-ray, neutron, and muon measurements. For each facility we describe the characteristics and identify the sources of the backgroundmore » fields encountered. Furthermore, the general understanding gained of background production mechanisms and their relationship to facility features will prove valuable for the planning of any sensitive measurement conducted therein.« less

  8. Acoustical gas core reactor with MHD power generation for burst power in a bimodal system

    NASA Astrophysics Data System (ADS)

    Dugan, E. T.; Jacobs, A. M.; Oliver, C. C.; Lear, W. E., Jr.

    Research is being conducted on gas core reactors for space nuclear power to establish the scientific feasibility and engineering validation of a reactor and energy conversion system that can significantly improve specific power, dynamic performance and system efficiency. Rapid achievement of burst mode (GWe) operation at core power densities of 1 kW/mL and reactor masses of a kg/MWt are research objectives; coupled with MHD conversion, system efficiencies of 40 percent for open cycle operation and heat rejection temperatures of 1500 K or higher for closed cycle operation are anticipated. The design of the gas core reactor/MHD generator configuration to directly produce pulsed electrical power, thereby alleviating external power conditioning requirements, is also a research objective.

  9. Power flow control using distributed saturable reactors

    DOEpatents

    Dimitrovski, Aleksandar D.

    2016-02-13

    A magnetic amplifier includes a saturable core having a plurality of legs. Control windings wound around separate legs are spaced apart from each other and connected in series in an anti-symmetric relation. The control windings are configured in such a way that a biasing magnetic flux arising from a control current flowing through one of the plurality of control windings is substantially equal to the biasing magnetic flux flowing into a second of the plurality of control windings. The flow of the control current through each of the plurality of control windings changes the reactance of the saturable core reactor by driving those portions of the saturable core that convey the biasing magnetic flux in the saturable core into saturation. The phasing of the control winding limits a voltage induced in the plurality of control windings caused by a magnetic flux passing around a portion of the saturable core.

  10. The neutronics studies of fusion fission hybrid power reactor

    SciTech Connect

    Zheng Youqi; Wu Hongchun; Zu Tiejun; Yang Chao; Cao Liangzhi

    2012-06-19

    In this paper, a series of neutronics analysis of hybrid power reactor is proposed. The ideas of loading different fuels in a modular-type fission blanket is analyzed, fitting different level of fusion developments, i.e., the current experimental power output, the level can be obtained in the coming future and the high-power fusion reactor like ITER. The energy multiplication of fission blankets and tritium breeding ratio are evaluated as the criterion of design. The analysis is implemented based on the D-type simplified model, aiming to find a feasible 1000MWe hybrid power reactor for 5 years' lifetime. Three patterns are analyzed: 1) for the low fusion power, the reprocessed fuel is chosen. The fuel with high plutonium content is loaded to achieve large energy multiplication. 2) For the middle fusion power, the spent fuel from PWRs can be used to realize about 30 times energy multiplication. 3) For the high fusion power, the natural uranium can be directly used and about 10 times energy multiplication can be achieved.

  11. Design considerations for an inertial confinement fusion reactor power plant

    SciTech Connect

    Massey, J.V.; Simpson, J.E.

    1981-08-10

    To further define the engineering and economic concerns for inertial confinement fusion reactors (ICR's), a conceptual design study was performed by Bechtel Group Incorporated under the direction of Lawrence Livermore National Laboratory (LLNL). The study examined alternatives to the LLNL HYLIFE concept and expanded the previous balance of plant design to incorporate information from recent liquid metal cooled fast breeder reactor (LMFBR) power plant studies. The majority of the effort was to incorporate present laser and target physics models into a reactor design with a low coolant flowrate and a high driver repetition rate. An example of such a design is the LLNL JADE concept. In addition to producing a power plant design for LLNL using the JADE example, Bechtel has also examined the applicability of the EAGLE (Energy Absorbing Gas Lithium Ejector) concept.

  12. Safety status of space radioisotope and reactor power sources

    NASA Technical Reports Server (NTRS)

    Bennett, Gary L.

    1990-01-01

    The current overall safety criterion for both radioisotope and reactor power sources is containment or immobilization in the case of a reentry accident. In addition, reactors are designed to remain subcritical under conditions of land impact or water immersion. A very extensive safety test and analysis program was completed on the radioisotope thermoelectric generators (RTGs) in use on the Galileo spacecraft and planned for use on the Ulysses spacecraft. The results of this work show that the RTGs will pose little or no risk for any credible accident. The SP-100 space nuclear reactor program has begun addressing its safety criteria, and the design is planned to be such as to ensure meeting the various safety criteria. Preliminary mission risk analyses on SP-100 show the expected value population dose from postulated accidents on the reference mission to be very small. It is concluded that the current US nuclear power sources are the safest flown.

  13. Reactors Power Balance Based on Compact Toroid

    NASA Astrophysics Data System (ADS)

    Romadanov, I.

    2013-10-01

    The power balance of the plasma source system based on compact toroid with a pulse mode of formation is considered. Developed model takes into account the time dependence of the processes, in a pulsed mode of operation of the system. Also magnetic configuration shape and nuclei energy distribution fluency were considered. Analytical solution of Grad-Shafranov equation was taken to determine the shape of the separatrix and magnetic fields into the configuration. For practical calculation, program was written. Code is able to calculates volume power reactions in the confined plasma, using as input the geometry of the magnetic field, the cross section of reaction rates and energy distribution of the nuclei.

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

    NASA Technical Reports Server (NTRS)

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

    1969-01-01

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

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

    NASA Astrophysics Data System (ADS)

    Buden, D.

    1993-08-01

    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.

  16. Systems aspects of a space nuclear reactor power system

    NASA Technical Reports Server (NTRS)

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

    1988-01-01

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

  17. Estimates of power requirements for a Manned Mars Rover powered by a nuclear reactor

    NASA Astrophysics Data System (ADS)

    Morley, Nicholas J.; El-Genk, Mohamed S.; Cataldo, Robert; Bloomfield, Harvey

    This paper assesses the power requirement for a Manned Mars Rover vehicle. Auxiliary power needs are fulfilled using a hybrid solar photovoltaic/regenerative fuel cell system, while the primary power needs are meet using an SP-100 type reactor. The primary electric power needs, which include 30-kW(e) net user power, depend on the reactor thermal power and the efficiency of the power conversion system. Results show that an SP-100 type reactor coupled to a Free Piston Stirling Engine yields the lowest total vehicle mass and lowest specific mass for the power system. The second lowest mass was for a SP-100 reactor coupled to a Closed Brayton Cycle using He/Xe as the working fluid. The specific mass of the nuclear reactor power system, including a man-rated radiation shield, ranged from 150-kg/kW(e) to 190-kg/KW(e) and the total mass of the Rover vehicle varied depend upon the cruising speed.

  18. Estimates of power requirements for a manned Mars rover powered by a nuclear reactor

    NASA Astrophysics Data System (ADS)

    Morley, Nicholas J.; El-Genk, Mohamed S.; Cataldo, Robert; Bloomfield, Harvey

    1991-01-01

    This paper assesses the power requirement for a Manned Mars Rover vehicle. Auxiliary power needs are fulfilled using a hybrid solar photovoltaic/regenerative fuel cell system, while the primary power needs are met using an SP-100 type reactor. The primary electric power needs, which include 30-kWe net user power, depend on the reactor thermal power and the efficiency of the power conversion system. Results show that an SP-100 type reactor coupled to a Free Piston Stirling Engine (FPSE) yields the lowest total vehicle mass and lowest specific mass for the power system. The second lowest mass was for a SP-100 reactor coupled to a Closed Brayton Cycle (CBC) using He/Xe as the working fluid. The specific mass of the nuclear reactor power systrem, including a man-rated radiation shield, ranged from 150-kg/kWe to 190-kg/kWe and the total mass of the Rover vehicle varied depend upon the cruising speed.

  19. Estimates of power requirements for a Manned Mars Rover powered by a nuclear reactor

    NASA Technical Reports Server (NTRS)

    Morley, Nicholas J.; El-Genk, Mohamed S.; Cataldo, Robert; Bloomfield, Harvey

    1991-01-01

    This paper assesses the power requirement for a Manned Mars Rover vehicle. Auxiliary power needs are fulfilled using a hybrid solar photovoltaic/regenerative fuel cell system, while the primary power needs are meet using an SP-100 type reactor. The primary electric power needs, which include 30-kW(e) net user power, depend on the reactor thermal power and the efficiency of the power conversion system. Results show that an SP-100 type reactor coupled to a Free Piston Stirling Engine yields the lowest total vehicle mass and lowest specific mass for the power system. The second lowest mass was for a SP-100 reactor coupled to a Closed Brayton Cycle using He/Xe as the working fluid. The specific mass of the nuclear reactor power system, including a man-rated radiation shield, ranged from 150-kg/kW(e) to 190-kg/KW(e) and the total mass of the Rover vehicle varied depend upon the cruising speed.

  20. A Gas-Cooled Reactor Surface Power System

    SciTech Connect

    Harms, G.A.; Lenard, R.X.; Lipinski, R.J.; Wright, S.A.

    1998-11-09

    A human outpost on Mars requires plentiful power to assure survival of the astronauts. Anywhere from 50 to 500 kW of electric power (kWe) will be needed, depending on the number of astronauts, level of scientific activity, and life- cycle closure desired. This paper describes a 250-kWe power system based on a gas-cooled nuclear reactor with a recuperated closed Brayton cycle conversion system. The design draws upon the extensive data and engineering experience developed under the various high-temperature gas cooled reactor programs and under the SP-100 program. The reactor core is similar in power and size to the research reactors found on numerous university campuses. The fuel is uranium nitide clad in Nb 1 %Zr, which has been extensively tested under the SP-I 00 program The fiel rods are arranged in a hexagonal array within a BeO block. The BeO softens the spectrum, allowing better use of the fbel and stabilizing the geometty against deformation during impact or other loadings. The system has a negative temperature feedback coefficient so that the power level will automatically follow a variable load without the need for continuous adjustment of control elements. Waste heat is removed by an air-cooled heat exchanger using cold Martian air. The amount of radioactivity in the reactor at launch is very small (less than a Curie, and about equal to a truckload of uranium ore). The system will need to be engineered so that criticality cannot occur for any launch accident. This system is also adaptable for electric propulsion or life-support during transit to and from Mars.

  1. A gas-cooled reactor surface power system

    NASA Astrophysics Data System (ADS)

    Lipinski, Ronald J.; Wright, Steven A.; Lenard, Roger X.; Harms, Gary A.

    1999-01-01

    A human outpost on Mars requires plentiful power to assure survival of the astronauts. Anywhere from 50 to 500 kW of electric power (kWe) will be needed, depending on the number of astronauts, level of scientific activity, and life-cycle closure desired. This paper describes a 250-kWe power system based on a gas-cooled nuclear reactor with a recuperated closed Brayton cycle conversion system. The design draws upon the extensive data and engineering experience developed under the various high-temperature gas cooled reactor programs and under the SP-100 program. The reactor core is similar in power and size to the research reactors found on numerous university campuses. The fuel is uranium nitride clad in Nb1%Zr, which has been extensively tested under the SP-100 program. The fuel rods are arranged in a hexagonal array within a BeO block. The BeO softens the spectrum, allowing better use of the fuel and stabilizing the geometry against deformation during impact or other loadings. The system has a negative temperature feedback coefficient so that the power level will automatically follow a variable load without the need for continuous adjustment of control elements. Waste heat is removed by an air-cooled heat exchanger using cold Martian air. The amount of radioactivity in the reactor at launch is very small (less than a Curie, and about equal to a truckload of uranium ore). The system will need to be engineered so that criticality can not occur for any launch accident. This system is also adaptable for electric propulsion or life-support during transit to and from Mars.

  2. A gas-cooled reactor surface power system

    SciTech Connect

    Lipinski, R.J.; Wright, S.A.; Lenard, R.X.; Harms, G.A.

    1999-01-01

    A human outpost on Mars requires plentiful power to assure survival of the astronauts. Anywhere from 50 to 500 kW of electric power (kWe) will be needed, depending on the number of astronauts, level of scientific activity, and life-cycle closure desired. This paper describes a 250-kWe power system based on a gas-cooled nuclear reactor with a recuperated closed Brayton cycle conversion system. The design draws upon the extensive data and engineering experience developed under the various high-temperature gas cooled reactor programs and under the SP-100 program. The reactor core is similar in power and size to the research reactors found on numerous university campuses. The fuel is uranium nitride clad in Nb1{percent}Zr, which has been extensively tested under the SP-100 program. The fuel rods are arranged in a hexagonal array within a BeO block. The BeO softens the spectrum, allowing better use of the fuel and stabilizing the geometry against deformation during impact or other loadings. The system has a negative temperature feedback coefficient so that the power level will automatically follow a variable load without the need for continuous adjustment of control elements. Waste heat is removed by an air-cooled heat exchanger using cold Martian air. The amount of radioactivity in the reactor at launch is very small (less than a Curie, and about equal to a truckload of uranium ore). The system will need to be engineered so that criticality can not occur for any launch accident. This system is also adaptable for electric propulsion or life-support during transit to and from Mars. {copyright} {ital 1999 American Institute of Physics.}

  3. A gas-cooled reactor surface power system

    SciTech Connect

    Lipinski, Ronald J.; Wright, Steven A.; Lenard, Roger X.; Harms, Gary A.

    1999-01-22

    A human outpost on Mars requires plentiful power to assure survival of the astronauts. Anywhere from 50 to 500 kW of electric power (kWe) will be needed, depending on the number of astronauts, level of scientific activity, and life-cycle closure desired. This paper describes a 250-kWe power system based on a gas-cooled nuclear reactor with a recuperated closed Brayton cycle conversion system. The design draws upon the extensive data and engineering experience developed under the various high-temperature gas cooled reactor programs and under the SP-100 program. The reactor core is similar in power and size to the research reactors found on numerous university campuses. The fuel is uranium nitride clad in Nb1%Zr, which has been extensively tested under the SP-100 program. The fuel rods are arranged in a hexagonal array within a BeO block. The BeO softens the spectrum, allowing better use of the fuel and stabilizing the geometry against deformation during impact or other loadings. The system has a negative temperature feedback coefficient so that the power level will automatically follow a variable load without the need for continuous adjustment of control elements. Waste heat is removed by an air-cooled heat exchanger using cold Martian air. The amount of radioactivity in the reactor at launch is very small (less than a Curie, and about equal to a truckload of uranium ore). The system will need to be engineered so that criticality can not occur for any launch accident. This system is also adaptable for electric propulsion or life-support during transit to and from Mars.

  4. High density operation for reactor-relevant power exhaust

    NASA Astrophysics Data System (ADS)

    Wischmeier, M.

    2015-08-01

    With increasing size of a tokamak device and associated fusion power gain an increasing power flux density towards the divertor needs to be handled. A solution for handling this power flux is crucial for a safe and economic operation. Using purely geometric arguments in an ITER-like divertor this power flux can be reduced by approximately a factor 100. Based on a conservative extrapolation of current technology for an integrated engineering approach to remove power deposited on plasma facing components a further reduction of the power flux density via volumetric processes in the plasma by up to a factor of 50 is required. Our current ability to interpret existing power exhaust scenarios using numerical transport codes is analyzed and an operational scenario as a potential solution for ITER like divertors under high density and highly radiating reactor-relevant conditions is presented. Alternative concepts for risk mitigation as well as strategies for moving forward are outlined.

  5. Gravity Scaling of a Power Reactor Water Shield

    SciTech Connect

    Reid, Robert S.; Pearson, J. Boise

    2008-01-21

    Water based reactor shielding is being considered as an affordable option for potential use on initial lunar surface reactor power systems. Heat dissipation in the shield from nuclear sources must be rejected by an auxillary thermal hydraulic cooling system. The mechanism for transferring heat through the shield is natural convection between the core surface and an array of thermosyphon radiator elements. Natural convection in a 100 kWt lunar surface reactor shield design has been previously evaluated at lower power levels (Pearson, 2006). The current baseline assumes that 5.5 kW are dissipated in the water shield, the preponderance on the core surface, but with some volumetric heating in the naturally circulating water as well. This power is rejected by a radiator located above the shield with a surface temperature of 370 K. A similarity analysis on a water-based reactor shield is presented examining the effect of gravity on free convection between a radiation shield inner vessel and a radiation shield outer vessel boundaries. Two approaches established similarity: 1) direct scaling of Rayleigh number equates gravity-surface heat flux products, 2) temperature difference between the wall and thermal boundary layer held constant on Earth and the Moon. Nussult number for natural convection (laminar and turbulent) is assumed of form Nu = CRa{sup n}. These combined results estimate similarity conditions under Earth and Lunar gravities. The influence of reduced gravity on the performance of thermosyphon heat pipes is also examined.

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

    SciTech Connect

    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

  7. Operating margin of Soviet RBMK-1000 nuclear power reactors

    SciTech Connect

    Adams, J.M.; Robinson, G.E. . Dept. of Nuclear Engineering); Hochreiter, L.E. )

    1991-12-01

    This paper reports on a coupled thermal- hydraulic analysis that is performed for the Soviet-designed RBMK-1000 nuclear power reactor to assess the operating margin to critical heat flux (CHF); the Chernobyl-4 reactor serves as the principal model for this study. Calculations are performed using a simplified subchannel analysis. The overall analysis involves an iterative search to determine the individual subchannel flow rates, and a boiling transition analysis is performed to obtain a measure of the core operating margin. The operating margin is determined via two distinct methods. The first involves a calculation of the core critical power ratio (CPR) using an empirically derived correlation that the Soviets developed expressly for the RBMK-1000. Additionally, various subchannel CHF correlations typical of those used in the design of nuclear-powered reactors in the United States are also employed. When the Soviet critical power correlation is used, the calculations carried out for both normal operating and reference overpower conditions result in CPRs of 1.115 and 1.019, respectively. In most cases, the subchannel CHF correlations indicate that additional operating margin over that calculated by the Soviet critical power correlation exists for this design.

  8. TFE fast driver reactor system for low-power applications

    NASA Astrophysics Data System (ADS)

    Van Hagan, Thomas H.; Lewis, Bryan R.; Bellis, Elizabeth A.; Fisher, Mike V.

    1992-01-01

    This paper addresses reactor design considerations for an in-core thermionic system concept proposed for emerging space power applications with electric power requirements in the 10 to 50-kWe range. At this power level an in-core thermionic rector core requires a combination of thermionic fuel elements (TFEs) and driver fuel elements to achieve nuclear criticality. A pumped liquid-metal loop cools the reactor, transporting the reject heat to a heat pipe radiator. The system concept is a straightforward derivative of the thermionic 100-kWe system designed during the SP-100 Phase 1 program. Combining existing thermionic technology with LMFBR fuel technology and pumped-loop waste heat removal defines a concept that has the advantages of reliability, scalability, technology maturity, and design flexibility.

  9. Space reactor power 1986 - A year of choices and transition

    NASA Technical Reports Server (NTRS)

    Wiley, R. L.; Verga, R. L.; Schnyer, A. D.; Sholtis, J. A., Jr.; Wahlquist, E. J.

    1986-01-01

    Both the SP-100 and Multimegawatt programs have made significant progress over the last year and that progress is the focus of this paper. In the SP-100 program the thermoelectric energy conversion concept powered by a compact, high-temperature, lithium-cooled, uranium-nitride-fueled fast spectrum reactor was selected for engineering development and ground demonstration testing at an electrical power level of 300 kilowatts. In the Multimegawatt program, activities moved from the planning phase into one of technology development and assessment with attendant preliminary definition and evaluation of power concepts against requirements of the Strategic Defense Initiative.

  10. 77 FR 8902 - Draft Regulatory Guide: Issuance, Availability Decommissioning of Nuclear Power Reactors

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-02-15

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

  11. 10 CFR 50.72 - Immediate notification requirements for operating nuclear power reactors.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... power reactors. 50.72 Section 50.72 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF... notification requirements for operating nuclear power reactors. (a) General requirements. 1 (1) Each nuclear... requirements for immediate notification of the NRC by licensed operating nuclear power reactors are...

  12. 10 CFR 50.72 - Immediate notification requirements for operating nuclear power reactors.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... power reactors. 50.72 Section 50.72 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF... notification requirements for operating nuclear power reactors. (a) General requirements. 1 (1) Each nuclear... requirements for immediate notification of the NRC by licensed operating nuclear power reactors are...

  13. 10 CFR 50.72 - Immediate notification requirements for operating nuclear power reactors.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... power reactors. 50.72 Section 50.72 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF... notification requirements for operating nuclear power reactors. (a) General requirements. 1 (1) Each nuclear... requirements for immediate notification of the NRC by licensed operating nuclear power reactors are...

  14. 10 CFR 73.58 - Safety/security interface requirements for nuclear power reactors.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... 10 Energy 2 2013-01-01 2013-01-01 false Safety/security interface requirements for nuclear power reactors. 73.58 Section 73.58 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PHYSICAL PROTECTION OF... requirements for nuclear power reactors. (a) Each operating nuclear power reactor licensee with a...

  15. 10 CFR 73.58 - Safety/security interface requirements for nuclear power reactors.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

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

  16. 10 CFR 73.58 - Safety/security interface requirements for nuclear power reactors.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

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

  17. 10 CFR 73.58 - Safety/security interface requirements for nuclear power reactors.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... 10 Energy 2 2014-01-01 2014-01-01 false Safety/security interface requirements for nuclear power reactors. 73.58 Section 73.58 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PHYSICAL PROTECTION OF... requirements for nuclear power reactors. (a) Each operating nuclear power reactor licensee with a...

  18. 10 CFR 73.58 - Safety/security interface requirements for nuclear power reactors.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... 10 Energy 2 2012-01-01 2012-01-01 false Safety/security interface requirements for nuclear power reactors. 73.58 Section 73.58 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PHYSICAL PROTECTION OF... requirements for nuclear power reactors. (a) Each operating nuclear power reactor licensee with a...

  19. Enabling autonomous control for space reactor power systems

    SciTech Connect

    Wood, R. T.

    2006-07-01

    The application of nuclear reactors for space power and/or propulsion presents some unique challenges regarding the operations and control of the power system. Terrestrial nuclear reactors employ varying degrees of human control and decision-making for operations and benefit from periodic human interaction for maintenance. In contrast, the control system of a space reactor power system (SRPS) employed for deep space missions must be able to accommodate unattended operations due to communications delays and periods of planetary occlusion while adapting to evolving or degraded conditions with no opportunity for repair or refurbishment. Thus, a SRPS control system must provide for operational autonomy. Oak Ridge National Laboratory (ORNL) has conducted an investigation of the state of the technology for autonomous control to determine the experience base in the nuclear power application domain, both for space and terrestrial use. It was found that control systems with varying levels of autonomy have been employed in robotic, transportation, spacecraft, and manufacturing applications. However, autonomous control has not been implemented for an operating terrestrial nuclear power plant nor has there been any experience beyond automating simple control loops for space reactors. Current automated control technologies for nuclear power plants are reasonably mature, and basic control for a SRPS is clearly feasible under optimum circumstances. However, autonomous control is primarily intended to account for the non optimum circumstances when degradation, failure, and other off-normal events challenge the performance of the reactor and near-term human intervention is not possible. Thus, the development and demonstration of autonomous control capabilities for the specific domain of space nuclear power operations is needed. This paper will discuss the findings of the ORNL study and provide a description of the concept of autonomy, its key characteristics, and a prospective

  20. Power Conversion Study for High Temperature Gas-Cooled Reactors

    SciTech Connect

    Chang Oh; Richard Moore; Robert Barner

    2005-05-01

    The Idaho National Laboratory (INL) is investigating a Brayton cycle efficiency improvement on a high temperature gas-cooled reactor (HTGR) as part of Generation-IV nuclear engineering research initiative. There are some technical issues to be resolved before the selection of the final design of the high temperature gascooled reactor, called as a Next Generation Nuclear Plant (NGNP), which is supposed to be built at the INEEL by year 2017. The technical issues are the selection of the working fluid, direct vs. indirect cycle, power cycle type, the optimized design in terms of a number of intercoolers, and others. In this paper, we investigated a number of working fluids for the power conversion loop, direct versus indirect cycle, the effect of intercoolers, and other thermal hydraulics issues. However, in this paper, we present part of the results we have obtained. HYSYS computer code was used along with a computer model developed using Visual Basic computer language.

  1. 10 CFR Appendix J to Part 50 - Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... Leakage Testing for Water-Cooled Power Reactors This appendix includes two options, A and B, either of..., and systems and components which penetrate containment of water-cooled power reactors, and establish... and feedwater piping and other systems which penetrate containment of direct-cycle boiling water...

  2. 10 CFR Appendix J to Part 50 - Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... Leakage Testing for Water-Cooled Power Reactors This appendix includes two options, A and B, either of..., and systems and components which penetrate containment of water-cooled power reactors, and establish... and feedwater piping and other systems which penetrate containment of direct-cycle boiling water...

  3. 10 CFR Appendix J to Part 50 - Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... Leakage Testing for Water-Cooled Power Reactors This appendix includes two options, A and B, either of..., and systems and components which penetrate containment of water-cooled power reactors, and establish... and feedwater piping and other systems which penetrate containment of direct-cycle boiling water...

  4. 10 CFR Appendix J to Part 50 - Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... Leakage Testing for Water-Cooled Power Reactors This appendix includes two options, A and B, either of..., and systems and components which penetrate containment of water-cooled power reactors, and establish... and feedwater piping and other systems which penetrate containment of direct-cycle boiling water...

  5. 10 CFR Appendix J to Part 50 - Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... Leakage Testing for Water-Cooled Power Reactors This appendix includes two options, A and B, either of..., and systems and components which penetrate containment of water-cooled power reactors, and establish... and feedwater piping and other systems which penetrate containment of direct-cycle boiling water...

  6. Technological implications of SNAP reactor power system development on future space nuclear power systems

    SciTech Connect

    Anderson, R.V.

    1982-11-16

    Nuclear reactor systems are one method of satisfying space mission power needs. The development of such systems must proceed on a path consistent with mission needs and schedules. This path, or technology roadmap, starts from the power system technology data base available today. Much of this data base was established during the 1960s and early 1970s, when government and industry developed space nuclear reactor systems for steady-state power and propulsion. One of the largest development programs was the Systems for Nuclear Auxiliary Power (SNAP) Program. By the early 1970s, a technology base had evolved from this program at the system, subsystem, and component levels. There are many implications of this technology base on future reactor power systems. A review of this base highlights the need for performing a power system technology and mission overview study. Such a study is currently being performed by Rockwell's Energy Systems Group for the Department of Energy and will assess power system capabilities versus mission needs, considering development, schedule, and cost implications. The end product of the study will be a technology roadmap to guide reactor power system development.

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

    SciTech Connect

    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.

  8. Autonomous Control Capabilities for Space Reactor Power Systems

    NASA Astrophysics Data System (ADS)

    Wood, Richard T.; Neal, John S.; Brittain, C. Ray; Mullens, James A.

    2004-02-01

    The National Aeronautics and Space Administration's (NASA's) Project Prometheus, the Nuclear Systems Program, is investigating a possible Jupiter Icy Moons Orbiter (JIMO) mission, which would conduct in-depth studies of three of the moons of Jupiter by using a space reactor power system (SRPS) to provide energy for propulsion and spacecraft power for more than a decade. Terrestrial nuclear power plants rely upon varying degrees of direct human control and interaction for operations and maintenance over a forty to sixty year lifetime. In contrast, an SRPS is intended to provide continuous, remote, unattended operation for up to fifteen years with no maintenance. Uncertainties, rare events, degradation, and communications delays with Earth are challenges that SRPS control must accommodate. Autonomous control is needed to address these challenges and optimize the reactor control design. In this paper, we describe an autonomous control concept for generic SRPS designs. The formulation of an autonomous control concept, which includes identification of high-level functional requirements and generation of a research and development plan for enabling technologies, is among the technical activities that are being conducted under the U.S. Department of Energy's Space Reactor Technology Program in support of the NASA's Project Prometheus. The findings from this program are intended to contribute to the successful realization of the JIMO mission.

  9. Autonomous Control Capabilities for Space Reactor Power Systems

    SciTech Connect

    Wood, Richard T.; Neal, John S.; Brittain, C. Ray; Mullens, James A.

    2004-02-04

    The National Aeronautics and Space Administration's (NASA's) Project Prometheus, the Nuclear Systems Program, is investigating a possible Jupiter Icy Moons Orbiter (JIMO) mission, which would conduct in-depth studies of three of the moons of Jupiter by using a space reactor power system (SRPS) to provide energy for propulsion and spacecraft power for more than a decade. Terrestrial nuclear power plants rely upon varying degrees of direct human control and interaction for operations and maintenance over a forty to sixty year lifetime. In contrast, an SRPS is intended to provide continuous, remote, unattended operation for up to fifteen years with no maintenance. Uncertainties, rare events, degradation, and communications delays with Earth are challenges that SRPS control must accommodate. Autonomous control is needed to address these challenges and optimize the reactor control design. In this paper, we describe an autonomous control concept for generic SRPS designs. The formulation of an autonomous control concept, which includes identification of high-level functional requirements and generation of a research and development plan for enabling technologies, is among the technical activities that are being conducted under the U.S. Department of Energy's Space Reactor Technology Program in support of the NASA's Project Prometheus. The findings from this program are intended to contribute to the successful realization of the JIMO mission.

  10. Gravity Scaling of a Power Reactor Water Shield

    NASA Technical Reports Server (NTRS)

    Reid, Robert S.; Pearson, J. Boise

    2008-01-01

    Water based reactor shielding is being considered as an affordable option for use on initial lunar surface power systems. Heat dissipation in the shield from nuclear sources must be rejected by an auxiliary thermal hydraulic cooling system. The mechanism for transferring heat through the shield is natural convection between the core surface and an array of thermosyphon radiator elements. Natural convection in a 100 kWt lunar surface reactor shield design has been previously evaluated at lower power levels (Pearson, 2007). The current baseline assumes that 5.5 kW are dissipated in the water shield, the preponderance on the core surface, but with some volumetric heating in the naturally circulating water as well. This power is rejected by a radiator located above the shield with a surface temperature of 370 K. A similarity analysis on a water-based reactor shield is presented examining the effect of gravity on free convection between a radiation shield inner vessel and a radiation shield outer vessel boundaries. Two approaches established similarity: 1) direct scaling of Rayleigh number equates gravity-surface heat flux products, 2) temperature difference between the wall and thermal boundary layer held constant on Earth and the Moon. Nussult number for natural convection (laminar and turbulent) is assumed of form Nu = CRa(sup n). These combined results estimate similarity conditions under Earth and Lunar gravities. The influence of reduced gravity on the performance of thermosyphon heat pipes is also examined.

  11. Supercritical Water Reactor Cycle for Medium Power Applications

    SciTech Connect

    BD Middleton; J Buongiorno

    2007-04-25

    Scoping studies for a power conversion system based on a direct-cycle supercritical water reactor have been conducted. The electric power range of interest is 5-30 MWe with a design point of 20 MWe. The overall design objective is to develop a system that has minimized physical size and performs satisfactorily over a broad range of operating conditions. The design constraints are as follows: Net cycle thermal efficiency {ge}20%; Steam turbine outlet quality {ge}90%; and Pumping power {le}2500 kW (at nominal conditions). Three basic cycle configurations were analyzed. Listed in order of increased plant complexity, they are: (1) Simple supercritical Rankine cycle; (2) All-supercritical Brayton cycle; and (3) Supercritical Rankine cycle with feedwater preheating. The sensitivity of these three configurations to various parameters, such as reactor exit temperature, reactor pressure, condenser pressure, etc., was assessed. The Thermoflex software package was used for this task. The results are as follows: (a) The simple supercritical Rankine cycle offers the greatest hardware simplification, but its high reactor temperature rise and reactor outlet temperature may pose serious problems from the viewpoint of thermal stresses, stability and materials in the core. (b) The all-supercritical Brayton cycle is not a contender, due to its poor thermal efficiency. (c) The supercritical Rankine cycle with feedwater preheating affords acceptable thermal efficiency with lower reactor temperature rise and outlet temperature. (d) The use of a moisture separator improves the performance of the supercritical Rankine cycle with feedwater preheating and allows for a further reduction of the reactor outlet temperature, thus it was selected for the next step. Preliminary engineering design of the supercritical Rankine cycle with feedwater preheating and moisture separation was performed. All major components including the turbine, feedwater heater, feedwater pump, condenser, condenser pump

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

    SciTech Connect

    M. L. Grossbeck J-P.A. Renier Tim Bigelow

    2003-09-30

    Burnable poisons are used in nuclear reactors to produce a more level distribution of power in the reactor core and to reduce to necessity for a large control system. An ideal burnable poison would burn at the same rate as the fuel. In this study, separation of neutron-absorbing isotopes was investigated in order to eliminate isotopes that remain as absorbers at the end of fuel life, thus reducing useful fuel life. The isotopes Gd-157, Dy-164, and Er-167 were found to have desirable properties. These isotopes were separated from naturally occurring elements by means of plasma separation to evaluate feasibility and cost. It was found that pure Gd-157 could save approximately $6 million at the end of four years. However, the cost of separation, using the existing facility, made separation cost- ineffective. Using a magnet with three times the field strength is expected to reduce the cost by a factor of ten, making isotopically separated burnable poisons a favorable method of increasing fuel life in commercial reactors, in particular Generation-IV reactors. The project also investigated various burnable poison configurations, and studied incorporation of metallic burnable poisons into fuel cladding.

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

    SciTech Connect

    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

  14. Inertial fusion energy power reactor fuel recovery system

    SciTech Connect

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

    2008-07-15

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

  15. Fast reactor power plant design having heat pipe heat exchanger

    DOEpatents

    Huebotter, P.R.; McLennan, G.A.

    1984-08-30

    The invention relates to a pool-type fission reactor power plant design having a reactor vessel containing a primary coolant (such as liquid sodium), and a steam expansion device powered by a pressurized water/steam coolant system. Heat pipe means are disposed between the primary and water coolants to complete the heat transfer therebetween. The heat pipes are vertically oriented, penetrating the reactor deck and being directly submerged in the primary coolant. A U-tube or line passes through each heat pipe, extended over most of the length of the heat pipe and having its walls spaced from but closely proximate to and generally facing the surrounding walls of the heat pipe. The water/steam coolant loop includes each U-tube and the steam expansion device. A heat transfer medium (such as mercury) fills each of the heat pipes. The thermal energy from the primary coolant is transferred to the water coolant by isothermal evaporation-condensation of the heat transfer medium between the heat pipe and U-tube walls, the heat transfer medium moving within the heat pipe primarily transversely between these walls.

  16. Fast reactor power plant design having heat pipe heat exchanger

    DOEpatents

    Huebotter, Paul R.; McLennan, George A.

    1985-01-01

    The invention relates to a pool-type fission reactor power plant design having a reactor vessel containing a primary coolant (such as liquid sodium), and a steam expansion device powered by a pressurized water/steam coolant system. Heat pipe means are disposed between the primary and water coolants to complete the heat transfer therebetween. The heat pipes are vertically oriented, penetrating the reactor deck and being directly submerged in the primary coolant. A U-tube or line passes through each heat pipe, extended over most of the length of the heat pipe and having its walls spaced from but closely proximate to and generally facing the surrounding walls of the heat pipe. The water/steam coolant loop includes each U-tube and the steam expansion device. A heat transfer medium (such as mercury) fills each of the heat pipes. The thermal energy from the primary coolant is transferred to the water coolant by isothermal evaporation-condensation of the heat transfer medium between the heat pipe and U-tube walls, the heat transfer medium moving within the heat pipe primarily transversely between these walls.

  17. Advanced thermionic reactors for surface nuclear power applications

    NASA Astrophysics Data System (ADS)

    Parlos, Alexander G.; Kent, Karl; Peddicord, Kenneth L.; Khan, Ehsan U.

    1991-09-01

    A preliminary feasibility study on a new concept for a highly compact space reactor power system is presented, consisting of in-core thermionic fuel elements and in-core heat pipes for passive core cooling. The reference fuel considered in this study is uranium carbide. The calculations reported include a neutronic design analysis using a 2D neutron transport model, as well as a simplified 1D thermal analysis of the reactor core, using a preliminary thermal sizing of the in-core heat pipes. Initial results indicate that the proposed core design is thermally and neutronically feasible, with a maximum steady-state fuel temperature below 2000 K. Alternate advanced fuels, such as various oxides of Am-242, result in exceedingly high fuel centerline temperatures because of the associated low thermal conductivities.

  18. Issues in the flight qualification of a space power reactor

    SciTech Connect

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

    1994-10-01

    This paper presents an overview of the Nuclear Electric Propulsion Space Test Program (NEPSTP). The program goals, the proposed mission, the spacecraft, and the Topaz II space nuclear power system are described. The subject of flight qualification is examined and the inherent difficulties of qualifying a space reactor are described. The differences between US and Russian flight qualification procedures are explored. A plan is then described that was developed to determine an appropriate flight qualification program for the Topaz II reactor to support a possible NEPSTP launch. Refocusing of the activities of the Ballistic Missile Defense Organization (BMDO), combined with budgetary pressures, forced the cancellation of the NEPSTP at the end of the 1993 fiscal year.

  19. PR-EDB: Power Reactor Embrittlement Database - Version 3

    SciTech Connect

    Wang, Jy-An John; Subramani, Ranjit

    2008-03-01

    The aging and degradation of light-water reactor pressure vessels is of particular concern because of their relevance to plant integrity and the magnitude of the expected irradiation embrittlement. The radiation embrittlement of reactor pressure vessel materials depends on many factors, such as neutron fluence, flux, and energy spectrum, irradiation temperature, and preirradiation material history and chemical compositions. These factors must be considered to reliably predict pressure vessel embrittlement and to ensure the safe operation of the reactor. Large amounts of data from surveillance capsules are needed to develop a generally applicable damage prediction model that can be used for industry standards and regulatory guides. Furthermore, the investigations of regulatory issues such as vessel integrity over plant life, vessel failure, and sufficiency of current codes, Standard Review Plans (SRPs), and Guides for license renewal can be greatly expedited by the use of a well-designed computerized database. The Power Reactor Embrittlement Database (PR-EDB) is such a comprehensive collection of data for U.S. designed commercial nuclear reactors. The current version of the PR-EDB lists the test results of 104 heat-affected-zone (HAZ) materials, 115 weld materials, and 141 base materials, including 103 plates, 35 forgings, and 3 correlation monitor materials that were irradiated in 321 capsules from 106 commercial power reactors. The data files are given in dBASE format and can be accessed with any personal computer using the Windows operating system. "User-friendly" utility programs have been written to investigate radiation embrittlement using this database. Utility programs allow the user to retrieve, select and manipulate specific data, display data to the screen or printer, and fit and plot Charpy impact data. The PR-EDB Version 3.0 upgrades Version 2.0. The package was developed based on the Microsoft .NET framework technology and uses Microsoft Access for

  20. Hydrogen Mitigation Strategy of the APR1400 Nuclear Power Plant for a Hypothetical Station Blackout Accident

    SciTech Connect

    Kim, Jongtae; Hong, Seong-Wan; Kim, Sang-Baik; Kim, Hee-Dong

    2005-06-15

    In order to analyze the hydrogen distribution during a hypothetical station blackout accident in the Korean next-generation Advanced Power Reactor 1400 (APR1400) containment, the three-dimensional computational fluid dynamics code GASFLOW was used. The source of the hydrogen and steam for the GASFLOW analysis was obtained from a MAAP calculation. The discharged water, steam, and hydrogen from the pressurizer are released into the water of the in-containment refueling water storage tank (IRWST). Most of the discharged steam is condensed in the IRWST water because of its subcooling, and dry hydrogen is released into the free volume of the IRWST; finally, it goes out to the annular compartment above the IRWST through the vent holes. From the GASFLOW analysis, it was found that the gas mixture in the IRWST becomes quickly nonflammable by oxygen starvation but the hydrogen is accumulated in the annular compartment because of the narrow ventilation gap between the operating deck and containment wall when the igniters installed in the IRWST are not operated. When the igniters installed in the APR1400 were turned on, a short period of burning occurred in the IRWST, and then the flame was extinguished by the oxygen starvation in the IRWST. The unburned hydrogen was released into the annular compartment and went up to the dome because no igniters are installed around the annular compartment in the base design of the APR1400. From this result, it could be concluded that the control of the hydrogen concentration is difficult for the base design. In this study design modifications are proposed and evaluated with GASFLOW in view of the hydrogen mitigation strategy.

  1. REACTOR

    DOEpatents

    Roman, W.G.

    1961-06-27

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

  2. Computer study of emergency shutdowns of a 60-kilowatt reactor Brayton space power system

    NASA Technical Reports Server (NTRS)

    Tew, R. C.; Jefferies, K. S.

    1974-01-01

    A digital computer study of emergency shutdowns of a 60-kWe reactor Brayton power system was conducted. Malfunctions considered were (1) loss of reactor coolant flow, (2) loss of Brayton system gas flow, (3)turbine overspeed, and (4) a reactivity insertion error. Loss of reactor coolant flow was the most serious malfunction for the reactor. Methods for moderating the reactor transients due to this malfunction are considered.

  3. Small space reactor power systems for unmanned solar system exploration missions

    NASA Technical Reports Server (NTRS)

    Bloomfield, Harvey S.

    1987-01-01

    A preliminary feasibility study of the application of small nuclear reactor space power systems to the Mariner Mark II Cassini spacecraft/mission was conducted. The purpose of the study was to identify and assess the technology and performance issues associated with the reactor power system/spacecraft/mission integration. The Cassini mission was selected because study of the Saturn system was identified as a high priority outer planet exploration objective. Reactor power systems applied to this mission were evaluated for two different uses. First, a very small 1 kWe reactor power system was used as an RTG replacement for the nominal spacecraft mission science payload power requirements while still retaining the spacecraft's usual bipropellant chemical propulsion system. The second use of reactor power involved the additional replacement of the chemical propulsion system with a small reactor power system and an electric propulsion system. The study also provides an examination of potential applications for the additional power available for scientific data collection. The reactor power system characteristics utilized in the study were based on a parametric mass model that was developed specifically for these low power applications. The model was generated following a neutronic safety and operational feasibility assessment of six small reactor concepts solicited from U.S. industry. This assessment provided the validation of reactor safety for all mission phases and generatad the reactor mass and dimensional data needed for the system mass model.

  4. Small space reactor power systems for unmanned solar system exploration missions

    SciTech Connect

    Bloomfield, H.S.

    1987-12-01

    A preliminary feasibility study of the application of small nuclear reactor space power systems to the Mariner Mark II Cassini spacecraft/mission was conducted. The purpose of the study was to identify and assess the technology and performance issues associated with the reactor power system/spacecraft/mission integration. The Cassini mission was selected because study of the Saturn system was identified as a high priority outer planet exploration objective. Reactor power systems applied to this mission were evaluated for two different uses. First, a very small 1 kWe reactor power system was used as an RTG replacement for the nominal spacecraft mission science payload power requirements while still retaining the spacecraft's usual bipropellant chemical propulsion system. The second use of reactor power involved the additional replacement of the chemical propulsion system with a small reactor power system and an electric propulsion system. The study also provides an examination of potential applications for the additional power available for scientific data collection. The reactor power system characteristics utilized in the study were based on a parametric mass model that was developed specifically for these low power applications. The model was generated following a neutronic safety and operational feasibility assessment of six small reactor concepts solicited from U.S. industry. This assessment provided the validation of reactor safety for all mission phases and generatad the reactor mass and dimensional data needed for the system mass model.

  5. Neutron dose estimation in a zero power nuclear reactor

    NASA Astrophysics Data System (ADS)

    Triviño, S.; Vedelago, J.; Cantargi, F.; Keil, W.; Figueroa, R.; Mattea, F.; Chautemps, A.; Santibañez, M.; Valente, M.

    2016-10-01

    This work presents the characterization and contribution of neutron and gamma components to the absorbed dose in a zero power nuclear reactor. A dosimetric method based on Fricke gel was implemented to evaluate the separation between dose components in the mixed field. The validation of this proposed method was performed by means of direct measurements of neutron flux in different positions using Au and Mg-Ni activation foils. Monte Carlo simulations were conversely performed using the MCNP main code with a dedicated subroutine to incorporate the exact complete geometry of the nuclear reactor facility. Once nuclear fuel elements were defined, the simulations computed the different contributions to the absorbed dose in specific positions inside the core. Thermal/epithermal contributions of absorbed dose were assessed by means of Fricke gel dosimetry using different isotopic compositions aimed at modifying the sensitivity of the dosimeter for specific dose components. Clear distinctions between gamma and neutron capture dose were obtained. Both Monte Carlo simulations and experimental results provided reliable estimations about neutron flux rate as well as dose rate during the reactor operation. Simulations and experimental results are in good agreement in every positions measured and simulated in the core.

  6. 75 FR 79423 - In the Matter of All Power Reactor Licensees and Research Reactor Licensees Who Transport Spent...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-12-20

    ... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY... All Power Reactor Licensees and Research Reactor Licensees Who Transport Spent Nuclear Fuel; Order... issued a specific license by the U.S. Nuclear Regulatory Commission (NRC or Commission) authorizing...

  7. 75 FR 70042 - In the Matter of All Power Reactor Licensees and Research Reactor Licensees Who Transport Spent...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-11-16

    ... From the Federal Register Online via the Government Publishing Office ] NUCLEAR REGULATORY... All Power Reactor Licensees and Research Reactor Licensees Who Transport Spent Nuclear Fuel; Order... been issued a specific license by the U.S. Nuclear Regulatory Commission (NRC or...

  8. Utilization of Minor Actinides (Np, Am, Cm) in Nuclear Power Reactor

    NASA Astrophysics Data System (ADS)

    Gerasimov, A.; Bergelson, B.; Tikhomirov, G.

    2014-06-01

    Calculation research of the utilization process of minor actinides (transmutation with use of power released) is performed for specialized power reactor of the VVER type operating on the level of electric power of 1000 MW. Five subsequent cycles are considered for the reactor with fuel elements containing minor actinides along with enriched uranium. It was shown that one specialized reactor for the one cycle (900 days) can utilize minor actinides from several VVER-1000 reactors without any technological and structural modifications. Power released because of minor actinide fission is about 4% with respect to the total power

  9. System aspects of a Space Nuclear Reactor Power System

    SciTech Connect

    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.

  10. 77 FR 38339 - Dairyland Power Cooperative, La Crosse Boiling Water Reactor Exemption From Certain Security...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-06-27

    ... COMMISSION Dairyland Power Cooperative, La Crosse Boiling Water Reactor Exemption From Certain Security Requirements 1.0 Background The La Crosse Boiling Water Reactor (LACBWR) is owned and was operated by the Dairyland Power Cooperative (DPC). The LACBWR was a nuclear power plant of nominal 50 Mw electrical...

  11. 77 FR 38338 - Dairyland Power Cooperative; La Crosse Boiling Water Reactor Exemption From Certain Security...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-06-27

    ... COMMISSION Dairyland Power Cooperative; La Crosse Boiling Water Reactor Exemption From Certain Security Requirements 1.0 Background The La Crosse Boiling Water Reactor (LACBWR) is owned and was operated by the Dairyland Power Cooperative (DPC). The LACBWR was a nuclear power plant of nominal 50 Mw electrical...

  12. 77 FR 3009 - Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Advanced Boiling Water Reactors

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-01-20

    ... COMMISSION Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Advanced Boiling Water Reactors..., ``Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Advanced Boiling Water Reactors.'' DATES... developed using this Catalog along with the Operator Licensing Examination Standards for Power...

  13. A preliminary investigation of the Topaz II reactor as a lunar surface power supply

    SciTech Connect

    Polansky, G.F.; Houts, M.G.

    1995-12-31

    Reactor power supplies offer many attractive characteristics for lunar surface applications. The Topaz II reactor resulted from an extensive development program in the former Soviet Union. Flight quality reactor units remain from this program and are currently under evaluation in the United States. This paper examines the potential for applying the Topaz II, originally developed to provide spacecraft power, as a lunar surface power supply.

  14. Core follow calculation with the nTRACER numerical reactor and verification using power reactor measurement data

    SciTech Connect

    Jung, Y. S.; Joo, H. G.; Yoon, J. I.

    2013-07-01

    The nTRACER direct whole core transport code employing the planar MOC solution based 3-D calculation method, the subgroup method for resonance treatment, the Krylov matrix exponential method for depletion, and a subchannel thermal/hydraulic calculation solver was developed for practical high-fidelity simulation of power reactors. Its accuracy and performance is verified by comparing with the measurement data obtained for three pressurized water reactor cores. It is demonstrated that accurate and detailed multi-physic simulation of power reactors is practically realizable without any prior calculations or adjustments. (authors)

  15. Reactor/Brayton power systems for nuclear electric spacecraft

    NASA Technical Reports Server (NTRS)

    Layton, J. P.

    1980-01-01

    Studies are currently underway to assess the technological feasibility of a nuclear-reactor-powered spacecraft propelled by electric thrusters. This vehicle would be capable of performing detailed exploration of the outer planets of the solar system during the remainder of this century. The purpose of this study was to provide comparative information on a closed cycle gas turbine power conversion system. The results have shown that the performance is very competitive and that a 400 kWe space power system is dimensionally compatible with a single Space Shuttle launch. Performance parameters of system mass and radiator area were determined for systems from 100 to 1000 kWe. A 400 kWe reference system received primary attention. The components of this system were defined and a conceptual layout was developed with encouraging results. The preliminary mass determination for the complete power system was very close to the desired goal of 20 kg/kWe. Use of more advanced technology (higher turbine inlet temperature) will substantially improve system performance characteristics.

  16. 10 CFR 50.60 - Acceptance criteria for fracture prevention measures for lightwater nuclear power reactors for...

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... lightwater nuclear power reactors for normal operation. 50.60 Section 50.60 Energy NUCLEAR REGULATORY... lightwater nuclear power reactors for normal operation. (a) Except as provided in paragraph (b) of this section, all light-water nuclear power reactors, other than reactor facilities for which...

  17. 10 CFR 50.60 - Acceptance criteria for fracture prevention measures for lightwater nuclear power reactors for...

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... lightwater nuclear power reactors for normal operation. 50.60 Section 50.60 Energy NUCLEAR REGULATORY... lightwater nuclear power reactors for normal operation. (a) Except as provided in paragraph (b) of this section, all light-water nuclear power reactors, other than reactor facilities for which...

  18. 10 CFR 50.60 - Acceptance criteria for fracture prevention measures for lightwater nuclear power reactors for...

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... lightwater nuclear power reactors for normal operation. 50.60 Section 50.60 Energy NUCLEAR REGULATORY... lightwater nuclear power reactors for normal operation. (a) Except as provided in paragraph (b) of this section, all light-water nuclear power reactors, other than reactor facilities for which...

  19. 10 CFR 50.60 - Acceptance criteria for fracture prevention measures for lightwater nuclear power reactors for...

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... lightwater nuclear power reactors for normal operation. 50.60 Section 50.60 Energy NUCLEAR REGULATORY... lightwater nuclear power reactors for normal operation. (a) Except as provided in paragraph (b) of this section, all light-water nuclear power reactors, other than reactor facilities for which...

  20. 10 CFR 50.60 - Acceptance criteria for fracture prevention measures for lightwater nuclear power reactors for...

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... lightwater nuclear power reactors for normal operation. 50.60 Section 50.60 Energy NUCLEAR REGULATORY... lightwater nuclear power reactors for normal operation. (a) Except as provided in paragraph (b) of this section, all light-water nuclear power reactors, other than reactor facilities for which...

  1. Test Results from a Direct Drive Gas Reactor Simulator Coupled to a Brayton Power Conversion Unit

    NASA Technical Reports Server (NTRS)

    Hervol, David S.; Briggs, Maxwell H.; Owen, Albert K.; Bragg-Sitton, Shannon M.; Godfroy, Thomas J.

    2010-01-01

    Component level testing of power conversion units proposed for use in fission surface power systems has typically been done using relatively simple electric heaters for thermal input. These heaters do not adequately represent the geometry or response of proposed reactors. As testing of fission surface power systems transitions from the component level to the system level it becomes necessary to more accurately replicate these reactors using reactor simulators. The Direct Drive Gas-Brayton Power Conversion Unit test activity at the NASA Glenn Research Center integrates a reactor simulator with an existing Brayton test rig. The response of the reactor simulator to a change in Brayton shaft speed is shown as well as the response of the Brayton to an insertion of reactivity, corresponding to a drum reconfiguration. The lessons learned from these tests can be used to improve the design of future reactor simulators which can be used in system level fission surface power tests.

  2. Power control of SAFE reactor using fuzzy logic

    NASA Astrophysics Data System (ADS)

    Irvine, Claude

    2002-01-01

    Controlling the 100 kW SAFE (Safe Affordable Fission Engine) reactor consists of design and implementation of a fuzzy logic process control system to regulate dynamic variables related to nuclear system power. The first phase of development concentrates primarily on system power startup and regulation, maintaining core temperature equilibrium, and power profile matching. This paper discusses the experimental work performed in those areas. Nuclear core power from the fuel elements is simulated using resistive heating elements while heat rejection is processed by a series of heat pipes. Both axial and radial nuclear power distributions are determined from neuronic modeling codes. The axial temperature profile of the simulated core is matched to the nuclear power profile by varying the resistance of the heating elements. The SAFE model establishes radial temperature profile equivalence by establishing 32 control zones as the nodal coordinates. Control features also allow for slow warm up, since complete shutoff can occur in the heat pipes if heat-source temperatures drop/rise below a certain minimum value, depending on the specific fluid and gas combination in the heat pipe. The entire system is expected to be self-adaptive, i.e., capable of responding to long-range changes in the space environment. Particular attention in the development of the fuzzy logic algorithm shall ensure that the system process remains at set point, virtually eliminating overshoot on start-up and during in-process disturbances. The controller design will withstand harsh environments and applications where it might come in contact with water, corrosive chemicals, radiation fields, etc. .

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

    SciTech Connect

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

    1991-09-01

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

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

    DOEpatents

    Honekamp, John R.; Fryer, Richard M.

    1978-01-01

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

  5. High-intensity power-resolved radiation imaging of an operational nuclear reactor.

    PubMed

    Beaumont, Jonathan S; Mellor, Matthew P; Villa, Mario; Joyce, Malcolm J

    2015-01-01

    Knowledge of the neutron distribution in a nuclear reactor is necessary to ensure the safe and efficient burnup of reactor fuel. Currently these measurements are performed by in-core systems in what are extremely hostile environments and in most reactor accident scenarios it is likely that these systems would be damaged. Here we present a compact and portable radiation imaging system with the ability to image high-intensity fast-neutron and gamma-ray fields simultaneously. This system has been deployed to image radiation fields emitted during the operation of a TRIGA test reactor allowing a spatial visualization of the internal reactor conditions to be obtained. The imaged flux in each case is found to scale linearly with reactor power indicating that this method may be used for power-resolved reactor monitoring and for the assay of ongoing nuclear criticalities in damaged nuclear reactors. PMID:26450669

  6. High-intensity power-resolved radiation imaging of an operational nuclear reactor.

    PubMed

    Beaumont, Jonathan S; Mellor, Matthew P; Villa, Mario; Joyce, Malcolm J

    2015-10-09

    Knowledge of the neutron distribution in a nuclear reactor is necessary to ensure the safe and efficient burnup of reactor fuel. Currently these measurements are performed by in-core systems in what are extremely hostile environments and in most reactor accident scenarios it is likely that these systems would be damaged. Here we present a compact and portable radiation imaging system with the ability to image high-intensity fast-neutron and gamma-ray fields simultaneously. This system has been deployed to image radiation fields emitted during the operation of a TRIGA test reactor allowing a spatial visualization of the internal reactor conditions to be obtained. The imaged flux in each case is found to scale linearly with reactor power indicating that this method may be used for power-resolved reactor monitoring and for the assay of ongoing nuclear criticalities in damaged nuclear reactors.

  7. High-intensity power-resolved radiation imaging of an operational nuclear reactor

    NASA Astrophysics Data System (ADS)

    Beaumont, Jonathan S.; Mellor, Matthew P.; Villa, Mario; Joyce, Malcolm J.

    2015-10-01

    Knowledge of the neutron distribution in a nuclear reactor is necessary to ensure the safe and efficient burnup of reactor fuel. Currently these measurements are performed by in-core systems in what are extremely hostile environments and in most reactor accident scenarios it is likely that these systems would be damaged. Here we present a compact and portable radiation imaging system with the ability to image high-intensity fast-neutron and gamma-ray fields simultaneously. This system has been deployed to image radiation fields emitted during the operation of a TRIGA test reactor allowing a spatial visualization of the internal reactor conditions to be obtained. The imaged flux in each case is found to scale linearly with reactor power indicating that this method may be used for power-resolved reactor monitoring and for the assay of ongoing nuclear criticalities in damaged nuclear reactors.

  8. High-intensity power-resolved radiation imaging of an operational nuclear reactor

    PubMed Central

    Beaumont, Jonathan S.; Mellor, Matthew P.; Villa, Mario; Joyce, Malcolm J.

    2015-01-01

    Knowledge of the neutron distribution in a nuclear reactor is necessary to ensure the safe and efficient burnup of reactor fuel. Currently these measurements are performed by in-core systems in what are extremely hostile environments and in most reactor accident scenarios it is likely that these systems would be damaged. Here we present a compact and portable radiation imaging system with the ability to image high-intensity fast-neutron and gamma-ray fields simultaneously. This system has been deployed to image radiation fields emitted during the operation of a TRIGA test reactor allowing a spatial visualization of the internal reactor conditions to be obtained. The imaged flux in each case is found to scale linearly with reactor power indicating that this method may be used for power-resolved reactor monitoring and for the assay of ongoing nuclear criticalities in damaged nuclear reactors. PMID:26450669

  9. System simulation of a multicell thermionic space power reactor

    NASA Astrophysics Data System (ADS)

    von Arx, Alan Vincent

    For many years, thermionic power has been considered for space application. The prominent feature of the power conversion system is that there are no moving parts. Although designs have been developed by various organizations, no comprehensive system models are known to exist which can simulate transient behavior of a multicell design nor is there a method to directly couple these models to other codes that can calculate variations in reactivity. Thus, a procedure has been developed to couple the performance calculations of a space nuclear reactor thermal/hydraulics code with a neutron diffusion code to analyze temperature feedback. Thermionic power is based on the thermionic emissions principle where free electrons in a conductor have sufficient energy to escape the surface. Kinetic energy is given to the electrons by heating the conductor. Specifically, a 48 kWe thermionic power converter system model has been developed and used to model startup and other transients. Less than 10% of the fuel heat is converted to electricity, and the rest is rejected to space via a heat pipe radiator. An electromagnetic pump circulates the liquid metal coolant. First, a startup transient model was developed which showed stable operation through ignition of the Thermionic Fuel Elements (TFEs) and thawing of the radiator heat pipes. Also, the model's capability was expanded to include two-phase heat transfer to model boiling using coupled mass and thermal energy conservation equations. The next step incorporated effects of reactivity feedback---showing that various mechanisms will prevent power and temperature run-up for a flow reduction scenario where the reactor control systems fail to respond. In particular, the Doppler effect was shown to counter a positive worth due to partial core voiding although steps must be taken to preclude film boiling in that high superheats will result in TFE failures. Finally, analysis of the core grid spacer location suggests it should be located at

  10. Non-Power Reactor Operator Licensing Examiner Standards. Revision 1

    SciTech Connect

    1995-06-01

    The Non-Power Reactor Operator Licensing Examiner Standards provide policy and guidance to NRC examiners and establish the procedures and practices for examining and licensing of applicants for NRC operator licenses pursuant to Part 55 of Title 10 of the Code of Federal Regulations (10 CFR 55). They are intended to assist NRC examiners and facility licensees to understand the examination process better and to provide for equitable and consistent administration of examinations to all applicants by NRC examiners. These standards are not a substitute for the operator licensing regulations and are subject to revision or other internal operator examination licensing policy changes. As appropriate, these standards will be revised periodically to accommodate comments and reflect new information or experience.

  11. Compact, high-power nuclear reactor systems based on small diameter particulate fuel

    NASA Astrophysics Data System (ADS)

    Powell, J. R.; Botts, T. E.

    Two compact, high-power nuclear reactor concepts are discussed. Both are gas-cooled cavity-type reactors which utilize particulate fuel of the type now used in HTGR reactors. Unshielded reactor volumes are on the order of one cubic meter. The Fixed Bed Reactor operating temperature is limited to 2500 K and the output power to 250 MW(e). In the Rotating Bed Reactor fuel is held within a rotating porous metal drum as a rotating fluidized bed. Rotating Bed Reactor outlet temperatures up to 3000 K and output power levels up to 1000 MW(e) are achievable. Both reactors can be brought up from stand by to full power in times on the order of a few seconds, due primarily to the short thermal time constant for the fuel particles. Turbine and MHD Brayton are the power conversion cycles of choice. Open cycle operation is generally favored for applications operating at less than 1000 sec of equivalent integrated full power. At power levels above 1 MW(e), the liquid droplet radiator is the favored means of heat rejection. Power system specific power levels of 10 kW(e)/kg (not including shield) appears to be quite feasible.

  12. Power monitoring in space nuclear reactors using silicon carbide radiation detectors

    NASA Technical Reports Server (NTRS)

    Ruddy, Frank H.; Patel, Jagdish U.; Williams, John G.

    2005-01-01

    Space reactor power monitors based on silicon carbide (SiC) semiconductor neutron detectors are proposed. Detection of fast leakage neutrons using SiC detectors in ex-core locations could be used to determine reactor power: Neutron fluxes, gamma-ray dose rates and ambient temperatures have been calculated as a function of distance from the reactor core, and the feasibility of power monitoring with SiC detectors has been evaluated at several ex-core locations. Arrays of SiC diodes can be configured to provide the required count rates to monitor reactor power from startup to full power Due to their resistance to temperature and the effects of neutron and gamma-ray exposure, SiC detectors can be expected to provide power monitoring information for the fill mission of a space reactor.

  13. A Basic LEGO Reactor Design for the Provision of Lunar Surface Power

    SciTech Connect

    John Darrell Bess

    2008-06-01

    A final design has been established for a basic Lunar Evolutionary Growth-Optimized (LEGO) Reactor using current and near-term technologies. The LEGO Reactor is a modular, fast-fission, heatpipe-cooled, clustered-reactor system for lunar-surface power generation. The reactor is divided into subcritical units that can be safely launched with lunar shipments from Earth, and then emplaced directly into holes drilled into the lunar regolith to form a critical reactor assembly. The regolith would not just provide radiation shielding, but serve as neutron-reflector material as well. The reactor subunits are to be manufactured using proven and tested materials for use in radiation environments, such as uranium-dioxide fuel, stainless-steel cladding and structural support, and liquid-sodium heatpipes. The LEGO Reactor system promotes reliability, safety, and ease of manufacture and testing at the cost of an increase in launch mass per overall rated power level and a reduction in neutron economy when compared to a single-reactor system. A single unshielded LEGO Reactor subunit has an estimated mass of approximately 448 kg and provides approximately 5 kWe. The overall envelope for a single subunit with fully extended radiator panels has a height of 8.77 m and a diameter of 0.50 m. Six subunits could provide sufficient power generation throughout the initial stages of establishing a lunar outpost. Portions of the reactor may be neutronically decoupled to allow for reduced power production during unmanned periods of base operations. During later stages of lunar-base development, additional subunits may be emplaced and coupled into the existing LEGO Reactor network, subject to lunar base power demand. Improvements in reactor control methods, fuel form and matrix, shielding, as well as power conversion and heat rejection techniques can help generate an even more competitive LEGO Reactor design. Further modifications in the design could provide power generative opportunities for

  14. Application of Molten Salt Reactor Technology to MMW In-Space NEP and Surface Power Missions

    NASA Technical Reports Server (NTRS)

    Patton, Bruce; Sorensen, Kirk; Rodgers, Stephen (Technical Monitor)

    2002-01-01

    Anticipated manned nuclear electric propulsion (NEP) and planetary surface power missions will require multimegawatt nuclear reactors that are lightweight, operationally robust, and scalable in power for widely varying scientific mission objectives. Molten salt reactor technology meets all of these requirements and offers an interesting alternative to traditional multimegawatt gas-cooled and liquid metal concepts.

  15. 10 CFR 50.36a - Technical specifications on effluents from nuclear power reactors.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... reactors. 50.36a Section 50.36a Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF PRODUCTION AND...; Ineligibility of Certain Applicants § 50.36a Technical specifications on effluents from nuclear power reactors..., including expected occurrences, as low as is reasonably achievable, each licensee of a nuclear power...

  16. 10 CFR 50.36a - Technical specifications on effluents from nuclear power reactors.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... reactors. 50.36a Section 50.36a Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF PRODUCTION AND...; Ineligibility of Certain Applicants § 50.36a Technical specifications on effluents from nuclear power reactors..., including expected occurrences, as low as is reasonably achievable, each licensee of a nuclear power...

  17. 10 CFR 50.36a - Technical specifications on effluents from nuclear power reactors.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... reactors. 50.36a Section 50.36a Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF PRODUCTION AND...; Ineligibility of Certain Applicants § 50.36a Technical specifications on effluents from nuclear power reactors..., including expected occurrences, as low as is reasonably achievable, each licensee of a nuclear power...

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

    SciTech Connect

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

    2004-08-11

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

  19. Computer simulation of magnetization-controlled shunt reactors for calculating electromagnetic transients in power systems

    SciTech Connect

    Karpov, A. S.

    2013-01-15

    A computer procedure for simulating magnetization-controlled dc shunt reactors is described, which enables the electromagnetic transients in electric power systems to be calculated. It is shown that, by taking technically simple measures in the control system, one can obtain high-speed reactors sufficient for many purposes, and dispense with the use of high-power devices for compensating higher harmonic components.

  20. 10 CFR 50.44 - Combustible gas control for nuclear power reactors.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... 10 Energy 1 2014-01-01 2014-01-01 false Combustible gas control for nuclear power reactors. 50.44 Section 50.44 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION... for nuclear power reactors. (a) Definitions—(1) Inerted atmosphere means a containment atmosphere...

  1. 10 CFR 50.44 - Combustible gas control for nuclear power reactors.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

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

  2. 10 CFR 50.44 - Combustible gas control for nuclear power reactors.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... 10 Energy 1 2013-01-01 2013-01-01 false Combustible gas control for nuclear power reactors. 50.44 Section 50.44 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION... for nuclear power reactors. (a) Definitions—(1) Inerted atmosphere means a containment atmosphere...

  3. 10 CFR 50.72 - Immediate notification requirements for operating nuclear power reactors.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... 10 Energy 1 2013-01-01 2013-01-01 false Immediate notification requirements for operating nuclear power reactors. 50.72 Section 50.72 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF... notification requirements for operating nuclear power reactors. (a) General requirements. 1 (1) Each...

  4. Capillary-Pumped Passive Reactor Concept for Space Nuclear Power

    SciTech Connect

    Dr. Thomas F. Lin; Dr. Thomas G. Hughes; Christopher G. Miller

    2008-05-30

    To develop the passively-cooled space reactor concept using the capillary-induced lithium flow, since molten lithium possesses a very favorable surface tension characteristic. In space where the gravitational field is minimal, the gravity-assisted natural convection cooling is not effective nor an option for reactor heat removal, the capillary induced cooling becomes an attractive means of providing reactor cooling.

  5. Plasma engineering studies for Tennessee Tokamak (TENTOK) fusion power reactor

    SciTech Connect

    Yokoyama, K.E.; Lacatski, J.T.; Miller, J.B.; Bryan, W.E.; King, P.W.; Santoro, R.T.; Uckan, N.A.; Shannon, T.E.

    1984-02-01

    This paper summarizes the results of the plasma engineering and systems analysis studies for the Tennessee Tokamak (TENTOK) fusion power reactor. TENTOK is a 3000-MW(t) central station power plant that uses deuterium-tritium fuel in a D-shaped tokamak plasma configuration with a double-null poloidal divertor. The major parameters are R/sub 0/ = 6.4 m, a = 1.6 m, sigma (elongation) = 1.65, (n) = 1.5 x 10/sup 20/ m/sup -3/, (T) = 15 keV, (..beta..) = 6%, B/sub T/ (on-axis) = 5.6 T, I/sub p/ = 8.5 MA, and wall loading = 3 MW/m/sup 2/. Detailed analyses are performed in the areas of (1) transport simulation using the one-and-one-half-dimensional (1-1/2-D) WHIST transport code, (2) equilibrium/poloidal field coil systems, (3) neutral beam and radiofrequency (rf) heating, and (4) pellet fueling. In addition, impurity control systems, diagnostics and controls, and possible microwave plasma preheating and steady-state current drive options are also considered. Some of the major features of TENTOK include rf heating in the ion cyclotron range of frequencies, superconducting equilibrium field coils outside the superconducting toroidal field coils, a double-null poloidal divertor for impurity control and alpha ash removal, and rf-assisted plasma preheating and current startup.

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

    SciTech Connect

    Renier, J.A.

    2002-04-17

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

  7. Coupled Monte Carlo neutronics and thermal hydraulics for power reactors

    SciTech Connect

    Bernnat, W.; Buck, M.; Mattes, M.; Zwermann, W.; Pasichnyk, I.; Velkov, K.

    2012-07-01

    The availability of high performance computing resources enables more and more the use of detailed Monte Carlo models even for full core power reactors. The detailed structure of the core can be described by lattices, modeled by so-called repeated structures e.g. in Monte Carlo codes such as MCNP5 or MCNPX. For cores with mainly uniform material compositions, fuel and moderator temperatures, there is no problem in constructing core models. However, when the material composition and the temperatures vary strongly a huge number of different material cells must be described which complicate the input and in many cases exceed code or memory limits. The second problem arises with the preparation of corresponding temperature dependent cross sections and thermal scattering laws. Only if these problems can be solved, a realistic coupling of Monte Carlo neutronics with an appropriate thermal-hydraulics model is possible. In this paper a method for the treatment of detailed material and temperature distributions in MCNP5 is described based on user-specified internal functions which assign distinct elements of the core cells to material specifications (e.g. water density) and temperatures from a thermal-hydraulics code. The core grid itself can be described with a uniform material specification. The temperature dependency of cross sections and thermal neutron scattering laws is taken into account by interpolation, requiring only a limited number of data sets generated for different temperatures. Applications will be shown for the stationary part of the Purdue PWR benchmark using ATHLET for thermal- hydraulics and for a generic Modular High Temperature reactor using THERMIX for thermal- hydraulics. (authors)

  8. Preapplication safety evaluation report for the Power Reactor Innovative Small Module (PRISM) liquid-metal reactor. Final report

    SciTech Connect

    Donoghue, J.E.; Donohew, J.N.; Golub, G.R.; Kenneally, R.M.; Moore, P.B.; Sands, S.P.; Throm, E.D.; Wetzel, B.A.

    1994-02-01

    This preapplication safety evaluation report (PSER) presents the results of the preapplication desip review for die Power Reactor Innovative Small Module (PRISM) liquid-mew (sodium)-cooled reactor, Nuclear Regulatory Commission (NRC) Project No. 674. The PRISM conceptual desip was submitted by the US Department of Energy in accordance with the NRC`s ``Statement of Policy for the Regulation of Advanced Nuclear Power Plants`` (51 Federal Register 24643). This policy provides for the early Commission review and interaction with designers and licensees. The PRISM reactor desip is a small, modular, pool-type, liquid-mew (sodium)-cooled reactor. The standard plant design consists of dim identical power blocks with a total electrical output rating of 1395 MWe- Each power block comprises three reactor modules, each with a thermal rating of 471 MWt. Each module is located in its own below-grade silo and is co to its own intermediate heat transport system and steam generator system. The reactors utilize a metallic-type fuel, a ternary alloy of U-Pu-Zr. The design includes passive reactor shutdown and passive decay heat removal features. The PSER is the NRC`s preliminary evaluation of the safety features in the PRISM design, including the projected research and development programs required to support the design and the proposed testing needs. Because the NRC review was based on a conceptual design, the PSER did not result in an approval of the design. Instead it identified certain key safety issues, provided some guidance on applicable licensing criteria, assessed the adequacy of the preapplicant`s research and development programs, and concluded that no obvious impediments to licensing the PRISM design had been identified.

  9. Diversion assumptions for high-powered research reactors. ISPO C-50 Phase 1

    SciTech Connect

    Binford, F.T.

    1984-01-01

    This study deals with diversion assumptions for high-powered research reactors -- specifically, MTR fuel; pool- or tank-type research reactors with light-water moderator; and water, beryllium, or graphite reflectors, and which have a power level of 25 MW(t) or more. The objective is to provide assistance to the IAEA in documentation of criteria and inspection observables related to undeclared plutonium production in the reactors described above, including: criteria for undeclared plutonium production, necessary design information for implementation of these criteria, verification guidelines including neutron physics and heat transfer, and safeguards measures to facilitate the detection of undeclared plutonium production at large research reactors.

  10. Benchmark Evaluation of the Medium-Power Reactor Experiment Program Critical Configurations

    SciTech Connect

    Margaret A. Marshall; John D. Bess

    2013-02-01

    A series of small, compact critical assembly (SCCA) experiments were performed in 1962-1965 at the Oak Ridge National Laboratory Critical Experiments Facility (ORCEF) for the Medium-Power Reactor Experiment (MPRE) program. The MPRE was a stainless-steel clad, highly enriched uranium (HEU)-O2 fuelled, BeO reflected reactor design to provide electrical power to space vehicles. Cooling and heat transfer were to be achieved by boiling potassium in the reactor core and passing vapor directly through a turbine. Graphite- and beryllium-reflected assemblies were constructed at ORCEF to verify the critical mass, power distribution, and other reactor physics measurements needed to validate reactor calculations and reactor physics methods. The experimental series was broken into three parts, with the third portion of the experiments representing the beryllium-reflected measurements. The latter experiments are of interest for validating current reactor design efforts for a fission surface power reactor. The entire series has been evaluated as acceptable benchmark experiments and submitted for publication in the International Handbook of Evaluated Criticality Safety Benchmark Experiments and in the International Handbook of Evaluated Reactor Physics Benchmark Experiments.

  11. 10 CFR 50.83 - Release of part of a power reactor facility or site for unrestricted use.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... release. Nuclear power reactor licensees seeking NRC approval shall— (1) Evaluate the effect of releasing... 10 Energy 1 2013-01-01 2013-01-01 false Release of part of a power reactor facility or site for... of a power reactor facility or site for unrestricted use. (a) Prior written NRC approval is...

  12. 10 CFR 50.83 - Release of part of a power reactor facility or site for unrestricted use.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... release. Nuclear power reactor licensees seeking NRC approval shall— (1) Evaluate the effect of releasing... 10 Energy 1 2014-01-01 2014-01-01 false Release of part of a power reactor facility or site for... of a power reactor facility or site for unrestricted use. (a) Prior written NRC approval is...

  13. 10 CFR 50.83 - Release of part of a power reactor facility or site for unrestricted use.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... release. Nuclear power reactor licensees seeking NRC approval shall— (1) Evaluate the effect of releasing... 10 Energy 1 2010-01-01 2010-01-01 false Release of part of a power reactor facility or site for... of a power reactor facility or site for unrestricted use. (a) Prior written NRC approval is...

  14. 10 CFR 50.83 - Release of part of a power reactor facility or site for unrestricted use.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... release. Nuclear power reactor licensees seeking NRC approval shall— (1) Evaluate the effect of releasing... 10 Energy 1 2011-01-01 2011-01-01 false Release of part of a power reactor facility or site for... of a power reactor facility or site for unrestricted use. (a) Prior written NRC approval is...

  15. 10 CFR 50.83 - Release of part of a power reactor facility or site for unrestricted use.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... release. Nuclear power reactor licensees seeking NRC approval shall— (1) Evaluate the effect of releasing... 10 Energy 1 2012-01-01 2012-01-01 false Release of part of a power reactor facility or site for... of a power reactor facility or site for unrestricted use. (a) Prior written NRC approval is...

  16. 10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... light-water nuclear power reactors. 50.46 Section 50.46 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC... Approvals § 50.46 Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors. (a)(1)(i) Each boiling or pressurized light-water nuclear power reactor fueled with uranium...

  17. 10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... light-water nuclear power reactors. 50.46 Section 50.46 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC... Approvals § 50.46 Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors. (a)(1)(i) Each boiling or pressurized light-water nuclear power reactor fueled with uranium...

  18. 10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... light-water nuclear power reactors. 50.46 Section 50.46 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC... Approvals § 50.46 Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors. (a)(1)(i) Each boiling or pressurized light-water nuclear power reactor fueled with uranium...

  19. 10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... light-water nuclear power reactors. 50.46 Section 50.46 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC... Approvals § 50.46 Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors. (a)(1)(i) Each boiling or pressurized light-water nuclear power reactor fueled with uranium...

  20. 10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... light-water nuclear power reactors. 50.46 Section 50.46 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC... Approvals § 50.46 Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors. (a)(1)(i) Each boiling or pressurized light-water nuclear power reactor fueled with uranium...

  1. APR1400 Reactivity Insertion Accident Analysis Using KNAP

    SciTech Connect

    Chang-Keun, Yang; Yo-Han, Kim; Chang-Kyung, Sung

    2006-07-01

    The Korea Electric Power Research Institute had decided to develop the new safety analysis code system for the Optimized Power Reactor 1000 (OPR1000) in Korea by the fund of the Ministry of Commerce, Industry and Energy. In this paper, some results of the Advanced Power Reactor 1400(APR1400) using the RETRAN code for some reactivity insertion accident are introduced to expand application from safety analysis experience of OPR1000. (authors)

  2. Space reactor/Stirling cycle systems for high power Lunar applications

    SciTech Connect

    Schmitz, P.D.; Mason, L.S.

    1994-09-01

    NASA`s Space Exploration Initiative (SEI) has proposed the use of high power nuclear power systems on the lunar surface as a necessary alternative to solar power. Because of the long lunar night ({approximately} 14 earth days) solar powered systems with the requisite energy storage in the form of regenerative fuel cells or batteries becomes prohibitively heavy at high power levels ({approximately} 100 kWe). At these high power levels nuclear power systems become an enabling technology for variety of missions. One way of producing power on the lunar surface is with an SP-100 class reactor coupled with Stirling power converters. In this study, analysis and characterization of the SP-100 class reactor coupled with Free Piston Stirling Power Conversion (FPSPC) system will be performed. Comparison of results with previous studies of other systems, particularly Brayton and Thermionic, are made.

  3. Heat pipe cooled reactors for multi-kilowatt space power supplies

    NASA Astrophysics Data System (ADS)

    Ranken, W. A.; Houts, M. G.

    Three nuclear reactor space power system designs are described that demonstrate how the use of high temperature heat pipes for reactor heat transport, combined with direct conversion of heat to electricity, can result in eliminating pumped heat transport loops for both primary reactor cooling and heat rejection. The result is a significant reduction in system complexity that leads to very low mass systems with high reliability, especially in the power range of 1 to 20 kWe. In addition to removing heat exchangers, electromagnetic pumps, and coolant expansion chambers, the heat pipe/direct conversion combination provides such capabilities as startup from the frozen state, automatic rejection of reactor decay heat in the event of emergency or accidental reactor shutdown, and the elimination of single point failures in the reactor cooling system. The power system designs described include a thermoelectric system that can produce 1 to 2 kWe, a bimodal modification of this system to increase its power level to 5 kWe and incorporate high temperature hydrogen propulsion capability, and a moderated thermionic reactor concept with 5 to 20 kWe power output that is based on beryllium modules that thermally couple cylindrical thermionic fuel elements (TFE's) to radiator heat pipes.

  4. Heat pipe cooled reactors for multi-kilowatt space power supplies

    SciTech Connect

    Ranken, W.A.; Houts, M.G.

    1995-01-01

    Three nuclear reactor space power system designs are described that demonstrate how the use of high temperature heat pipes for reactor heat transport, combined with direct conversion of heat to electricity, can result in eliminating pumped heat transport loops for both primary reactor cooling and heat rejection. The result is a significant reduction in system complexity that leads to very low mass systems with high reliability, especially in the power range of 1 to 20 kWe. In addition to removing heat exchangers, electromagnetic pumps, and coolant expansion chambers, the heat pipe/direct conversion combination provides such capabilities as startup from the frozen state, automatic rejection of reactor decay heat in the event of emergency or accidental reactor shutdown, and the elimination of single point failures in the reactor cooling system. The power system designs described include a thermoelectric system that can produce 1 to 2 kWe, a bimodal modification of this system to increase its power level to 5 kWe and incorporate high temperature hydrogen propulsion capability, and a moderated thermionic reactor concept with 5 to 20 kWe power output that is based on beryllium modules that thermally couple cylindrical thermionic fuel elements (TFEs) to radiator heat pipes.

  5. Progress in space nuclear reactor power systems technology development - The SP-100 program

    NASA Technical Reports Server (NTRS)

    Davis, H. S.

    1984-01-01

    Activities related to the development of high-temperature compact nuclear reactors for space applications had reached a comparatively high level in the U.S. during the mid-1950s and 1960s, although only one U.S. nuclear reactor-powered spacecraft was actually launched. After 1973, very little effort was devoted to space nuclear reactor and propulsion systems. In February 1983, significant activities toward the development of the technology for space nuclear reactor power systems were resumed with the SP-100 Program. Specific SP-100 Program objectives are partly related to the determination of the potential performance limits for space nuclear power systems in 100-kWe and 1- to 100-MW electrical classes. Attention is given to potential missions and applications, regimes of possible space power applicability, safety considerations, conceptual system designs, the establishment of technical feasibility, nuclear technology, materials technology, and prospects for the future.

  6. Digital computer study of nuclear reactor thermal transients during startup of 60-kWe Brayton power conversion system

    NASA Technical Reports Server (NTRS)

    Jefferies, K. S.; Tew, R. C.

    1974-01-01

    A digital computer study was made of reactor thermal transients during startup of the Brayton power conversion loop of a 60-kWe reactor Brayton power system. A startup procedure requiring the least Brayton system complication was tried first; this procedure caused violations of design limits on key reactor variables. Several modifications of this procedure were then found which caused no design limit violations. These modifications involved: (1) using a slower rate of increase in gas flow; (2) increasing the initial reactor power level to make the reactor respond faster; and (3) appropriate reactor control drum manipulation during the startup transient.

  7. Hot zero power reactor calculations using the Insilico code

    NASA Astrophysics Data System (ADS)

    Hamilton, Steven P.; Evans, Thomas M.; Davidson, Gregory G.; Johnson, Seth R.; Pandya, Tara M.; Godfrey, Andrew T.

    2016-06-01

    In this paper we describe the reactor physics simulation capabilities of the Insilico code. A description of the various capabilities of the code is provided, including detailed discussion of the geometry, meshing, cross section processing, and neutron transport options. Numerical results demonstrate that Insilico using an SPN solver with pin-homogenized cross section generation is capable of delivering highly accurate full-core simulation of various pressurized water reactor problems. Comparison to both Monte Carlo calculations and measured plant data is provided.

  8. Advanced-power-reactor design concepts and performance characteristics

    NASA Technical Reports Server (NTRS)

    Davison, H. W.; Kirchgessner, T. A.; Springborn, R. H.; Yacobucci, H. G.

    1974-01-01

    Five reactor cooling concepts which allow continued reactor operation following a single rupture of the coolant system are presented for application with the APR. These concepts incorporate convective cooling, double containment, or heat pipes to ensure operation after a coolant line rupture. Based on an evaluation of several control system concepts, a molybdenum clad, beryllium oxide sliding reflector located outside the pressure vessel is recommended.

  9. Gravity Scaling of a Power Reactor Water Shield

    NASA Technical Reports Server (NTRS)

    Reid, Robert S.; Pearson, J. Boise

    2007-01-01

    A similarity analysis on a water-based reactor shield examined the effect of gravity on free convection between a reactor shield inner and outer vessel boundaries. Two approaches established similarity between operation on the Earth and the Moon: 1) direct scaling of Rayleigh number equating gravity-surface heat flux products, 2) temperature difference between the wall and thermal boundary layer held constant. Nusselt number for natural convection (laminar and turbulent) is assumed of form Nu = CRa(sup n).

  10. Preliminary design of reactor power systems for the manned space base.

    NASA Technical Reports Server (NTRS)

    Mckhann, G. G.; Coggi, J. V.; Diamond, S. D.

    1972-01-01

    The results of design integration studies of uranium-zirconium hydride (UZr-Hx) reactor power systems for the NASA space base study program are presented. The power conversion systems investigated include the Brayton cycle, the organic Rankine cycle, the SNAP-8 mercury Rankine cycle, and thermoelectric (PbTe). The proposed space base has a 10-year life and requires 100 kWe of power. Two 50-kWe power systems with a nominal replacement life of 5 years are utilized. Parametric design data such as life, weight, radiator area, reactor outlet-temperature, reactor thermal power, and power conversion system efficiency are presented and used for the design and integration of the system with the space base.

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

    DOEpatents

    Hampel, Viktor E.

    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.

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

    DOEpatents

    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.

  13. 77 FR 74697 - Meeting of the ACRS, Subcommittee on U.S. Evolutionary Power Reactor; Notice of Meeting

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-12-17

    ... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS) Meeting of the ACRS, Subcommittee on U.S. Evolutionary Power Reactor; Notice of Meeting The ACRS Subcommittee on U.S. Evolutionary Power Reactor (U.S. EPR) will hold a meeting on January 17,...

  14. Sodium coolant purification systems for a nuclear power station equipped with a BN-1200 reactor

    NASA Astrophysics Data System (ADS)

    Alekseev, V. V.; Kovalev, Yu. P.; Kalyakin, S. G.; Kozlov, F. A.; Kumaev, V. Ya.; Kondrat'ev, A. S.; Matyukhin, V. V.; Pirogov, E. P.; Sergeev, G. P.; Sorokin, A. P.; Torbenkova, I. Yu.

    2013-05-01

    Both traditional coolant purification methods (by means of traps and sorbents for removing cesium), the use of which supported successful operation of nuclear power installations equipped with fast-neutron reactors with a sodium coolant, and the possibility of removing oxygen from sodium through the use of hot traps are analyzed in substantiating the purification system for a nuclear power station equipped with a BN-1200 reactor. It is shown that a cold trap built into the reactor vessel must be a mandatory component of the reactor plant primary coolant circuit's purification system. The use of hot traps allows oxygen to be removed from the sodium coolant down to permissible concentrations when the nuclear power station operates in its rated mode. The main lines of works aimed at improving the performance characteristics of cold traps are suggested based on the results of performed investigations.

  15. Power distributions in fresh and depleted LEU and HEU cores of the MITR reactor.

    SciTech Connect

    Wilson, E.H.; Horelik, N.E.; Dunn, F.E.; Newton, T.H., Jr.; Hu, L.; Stevens, J.G.

    2012-04-04

    The Massachusetts Institute of Technology Reactor (MITR-II) is a research reactor in Cambridge, Massachusetts designed primarily for experiments using neutron beam and in-core irradiation facilities. It delivers a neutron flux comparable to current LWR power reactors in a compact 6 MW core using Highly Enriched Uranium (HEU) fuel. In the framework of its non-proliferation policies, the international community presently aims to minimize the amount of nuclear material available that could be used for nuclear weapons. In this geopolitical context, most research and test reactors both domestic and international have started a program of conversion to the use of Low Enriched Uranium (LEU) fuel. A new type of LEU fuel based on an alloy of uranium and molybdenum (UMo) is expected to allow the conversion of U.S. domestic high performance reactors like the MITR-II reactor. Toward this goal, core geometry and power distributions are presented. Distributions of power are calculated for LEU cores depleted with MCODE using an MCNP5 Monte Carlo model. The MCNP5 HEU and LEU MITR models were previously compared to experimental benchmark data for the MITR-II. This same model was used with a finer spatial depletion in order to generate power distributions for the LEU cores. The objective of this work is to generate and characterize a series of fresh and depleted core peak power distributions, and provide a thermal hydraulic evaluation of the geometry which should be considered for subsequent thermal hydraulic safety analyses.

  16. Thermonuclear inverse magnetic pumping power cycle for stellarator reactor

    DOEpatents

    Ho, Darwin D.; Kulsrud, Russell M.

    1991-01-01

    The plasma column in a stellarator is compressed and expanded alternatively in minor radius. First a plasma in thermal balance is compressed adiabatically. The volume of the compressed plasma is maintained until the plasma reaches a new thermal equilibrium. The plasma is then expanded to its original volume. As a result of the way a stellarator works, the plasma pressure during compression is less than the corresponding pressure during expansion. Therefore, negative work is done on the plasma over a complete cycle. This work manifests itself as a back-voltage in the toroidal field coils. Direct electrical energy is obtained from this voltage. Alternatively, after the compression step, the plasma can be expanded at constant pressure. The cycle can be made self-sustaining by operating a system of two stellarator reactors in tandem. Part of the energy derived from the expansion phase of a first stellarator reactor is used to compress the plasma in a second stellarator reactor.

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

    SciTech Connect

    Myers, Carl W; Elkins, Ned Z

    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.

  18. Nuclear reactor descriptions for space power systems analysis

    NASA Technical Reports Server (NTRS)

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

    1972-01-01

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

  19. DynMo: Dynamic Simulation Model for Space Reactor Power Systems

    NASA Astrophysics Data System (ADS)

    El-Genk, Mohamed; Tournier, Jean-Michel

    2005-02-01

    A Dynamic simulation Model (DynMo) for space reactor power systems is developed using the SIMULINK® platform. DynMo is modular and could be applied to power systems with different types of reactors, energy conversion, and heat pipe radiators. This paper presents a general description of DynMo-TE for a space power system powered by a Sectored Compact Reactor (SCoRe) and that employs off-the-shelf SiGe thermoelectric converters. SCoRe is liquid metal cooled and designed for avoidance of a single point failure. The reactor core is divided into six equal sectors that are neutronically, but not thermal-hydraulically, coupled. To avoid a single point failure in the power system, each reactor sector has its own primary and secondary loops, and each loop is equipped with an electromagnetic (EM) pump. A Power Conversion assembly (PCA) and a Thermoelectric Conversion Assembly (TCA) of the primary and secondary EM pumps thermally couple each pair of a primary and a secondary loop. The secondary loop transports the heat rejected by the PCA and the pumps TCA to a rubidium heat pipes radiator panel. The primary loops transport the thermal power from the reactor sector to the PCAs for supplying a total of 145-152 kWe to the load at 441-452 VDC, depending on the selections of the primary and secondary liquid metal coolants. The primary and secondary coolant combinations investigated are lithium (Li)/Li, Li/sodium (Na), Na-Na, Li/NaK-78 and Na/NaK-78, for which the reactor exit temperature is kept below 1250 K. The results of a startup transient of the system from an initial temperature of 500 K are compared and discussed.

  20. DynMo: Dynamic Simulation Model for Space Reactor Power Systems

    SciTech Connect

    El-Genk, Mohamed; Tournier, Jean-Michel

    2005-02-06

    A Dynamic simulation Model (DynMo) for space reactor power systems is developed using the SIMULINK registered platform. DynMo is modular and could be applied to power systems with different types of reactors, energy conversion, and heat pipe radiators. This paper presents a general description of DynMo-TE for a space power system powered by a Sectored Compact Reactor (SCoRe) and that employs off-the-shelf SiGe thermoelectric converters. SCoRe is liquid metal cooled and designed for avoidance of a single point failure. The reactor core is divided into six equal sectors that are neutronically, but not thermal-hydraulically, coupled. To avoid a single point failure in the power system, each reactor sector has its own primary and secondary loops, and each loop is equipped with an electromagnetic (EM) pump. A Power Conversion assembly (PCA) and a Thermoelectric Conversion Assembly (TCA) of the primary and secondary EM pumps thermally couple each pair of a primary and a secondary loop. The secondary loop transports the heat rejected by the PCA and the pumps TCA to a rubidium heat pipes radiator panel. The primary loops transport the thermal power from the reactor sector to the PCAs for supplying a total of 145-152 kWe to the load at 441-452 VDC, depending on the selections of the primary and secondary liquid metal coolants. The primary and secondary coolant combinations investigated are lithium (Li)/Li, Li/sodium (Na), Na-Na, Li/NaK-78 and Na/NaK-78, for which the reactor exit temperature is kept below 1250 K. The results of a startup transient of the system from an initial temperature of 500 K are compared and discussed.

  1. SVBR-100 module-type fast reactor of the IV generation for regional power industry

    NASA Astrophysics Data System (ADS)

    Zrodnikov, A. V.; Toshinsky, G. I.; Komlev, O. G.; Stepanov, V. S.; Klimov, N. N.

    2011-08-01

    In the report the following is presented: basic conceptual provisions of the innovative nuclear power technology (NPT) based on modular fast reactors (FR) SVBR-100, summarized results of calculations of the reactor, analysis of the opportunities of multi-purpose application of such reactor facilities (RF) including export potentials with due account of nonproliferation requirements. The most important features of the proposed NPT analyzed in the report are as follows: (1) integral (monoblock) arrangement of the primary circuit equipment with entire elimination of the primary circuit pipelines and valves that considerably reduces the construction and assembly works period and coupling with high boiling point of lead-bismuth coolant (LBC) deterministically eliminates accidents of the LOCA type, (2) option for 100 MWe power and dimensions of the reactor provide: on the one hand, an opportunity to transport the reactor monoblock in factory-readiness by railway as well as other kinds of transport, on the other hand, core breeding ratio (CBR) exceeds 1 while MOX-fuel is used. The preferable area of application of RF SVBR-100 is regional and small power requiring power-units of electric power in a range of (100-600) MW, which could be used for cogeneration-based district heating while locating them nearby cities as well as for generation of electric power in a mode of load tracking in the regions with low network systems.

  2. 10 CFR 50.36a - Technical specifications on effluents from nuclear power reactors.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... 10 Energy 1 2013-01-01 2013-01-01 false Technical specifications on effluents from nuclear power reactors. 50.36a Section 50.36a Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF PRODUCTION AND...; Ineligibility of Certain Applicants § 50.36a Technical specifications on effluents from nuclear power...

  3. 10 CFR 50.36a - Technical specifications on effluents from nuclear power reactors.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... 10 Energy 1 2012-01-01 2012-01-01 false Technical specifications on effluents from nuclear power reactors. 50.36a Section 50.36a Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF PRODUCTION AND...; Ineligibility of Certain Applicants § 50.36a Technical specifications on effluents from nuclear power...

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

    DOEpatents

    LeVert, Francis E.; Cox, Samson A.

    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.

  5. A computer program to determine the specific power of prismatic-core reactors

    SciTech Connect

    Dobranich, D.

    1987-05-01

    A computer program has been developed to determine the maximum specific power for prismatic-core reactors as a function of maximum allowable fuel temperature, core pressure drop, and coolant velocity. The prismatic-core reactors consist of hexagonally shaped fuel elements grouped together to form a cylindrically shaped core. A gas coolant flows axially through circular channels within the elements, and the fuel is dispersed within the solid element material either as a composite or in the form of coated pellets. Different coolant, fuel, coating, and element materials can be selected to represent different prismatic-core concepts. The computer program allows the user to divide the core into any arbitrary number of axial levels to account for different axial power shapes. An option in the program allows the automatic determination of the core height that results in the maximum specific power. The results of parametric specific power calculations using this program are presented for various reactor concepts.

  6. Evaluation of Launch Accident Safety Options for Low-Power Surface Reactors

    NASA Astrophysics Data System (ADS)

    Fung Poon, Cindy; Poston, David I.

    2006-01-01

    Safety options for surface reactors of less than 800 kW (thermal power) are analyzed. The concepts under consideration are heat pipe cooled reactors fueled with either uranium nitride or uranium dioxide. This study investigates the impact of launch accident criteria on the system mass, while ensuring the mechanical integrity and reliability of the system through launch accident scenarios. The four criticality scenarios analyzed for shutdown determination are dry sand surround with reflectors stripped, water submersion on concrete, water submersion with all control drums in, and the nominal shutdown reactor condition. Additionally the following two operational criteria are analyzed: reactor is warm and swelled, and reactor is warm and swelled with one drum in (where swelled includes both thermal mechanical expansion and irradiation induced swelling of the fuel).

  7. Evaluation of Launch Accident Safety Options for Low-Power Surface Reactors

    SciTech Connect

    Fung Poon, Cindy; Poston, David I.

    2006-01-20

    Safety options for surface reactors of less than 800 kW (thermal power) are analyzed. The concepts under consideration are heat pipe cooled reactors fueled with either uranium nitride or uranium dioxide. This study investigates the impact of launch accident criteria on the system mass, while ensuring the mechanical integrity and reliability of the system through launch accident scenarios. The four criticality scenarios analyzed for shutdown determination are dry sand surround with reflectors stripped, water submersion on concrete, water submersion with all control drums in, and the nominal shutdown reactor condition. Additionally the following two operational criteria are analyzed: reactor is warm and swelled, and reactor is warm and swelled with one drum in (where swelled includes both thermal mechanical expansion and irradiation induced swelling of the fuel)

  8. Subcritical power reactor with irradiation by a beam of accelerated protons

    SciTech Connect

    Ado, Yu.M.; Kryuchkov, V.P.; Lebedev, V.N.

    1995-04-01

    The physical and economic aspects of constructing a reactivity accident-free nuclear reactor are discussed. The approach described is based on uranium fission in a deeply subcritical reactor in which the chain reaction is initiated by an external source of neutrons, thus eliminating runaway. Protons are assumed to be the primary particles because the accelerator technology is best developed for this method of irradiation. A subcritical reactor and a high-power proton accelerator is determined to be sound in principle, and has the advantages of eliminating runaway accidents, decreasing fuel costs, higher efficiency due to increased intervals between fuel loadings, and controlling the reactor power and shielding by changing the beam current of the accelerator. 29 refs., 8 figs., 2 tabs.

  9. Power generation from nuclear reactors in aerospace applications

    SciTech Connect

    English, R.E.

    1982-01-01

    Power generation in nuclear powerplants in space is addressed. In particular, the states of technology of the principal competitive concepts for power generation are assessed. The possible impact of power conditioning on power generation is also discussed. For aircraft nuclear propulsion, the suitability of various technologies is cursorily assessed for flight in the Earth's atmosphere. A program path is suggested to ease the conditions of first use of aircraft nuclear propulsion.

  10. Power Generation from Nuclear Reactors in Aerospace Applications

    NASA Technical Reports Server (NTRS)

    English, Robert E.

    1982-01-01

    Power generation in nuclear powerplants in space is addressed. In particular, the states of technology of the principal competitive concepts for power generation are assessed. The possible impact of power conditioning on power generation is also discussed. For aircraft nuclear propulsion, the suitability of various technologies is cursorily assessed for flight in the Earth's atmosphere; a program path is suggested to ease the conditions of first use of aircraft nuclear propulsion.

  11. Nuclear Reactors for Space Power, Understanding the Atom Series.

    ERIC Educational Resources Information Center

    Corliss, William R.

    The historical development of rocketry and nuclear technology includes a specific description of Systems for Nuclear Auxiliary Power (SNAP) programs. Solar cells and fuel cells are considered as alternative power supplies for space use. Construction and operation of space power plants must include considerations of the transfer of heat energy to…

  12. Fuel supply of nuclear power industry with the introduction of fast reactors

    NASA Astrophysics Data System (ADS)

    Muraviev, E. V.

    2014-12-01

    The results of studies conducted for the validation of the updated development strategy for nuclear power industry in Russia in the 21st century are presented. Scenarios with different options for the reprocessing of spent fuel of thermal reactors and large-scale growth of nuclear power industry based on fast reactors of inherent safety with a breeding ratio of ˜1 in a closed nuclear fuel cycle are considered. The possibility of enhanced fuel breeding in fast reactors is also taken into account in the analysis. The potential to establish a large-scale nuclear power industry that covers 100% of the increase in electric power requirements in Russia is demonstrated. This power industry may be built by the end of the century through the introduction of fast reactors (replacing thermal ones) with a gross uranium consumption of up to ˜1 million t and the termination of uranium mining even if the reprocessing of spent fuel of thermal reactors is stopped or suffers a long-term delay.

  13. Enhanced situation awareness and decision making for an intelligent reconfigurable reactor power controller

    SciTech Connect

    Kenney, S.J.; Edwards, R.M.

    1996-07-01

    A Learning Automata based intelligent reconfigurable controller has been adapted for use as a reactor power controller to achieve improved reactor temperature performance. The intelligent reconfigurable controller is capable of enforcing either a classical or an optimal reactor power controller based on control performance feedback. Four control performance evaluation measures: dynamically estimated average quadratic temperature error, power, rod reactivity and rod reactivity rate were developed to provide feedback to the control decision component of the intelligent reconfigurable controller. Fuzzy Logic and Neural Network controllers have been studied for inclusion in the bank of controllers that form the intermediate level of an enhanced intelligent reconfigurable reactor power controller (IRRPC). The increased number of alternatives available to the supervisory level of the IRRPC requires enhanced situation awareness. Additional performance measures have been designed and a method for synthesizing them into a single indication of the overall performance of the currently enforced reactor power controller has been conceptualized. Modification of the reward/penalty scheme implemented in the existing IRRPC to increase the quality of the supervisory level decision process has been studied. The logogen model of human memory (Morton, 1969) and individual controller design information could be used to allocate reward to the most appropriate controller. Methods for allocating supervisory level attention were also studied with the goal of maximizing learning rate.

  14. Hot zero power reactor calculations using the Insilico code

    DOE PAGESBeta

    Hamilton, Steven P.; Evans, Thomas M.; Davidson, Gregory G.; Johnson, Seth R.; Pandya, Tara M.; Godfrey, Andrew T.

    2016-03-18

    In this paper we describe the reactor physics simulation capabilities of the insilico code. A description of the various capabilities of the code is provided, including detailed discussion of the geometry, meshing, cross section processing, and neutron transport options. Numerical results demonstrate that the insilico SPN solver with pin-homogenized cross section generation is capable of delivering highly accurate full-core simulation of various PWR problems. Comparison to both Monte Carlo calculations and measured plant data is provided.

  15. Ultrahigh temperature vapor core reactor-MHD system for space nuclear electric power

    NASA Technical Reports Server (NTRS)

    Maya, Isaac; Anghaie, Samim; Diaz, Nils J.; Dugan, Edward T.

    1991-01-01

    The conceptual design of a nuclear space power system based on the ultrahigh temperature vapor core reactor with MHD energy conversion is presented. This UF4 fueled gas core cavity reactor operates at 4000 K maximum core temperature and 40 atm. Materials experiments, conducted with UF4 up to 2200 K, demonstrate acceptable compatibility with tungsten-molybdenum-, and carbon-based materials. The supporting nuclear, heat transfer, fluid flow and MHD analysis, and fissioning plasma physics experiments are also discussed.

  16. Materials technology for an advanced space power nuclear reactor concept: Program summary

    NASA Technical Reports Server (NTRS)

    Gluyas, R. E.; Watson, G. K.

    1975-01-01

    The results of a materials technology program for a long-life (50,000 hr), high-temperature (950 C coolant outlet), lithium-cooled, nuclear space power reactor concept are reviewed and discussed. Fabrication methods and compatibility and property data were developed for candidate materials for fuel pins and, to a lesser extent, for potential control systems, reflectors, reactor vessel and piping, and other reactor structural materials. The effects of selected materials variables on fuel pin irradiation performance were determined. The most promising materials for fuel pins were found to be 85 percent dense uranium mononitride (UN) fuel clad with tungsten-lined T-111 (Ta-8W-2Hf).

  17. Movable-molybdenum-reflector reactivity experiments for control studies of compact space power reactor concepts

    NASA Technical Reports Server (NTRS)

    Fox, T. A.

    1973-01-01

    An experimental reflector reactivity study was made with a compact cylindrical reactor using a uranyl fluoride - water fuel solution. The reactor was axially unreflected and radially reflected with segments of molybdenum. The reflector segments were displaced incrementally in both the axial and radial dimensions, and the shutdown of each configuration was measured by using the pulsed-neutron source technique. The reactivity effects for axial and radial displacement of reflector segments are tabulated separately and compared. The experiments provide data for control-system studies of compact-space-power-reactor concepts.

  18. 78 FR 64029 - Cost-Benefit Analysis for Radwaste Systems for Light-Water-Cooled Nuclear Power Reactors

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-10-25

    ... COMMISSION Cost-Benefit Analysis for Radwaste Systems for Light-Water-Cooled Nuclear Power Reactors AGENCY... Systems for Light-Water-Cooled Nuclear Power Reactors,'' in which the NRC made editorial corrections and... analysis for liquid and gaseous radwaste system components for light water nuclear power...

  19. Discharge Characteristics of Series Surface/Packed-Bed Discharge Reactor Diven by Bipolar Pulsed Power

    NASA Astrophysics Data System (ADS)

    Hu, Jian; Jiang, Nan; Li, Jie; Shang, Kefeng; Lu, Na; Wu, Yan; Mizuno, Akira

    2016-03-01

    The discharge characteristics of the series surface/packed-bed discharge (SSPBD) reactor driven by bipolar pulse power were systemically investigated in this study. In order to evaluate the advantages of the SSPBD reactor, it was compared with traditional surface discharge (SD) reactor and packed-bed discharge (PBD) reactor in terms of the discharge voltage, discharge current, and ozone formation. The SSPBD reactor exhibited a faster rising time and lower tail voltage than the SD and PBD reactors. The distribution of the active species generated in different discharge regions of the SSPBD reactor was analyzed by optical emission spectra and ozone analysis. It was found that the packed-bed discharge region (3.5 mg/L), rather than the surface discharge region (1.3 mg/L) in the SSPBD reactor played a more important role in ozone generation. The optical emission spectroscopy analysis indicated that more intense peaks of the active species (e.g. N2 and OI) in the optical emission spectra were observed in the packed-bed region. supported by National Natural Science Foundation of China (No. 51177007), the Joint Funds of National Natural Science Foundation of China (No. U1462105), and Dalian University of Technology Fundamental Research Fund of China (No. DUT15RC(3)030)

  20. Analysis of N-16 concentration in primary cooling system of AP1000 power reactor

    SciTech Connect

    Rohanda, Anis; Waris, Abdul

    2015-04-16

    Nitrogen-16 (N-16) is one of the radiation safety parameter on the primary reactor system. The activation product, N-16, is the predominant contributor to the activity in the reactor coolant system during reactor operation. N-16 is activation product derived from activation of O-16 with fast neutron based on {sup 16}O(n,p){sup 16}N reaction. Thus study is needed and it performs to determine N-16 concentration in reactor coolant (primary coolant) in supporting radiation safety. One of the way is using analytical methode based on activation and redecay princip to obtain N-16 concentration. The analysis was performed on the configuration basis and operational of Westinghouse AP1000 power reactor in several monitoring points at coolant reactor system. The results of the calculation of N-16 concentration at the core outlet, reactor vessel outlet, pressurizer line, inlet and outlet of steam generators, primary pumps, reactor vessels inlet and core inlet are: 281, 257, 255, 250, 145, 142, 129 and 112 µCi/gram respectively. The results of analysis compared with AP1000 design control document as standard values. The verification showed very high accuracy comparation between analytical results and standard values.

  1. Analysis of N-16 concentration in primary cooling system of AP1000 power reactor

    NASA Astrophysics Data System (ADS)

    Rohanda, Anis; Waris, Abdul

    2015-04-01

    Nitrogen-16 (N-16) is one of the radiation safety parameter on the primary reactor system. The activation product, N-16, is the predominant contributor to the activity in the reactor coolant system during reactor operation. N-16 is activation product derived from activation of O-16 with fast neutron based on 16O(n,p)16N reaction. Thus study is needed and it performs to determine N-16 concentration in reactor coolant (primary coolant) in supporting radiation safety. One of the way is using analytical methode based on activation and redecay princip to obtain N-16 concentration. The analysis was performed on the configuration basis and operational of Westinghouse AP1000 power reactor in several monitoring points at coolant reactor system. The results of the calculation of N-16 concentration at the core outlet, reactor vessel outlet, pressurizer line, inlet and outlet of steam generators, primary pumps, reactor vessels inlet and core inlet are: 281, 257, 255, 250, 145, 142, 129 and 112 µCi/gram respectively. The results of analysis compared with AP1000 design control document as standard values. The verification showed very high accuracy comparation between analytical results and standard values.

  2. Direct Estimation of Power Distribution in Reactors for Nuclear Thermal Space Propulsion

    NASA Astrophysics Data System (ADS)

    Aldemir, Tunc; Miller, Don W.; Burghelea, Andrei

    2004-02-01

    A recently proposed constant temperature power sensor (CTPS) has the capability to directly measure the local power deposition rate in nuclear reactor cores proposed for space thermal propulsion. Such a capability reduces the uncertainties in the estimated power peaking factors and hence increases the reliability of the nuclear engine. The CTPS operation is sensitive to the changes in the local thermal conditions. A procedure is described for the automatic on-line calibration of the sensor through estimation of changes in thermal .conditions.

  3. The SP-100 space reactor as a power source for Mars exploration missions

    NASA Technical Reports Server (NTRS)

    Isenberg, Lon; Heller, Jack A.

    1989-01-01

    This paper argues that many of the power requirements of complex, relatively long-duration space missions such as the exploration of Mars may best be met through the use of power systems which use nuclear reactors as a thermal energy source. The development of such a power system, the SP-100, and its application in Mars mission scenarios is described. The missions addressed include a freighter mission and a mission involving exploration of the Martian surface.

  4. Direct Estimation of Power Distribution in Reactors for Nuclear Thermal Space Propulsion

    SciTech Connect

    Aldemir, Tunc; Miller, Don W.; Burghelea, Andrei

    2004-02-04

    A recently proposed constant temperature power sensor (CTPS) has the capability to directly measure the local power deposition rate in nuclear reactor cores proposed for space thermal propulsion. Such a capability reduces the uncertainties in the estimated power peaking factors and hence increases the reliability of the nuclear engine. The CTPS operation is sensitive to the changes in the local thermal conditions. A procedure is described for the automatic on-line calibration of the sensor through estimation of changes in thermal conditions.

  5. Calculated power distribution of a thermionic, beryllium oxide reflected, fast-spectrum reactor

    NASA Technical Reports Server (NTRS)

    Mayo, W.; Lantz, E.

    1973-01-01

    A procedure is developed and used to calculate the detailed power distribution in the fuel elements next to a beryllium oxide reflector of a fast-spectrum, thermionic reactor. The results of the calculations show that, although the average power density in these outer fuel elements is not far from the core average, the power density at the very edge of the fuel closest to the beryllium oxide is about 1.8 times the core avearge.

  6. Fusion-power-core design of a Compact Reversed-Field Pinch Reactor (CRFPR)

    NASA Astrophysics Data System (ADS)

    Copenhaver, C.; Schnurr, N. M.; Krakowski, R. A.; Hagenson, R. L.; Mynard, R. C.; Cappiello, C.; Lujan, R. E.; Davidson, J. W.; Chaffee, A. D.; Battat, M. E.

    A conceptual design of a fusion power core (FPC, i.e., plasma chamber, first wall, blanket, shield, coils) based on a Reversed-Field Pinch (RFP) has been completed. After a brief statement of rationale and description of the reactor configuraton, the FPC integration is described in terms of power balance, thermal-hydraulics, and mechanical design. The engineering versatility, promise, and problems of this high-power-density approach to fusion are addressed.

  7. Conceptual Design of a 100-kWe Space Nuclear Reactor Power System with High-Power AMTEC

    NASA Astrophysics Data System (ADS)

    El-Genk, Mohamed S.; Tournier, Jean-Michel

    2003-01-01

    Alkali Metal Thermal-to-Electric Conversion (AMTEC), although currently at a Technology Readiness Level-3 (TRL-3), has an excellent potential for use in Space Nuclear Reactor Power (SNRP) systems for NASA's deep-space exploration missions. In addition to operating at a conversion efficiency > 20%, representing the highest fraction (> 60%) of Carnot efficiency of all other static and dynamic conversion technology options, the relatively high heat rejection radiator temperature (650-700 K) reduces the size and mass of the radiator and of the SNRP system. A high-power AMTEC unit design has been developed and optimized for operating at reactor exit temperatures <= 1180 K and radiator temperature <= 680 K. Depending on the reactor exit temperature, the nominal electrical power of the AMTEC unit, measuring 594 mm × 410 mm × 115 mm and weighting 44.3 kg, could be as high as 5.6 kWe, with a margin of >= 5% for an additional load-following increase. A conceptual design of a 100 kWe SNRP system with these high-power AMTEC units is developed and presented in this paper. The total mass of major subsystems, including the converters, nuclear reactor, shadow radiation shield, and radiator, is calculated and compared with that for the SP-100. Despite the large specific mass of the AMTEC units compared to the SiGe thermoelectrics in the SP-100 system, the lower masses of the reactor, radiation shield, and radiator make the present AMTEC-SNRP system > 26% lighter, for the same electrical power. An optimized AMTEC-SNRP system could potentially operate at a specific power > 30 We/kg (or specific mass < 33 kg/kWe), use non-refractory structures of super-steel alloys with well-know properties, relatively low density, low Ductile-To-Brittle (DTB) transition temperatures, and good compatibility with space and planetary environments containing CO2 and oxygen. The radiator area for the baseline 100 kWe AMTEC SNRP system is < 27 m2, which, together with operating the potassium heat pipes

  8. Analysis and evaluation of ZPPR (Zero Power Physics Reactor) critical experiments for a 100 kilowatt-electric space reactor

    SciTech Connect

    McFarlane, H.F.; Collins, P.J.; Carpenter, S.G.; Olsen, D.N.; Smith, D.M.; Schaefer, R.W. ); Doncals, R.A.; Andre, S.V.; Porter, C.A. ); Cowan, C.L; Stewart, S.L.; Protsik, R. . Astro Space Div.)

    1990-01-01

    ZPPR critical experiments were used for physics testing the reactor design of the SP-100, a 100-kW thermoelectric LMR that is being developed to provide electrical power for space applications. These tests validated all key physics characteristics of the design, including the ultimate safety in the event of a launch or re-entry accident. Both the experiments and the analysis required the use of techniques not previously applied to fast reactor designs. A few significant discrepancies between the experimental and calculated results leave opportunities for further optimization. An initial investigation has been made into application of the ZPPR-20 results, along with those of other relevant integral data, to the SP-100 design. 13 refs., 5 figs., 7 tabs.

  9. Design Concept for a Nuclear Reactor-Powered Mars Rover

    NASA Technical Reports Server (NTRS)

    Elliott, John; Poston, Dave; Lipinski, Ron

    2007-01-01

    A report presents a design concept for an instrumented robotic vehicle (rover) to be used on a future mission of exploration of the planet Mars. The design incorporates a nuclear fission power system to provide long range, long life, and high power capabilities unachievable through the use of alternative solar or radioisotope power systems. The concept described in the report draws on previous rover designs developed for the 2009 Mars Science laboratory (MSL) mission to minimize the need for new technology developments.

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

    NASA Technical Reports Server (NTRS)

    George, Jeffrey A.

    1991-01-01

    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. Shield materials recommended for space power nuclear reactors

    NASA Technical Reports Server (NTRS)

    Kaszubinski, L. J.

    1973-01-01

    Lithium hydride is recommended for neutron attenuation and depleted uranium is recommended for gamma ray attenuation. For minimum shield weights these materials must be arranged in alternate layers to attenuate the secondary gamma rays efficiently. In the regions of the shield near the reactor, where excessive fissioning occurs in the uranium, a tungsten alloy is used instead. Alloys of uranium such as either the U-0.5Ti or U-8Mo are available to accommodate structural requirements. The zone-cooled casting process is recommended for lithium hydride fabrication. Internal honeycomb reinforcement to control cracks in the lithium hydride is recommended.

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

    SciTech Connect

    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.

  13. Lunar in-core thermionic nuclear reactor power system conceptual design

    NASA Technical Reports Server (NTRS)

    Mason, Lee S.; Schmitz, Paul C.; Gallup, Donald R.

    1991-01-01

    This paper presents a conceptual design of a lunar in-core thermionic reactor power system. The concept consists of a thermionic reactor located in a lunar excavation with surface mounted waste heat radiators. The system was integrated with a proposed lunar base concept representative of recent NASA Space Exploration Initiative studies. The reference mission is a permanently-inhabited lunar base requiring a 550 kWe, 7 year life central power station. Performance parameters and assumptions were based on the Thermionic Fuel Element (TFE) Verification Program. Five design cases were analyzed ranging from conservative to advanced. The cases were selected to provide sensitivity effects on the achievement of TFE program goals.

  14. Rotating-bed reactor as a power source for EM gun applications

    SciTech Connect

    Powell, J.; Botts, T.; Stickley, C.M.; Meth, S.

    1980-01-01

    Electromagnetic gun applications of the Rotating Bed Reactor (RBR) are examined. The RBR is a compact (approx. 1 m/sup 3/), (up to several thousand MW(th)), high-power reactor concept, capable of producing a high-temperature (up to approx. 300/sup 0/K) gas stream with a MHD generator coupled to it, the RBR can generate electric power (up to approx. 1000 MW(e)) in the pulsed or cw modes. Three EM gun applications are investigated: a rail gun thruster for orbit transfer, a rapid-fire EM gun for point defense, and a direct ground-to-space launch. The RBR appears suitable for all applications.

  15. High Efficiency Nuclear Power Plants using Liquid Fluoride Thorium Reactor Technology

    NASA Technical Reports Server (NTRS)

    Juhasz, Albert J.; Rarick, Richard A.; Rangarajan, Rajmohan

    2009-01-01

    An overall system analysis approach is used to propose potential conceptual designs of advanced terrestrial nuclear power plants based on Oak Ridge National Laboratory (ORNL) Molten Salt Reactor (MSR) experience and utilizing Closed Cycle Gas Turbine (CCGT) thermal-to-electric energy conversion technology. In particular conceptual designs for an advanced 1 GWe power plant with turbine reheat and compressor intercooling at a 950 K turbine inlet temperature (TIT), as well as near term 100 MWe demonstration plants with TITS of 950 K and 1200 K are presented. Power plant performance data were obtained for TITS ranging from 650 to 1300 K by use of a Closed Brayton Cycle (CBC) systems code which considered the interaction between major sub-systems, including the Liquid Fluoride Thorium Reactor (LFTR), heat source and heat sink heat exchangers, turbo -generator machinery, and an electric power generation and transmission system. Optional off-shore submarine installation of the power plant is a major consideration.

  16. High Efficiency Nuclear Power Plants Using Liquid Fluoride Thorium Reactor Technology

    NASA Technical Reports Server (NTRS)

    Juhasz, Albert J.; Rarick, Richard A.; Rangarajan, Rajmohan

    2009-01-01

    An overall system analysis approach is used to propose potential conceptual designs of advanced terrestrial nuclear power plants based on Oak Ridge National Laboratory (ORNL) Molten Salt Reactor (MSR) experience and utilizing Closed Cycle Gas Turbine (CCGT) thermal-to-electric energy conversion technology. In particular conceptual designs for an advanced 1 GWe power plant with turbine reheat and compressor intercooling at a 950 K turbine inlet temperature (TIT), as well as near term 100 MWe demonstration plants with TITs of 950 and 1200 K are presented. Power plant performance data were obtained for TITs ranging from 650 to 1300 K by use of a Closed Brayton Cycle (CBC) systems code which considered the interaction between major sub-systems, including the Liquid Fluoride Thorium Reactor (LFTR), heat source and heat sink heat exchangers, turbo-generator machinery, and an electric power generation and transmission system. Optional off-shore submarine installation of the power plant is a major consideration.

  17. Loop system with gas control of the power production in the MR reactor

    SciTech Connect

    Andreev, V.I.; Kolyadin, V.I.; Smirnov, A.I.; Yakovlev, V.V.

    1987-01-01

    An unsolved problem in reactor design is premature fuel-pin failure on account of mechanical interaction between the fuel and the sheath under nonstationary operating conditions. To examine the effects of this interaction on the viability, the authors have built an experimental system with the MR reactor. To provide for varying the power production over wide ranges, gas regulation based on /sup 3/He as neutron absorber is used. A map of core loading in the MR reactor is provided and variation in power in the experimental fuel assembly in accordance with /sup 3/He pressure and control location is shown. A structural diagram shows the reactor apparatus with gas power control in the experimental pin assembly. The relative changes in channel power in relation to neutron absorber pressure in GCU in channel 1-4 are presented. The results are offered on the power variation in the experimental assembly and reactivity as functions of /sup 3/He pressure in the GCU, together with the calculated data.

  18. Triga Mark III Reactor Operating Power and Neutron Flux Study by Nuclear Track Methodology

    NASA Astrophysics Data System (ADS)

    Espinosa, G.; Golzarri, J. I.; Raya-Arredondo, R.; Cruz-Galindo, S.; Sajo-Bohus, L.

    The operating power of a TRIGA Mark III reactor was studied using Nuclear Track Methodology (NTM). The facility has a Highly Enriched Uranium core that provides a neutron flux of around 2 x 1012 n cm-2 s-1 in the TO-2 irradiation channel. The detectors consisted of a Landauer® CR-39 (allyl diglycol polycarbonate) chip covered with a 3 mm Plexiglas® converter. After irradiation, the detectors were chemically etched in a 6.25M-KOH solution at 60±1 °C for 6 h. Track density was determined by a custom-made Digital Image Analysis System. The results show a direct proportionality between reactor power and average nuclear track density for powers in the range 0.1-7 kW. Data reproducibility and relatively low uncertainty (±3%) were achieved. NTM is a simple, fast and reliable technique that can serve as a complementary procedure to measure reactor operating power. It offers the possibility of calibrating the neutron flux density in any low power reactor.

  19. Installation of the Light-Water Breeder Reactor at the Shippingport Atomic Power Station (LWBR Development Program)

    SciTech Connect

    Massimino, R.J.; Williams, D.A.

    1983-05-01

    This report summarizes the refueling operations performed to install a Light Water Breeder Reactor (LWBR) core into the existing pressurized water reactor vessel at the Shippingport Atomic Power Station. Detailed descriptions of the major installation operations (e.g., primary system preconditioning, fuel installation, pressure boundary seal welding) are included as appendices to this report; these operations are of technical interest to any reactor servicing operation, whether the reactor is a breeder or a conventional light water non-breeder core.

  20. Split-core heat-pipe reactors for out-of-pile thermionic power systems.

    NASA Technical Reports Server (NTRS)

    Niederauer, G.; Lantz, E.; Breitweiser, R.

    1971-01-01

    Description of the concept of splitting a heat-pipe reactor for out-of-core thermionics into two identical halves and using the resulting center gap for reactivity control. Short Li-W reactor heat pipes penetrate the axial reflectors and form a heat exchanger with long heat pipes which wind through the shield to the thermionic diodes. With one reactor half anchored to the shield, the other is attached to a long arm with a pivot behind the shield and swings through a small arc for reactivity control. A safety shim prevents large reactivity inputs, and a fueled control arm drive shaft acts as a power stabilizer. Reactors fueled with U-235C and with U-233C have been studied.-

  1. Partial site release at a power reactor facility.

    PubMed

    Darman, Joseph; Whitney, Michael; Dubiel, Richard

    2004-01-01

    U.S. NRC licensed facilities undergoing decommissioning may wish to remove portions of their site from the jurisdiction of their license, prior to final license termination. The method of partial site release, relevant to radiological conditions, described herein employs NUREG-1505 methodology for demonstrating indistinguishability from background. The partial site release process was also informed by NRC Regulatory Issue Summary 2000-19 "Partial Release of Reactor Site for Unrestricted Use Before NRC Approval of the License Termination Plan." However, the focus of this discussion is the radiological aspects of partial site release, relevant to the implementation of NUREG-1505 methodology for demonstrating indistinguishability from background, based on the 137Cs concentrations at the site and a suitable background reference area. This type of approach was found acceptable by the NRC, and the partial site release was granted. PMID:14695009

  2. Novel, Integrated Reactor / Power Conversion System (LMR-AMTEC)

    SciTech Connect

    Pablo Rubiolo, Principal Investigator

    2003-03-21

    The main features of this project were the development of a long life (up to 10 years) Liquid Metal Reactor (LMR) and a static conversion subsystem comprising an Alkali Metal Thermal-to-Electric (AMTEC) topping cycle and a ThermoElectric (TE) Bottom cycle. Various coupling options of the LMR with the energy conversion subsystem were explored and, base in the performances found in this analysis, an Indirect Coupling (IC) between the LMR and the AMTEC/TE converters with Alkali Metal Boilers (AMB) was chosen as the reference design. The performance model of the fully integrated sodium-and potassium-AMTEC/TE converters shows that a combined conversion efficiency in excess of 30% could be achieved by the plant. (B204)

  3. SUSEE: A Compact, Lightweight Space Nuclear Power System Using Present Water Reactor Technology

    SciTech Connect

    Maise, George; Powell, James; Paniagua, John

    2006-01-20

    The SUSEE space reactor system uses existing nuclear fuels and the standard steam cycle to generate electrical and thermal power for a wide range of in-space and surface applications, including manned bases, sub-surface mobile probes to explore thick ice deposits on Mars and the Jovian moons, and mobile rovers. SUSEE cycle efficiency, thermal to electric, ranges from {approx}20 to 24%, depending on operating parameters. Rejection of waste heat is by a lightweight condensing radiator that can be launched as a compact rolled-up package and deployed into flat panels when appropriate. The 50 centimeter diameter SUSEE reactor can provide power over the range of 10 kW(e) to 1 MW(e) for a period of 10 years. Higher power outputs are possible using slightly larger reactors. System specific weight (reactor, turbine, generator, piping, and radiator) is {approx}3 kg/kW(e). Two SUSEE reactor options are described, based on the existing Zr/O2 cermet and the UH3/ZrH2 TRIGA nuclear fuels.

  4. Solid-Core, Gas-Cooled Reactor for Space and Surface Power

    NASA Astrophysics Data System (ADS)

    King, Jeffrey C.; El-Genk, Mohamed S.

    2006-01-01

    The solid-core, gas-cooled, Submersion-Subcritical Safe Space (S∧4) reactor is developed for future space power applications and avoidance of single point failures. The Mo-14%Re reactor core is loaded with uranium nitride fuel in enclosed cavities, cooled by He-30%Xe, and sized to provide 550 kWth for seven years of equivalent full power operation. The beryllium oxide reflector disassembles upon impact on water or soil. In addition to decreasing the reactor and shadow shield mass, Spectral Shift Absorber (SSA) materials added to the reactor core ensure that it remains subcritical in the worst-case submersion accident. With a 0.1 mm thick boron carbide coating on the outside surface of the core block and 0.25 mm thick iridium sleeves around the fuel stacks, the reflector outer diameter is 43.5 cm and the combined reactor and shadow shield mass is 935.1 kg. With 12.5 atom% gadolinium-155 added to the fuel, 2.0 mm diameter gadolinium-155 sesquioxide intersititial pins, and a 0.1 mm thick gadolinium-155 sesquioxide coating, the S∧4 reactor has a slightly smaller reflector outer diameter of 43.0 cm, and a total reactor and shield mass of 901.7 kg. With 8.0 atom% europium-151 added to the fuel, 2.0 mm diameter europium-151 sesquioxide interstitial pins, and a 0.1 mm thick europium-151 sesquioxide coating, the reflector's outer diameter and the total reactor and shield mass are further reduced to 41.5 cm and 869.2 kg, respectively.

  5. Solid-Core, Gas-Cooled Reactor for Space and Surface Power

    SciTech Connect

    King, Jeffrey C.; El-Genk, Mohamed S.

    2006-01-20

    The solid-core, gas-cooled, Submersion-Subcritical Safe Space (S and 4) reactor is developed for future space power applications and avoidance of single point failures. The Mo-14%Re reactor core is loaded with uranium nitride fuel in enclosed cavities, cooled by He-30%Xe, and sized to provide 550 kWth for seven years of equivalent full power operation. The beryllium oxide reflector disassembles upon impact on water or soil. In addition to decreasing the reactor and shadow shield mass, Spectral Shift Absorber (SSA) materials added to the reactor core ensure that it remains subcritical in the worst-case submersion accident. With a 0.1 mm thick boron carbide coating on the outside surface of the core block and 0.25 mm thick iridium sleeves around the fuel stacks, the reflector outer diameter is 43.5 cm and the combined reactor and shadow shield mass is 935.1 kg. With 12.5 atom% gadolinium-155 added to the fuel, 2.0 mm diameter gadolinium-155 sesquioxide intersititial pins, and a 0.1 mm thick gadolinium-155 sesquioxide coating, the S and 4 reactor has a slightly smaller reflector outer diameter of 43.0 cm, and a total reactor and shield mass of 901.7 kg. With 8.0 atom% europium-151 added to the fuel, 2.0 mm diameter europium-151 sesquioxide interstitial pins, and a 0.1 mm thick europium-151 sesquioxide coating, the reflector's outer diameter and the total reactor and shield mass are further reduced to 41.5 cm and 869.2 kg, respect0011ive.

  6. Advanced Power Conversion Efficiency in Inventive Plasma for Hybrid Toroidal Reactor

    NASA Astrophysics Data System (ADS)

    Hançerlioğullari, Aybaba; Cini, Mesut; Güdal, Murat

    2013-08-01

    Apex hybrid reactor has a good potential to utilize uranium and thorium fuels in the future. This toroidal reactor is a type of system that facilitates the occurrence of the nuclear fusion and fission events together. The most important feature of hybrid reactor is that the first wall surrounding the plasma is liquid. The advantages of utilizing a liquid wall are high power density capacity good power transformation productivity, the magnitude of the reactor's operational duration, low failure percentage, short maintenance time and the inclusion of the system's simple technology and material. The analysis has been made using the MCNP Monte Carlo code and ENDF/B-V-VI nuclear data. Around the fusion chamber, molten salts Flibe (LI2BeF4), lead-lithium (PbLi), Li-Sn, thin-lityum (Li20Sn80) have used as cooling materials. APEX reactor has modeled in the torus form by adding nuclear materials of low significance in the specified percentages between 0 and 12 % to the molten salts. In this study, the neutronic performance of the APEX fusion reactor using various molten salts has been investigated. The nuclear parameters of Apex reactor has been searched for Flibe (LI2BeF4) and Li-Sn, for blanket layers. In case of usage of the Flibe (LI2BeF4), PbLi, and thin-lityum (Li20Sn80) salt solutions at APEX toroidal reactors, fissile material production per source neutron, tritium production speed, total fission rate, energy reproduction factor has been calculated, the results obtained for both salt solutions are compared.

  7. Space and Terrestrial Power System Integration Optimization Code BRMAPS for Gas Turbine Space Power Plants With Nuclear Reactor Heat Sources

    NASA Technical Reports Server (NTRS)

    Juhasz, Albert J.

    2007-01-01

    In view of the difficult times the US and global economies are experiencing today, funds for the development of advanced fission reactors nuclear power systems for space propulsion and planetary surface applications are currently not available. However, according to the Energy Policy Act of 2005 the U.S. needs to invest in developing fission reactor technology for ground based terrestrial power plants. Such plants would make a significant contribution toward drastic reduction of worldwide greenhouse gas emissions and associated global warming. To accomplish this goal the Next Generation Nuclear Plant Project (NGNP) has been established by DOE under the Generation IV Nuclear Systems Initiative. Idaho National Laboratory (INL) was designated as the lead in the development of VHTR (Very High Temperature Reactor) and HTGR (High Temperature Gas Reactor) technology to be integrated with MMW (multi-megawatt) helium gas turbine driven electric power AC generators. However, the advantages of transmitting power in high voltage DC form over large distances are also explored in the seminar lecture series. As an attractive alternate heat source the Liquid Fluoride Reactor (LFR), pioneered at ORNL (Oak Ridge National Laboratory) in the mid 1960's, would offer much higher energy yields than current nuclear plants by using an inherently safe energy conversion scheme based on the Thorium --> U233 fuel cycle and a fission process with a negative temperature coefficient of reactivity. The power plants are to be sized to meet electric power demand during peak periods and also for providing thermal energy for hydrogen (H2) production during "off peak" periods. This approach will both supply electric power by using environmentally clean nuclear heat which does not generate green house gases, and also provide a clean fuel H2 for the future, when, due to increased global demand and the decline in discovering new deposits, our supply of liquid fossil fuels will have been used up. This is

  8. Estimates of the financial consequences of nuclear-power-reactor accidents

    SciTech Connect

    Strip, D.R.

    1982-09-01

    This report develops preliminary techniques for estimating the financial consequences of potential nuclear power reactor accidents. Offsite cost estimates are based on CRAC2 calculations. Costs are assigned to health effects as well as property damage. Onsite costs are estimated for worker health effects, replacement power, and cleanup costs. Several classes of costs are not included, such as indirect costs, socio-economic costs, and health care costs. Present value discounting is explained and then used to calculate the life cycle cost of the risks of potential reactor accidents. Results of the financial consequence estimates for 156 reactor-site combinations are summarized, and detailed estimates are provided in an appendix. The results indicate that, in general, onsite costs dominate the consequences of potential accidents.

  9. Manned mars rover powered by a nuclear reactor; Radiation shield analysis

    SciTech Connect

    Morley, N.J.; El-Genk, M. . Dept. of Chemical and Nuclear Engineering)

    1992-08-01

    This paper discusses a key element in the conceptual design of a nuclear reactor power system for a manned Mars rover is the analysis, design, and integration of the radiation shield. A shield analysis is carried out to characterize the thickness and spacing of shield layers to provide the minimum mass configuration that meets a dose rate requirement of 300 mSv/yr. The analysis utilizes a two-dimensional transport code to model the reactor and to provide a source term that is subsequently used to calculate dose rates as a function of reactor power level and shield layer thickness. Results show that a multilayered tungsten and lithium hydride (LiH) shield would satisfy the dose rate limit of 300 mSv/yr (30 rem/yr) to the rover crew. The position of two tungsten and LiH layers is varied to minimize secondary gamma-ray production and to optimize shield mass.

  10. Insights from Investigations of In-Vessel Retention for High Powered Reactors

    SciTech Connect

    Joy L. Rempe

    2005-10-01

    In a three-year U.S. - Korean International Nuclear Energy Research Initiative (INERI), state-of-the-art analytical tools and key U.S. and Korean experimental facilities were used to explore two options, enhanced ERVC performance and the use of internal core catchers, that have the potential to increase the margin for in-vessel retention (IVR) in high power reactors (up to 1500 MWe). This increased margin has the potential to improve plant economics (owing to reduced regulatory requirements) and increase public acceptance (owing to reduced plant risk). Although this program focused upon the Korean Advanced Power Reactor -- 1400 MWe (APR 1400) design, recommentations were developed so that they can easily be applied to a wide range of existing and advanced reactor designs. This paper summarizes new data gained for evaluating the margin associated with various options investigated in this program. Insights from analyses completed with this data are also highlighted.

  11. Study of reactor Brayton power systems for nuclear electric spacecraft

    NASA Technical Reports Server (NTRS)

    1979-01-01

    The feasibility of using Brayton power systems for nuclear electric spacecraft was investigated. The primary performance parameters of systems mass and radiator area were determined for systems from 100 to 1000 kW sub e. Mathematical models of all system components were used to determine masses and volumes. Two completely independent systems provide propulsion power so that no single-point failure can jeopardize a mission. The waste heat radiators utilize armored heat pipes to limit meteorite puncture. The armor thickness was statistically determined to achieve the required probability of survival. A 400 kW sub e reference system received primary attention as required by the contract. The components of this system were defined and a conceptual layout was developed with encouraging results. An arrangement with redundant Brayton power systems having a 1500 K (2240 F) turbine inlet temperature was shown to be compatible with the dimensions of the space shuttle orbiter payload bay.

  12. Global radioxenon emission inventory based on nuclear power reactor reports.

    PubMed

    Kalinowski, Martin B; Tuma, Matthias P

    2009-01-01

    Atmospheric radioactivity is monitored for the verification of the Comprehensive Nuclear-Test-Ban Treaty, with xenon isotopes 131mXe, 133Xe, 133mXe and 135Xe serving as important indicators of nuclear explosions. The treaty-relevant interpretation of atmospheric concentrations of radioxenon is enhanced by quantifying radioxenon emissions released from civilian facilities. This paper presents the first global radioxenon emission inventory for nuclear power plants, based on North American and European emission reports for the years 1995-2005. Estimations were made for all power plant sites for which emission data were unavailable. According to this inventory, a total of 1.3PBq of radioxenon isotopes are released by nuclear power plants as continuous or pulsed emissions in a generic year.

  13. Power flattening on modified CANDLE small long life gas-cooled fast reactor

    NASA Astrophysics Data System (ADS)

    Monado, Fiber; Su'ud, Zaki; Waris, Abdul; Basar, Khairul; Ariani, Menik; Sekimoto, Hiroshi

    2014-09-01

    Gas-cooled Fast Reactor (GFR) is one of the candidates of next generation Nuclear Power Plants (NPPs) that expected to be operated commercially after 2030. In this research conceptual design study of long life 350 MWt GFR with natural uranium metallic fuel as fuel cycle input has been performed. Modified CANDLE burn-up strategy with first and second regions located near the last region (type B) has been applied. This reactor can be operated for 10 years without refuelling and fuel shuffling. Power peaking reduction is conducted by arranging the core radial direction into three regions with respectively uses fuel volume fraction 62.5%, 64% and 67.5%. The average power density in the modified core is about 82 Watt/cc and the power peaking factor decreased from 4.03 to 3.43.

  14. Effect of Void Distribution Parameter and Axial Power Profile on Boiling Water Reactor Bifurcation Characteristics

    SciTech Connect

    Bragt, D.D.B. van; Rizwan-uddin; Hagen, T.H.J.J. van der

    2000-02-15

    Bifurcation analyses of the impact of the void distribution parameter C{sub 0} and the axial power profile on the stability of boiling water reactors (BWRs) are reported. Bifurcation characteristics of heated channels (without nuclear feedback) appear to be very sensitive to the axial power profile. A turning point bifurcation was detected for a (symmetrically) peaked axial power profile. This kind of bifurcation does not occur for a uniformly heated channel.Both supercritical and subcritical Hopf bifurcations were encountered in a (nuclear-coupled) reactor system, depending on the strength of the void reactivity feedback. Subcritical bifurcations become less likely to occur as C{sub 0} is significantly larger than unity. In BWRs with a strong nuclear feedback, the oscillation amplitude of limit cycles caused by a supercritical bifurcation is very sensitive to both C{sub 0} and the axial power profile.

  15. Power flattening on modified CANDLE small long life gas-cooled fast reactor

    SciTech Connect

    Monado, Fiber; Su'ud, Zaki; Waris, Abdul; Basar, Khairul; Ariani, Menik; Sekimoto, Hiroshi

    2014-09-30

    Gas-cooled Fast Reactor (GFR) is one of the candidates of next generation Nuclear Power Plants (NPPs) that expected to be operated commercially after 2030. In this research conceptual design study of long life 350 MWt GFR with natural uranium metallic fuel as fuel cycle input has been performed. Modified CANDLE burn-up strategy with first and second regions located near the last region (type B) has been applied. This reactor can be operated for 10 years without refuelling and fuel shuffling. Power peaking reduction is conducted by arranging the core radial direction into three regions with respectively uses fuel volume fraction 62.5%, 64% and 67.5%. The average power density in the modified core is about 82 Watt/cc and the power peaking factor decreased from 4.03 to 3.43.

  16. Nuclear reactor power as applied to a space-based radar mission

    NASA Technical Reports Server (NTRS)

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

    1988-01-01

    The SP-100 Project was established to develop and demonstrate feasibility of a space reactor power system (SRPS) at power levels of 10's of kilowatts to a megawatt. To help determine systems requirements for the SRPS, a mission and spacecraft were examined which utilize this power system for a space-based radar to observe moving objects. Aspects of the mission and spacecraft bearing on the power system were the primary objectives of this study; performance of the radar itself was not within the scope. The study was carried out by the Systems Design Audit Team of the SP-100 Project.

  17. Requirements for a common nuclear propulsion and power reactor for human exploration missions to Mars

    SciTech Connect

    Cataldo, Robert L.; Borowski, Stanley K.

    1998-01-15

    Requirements for propulsion and power systems capable of achieving a safe, reliable, robust and affordable human Mars exploration mission have been identified. Nuclear systems have been identified that can meet the challenges of short trip times, reduced number of launch vehicles, potential for 'all propulsive' maneuvers, abundant in-space power and low mass, volume and deployed area, and energy rich surface power. Reduced total systems cost will also be mandatory to achieve affordable human exploration of Mars. Hence, it is desirable to design a space propulsion and surface power reactor with the greatest degree of commonality as possible with the goal of reducing total system costs.

  18. Gaseous-fuel nuclear reactor research for multimegawatt power in space

    NASA Technical Reports Server (NTRS)

    Thom, K.; Schneider, R. T.; Helmick, H. H.

    1977-01-01

    In the gaseous-fuel reactor concept, the fissile material is contained in a moderator-reflector cavity and exists in the form of a flowing gas or plasma separated from the cavity walls by means of fluid mechanical forces. Temperatures in excess of structural limitations are possible for low-specific-mass power and high-specific-impulse propulsion in space. Experiments have been conducted with a canister filled with enriched UF6 inserted into a beryllium-reflected cavity. A theoretically predicted critical mass of 6 kg was measured. The UF6 was also circulated through this cavity, demonstrating stable reactor operation with the fuel in motion. Because the flowing gaseous fuel can be continuously processed, the radioactive waste in this type of reactor can be kept small. Another potential of fissioning gases is the possibility of converting the kinetic energy of fission fragments directly into coherent electromagnetic radiation, the nuclear pumping of lasers. Numerous nuclear laser experiments indicate the possibility of transmitting power in space directly from fission energy. The estimated specific mass of a multimegawatt gaseous-fuel reactor power system is from 1 to 5 kg/kW while the companion laser-power receiver station would be much lower in specific mass.

  19. Nuclear reactor power as applied to a space-based radar mission

    NASA Technical Reports Server (NTRS)

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

    1988-01-01

    A space-based radar mission and spacecraft are examined to determine system requirements for a 300 kWe space nuclear reactor power system. The spacecraft configuration and its orbit, launch vehicle, and propulsion are described. Mission profiles are addressed, and storage in assembly orbit is considered. Dynamics and attitude control and the problems of nuclear and thermal radiation are examined.

  20. Introduction to Nuclear Fusion Power and the Design of Fusion Reactors. An Issue-Oriented Module.

    ERIC Educational Resources Information Center

    Fillo, J. A.

    This three-part module focuses on the principles of nuclear fusion and on the likely nature and components of a controlled-fusion power reactor. The physical conditions for a net energy release from fusion and two approaches (magnetic and inertial confinement) which are being developed to achieve this goal are described. Safety issues associated…

  1. Thermal-hydraulics and safety analysis of sectored compact reactor for lunar surface power

    SciTech Connect

    Schriener, T. M.; El-Genk, M. S.

    2012-07-01

    The liquid NaK-cooled, fast-neutron spectrum, Sectored Compact Reactor (SCoRe-N 5) concept has been developed at the Univ. of New Mexico for lunar surface power applications. It is loaded with highly enriched UN fuel pins in a triangular lattice, and nominally operates at exit and inlet coolant temperatures of 850 K and 900 K. This long-life reactor generates up to 1 MWth continuously for {>=} 20 years. To avoid a single point failure in reactor cooling, the core is divided into 6 sectors that are neutronically and thermally coupled, but hydraulically independent. This paper performs a 3-D the thermal-hydraulic analysis of SCoRe--N 5 at nominal operation temperatures and a power level of 1 MWth. In addition, the paper investigates the potential of continuing reactor operation at a lower power in the unlikely event that one sector in the core experiences a loss of coolant (LOC). Redesigning the core with a contiguous steel matrix enhances the cooling of the sector experiencing a LOC. Results show that with a core sector experiencing a LOC, SCORE-N 5 could continue operating safely at a reduced power of 166.6 kWth. (authors)

  2. A modular gas-cooled cermet reactor system for planetary base power

    NASA Astrophysics Data System (ADS)

    Jahshan, S. N.; Borkowski, J. A.

    1992-10-01

    Fission nuclear power is foreseen as the source for electricity in planetary colonization and exploration. This report discusses a six module, gas-cooled, cermet-fueled reactor which is proposed to meet the design objectives. The highly enriched core is compact and can operate at high temperatures for long periods of time. The helium coolant powers six modular Brayton cycles that compare favorably with the SP-100-based Brayton cycle.

  3. A modular gas-cooled cermet reactor system for planetary base power

    SciTech Connect

    Jahshan, S.N.; Borkowski, J.A. )

    1993-01-15

    Fission nuclear power is foreseen as the source for electricity in planetary colonization and exploration. A six module gas-cooled, cermet-fueled reactor is proposed that can meet the design objectives. The highly enriched core is compact and can operate at high temperature for a long life. The helium coolant powers six modular Brayton cycles that compare favorably with the SP-100-based Brayton cycle.

  4. Dynamic neutronic and stability analysis of a burst mode, single cavity gas core reactor Brayton cycle space power system

    NASA Astrophysics Data System (ADS)

    Dugan, Edward T.; Kutikkad, Kiratadas

    The conceptual, burst-mode gaseous-core reactor (GCR) space nuclear power system presently subjected to reactor-dynamics and system stability studies operates on a closed Brayton cycle, via disk MHD generator for energy conversion. While the gaseous fuel density power coefficient of reactivity is found to be capable of rapidly stabilizing the GCR system, the power of this feedback renders standard external reactivity insertions inadequate for significant power-level changes during normal operation.

  5. REACTOR

    DOEpatents

    Christy, R.F.

    1961-07-25

    A means is described for co-relating the essential physical requirements of a fission chain reaction in order that practical, compact, and easily controllable reactors can be built. These objects are obtained by employing a composition of fissionsble isotope and moderator in fluid form in which the amount of fissionsble isotcpe present governs the reaction. The size of the reactor is no longer a critical factor, the new criterion being the concentration of the fissionable isotope.

  6. REACTOR

    DOEpatents

    Szilard, L.

    1963-09-10

    A breeder reactor is described, including a mass of fissionable material that is less than critical with respect to unmoderated neutrons and greater than critical with respect to neutrons of average energies substantially greater than thermal, a coolant selected from sodium or sodium--potassium alloys, a control liquid selected from lead or lead--bismuth alloys, and means for varying the quantity of control liquid in the reactor. (AEC)

  7. EPRI`s nuclear power plant instrumentation and control program and its applicability to advanced reactors

    SciTech Connect

    Naser, J.; Torok, R.; Wilkinson, D.

    1997-12-01

    I&C systems in nuclear power plants need to be upgraded over the lifetime of the plant in a reliable and cost-effective manner to replace obsolete equipment, to reduce O&M costs, to improve plant performance, and to maintain safety. This applies to operating plants now and will apply to advanced reactors in the future. The major drivers for the replacement of the safety, control, and information systems in nuclear power plants are the obsolescence of the existing hardware and the need for more cost-effective power production. Competition between power producers is dictating more cost-effective power production. The increasing O&M costs to maintain systems experiencing obsolescence problems is counter to the needs for more cost-effective power production and improved competitiveness. This need for increased productivity applies to government facilities as well as commercial plants. Increasing competition will continue to be a major factor in the operation of both operating plants and advanced reactors. It will continue to dictate the need for improved productivity and cost-effectiveness. EPRI and its member nuclear utilities are working together on an industry wide I&C Program to address I&C issues and to develop cost-effective solutions. A majority of the I&C products and demonstrations being developed under this program will benefit advanced reactors in both the design and operational phases of their life cycle as well as it will benefit existing plants. 20 refs.

  8. Nuclear reactors using fine-particulate fuel for primary power in space

    SciTech Connect

    Botts, T.E.; Powell, J.R.; Usher, J.L.; Horn, F.L.

    1982-01-01

    Large future power requirements in space, include power beaming to earth, airplanes, and solar-powered satellites in eclipse; industrial processing; and space colonies. The Rotating Bed Nuclear Reactor (RBR) and Fixed Bed Reactor (FBR) are multi-megawatt power systems which are light, compact and suited to operation in space. Both are cavity reactors, with an annular fuel region (e.g., a bed of 500 ..mu.. HTGR fuel particulates made of UC with ceramic coating) surrounded by a reflector that moderates fast neutrons from the /sup 235/U fuel. A porous metal drum holds the fuel. In the RBR, rotation of the drum allows the particulate fuel bed to fluidize as cooling gas passes through. In the FBR, an inner porous carbon drum holds the packed fuel bed, which is not fluidized. The RBR and FBR have many important features for space nuclear power: very high power density (up to thousands of MW(th)/m/sup 3/ of fuel); very small size and weight, excellent thermal shock and fatigue resistance; short start/stop times (sec); high gas outlet temperatures (to 3000/sup 0/K), good neutron economy, low critical mass; and simple/reliable construction.

  9. Space Molten Salt Reactor Concept for Nuclear Electric Propulsion and Surface Power

    NASA Astrophysics Data System (ADS)

    Eades, M.; Flanders, J.; McMurray, N.; Denning, R.; Sun, X.; Windl, W.; Blue, T.

    Students at The Ohio State University working under the NASA Steckler Grant sought to investigate how molten salt reactors with fissile material dissolved in a liquid fuel medium can be applied to space applications. Molten salt reactors of this kind, built for non-space applications, have demonstrated high power densities, high temperature operation without pressurization, high fuel burn up and other characteristics that are ideal for space fission systems. However, little research has been published on the application of molten salt reactor technology to space fission systems. This paper presents a conceptual design of the Space Molten Salt Reactor (SMSR), which utilizes molten salt reactor technology for Nuclear Electric Propulsion (NEP) and surface power at the 100 kWe to 15 MWe level. Central to the SMSR design is a liquid mixture of LiF, BeF2 and highly enriched U235F4 that acts as both fuel and core coolant. In brief, some of the positive characteristics of the SMSR are compact size, simplified core design, high fuel burn up percentages, proliferation resistant features, passive safety mechanisms, a considerable body of previous research, and the possibility for flexible mission architecture.

  10. Feasibility Study of Supercritical Light Water Cooled Fast Reactors for Actinide Burning and Electric Power Production

    SciTech Connect

    Mac Donald, Philip Elsworth; Buongiorno, Jacopo; Davis, Cliff Bybee; Weaver, Kevan Dean

    2002-01-01

    The use of supercritical temperature and pressure light water as the coolant in a direct-cycle nuclear reactor offers potential for considerable plant simplification and consequent capital and O&M cost reduction compared with current light water reactor (LWR) designs. Also, given the thermodynamic conditions of the coolant at the core outlet (i.e. temperature and pressure beyond the water critical point), very high thermal efficiencies of the power conversion cycle are possible (i.e. up to 46%). Because no change of phase occurs in the core, the need for steam separators and dryers as well as for BWR-type recirculation pumps is eliminated, which, for a given reactor power, results in a substantially shorter reactor vessel than the current BWRs. Furthermore, in a direct cycle the steam generators are not needed. If a tight fuel rod lattice is adopted, it is possible to significantly reduce the neutron moderation and attain fast neutron energy spectrum conditions. In this project a supercritical water reactor concept with a simple, blanket-free, pancake-shaped core will be developed. This type of core can make use of either fertile or fertile-free fuel and retain the hard spectrum to effectively burn plutonium and minor actinides from LWR spent fuel while efficiently generating electricity.

  11. Optimum Reflector Configurations for Minimizing Fission Power Peaking in a Lithium-Cooled, Liquid-Metal Reactor with Sliding Reflectors

    SciTech Connect

    Fensin, Michael L.; Poston, David I.

    2005-02-06

    Many design constraints limit the development of a space fission power system optimized for fuel performance, system reliability, and mission cost. These design constraints include fuel mass provisions to meet cycle-length requirements, fuel centerline and clad temperatures, and clad creep from fission gas generation. Decreasing the fission power peaking of the reactor system enhances all of the mentioned parameters. This design study identifies the cause, determines the reflector configurations for reactor criticality, and generates worth curves for minimized fission-power-peaking configuration in a lithium-cooled liquid-metal reactor that uses sliding reflectors. Because of the characteristics of the core axial power distribution and axial power distortions inherent to the sliding reflector design, minimizing the power peaking of the reactor involves placing the reflectors in a position that least distorts the axial power distribution. The views expressed in this document are those of the author and do not necessarily reflect agreement by the Government.

  12. A neutron tomography facility at a low power research reactor

    NASA Astrophysics Data System (ADS)

    Koerner, S.; Schillinger, B.; Vontobel, P.; Rauch, H.

    2001-09-01

    Neutron radiography (NR) provides a very efficient tool in the field of non-destructive testing as well as for many applications in fundamental research. A neutron beam penetrating a specimen is attenuated by the sample material and detected by a two-dimensional (2D) imaging device. The image contains information about materials and structure inside the sample because neutrons are attenuated according to the basic law of radiation attenuation. Contrary to X-rays, neutrons can be attenuated by some light materials, as for example, hydrogen and boron, but penetrate many heavy materials. Therefore, NR can yield important information not obtainable by more traditional methods. Nevertheless, there are many aspects of structure, both quantitative and qualitative, that are not accessible from 2D transmission images. Hence, there is an interest in three-dimensional neutron imaging. At the 250 kW TRIGA Mark II reactor of the Atominstitut in Austria a neutron tomography facility has been installed. The neutron flux at this beam position is 1.3×10 5 neutrons/cm 2 s and the beam diameter is 8 cm. For a 3D tomographic reconstruction of the sample interior, transmission images of the object taken from different view angles are required. Therefore, a rotary table driven by a step motor connected to a computerized motion control system has been installed at the sample position. In parallel a suitable electronic imaging device based on a neutron sensitive scintillator screen and a CCD-camera has been designed. It can be controlled by a computer in order to synchronize the software of the detector and of the rotary table with the aim of an automation of measurements. Reasonable exposure times can get as low as 20 s per image. This means that a complete tomography of a sample can be performed within one working day. Calculation of the 3D voxel array is made by using the filtered backprojection algorithm.

  13. Utilization of the Philippine Research Reactor as a training facility for nuclear power plant operators

    SciTech Connect

    Palabrica, R.J.

    1981-01-01

    The Philippines has a 1-MW swimming-pool reactor facility operated by the Philippine Atomic Energy Commission (PAEC). The reactor is light-water moderated and cooled, graphite reflected, and fueled with 90% enriched uranium. Since it became critical in 1963 it has been utilized for research, radioisotope production, and training. It was used initially in the training of PAEC personnel and other research institutions and universities. During the last few years, however, it has played a key role in training personnel for the Philippine Nuclear Power Project (PNPP).

  14. Space power reactor in-core thermionic multicell evolutionary (S-prime) design

    NASA Astrophysics Data System (ADS)

    Determan, William R.; Van Hagan, Tom H.

    1993-01-01

    A 5- to 40-kWe moderated in-core thermionic space nuclear power system (TI-SNPS) concept was developed to address the TI-SNPS program requirements. The 40-kWe baseline design uses multicell Thermionic Fuel Elements (TFEs) in a zirconium hydride moderated reactor to achieve a specific mass of 18.2 We/kg and a net end-of-mission (EOM) efficiency of 8.2%. The reactor is cooled with a single NaK-78 pumped loop, which rejects the heat through a 24 m2 heat pipe space radiator.

  15. Space power reactor in-core thermionic multicell evolutionary (S-prime) design

    SciTech Connect

    Determan, W.R. ); Van Hagan, T.H. )

    1993-01-20

    A 5- to 40-kWe moderated in-core thermionic space nuclear power system (TI-SNPS) concept was developed to address the TI-SNPS program requirements. The 40-kWe baseline design uses multicell Thermionic Fuel Elements (TFEs) in a zirconium hydride moderated reactor to achieve a specific mass of 18.2 We/kg and a net end-of-mission (EOM) efficiency of 8.2%. The reactor is cooled with a single NaK-78 pumped loop, which rejects the heat through a 24 m[sup 2] heat pipe space radiator.

  16. Nuclear design of the burst power ultrahigh temperature UF4 vapor core reactor system

    NASA Astrophysics Data System (ADS)

    Kahook, Samer D.; Dugan, Edward T.

    1991-01-01

    Static and dynamic neutronic analyses are being performed, as part of an integrated series of studies, on an innovative burst power UF4 Ultrahigh Temperature Vapor Core Reactor (UTVR)/Disk Magnetohydrodynamic (MHD) generator for space nuclear power applications. This novel reactor concept operates on a direct, closed Rankine cycle in the burst power mode (hundreds of MWe for thousands of seconds). The fuel/working fluid is a mixture of UF4 and metal fluoride. Preliminary calculations indicate high overall system efficiencies (≊20%), small radiator size (≊5 m2/MWe), and high specific power (≊5 kWe/kg). Neutronic analysis has revealed a number of attractive features for this novel reactor concept. These include some unique and very effective inherent negative reactivity control mechanisms such as the vapor-fuel density power coefficient of reactivity, the direct neutronic coupling among the multiple fissioning core regions (the central vapor core and the surrounding boiler columns), and the mass flow coupling feedback between the fissioning cores.

  17. Experimental power density distribution benchmark in the TRIGA Mark II reactor

    SciTech Connect

    Snoj, L.; Stancar, Z.; Radulovic, V.; Podvratnik, M.; Zerovnik, G.; Trkov, A.; Barbot, L.; Domergue, C.; Destouches, C.

    2012-07-01

    In order to improve the power calibration process and to benchmark the existing computational model of the TRIGA Mark II reactor at the Josef Stefan Inst. (JSI), a bilateral project was started as part of the agreement between the French Commissariat a l'energie atomique et aux energies alternatives (CEA) and the Ministry of higher education, science and technology of Slovenia. One of the objectives of the project was to analyze and improve the power calibration process of the JSI TRIGA reactor (procedural improvement and uncertainty reduction) by using absolutely calibrated CEA fission chambers (FCs). This is one of the few available power density distribution benchmarks for testing not only the fission rate distribution but also the absolute values of the fission rates. Our preliminary calculations indicate that the total experimental uncertainty of the measured reaction rate is sufficiently low that the experiments could be considered as benchmark experiments. (authors)

  18. 14C content in vegetation in the vicinities of Brazilian nuclear power reactors.

    PubMed

    Dias, Cíntia Melazo; Santos, Roberto Ventura; Stenström, Kristina; Nícoli, Iêda Gomes; Skog, Göran; da Silveira Corrêa, Rosangela

    2008-07-01

    (14)C specific activities were measured in grass samples collected around Brazilian nuclear power reactors. The specific activity values varied between 227 and 299 Bq/kg C. Except for two samples which showed (14)C specific activities 22% above background values, half of the samples showed background specific activities, and the other half had a (14)C excess of 1-18%. The highest specific activities were found close to the nuclear power plants and along the main wind directions (NE and NNE). The activity values were found to decrease with increasing distance from the reactors. The unexpectedly high (14)C excess values found in two samples were related to the local topography, which favors (14)C accumulation and limits the dispersion of the plume. The results indicate a clear (14)C anthropogenic signal within 5 km around the nuclear power plants which is most prominent along northeastwards, the prevailing wind direction.

  19. Data bases for rapid response to power reactor problems

    SciTech Connect

    Maskewitz, B.F.

    1980-01-01

    The urgency of the TMI-2 incident demanded prompt answers to an imperious situation. In responding to these challenging circumstances, both government and industry recognized deficiencies in both availability of essential retrievable data and calculational capabilities designed to respond immediately to actual abnormal events. Each responded by initiating new programs to provide a remedy for the deficiencies and to generally improve all safety measures in the nuclear power industry. Many data bases and information centers offer generic data and other technology resources which are generally useful in support of nuclear safety programs. A few centers can offer rapid access to calculational methods and associated data and more will make an effort to do so. As a beneficial spin-off from the lessons learned from TMI-2, more technical effort and financial resources will be devoted to the prevention of accidents, and to improvement of safety measures in the immediate future and for long term R and D programs by both government and the nuclear power industry.

  20. Mars, the Moon, and the Ends of the Earth: Autonomy for Small Reactor Power Systems

    SciTech Connect

    Wood, Richard Thomas

    2008-01-01

    In recent years, the National Aeronautics and Space Administration (NASA) has been considering deep space missions that utilize a small-reactor power system (SRPS) to provide energy for propulsion and spacecraft power. Additionally, application of SRPS modules as a planetary power source is being investigated to enable a continuous human presence for nonpolar lunar sites and on Mars. A SRPS can supply high-sustained power for space and surface applications that is both reliable and mass efficient. The use of small nuclear reactors for deep space or planetary missions presents some unique challenges regarding the operations and control of the power system. Current-generation terrestrial nuclear reactors employ varying degrees of human control and decision-making for operations and benefit from periodic human interaction for maintenance. In contrast, the control system of a SRPS employed for deep space missions must be able to accommodate unattended operations due to communications delays and periods of planetary occlusion while adapting to evolving or degraded conditions with no opportunity for repair or refurbishment. While surface power systems for planetary outposts face less extreme delays and periods of isolation and may benefit from limited maintenance capabilities, considerations such as human safety, resource limitations and usage priorities, and economics favor minimizing direct, continuous human interaction with the SRPS for online, dedicated power system management. Thus, a SRPS control system for space or planetary missions must provide capabilities for operational autonomy. For terrestrial reactors, large-scale power plants remain the preferred near-term option for nuclear power generation. However, the desire to reduce reliance on carbon-emitting power sources in developing countries may lead to increased consideration of SRPS modules for local power generation in remote regions that are characterized by emerging, less established infrastructures

  1. A gas-cooled cermet reactor system for planetary base power

    SciTech Connect

    Jahshan, S.N.; Borkowski, J.A.

    1992-01-01

    Fission nuclear power is foreseen as the source for electricity in colonization exploration. A gas-cooled, cermet-fueled reactor is proposed that can meet many of the design objectives. The highly enriched core is compact and can operate at high temperature for a long life. The helium coolant powers a Brayton cycle that compares well with the SP-100-based Brayton cycle. The power cycle can be upgraded further under certain siting-related conditions by the addition of a low temperature Rankine cycle.

  2. A gas-cooled cermet reactor system for planetary base power

    SciTech Connect

    Jahshan, S.N.; Borkowski, J.A.

    1992-08-01

    Fission nuclear power is foreseen as the source for electricity in colonization exploration. A gas-cooled, cermet-fueled reactor is proposed that can meet many of the design objectives. The highly enriched core is compact and can operate at high temperature for a long life. The helium coolant powers a Brayton cycle that compares well with the SP-100-based Brayton cycle. The power cycle can be upgraded further under certain siting-related conditions by the addition of a low temperature Rankine cycle.

  3. Applicability of 100kWe-class of space reactor power systems to NASA manned space station missions

    NASA Technical Reports Server (NTRS)

    Silverman, S. W.; Willenberg, H. J.; Robertson, C.

    1985-01-01

    An assessment is made of a manned space station operating with sufficiently high power demands to require a multihundred kilowatt range electrical power system. The nuclear reactor is a competitor for supplying this power level. Load levels were selected at 150kWe and 300kWe. Interactions among the reactor electrical power system, the manned space station, the space transportation system, and the mission were evaluated. The reactor shield and the conversion equipment were assumed to be in different positions with respect to the station; on board, tethered, and on a free flyer platform. Mission analyses showed that the free flyer concept resulted in unacceptable costs and technical problems. The tethered reactor providing power to an electrolyzer for regenerative fuel cells on the space station, results in a minimum weight shield and can be designed to release the reactor power section so that it moves to a high altitude orbit where the decay period is at least 300 years. Placing the reactor on the station, on a structural boom is an attractive design, but heavier than the long tethered reactor design because of the shield weight for manned activity near the reactor.

  4. Mini-cavity plasma core reactors for dual-mode space nuclear power/propulsion systems. M.S. Thesis

    NASA Technical Reports Server (NTRS)

    Chow, S.

    1976-01-01

    A mini-cavity plasma core reactor is investigated for potential use in a dual-mode space power and propulsion system. In the propulsive mode, hydrogen propellant is injected radially inward through the reactor solid regions and into the cavity. The propellant is heated by both solid driver fuel elements surrounding the cavity and uranium plasma before it is exhausted out the nozzle. The propellant only removes a fraction of the driver power, the remainder is transferred by a coolant fluid to a power conversion system, which incorporates a radiator for heat rejection. Neutronic feasibility of dual mode operation and smaller reactor sizes than those previously investigated are shown to be possible. A heat transfer analysis of one such reactor shows that the dual-mode concept is applicable when power generation mode thermal power levels are within the same order of magnitude as direct thrust mode thermal power levels.

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

    SciTech Connect

    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.

  6. The U.S.-Russian joint studies on using power reactors to disposition surplus weapon plutonium as spent fuel

    SciTech Connect

    Chebeskov, A.; Kalashnikov, A.; Bevard, B.; Moses, D.; Pavlovichev, A.

    1997-09-01

    In 1996, the US and the Russian Federation completed an initial joint study of the candidate options for the disposition of surplus weapons plutonium in both countries. The options included long term storage, immobilization of the plutonium in glass or ceramic for geologic disposal, and the conversion of weapons plutonium to spent fuel in power reactors. For the latter option, the US is only considering the use of existing light water reactors (LWRs) with no new reactor construction for plutonium disposition, or the use of Canadian deuterium uranium (CANDU) heavy water reactors. While Russia advocates building new reactors, the cost is high, and the continuing joint study of the Russian options is considering only the use of existing VVER-1000 LWRs in Russia and possibly Ukraine, the existing BN-60O fast neutron reactor at the Beloyarsk Nuclear Power Plant in Russia, or the use of the Canadian CANDU reactors. Six of the seven existing VVER-1000 reactors in Russia and the eleven VVER-1000 reactors in Ukraine are all of recent vintage and can be converted to use partial MOX cores. These existing VVER-1000 reactors are capable of converting almost 300 kg of surplus weapons plutonium to spent fuel each year with minimum nuclear power plant modifications. Higher core loads may be achievable in future years.

  7. Power beaming to space using a nuclear reactor-pumped laser

    SciTech Connect

    Lipinski, R.J.; Monroe, D.K.; Pickard, P.S.

    1993-10-01

    The present political and environmental climate may slow the inevitable direct utilization of nuclear power in space. In the meantime, there is another approach for using nuclear energy for space power. That approach is to let nuclear energy generate a laser beam in a ground-based nuclear reactor-pumped laser (RPL), and then beam the optical energy into space. Potential space applications for a ground-based RPL include (1) illuminating geosynchronous communication satellites in the earth`s shadow to extend their lives, (2) beaming power to orbital transfer vehicles, (3) providing power (from earth) to a lunar base during the long lunar night, and (4) removing space debris. FALCON is a high-power, steady-state, nuclear reactor-pumped laser (RPL) concept that is being developed by the Department of Energy with Sandia National Laboratories as the lead laboratory. The FALCON program has experimentally demonstrated reactor-pumped lasing in various mixtures of xenon, argon, neon, and helium at wavelengths of 0.585, 0.703, 0.725, 1.271, 1.733, 1.792, 2.032, 2.63, 2.65, and 3.37 {mu}m with intrinsic efficiency as high as 2.5%. Frequency-doubling the 1.733{minus}{mu}m line would yield a good match for photovoltaic arrays at 0.867 {mu}m. Preliminary designs of an RPL suitable for power beaming have been completed. The MWclass laser is fairly simple in construction, self-powered, closed-cycle (no exhaust gases), and modular. This paper describes the FALCON program accomplishments and power-beaming applications.

  8. ANALYSIS OF A HIGH TEMPERATURE GAS-COOLED REACTOR POWERED HIGH TEMPERATURE ELECTROLYSIS HYDROGEN PLANT

    SciTech Connect

    M. G. McKellar; E. A. Harvego; A. M. Gandrik

    2010-11-01

    An updated reference design for a commercial-scale high-temperature electrolysis (HTE) plant for hydrogen production has been developed. The HTE plant is powered by a high-temperature gas-cooled reactor (HTGR) whose configuration and operating conditions are based on the latest design parameters planned for the Next Generation Nuclear Plant (NGNP). The current HTGR reference design specifies a reactor power of 600 MWt, with a primary system pressure of 7.0 MPa, and reactor inlet and outlet fluid temperatures of 322°C and 750°C, respectively. The reactor heat is used to produce heat and electric power to the HTE plant. A Rankine steam cycle with a power conversion efficiency of 44.4% was used to provide the electric power. The electrolysis unit used to produce hydrogen includes 1.1 million cells with a per-cell active area of 225 cm2. The reference hydrogen production plant operates at a system pressure of 5.0 MPa, and utilizes a steam-sweep system to remove the excess oxygen that is evolved on the anode (oxygen) side of the electrolyzer. The overall system thermal-to-hydrogen production efficiency (based on the higher heating value of the produced hydrogen) is 42.8% at a hydrogen production rate of 1.85 kg/s (66 million SCFD) and an oxygen production rate of 14.6 kg/s (33 million SCFD). An economic analysis of this plant was performed with realistic financial and cost estimating The results of the economic analysis demonstrated that the HTE hydrogen production plant driven by a high-temperature helium-cooled nuclear power plant can deliver hydrogen at a competitive cost. A cost of $3.03/kg of hydrogen was calculated assuming an internal rate of return of 10% and a debt to equity ratio of 80%/20% for a reactor cost of $2000/kWt and $2.41/kg of hydrogen for a reactor cost of $1400/kWt.

  9. A comparative risk assessment for the Russian V213 power reactor

    SciTech Connect

    Marshall, T.D.; Hockenbury, R.W.; Honey, J.A.; Cadwallader, L.C.

    1996-04-01

    Probabilistic risk assessment methodology is applied to generate an evaluation of the relative likelihood of safe recovery following selected pressurized water reactor (PWR) design basis accidents for a Russian V213 nuclear power reactor. US-designed PWRs similar to the V213 are used for reference and comparison. This V213 risk assessment is based on comparison analyses of the following aspects: accident progression event tree success paths for typical PWR accident initiating events, safety aspects in reactor design, and perceived performance of reactor safety systems. The four initiating events considered here are: loss of offsite power with station blackout, large-break loss-of-coolant accident (LOCA), medium-break LOCA, and small-break LOCA. The success probabilities for the V213 reaching a non-core-damage state after the onset of the selected initiating events are calculated for two scenarios: (a) using actual component reliability data from US PWRs and (b) assuming common component reliability data. US PWR component reliability data are used based of the unavailability of such data for the V213 at the time of the analyses. While the use of US PWR data in this risk assessment of the V213 does strongly infer V213 comparability to US plants, the risk assessment using common component reliability does not have such a stringent limitation and is thus a separate scoping assessment of the V213 engineered safety systems. The results of the analyses suggest that the V213 has certain design features that significantly improve the reactor`s safety margin for the selected initiating events and that the V213 design has a relative risk of core damage for selected initiating events that is at least comparable to US PWRs. It is important to realize that these analyses are of a scoping nature and may be significantly influenced by important risk factors such as V213 operator training, quality control, and maintenance procedures.

  10. Advanced Fusion Reactors for Space Propulsion and Power Systems

    SciTech Connect

    Chapman, John J.

    2011-06-15

    In recent years the methodology proposed for conversion of light elements into energy via fusion has made steady progress. Scientific studies and engineering efforts in advanced fusion systems designs have introduced some new concepts with unique aspects including consideration of Aneutronic fuels. The plant parameters for harnessing aneutronic fusion appear more exigent than those required for the conventional fusion fuel cycle. However aneutronic fusion propulsion plants for Space deployment will ultimately offer the possibility of enhanced performance from nuclear gain as compared to existing ionic engines as well as providing a clean solution to Planetary Protection considerations and requirements. Proton triggered 11Boron fuel (p- 11B) will produce abundant ion kinetic energy for In-Space vectored thrust. Thus energetic alpha particles' exhaust momentum can be used directly to produce high Isp thrust and also offer possibility of power conversion into electricity. p-11B is an advanced fusion plant fuel with well understood reaction kinematics but will require some new conceptual thinking as to the most effective implementation.

  11. Advanced Fusion Reactors for Space Propulsion and Power Systems

    NASA Technical Reports Server (NTRS)

    Chapman, John J.

    2011-01-01

    In recent years the methodology proposed for conversion of light elements into energy via fusion has made steady progress. Scientific studies and engineering efforts in advanced fusion systems designs have introduced some new concepts with unique aspects including consideration of Aneutronic fuels. The plant parameters for harnessing aneutronic fusion appear more exigent than those required for the conventional fusion fuel cycle. However aneutronic fusion propulsion plants for Space deployment will ultimately offer the possibility of enhanced performance from nuclear gain as compared to existing ionic engines as well as providing a clean solution to Planetary Protection considerations and requirements. Proton triggered 11Boron fuel (p- 11B) will produce abundant ion kinetic energy for In-Space vectored thrust. Thus energetic alpha particles "exhaust" momentum can be used directly to produce high ISP thrust and also offer possibility of power conversion into electricity. p- 11B is an advanced fusion plant fuel with well understood reaction kinematics but will require some new conceptual thinking as to the most effective implementation.

  12. TFTR (Tokamak Fusion Test Reactor) neutral beam injected power measurement

    SciTech Connect

    Kamperschroer, J.H.; Grisham, L.R.; Dudek, L.E.; Gammel, G.M.; Johnson, G.A.; Kugel, H.W.; Lagin, L.; O'Connor, T.E.; Shah, P.A.; Sichta, P.

    1989-05-01

    Energy flow within TFTR neutral beamlines is measured with a waterfall calorimetry system capable of simultaneously measuring the energy deposited within four heating beamlines (three ion sources each), or of measuring the energy deposited in a separate neutral beam test stand. Of the energy extracted from the ion source in the well instrumented test stand, 99.5 +- 3.5% can be accounted for. When the ion deflection magnet is energized, however, 6.5% of the extracted energy is lost. This loss is attributed to a spray of devious particles onto unmonitored surfaces. A 30% discrepancy is also observed between energy measurements on the internal beamline calorimeter and energy measurements on a calorimeter located in the test stand target chamber. Particle reflection from the flat plate calorimeter in the target chamber, which the incident beam strikes at a near-grazing angle of 12/degree/, is the primary loss of this energy. A slight improvement in energy accountability is observed as the beam pulse length is increased. This improvement is attributed to systematic error in the sensitivity of the energy measurement to small fluctuations on the supply water temperature. An overall accuracy of 15% is estimated for the total power injected into TFTR. Contributions to this error are uncertainties in the beam neutralization efficiency, reionization and beam scrape-off in the drift duct, and fluctuations in the temperature of the supply water. 28 refs., 9 figs., 1 tab.

  13. Thermal analysis of heat and power plant with high temperature reactor and intermediate steam cycle

    NASA Astrophysics Data System (ADS)

    Fic, Adam; Składzień, Jan; Gabriel, Michał

    2015-03-01

    Thermal analysis of a heat and power plant with a high temperature gas cooled nuclear reactor is presented. The main aim of the considered system is to supply a technological process with the heat at suitably high temperature level. The considered unit is also used to produce electricity. The high temperature helium cooled nuclear reactor is the primary heat source in the system, which consists of: the reactor cooling cycle, the steam cycle and the gas heat pump cycle. Helium used as a carrier in the first cycle (classic Brayton cycle), which includes the reactor, delivers heat in a steam generator to produce superheated steam with required parameters of the intermediate cycle. The intermediate cycle is provided to transport energy from the reactor installation to the process installation requiring a high temperature heat. The distance between reactor and the process installation is assumed short and negligable, or alternatively equal to 1 km in the analysis. The system is also equipped with a high temperature argon heat pump to obtain the temperature level of a heat carrier required by a high temperature process. Thus, the steam of the intermediate cycle supplies a lower heat exchanger of the heat pump, a process heat exchanger at the medium temperature level and a classical steam turbine system (Rankine cycle). The main purpose of the research was to evaluate the effectiveness of the system considered and to assess whether such a three cycle cogeneration system is reasonable. Multivariant calculations have been carried out employing the developed mathematical model. The results have been presented in a form of the energy efficiency and exergy efficiency of the system as a function of the temperature drop in the high temperature process heat exchanger and the reactor pressure.

  14. Apparatus and method for closed-loop control of reactor power in minimum time

    DOEpatents

    Bernard, Jr., John A.

    1988-11-01

    Closed-loop control law for altering the power level of nuclear reactors in a safe manner and without overshoot and in minimum time. Apparatus is provided for moving a fast-acting control element such as a control rod or a control drum for altering the nuclear reactor power level. A computer computes at short time intervals either the function: .rho.=(.beta.-.rho.).omega.-.lambda..sub.e '.rho.-.SIGMA..beta..sub.i (.lambda..sub.i -.lambda..sub.e ')+l* .omega.+l* [.omega..sup.2 +.lambda..sub.e '.omega.] or the function: .rho.=(.beta.-.rho.).omega.-.lambda..sub.e .rho.-(.lambda..sub.e /.lambda..sub.e)(.beta.-.rho.)+l* .omega.+l* [.omega..sup.2 +.lambda..sub.e .omega.-(.lambda..sub.e /.lambda..sub.e).omega.] These functions each specify the rate of change of reactivity that is necessary to achieve a specified rate of change of reactor power. The direction and speed of motion of the control element is altered so as to provide the rate of reactivity change calculated using either or both of these functions thereby resulting in the attainment of a new power level without overshoot and in minimum time. These functions are computed at intervals of approximately 0.01-1.0 seconds depending on the specific application.

  15. On the fusion triple product and fusion power gain of tokamak pilot plants and reactors

    NASA Astrophysics Data System (ADS)

    Costley, A. E.

    2016-06-01

    The energy confinement time of tokamak plasmas scales positively with plasma size and so it is generally expected that the fusion triple product, nTτ E, will also increase with size, and this has been part of the motivation for building devices of increasing size including ITER. Here n, T, and τ E are the ion density, ion temperature and energy confinement time respectively. However, tokamak plasmas are subject to operational limits and two important limits are a density limit and a beta limit. We show that when these limits are taken into account, nTτ E becomes almost independent of size; rather it depends mainly on the fusion power, P fus. In consequence, the fusion power gain, Q fus, a parameter closely linked to nTτ E is also independent of size. Hence, P fus and Q fus, two parameters of critical importance in reactor design, are actually tightly coupled. Further, we find that nTτ E is inversely dependent on the normalised beta, β N; an unexpected result that tends to favour lower power reactors. Our findings imply that the minimum power to achieve fusion reactor conditions is driven mainly by physics considerations, especially energy confinement, while the minimum device size is driven by technology and engineering considerations. Through dedicated R&D and parallel developments in other fields, the technology and engineering aspects are evolving in a direction to make smaller devices feasible.

  16. An evaluation of the ecological consequences of partial-power operation of the K Reactor, SRS

    SciTech Connect

    Gladden, J.B.; Mackey, H.E.; Paller, M.H.; Specht, W.L.; Wike, L.D.; Wilde, E.W.

    1991-06-01

    The K Reactor at the Savannah River Site (SRS) shut-down in spring 1988 for maintenance and safety upgrades. Since that time the receiving stream for thermal effluent, Indian Grave Branch and Pen Branch, have undergone a pattern of post-thermal recovery that is typical of other SRS streams following removal of thermal stress. Divesity of fish and aquatic macroinvertebrate communities has increased and available habitats have been colonized by numerous species of herbaceous and woody plants. K Reactor is scheduled to resume operation in 1991 and operate through 1992 without a cooling tower to cool the discharge. It is likely that the reactor will operate at approximately one-third to one-half of full power (800--1200 MW thermal) during this period and effluent temperatures will be substantially lower than earlier operation at full power. Monthly average discharge temperatures at half-power operation will range from approximately 42{degrees}C in winter to 49{degrees}C in summer. The volume of water discharged will not be affected by altered power levels and will average approximately 10--11 m{sup 3}/s. The ecological consequences of this mode of operation on the Indian Grave/Pen Branch stream system have been evaluated.

  17. Testing of an Integrated Reactor Core Simulator and Power Conversion System with Simulated Reactivity Feedback

    NASA Technical Reports Server (NTRS)

    Bragg-Sitton, Shannon M.; Hervol, David S.; Godfroy, Thomas J.

    2010-01-01

    A Direct Drive Gas-Cooled (DDG) reactor core simulator has been coupled to a Brayton Power Conversion Unit (BPCU) for integrated system testing at NASA Glenn Research Center (GRC) in Cleveland, Ohio. This is a closed-cycle system that incorporates an electrically heated reactor core module, turboalternator, recuperator, and gas cooler. Nuclear fuel elements in the gas-cooled reactor design are replaced with electric resistance heaters to simulate the heat from nuclear fuel in the corresponding fast spectrum nuclear reactor. The thermodynamic transient behavior of the integrated system was the focus of this test series. In order to better mimic the integrated response of the nuclear-fueled system, a simulated reactivity feedback control loop was implemented. Core power was controlled by a point kinetics model in which the reactivity feedback was based on core temperature measurements; the neutron generation time and the temperature feedback coefficient are provided as model inputs. These dynamic system response tests demonstrate the overall capability of a non-nuclear test facility in assessing system integration issues and characterizing integrated system response times and response characteristics.

  18. Testing of an Integrated Reactor Core Simulator and Power Conversion System with Simulated Reactivity Feedback

    NASA Technical Reports Server (NTRS)

    Bragg-Sitton, Shannon M.; Hervol, David S.; Godfroy, Thomas J.

    2009-01-01

    A Direct Drive Gas-Cooled (DDG) reactor core simulator has been coupled to a Brayton Power Conversion Unit (BPCU) for integrated system testing at NASA Glenn Research Center (GRC) in Cleveland, OH. This is a closed-cycle system that incorporates an electrically heated reactor core module, turbo alternator, recuperator, and gas cooler. Nuclear fuel elements in the gas-cooled reactor design are replaced with electric resistance heaters to simulate the heat from nuclear fuel in the corresponding fast spectrum nuclear reactor. The thermodynamic transient behavior of the integrated system was the focus of this test series. In order to better mimic the integrated response of the nuclear-fueled system, a simulated reactivity feedback control loop was implemented. Core power was controlled by a point kinetics model in which the reactivity feedback was based on core temperature measurements; the neutron generation time and the temperature feedback coefficient are provided as model inputs. These dynamic system response tests demonstrate the overall capability of a non-nuclear test facility in assessing system integration issues and characterizing integrated system response times and response characteristics.

  19. Analysis of closed cycle megawatt class space power systems with nuclear reactor heat sources

    NASA Technical Reports Server (NTRS)

    Juhasz, A. J.; Jones, B. I.

    1987-01-01

    The analysis and integration studies of multimegawatt nuclear power conversion systems for potential SDI applications is presented. A study is summarized which considered 3 separate types of power conversion systems for steady state power generation with a duty requirement of 1 yr at full power. The systems considered are based on the following conversion cycles: direct and indirect Brayton gas turbine, direct and indirect liquid metal Rankine, and in core thermionic. A complete mass analysis was performed for each system at power levels ranging from 1 to 25 MWe for both heat pipe and liquid droplet radiator options. In the modeling of common subsystems, reactor and shield calculations were based on multiparameter correlation and an in-house analysis for the heat rejection and other subsystems.

  20. Status of R&D activities on materials for fusion power reactors

    NASA Astrophysics Data System (ADS)

    Baluc, N.; Abe, K.; Boutard, J. L.; Chernov, V. M.; Diegele, E.; Jitsukawa, S.; Kimura, A.; Klueh, R. L.; Kohyama, A.; Kurtz, R. J.; Lässer, R.; Matsui, H.; Möslang, A.; Muroga, T.; Odette, G. R.; Tran, M. Q.; van der Schaaf, B.; Wu, Y.; Yu, J.; Zinkle, S. J.

    2007-10-01

    Current R&D activities on materials for fusion power reactors are mainly focused on plasma facing, structural and tritium breeding materials for plasma facing (first wall, divertor) and breeding blanket components. Most of these activities are being performed in Europe, Japan, the People's Republic of China, Russia and the USA. They relate to the development of new high temperature, radiation resistant materials, the development of coatings that will act as erosion, corrosion, permeation and/or electrical/MHD barriers, characterization of candidate materials in terms of mechanical and physical properties, assessment of irradiation effects, compatibility experiments, development of reliable joints, and development and/or validation of design rules. Priorities defined worldwide in the field of materials for fusion power reactors are summarized, as well as the main achievements obtained during the last few years and the near-term perspectives in the different investigation areas.

  1. Rotating ned reactor as a power source for em gun applications

    SciTech Connect

    Powell, J.; Botts, T.; Stickley, C.M.; Meth, S.

    1982-01-01

    Electromagnetic gun applications of the Rotating Bed Reactor (RBR) are examined. The RBR is a compact (approximately 1 m/sup 3/), (up to several thousand MW(th), high-power reactor concept, capable of producing a high-temperature (up to approximately 3000/degree/K) gas stream with a MHD generator coupled to it. The RBR can generate electric power (up to approximately 1000 MW(e)) in the pulsed or cw modes. Three EM gun applications are investigated: a rail gun thruster for orbit transfer, a rapid-fire EM gun for point defense, and a direct ground-to-space launch. The RBR appears suitable for all applications. 4 refs.

  2. Design and Build of Reactor Simulator for Fission Surface Power Technology Demonstrator Unit

    NASA Astrophysics Data System (ADS)

    Godfroy, T.; Dickens, R.; Houts, M.; Pearson, B.; Webster, K.; Gibson, M.; Qualls, L.; Poston, D.; Werner, J.; Radel, R.

    The Nuclear Systems Team at Marshall Space Flight Center (MSFC) focuses on technology development for state of the art capability in non-nuclear testing of nuclear system and Space Nuclear Power for fission reactor systems for lunar and mars surface power generation as well as radioisotope power systems for both spacecraft and surface applications. Currently being designed and developed is a reactor simulator (RxSim) for incorporation into the Technology Demonstrator Unit (TDU) for the Fission Surface Power System (FSPS) Program which is supported by multiple national laboratories and NASA centers. The ultimate purpose of the RxSim is to provide heated NaK to a pair of Stirling engines in the TDU. The RxSim includes many different systems, components, and instrumentation that have been developed at MSFC while working with pumped NaK systems and in partnership with the national laboratories and NASA centers. The main components of the RxSim are a core, a pump, a heat exchanger (to mimic the thermal load of the Stirling engines), and a flow meter when being tested at MSFC. When tested at GRC the heat exchanger will be replaced with a Stirling power conversion engine. Additional components include storage reservoirs, expansion volumes, overflow catch tanks, safety and support hardware, instrumenta- tion (temperature, pressure, flow) data collection, and power supplies. This paper will discuss the design and current build status of the RxSim for delivery to GRC in early 2012.

  3. Design and Build of Reactor Simulator for Fission Surface Power Technology Demonstrator Unit

    NASA Technical Reports Server (NTRS)

    Godfroy, Thomas; Dickens, Ricky; Houts, Michael; Pearson, Boise; Webster, Kenny; Gibson, Marc; Qualls, Lou; Poston, Dave; Werner, Jim; Radel, Ross

    2011-01-01

    The Nuclear Systems Team at NASA Marshall Space Flight Center (MSFC) focuses on technology development for state of the art capability in non-nuclear testing of nuclear system and Space Nuclear Power for fission reactor systems for lunar and Mars surface power generation as well as radioisotope power systems for both spacecraft and surface applications. Currently being designed and developed is a reactor simulator (RxSim) for incorporation into the Technology Demonstrator Unit (TDU) for the Fission Surface Power System (FSPS) Program, which is supported by multiple national laboratories and NASA centers. The ultimate purpose of the RxSim is to provide heated NaK to a pair of Stirling engines in the TDU. The RxSim includes many different systems, components, and instrumentation that have been developed at MSFC while working with pumped NaK systems and in partnership with the national laboratories and NASA centers. The main components of the RxSim are a core, a pump, a heat exchanger (to mimic the thermal load of the Stirling engines), and a flow meter for tests at MSFC. When tested at NASA Glenn Research Center (GRC) the heat exchanger will be replaced with a Stirling power conversion engine. Additional components include storage reservoirs, expansion volumes, overflow catch tanks, safety and support hardware, instrumentation (temperature, pressure, flow) for data collection, and power supplies. This paper will discuss the design and current build status of the RxSim for delivery to GRC in early 2012.

  4. Lunar electric power systems utilizing the SP-100 reactor coupled to dynamic conversion systems

    NASA Technical Reports Server (NTRS)

    Harty, Richard B.; Durand, Richard E.

    1993-01-01

    An integration study was performed by Rocketdyne under contract to NASA-LeRC. The study was concerned with coupling an SP-0100 reactor to either a Brayton or Stirling power conversion system. The application was for a surface power system to supply power requirements to a lunar base. A power level of 550 kWe was selected based on the NASA Space Exploration Initiative 90-day study. Reliability studies were initially performed to determine optimum power conversion redundancy. This study resulted in selecting three operating engines and one stand-by unit. Integration design studies indicated that either the Brayton or Stirling power conversion systems could be integrated with the PS-100 reactor. The Stirling system had an integration advantage because of smaller piping size and fewer components. The Stirling engine, however, is more complex and heavier than the Brayton rotating unit, which tends to off-set the Stirling integration advantage. From a performance consideration, the Brayton had a 9 percent mass advantage, and the Stirling had a 50 percent radiator advantage.

  5. Tritium activities in Canada supporting CANDU{sup R} nuclear power reactors

    SciTech Connect

    Miller, J. M.

    2008-07-15

    An overview of the various Canadian tritium research and operational activities supporting the development, refurbishment and operation of CANDU{sup R} nuclear power reactors is presented. These activities encompass tritium health and safety, tritium in the environment, tritium interaction with materials, and tritium processing, and relate to both supporting R and D advances as well as operational best practices. The collective results of these activities contribute to our goals of improving worker and public safety, and operational efficiency. (authors)

  6. Removal of toxic uranium from synthetic nuclear power reactor effluents using uranyl ion imprinted polymer particles.

    PubMed

    Preetha, Chandrika Ravindran; Gladis, Joseph Mary; Rao, Talasila Prasada; Venkateswaran, Gopala

    2006-05-01

    Major quantities of uranium find use as nuclear fuel in nuclear power reactors. In view of the extreme toxicity of uranium and consequent stringent limits fixed by WHO and various national governments, it is essential to remove uranium from nuclear power reactor effluents before discharge into environment. Ion imprinted polymer (IIP) materials have traditionally been used for the recovery of uranium from dilute aqueous solutions prior to detection or from seawater. We now describe the use of IIP materials for selective removal of uranium from a typical synthetic nuclear power reactor effluent. The IIP materials were prepared for uranyl ion (imprint ion) by forming binary salicylaldoxime (SALO) or 4-vinylpyridine (VP) or ternary SALO-VP complexes in 2-methoxyethanol (porogen) and copolymerizing in the presence of styrene (monomer), divinylbenzene (cross-linking monomer), and 2,2'-azobisisobutyronitrile (initiator). The resulting materials were then ground and sieved to obtain unleached polymer particles. Leached IIP particles were obtained by leaching the imprint ions with 6.0 M HCl. Control polymer particles were also prepared analogously without the imprint ion. The IIP particles obtained with ternary complex alone gave quantitative removal of uranyl ion in the pH range 3.5-5.0 with as low as 0.08 g. The retention capacity of uranyl IIP particles was found to be 98.50 mg/g of polymer. The present study successfully demonstrates the feasibility of removing uranyl ions selectively in the range 5 microg - 300 mg present in 500 mL of synthetic nuclear power reactor effluent containing a host of other inorganic species.

  7. Survey of Remote Area Monitoring Systems at U.S. Light-Water-Cooled Power Reactors

    SciTech Connect

    Kathren, R. L.; Mileham, A. P.

    1982-04-01

    A study was made of the capabilities and operating practices, including calibration, of remote area monitoring (RAM) systems at light-water-cooled power reactors in the United States. The information was obtained by mail questionaire. Specific design capabilities, including range, readout and alarm features are documented along with the numbers and location of detectors, calibration and operational procedures. Comments of respondents regarding RAM systems are also included.

  8. Experimental Evaluation of a Water Shield for a Surface Power Reactor

    NASA Technical Reports Server (NTRS)

    Pearson, J. Boise; Reid, Robert S.

    2006-01-01

    As part of the Vision for Space Exploration the end of the next decade will bring man back to the surface of the moon. One of the most critical issues for the establishment of human presence on the moon will be the availability of compact power sources. The establishment of man on the moon will require power from greater than 10's of kWt's in follow on years. Nuclear reactors are extremely we11 suited to meet the needs for power generation on the lunar or Martian surface. reactor system. Several competing concepts exist for lightweight, safe, robust shielding systems such as a water shield, lithium hydride (LiH), Boron Carbide, and others. Water offers several potential advantages, including reduced cost, reduced technical risk, and reduced mass. Water has not typically been considered for space reactor applications because of the need for gravity to remove the potential for radiation streaming paths. The water shield concept relies on predictions of passive circulation of the shield water by natural convection to adequately cool the shield. This prediction needs to be experimentally evaluated, especially for shields with complex geometries. MSFC has developed the experience and fac necessary to do this evaluation in the Early Flight Fission - Test Facility (EFF-TF).

  9. Preliminary Evaluation of Convective Heat Transfer in a Water Shield for a Surface Power Reactor

    NASA Technical Reports Server (NTRS)

    Pearson J. Boise; Reid, Robert S.

    2007-01-01

    As part of the Vision for Space Exploration, the end of the next decade will bring man back to the surface of the moon. A crucial issue for the establishment of human presence on the moon will be the availability of compact power sources. This presence could require greater than 10's of kWt's in follow on years. Nuclear reactors are well suited to meet the needs for power generation on the lunar or Martian surface. Radiation shielding is a key component of any surface power reactor system. Several competing concepts exist for lightweight, safe, robust shielding systems such as a water shield, lithium hydride (LiH), and boron carbide. Water offers several potential advantages, including reduced cost, reduced technical risk, and reduced mass. Water has not typically been considered for space reactor applications because of the need for gravity to fix the location of any vapor that could form radiation streaming paths. The water shield concept relies on the predictions of passive circulation of the shield water by natural convection to adequately cool the shield. This prediction needs to be experimentally evaluated, especially for shields with complex geometries. NASA Marshall Space Flight Center has developed the experience and facilities necessary to do this evaluation in its Early Flight Fission - Test Facility (EFF-TF).

  10. Power Distribution Analysis for the ORNL High Flux Isotope Reactor Critical Experiment 3

    SciTech Connect

    Chandler, David; Primm, Trent; Maldonado, G Ivan

    2010-01-01

    The mission of the Reduced Enrichment for Research and Test Reactors Program is to minimize and, to the extent possible, eliminate the use of highly enriched uranium (HEU) in civilian nuclear applications by working to convert research and test reactors, as well as radioisotope production processes, to low-enriched uranium (LEU) fuel and targets. Oak Ridge National Laboratory (ORNL) is currently reviewing the design bases and key operating criteria including fuel operating parameters, enrichment-related safety analyses, fuel performance, and fuel fabrication in regard to converting the fuel of the High Flux Isotope Reactor (HFIR) from HEU to LEU. The purpose of this study is to validate Monte Carlo methods currently in use for conversion analyses. The methods have been validated for the prediction offlux values in the reactor target, reflector, and beam tubes, but this study focuses on the prediction of the power density profile in the core. Power distributions were calculated in the fuel elements of the HFIR, a research reactor at ORNL, via MCNP and were compared to experimentally obtained data. This study was performed to validate Monte Carlo methods for power density calculations and to observe biases. A current three-dimensional MCNP model was modified to replicate the 1965 HFIR Critical Experiment 3 (HFIRCE-3). In this experiment, the power profile was determined by counting the gamma activity at selected locations in the core. 'Foils' (chunks of fuel meat and clad) were punched out of the fuel elements in HFIRCE-3 following irradiation, and experimental relative power densities were obtained by measuring the activity of these foils and comparing each foil's activity to the activity of a normalizing foil. This analysis consisted of calculating corresponding activities by inserting volume tallies into the modified MCNP model to represent the punchings. The average fission density was calculated for each foil location and then normalized to the reference foil

  11. Conceptual Design of HP-STMCs Space Reactor Power System for 110 kWe

    NASA Astrophysics Data System (ADS)

    El-Genk, Mohamed S.; Tournier, Jean-Michel

    2004-02-01

    A conceptual design of a Heat Pipe-Segmented Thermoelectric Module Converters (HP-STMCs) space reactor power system (SRPS) for a net power of 110 kWe is developed. The parametric analysis changed the number of radiator's potassium heat pipes from 224 to 336 and calculated the effects on the operation parameters and total mass of the system. The reactor has a hexagonal core comprised of 126 heat pipe modules, each consists of three UN, 1.5 cm OD fuel pins brazed to a central lithium heat pipe of identical diameter. The Re cladding of the fuel pins is brazed along the active core length to the lithium heat pipe using 6 Re tri-cusps. The reactor control is accomplished using 12 B4C/BeO control drums, a large diameter one on each side of the hexagonal core and a small diameter one at each corner. The control drums are placed within the radial BeO reflector (7.1-9.1 cm thick). The fuel pin peak-to-average power ratio in the reactor core is 1.12-1.19. Despite its very high density and fabrication challenge, using rhenium structure in the reactor core is necessary for three main reasons: (a) the high reactor temperature (>= 1500 K) (b) excellent compatibility with the UN fuel and lithium; (c) to cause a spectrum shift that ensures having sufficient negative reactivity margin during a water submersion accident. The reference HP-STMC system with 324, 2.42-3.03 cm OD potassium heat pipes in the radiator is 9.60 m long and has a cone angle of 30°. The nominal operation of the reactor's lithium heat pipes and of the radiator's potassium heat pipes is at or below ~ 45% of the prevailing wicking and sonic limit, respectively. The masses of the reactor and radiation shadow shield are 753.7 kg and 999.5 kg, respectively; the average heat pipes temperature in the reactor is 1513 K; the mass of the reactor's lithium heat pipes with a C-C finned condenser that is 1.5 m long is 516.1 kg; the mass of the radiator is 557.5 kg, with an outer surface area of 87 m2 (6.41 kg/m2) and

  12. Conceptual Design of HP-STMCs Space Reactor Power System for 110 kWe

    SciTech Connect

    El-Genk, Mohamed S.; Tournier, Jean-Michel

    2004-02-04

    A conceptual design of a Heat Pipe-Segmented Thermoelectric Module Converters (HP-STMCs) space reactor power system (SRPS) for a net power of 110 kWe is developed. The parametric analysis changed the number of radiator's potassium heat pipes from 224 to 336 and calculated the effects on the operation parameters and total mass of the system. The reactor has a hexagonal core comprised of 126 heat pipe modules, each consists of three UN, 1.5 cm OD fuel pins brazed to a central lithium heat pipe of identical diameter. The Re cladding of the fuel pins is brazed along the active core length to the lithium heat pipe using 6 Re tri-cusps. The reactor control is accomplished using 12 B4C/BeO control drums, a large diameter one on each side of the hexagonal core and a small diameter one at each corner. The control drums are placed within the radial BeO reflector (7.1-9.1 cm thick). The fuel pin peak-to-average power ratio in the reactor core is 1.12-1.19. Despite its very high density and fabrication challenge, using rhenium structure in the reactor core is necessary for three main reasons: (a) the high reactor temperature ({>=} 1500 K); (b) excellent compatibility with the UN fuel and lithium; (c) to cause a spectrum shift that ensures having sufficient negative reactivity margin during a water submersion accident. The reference HP-STMC system with 324, 2.42-3.03 cm OD potassium heat pipes in the radiator is 9.60 m long and has a cone angle of 30 deg. The nominal operation of the reactor's lithium heat pipes and of the radiator's potassium heat pipes is at or below {approx} 45% of the prevailing wicking and sonic limit, respectively. The masses of the reactor and radiation shadow shield are 753.7 kg and 999.5 kg, respectively; the average heat pipes temperature in the reactor is 1513 K; the mass of the reactor's lithium heat pipes with a C-C finned condenser that is 1.5 m long is 516.1 kg; the mass of the radiator is 557.5 kg, with an outer surface area of 87 m2 (6.41 kg/m2

  13. Feasibility of Ground Testing a Moon and Mars Surface Power Reactor in EBR-II

    NASA Astrophysics Data System (ADS)

    Morton, Sheryl L.; Baily, Carl E.; Hill, Thomas J.; Werner, James E.

    2006-01-01

    Ground testing of a surface fission power system would be necessary to verify the design and validate reactor performance to support safe and sustained human exploration of the Moon and Mars. The Idaho National Laboratory (INL) has several facilities that could be adapted to support a ground test. This paper focuses on the feasibility of ground testing at the Experimental Breeder Reactor II (EBR-II) facility and using other INL existing infrastructure to support such a test. This brief study concludes that the INL EBR-II facility and supporting infrastructure are a viable option for ground testing the surface power system. It provides features and attributes that offer advantages to locating and performing ground testing at this site, and it could support the National Aeronautics and Space Administration schedules for human exploration of the Moon. This study used the initial concept examined by the U.S. Department of Energy Inter-laboratory Design and Analysis Support Team for surface power, a low-temperature, liquid-metal, three-loop Brayton power system. With some facility modification, the EBR-II can safely house a test chamber and perform long-term testing of the space reactor power system. The INL infrastructure is available to receive and provide bonded storage for special nuclear materials. Facilities adjacent to EBR-II can provide the clean room environment needed to assemble and store the test article assembly, disassemble the power system at the conclusion of testing, and perform posttest examination. Capability for waste disposal is also available at the INL.

  14. Feasibility of Ground Testing a Moon and Mars Surface Power Reactor in EBR-II

    SciTech Connect

    Sheryl Morton; Carl Baily; Tom Hill; Jim Werner

    2006-02-01

    Ground testing of a surface fission power system would be necessary to verify the design and validate reactor performance to support safe and sustained human exploration of the Moon and Mars. The Idaho National Laboratory (INL) has several facilities that could be adapted to support a ground test. This paper focuses on the feasibility of ground testing at the Experimental Breeder Reactor II (EBR-II) facility and using other INL existing infrastructure to support such a test. This brief study concludes that the INL EBR-II facility and supporting infrastructure are a viable option for ground testing the surface power system. It provides features and attributes that offer advantages to locating and performing ground testing at this site, and it could support the National Aeronautics and Space Administration schedules for human exploration of the Moon. This study used the initial concept examined by the U.S. Department of Energy Inter-laboratory Design and Analysis Support Team for surface power, a lowtemperature, liquid-metal, three-loop Brayton power system. With some facility modification, the EBR-II can safely house a test chamber and perform long-term testing of the space reactor power system. The INL infrastructure is available to receive and provide bonded storage for special nuclear materials. Facilities adjacent to EBR-II can provide the clean room environment needed to assemble and store the test article assembly, disassemble the power system at the conclusion of testing, and perform posttest examination. Capability for waste disposal is also available at the INL.

  15. Feasibility of Ground Testing a Moon and Mars Surface Power Reactor in EBR-II

    SciTech Connect

    Morton, Sheryl L.; Baily, Carl E.; Hill, Thomas J.; Werner, James E.

    2006-01-20

    Ground testing of a surface fission power system would be necessary to verify the design and validate reactor performance to support safe and sustained human exploration of the Moon and Mars. The Idaho National Laboratory (INL) has several facilities that could be adapted to support a ground test. This paper focuses on the feasibility of ground testing at the Experimental Breeder Reactor II (EBR-II) facility and using other INL existing infrastructure to support such a test. This brief study concludes that the INL EBR-II facility and supporting infrastructure are a viable option for ground testing the surface power system. It provides features and attributes that offer advantages to locating and performing ground testing at this site, and it could support the National Aeronautics and Space Administration schedules for human exploration of the Moon. This study used the initial concept examined by the U.S. Department of Energy Inter-laboratory Design and Analysis Support Team for surface power, a low-temperature, liquid-metal, three-loop Brayton power system. With some facility modification, the EBR-II can safely house a test chamber and perform long-term testing of the space reactor power system. The INL infrastructure is available to receive and provide bonded storage for special nuclear materials. Facilities adjacent to EBR-II can provide the clean room environment needed to assemble and store the test article assembly, disassemble the power system at the conclusion of testing, and perform posttest examination. Capability for waste disposal is also available at the INL.

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

    SciTech Connect

    Devgun, Jas S.

    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)

  17. Development of Liquid-Vapor Core Reactors with MHD Generator for Space Power and Propulsion Applications

    SciTech Connect

    Samim Anghaie

    2002-08-13

    Any reactor that utilizes fuel consisting of a fissile material in a gaseous state may be referred to as a gaseous core reactor (GCR). Studies on GCRs have primarily been limited to the conceptual phase, mostly due to budget cuts and program cancellations in the early 1970's. A few scientific experiments have been conducted on candidate concepts, primarily of static pressure fissile gas filling a cylindrical or spherical cavity surrounded by a moderating shell, such as beryllium, heavy water, or graphite. The main interest in this area of nuclear power generation is for space applications. The interest in space applications has developed due to the promise of significant enhancement in fuel utilization, safety, plant efficiency, special high-performance features, load-following capabilities, power conversion optimization, and other key aspects of nuclear power generation. The design of a successful GCR adapted for use in space is complicated. The fissile material studied in the pa st has been in a fluorine compound, either a tetrafluoride or a hexafluoride. Both of these molecules have an impact on the structural material used in the making of a GCR. Uranium hexafluoride as a fuel allows for a lower operating temperature, but at temperatures greater than 900K becomes essentially impossible to contain. This difficulty with the use of UF6 has caused engineers and scientists to use uranium tetrafluoride, which is a more stable molecule but has the disadvantage of requiring significantly higher operating temperatures. Gas core reactors have traditionally been studied in a steady state configuration. In this manner a fissile gas and working fluid are introduced into the core, called a cavity, that is surrounded by a reflector constructed of materials such as Be or BeO. These reactors have often been described as cavity reactors because the density of the fissile gas is low and criticality is achieved only by means of the reflector to reduce neutron leakage from the core

  18. Thermal and neutron-physical features of the nuclear reactor for a power pulsation plant for space applications

    NASA Astrophysics Data System (ADS)

    Gordeev, É. G.; Kaminskii, A. S.; Konyukhov, G. V.; Pavshuk, V. A.; Turbina, T. A.

    2012-05-01

    We have explored the possibility of creating small-size reactors with a high power output with the provision of thermal stability and nuclear safety under standard operating conditions and in emergency situations. The neutron-physical features of such a reactor have been considered and variants of its designs preserving the main principles and approaches of nuclear rocket engine technology are presented.

  19. High Temperature Water Heat Pipes Radiator for a Brayton Space Reactor Power System

    NASA Astrophysics Data System (ADS)

    El-Genk, Mohamed S.; Tournier, Jean-Michel

    2006-01-01

    A high temperature water heat pipes radiator design is developed for a space power system with a sectored gas-cooled reactor and three Closed Brayton Cycle (CBC) engines, for avoidance of single point failures in reactor cooling and energy conversion and rejection. The CBC engines operate at turbine inlet and exit temperatures of 1144 K and 952 K. They have a net efficiency of 19.4% and each provides 30.5 kWe of net electrical power to the load. A He-Xe gas mixture serves as the turbine working fluid and cools the reactor core, entering at 904 K and exiting at 1149 K. Each CBC loop is coupled to a reactor sector, which is neutronically and thermally coupled, but hydraulically decoupled to the other two sectors, and to a NaK-78 secondary loop with two water heat pipes radiator panels. The segmented panels each consist of a forward fixed segment and two rear deployable segments, operating hydraulically in parallel. The deployed radiator has an effective surface area of 203 m2, and when the rear segments are folded, the stowed power system fits in the launch bay of the DELTA-IV Heavy launch vehicle. For enhanced reliability, the water heat pipes operate below 50% of their wicking limit; the sonic limit is not a concern because of the water, high vapor pressure at the temperatures of interest (384 - 491 K). The rejected power by the radiator peaks when the ratio of the lengths of evaporator sections of the longest and shortest heat pipes is the same as that of the major and minor widths of the segments. The shortest and hottest heat pipes in the rear segments operate at 491 K and 2.24 MPa, and each rejects 154 W. The longest heat pipes operate cooler (427 K and 0.52 MPa) and because they are 69% longer, reject more power (200 W each). The longest and hottest heat pipes in the forward segments reject the largest power (320 W each) while operating at ~ 46% of capillary limit. The vapor temperature and pressure in these heat pipes are 485 K and 1.97 MPa. By contrast, the

  20. Flux stability and power control in the Soviet RBMK-1000 reactors

    SciTech Connect

    Meriwether, G.H.; McNeece, J.P.

    1993-08-01

    As a result of the Chernobyl accident, the Soviets have studied and implemented various design changes to improve the safety of the RBMK reactors. The safety measurements include modifications of the control rod configuration, fuel enrichment increase from 2.0 to 2.4 weight percent U-235, and installation of additional supplemental absorbers. The purpose of this study is to investigate the effects of increased fuel enrichment, different control rod positions, and supplemental absorber loadings on reactivity control, power distribution within the large RBMK core, and relative stability against power oscillations.

  1. Review of the Tri-Agency Space Nuclear Reactor Power System Technology Program

    NASA Technical Reports Server (NTRS)

    Ambrus, J. H.; Wright, W. E.; Bunch, D. F.

    1984-01-01

    The Space Nuclear Reactor Power System Technology Program designated SP-100 was created in 1983 by NASA, the U.S. Department of Defense, and the Defense Advanced Research Projects Agency. Attention is presently given to the development history of SP-100 over the course of its first year, in which it has been engaged in program objectives' definition, the analysis of civil and military missions, nuclear power system functional requirements' definition, concept definition studies, the selection of primary concepts for technology feasibility validation, and the acquisition of initial experimental and analytical results.

  2. High Temperature Water Heat Pipes Radiator for a Brayton Space Reactor Power System

    SciTech Connect

    El-Genk, Mohamed S.; Tournier, Jean-Michel

    2006-01-20

    A high temperature water heat pipes radiator design is developed for a space power system with a sectored gas-cooled reactor and three Closed Brayton Cycle (CBC) engines, for avoidance of single point failures in reactor cooling and energy conversion and rejection. The CBC engines operate at turbine inlet and exit temperatures of 1144 K and 952 K. They have a net efficiency of 19.4% and each provides 30.5 kWe of net electrical power to the load. A He-Xe gas mixture serves as the turbine working fluid and cools the reactor core, entering at 904 K and exiting at 1149 K. Each CBC loop is coupled to a reactor sector, which is neutronically and thermally coupled, but hydraulically decoupled to the other two sectors, and to a NaK-78 secondary loop with two water heat pipes radiator panels. The segmented panels each consist of a forward fixed segment and two rear deployable segments, operating hydraulically in parallel. The deployed radiator has an effective surface area of 203 m2, and when the rear segments are folded, the stowed power system fits in the launch bay of the DELTA-IV Heavy launch vehicle. For enhanced reliability, the water heat pipes operate below 50% of their wicking limit; the sonic limit is not a concern because of the water, high vapor pressure at the temperatures of interest (384 - 491 K). The rejected power by the radiator peaks when the ratio of the lengths of evaporator sections of the longest and shortest heat pipes is the same as that of the major and minor widths of the segments. The shortest and hottest heat pipes in the rear segments operate at 491 K and 2.24 MPa, and each rejects 154 W. The longest heat pipes operate cooler (427 K and 0.52 MPa) and because they are 69% longer, reject more power (200 W each). The longest and hottest heat pipes in the forward segments reject the largest power (320 W each) while operating at {approx} 46% of capillary limit. The vapor temperature and pressure in these heat pipes are 485 K and 1.97 MPa. By

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

    SciTech Connect

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

    1991-01-01

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

  4. [A new correction method for radionuclide formation in neutron activation analysis using a reactor power meter coupled with a microcomputer].

    PubMed

    Hirai, S; Yoshino, Y; Suzuki, S; Horiuchi, N

    1982-05-01

    Neutron flux and irradiation time should be accurately known in neutron activation analysis using very short lived nuclides in which conventional monitoring methods i.e., a comparator method, flux monitor method and so on cannot be used satisfactorily. Especially, fluctuation of neutron flux has not been corrected. We noted a change of reactor power at a pneumatic operation, and found out a new correction method for its correction in activation analysis. In our small nuclear reactor, TRIGA-II, the reactor power increased rapidly a few % when a pneumatic-operated capsule arrived at a core of the reactor, and decreased when the capsule left from the core. If the duration between these two changes of the reactor power is equal to the irradiation time, and that the reactor power is proportional to the neutron flux, we can regard an activity formation as a time integration of the reactor power. Then, the correction system was made of a reactor power meter, a V-F converter (voltage to frequency converter), a clock time, a counter, a microcomputer, electric circuits and so on. The signal of the reactor power during the irradiation was counted through the V-F converter, and was accumulated in a memory of the microcomputer. The neutron fluence was calculated in this microcomputer. This method was examined by means of activation of copper and selenium standard samples by 9-11 sec irradiations. The observed activity involved +/- 10% error. However, the error in the corrected activity was decreased to a few % using this correction method. As a result, we found that this method can be used to obtain accurate value for radionuclide formation.

  5. Formulation and experimental evaluation of closed-form control laws for the rapid maneuvering of reactor neutronic power

    SciTech Connect

    Bernard, J.A. . Nuclear Reactor Lab.)

    1989-09-01

    This report describes both the theoretical development and the experimental evaluation of a novel, robust methodology for the time-optimal adjustment of a reactor's neutronic power under conditions of closed-loop digital control. Central to the approach are the MIT-SNL Period-Generated Minimum Time Control Laws' which determine the rate at which reactivity should be changed in order to cause a reactor's neutronic power to conform to a specified trajectory. Using these laws, reactor power can be safely raised by five to seven orders of magnitude in a few seconds. The MIT-SNL laws were developed to facilitate rapid increases of neutronic power on spacecraft reactors operating in an SDI environment. However, these laws are generic and have other applications including the rapid recovery of research and test reactors subsequent to an unanticipated shutdown, power increases following the achievement of criticality on commercial reactors, power adjustments on commercial reactors so as to minimize thermal stress, and automated startups. The work reported here was performed by the Massachusetts Institute of Technology under contract to the Sandia National Laboratories. Support was also provided by the US Department of Energy's Division of University and Industry Programs. The work described in this report is significant in that a novel solution to the problem of time-optimal control of neutronic power was identified, in that a rigorous description of a reactor's dynamics was derived in that the rate of change of reactivity was recognized as the proper control signal, and in that extensive experimental trials were conducted of these newly developed concepts on actual nuclear reactors. 43 refs., 118 figs., 11 tabs.

  6. 76 FR 17160 - Office of New Reactors; Final Interim Staff Guidance on the Review of Nuclear Power Plant Designs...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-03-28

    ... COMMISSION Office of New Reactors; Final Interim Staff Guidance on the Review of Nuclear Power Plant Designs Using a Gas Turbine Driven Standby Emergency Alternating Current Power System AGENCY: Nuclear Regulatory... Guidance (ISG) DC/COL-ISG-021 titled ``Interim Staff Guidance on the Review of Nuclear Power Plant...

  7. An RF-powered micro-reactor for the detection of astrobiological target molecules on planetary bodies

    NASA Astrophysics Data System (ADS)

    Scott, Valerie J.; Tse, Margaret; Shearn, Michael J.; Siegel, Peter H.; Amashukeli, Xenia

    2012-08-01

    We describe a sample-processing micro-reactor that utilizes 60 GHz RF radiation with approximately 730 mW of output power. The instrument design and performance characterization are described and then illustrated with modeling and experimental studies. The micro-reactor's efficiency on affecting hydrolysis of chemical bonds similar to those within large complex molecules was demonstrated: a disaccharide—sucrose—was hydrolyzed completely under micro-reactor conditions. The products of the micro-reactor-facilitated hydrolysis were analyzed using mass spectroscopy and proton nuclear magnetic resonance analytical techniques.

  8. An RF-powered micro-reactor for the detection of astrobiological target molecules on planetary bodies.

    PubMed

    Scott, Valerie J; Tse, Margaret; Shearn, Michael J; Siegel, Peter H; Amashukeli, Xenia

    2012-08-01

    We describe a sample-processing micro-reactor that utilizes 60 GHz RF radiation with approximately 730 mW of output power. The instrument design and performance characterization are described and then illustrated with modeling and experimental studies. The micro-reactor's efficiency on affecting hydrolysis of chemical bonds similar to those within large complex molecules was demonstrated: a disaccharide-sucrose-was hydrolyzed completely under micro-reactor conditions. The products of the micro-reactor-facilitated hydrolysis were analyzed using mass spectroscopy and proton nuclear magnetic resonance analytical techniques.

  9. ORNL R and D on advanced small and medium power reactors: Selected topics

    SciTech Connect

    White, J.D.; Trauger, D.B.

    1988-01-01

    From 1984-1985, ORNL studied several innovative small and medium power nuclear concepts with respect to viability. Criteria for assessment of market attractiveness were developed and are described here. Using these criteria and descriptions of selected advanced reactor concepts, and assessment of their projected market viability in the time period 2000-2010 was made. All of these selected concepts could be considered as having the potential for meeting the criteria but, in most cases, considerable RandD would be required to reduce uncertainties. This work and later studies of safety and licensing of advanced, passively safe reactor concepts by ORNL are described. The results of these studies are taken into account in most of the current (FY 1989) work at ORNL on advanced reactors. A brief outline of this current work is given. One of the current RandD efforts at ORNL which addresses the operability and safety of advanced reactors is the Advanced Controls Program. Selected topics from this Program are described. 13 refs., 1 fig.

  10. Knowledges and abilities catalog for nuclear power plant operators: Savannah River Site (SRS) production reactors

    SciTech Connect

    Not Available

    1990-06-20

    The Knowledges and Abilities Catalog for Nuclear Power Plant Operations: Savannah River Site (SRS) Production Reactors, provides the basis for the development of content-valid certification examinations for Senior Reactor Operators (SROs) and Central Control Room Supervisors (SUP). The position of Shift Technical Engineer (STE) has been included in the catalog for completeness. This new SRS reactor operating shift crew position is held by an individual holding a CCR Supervisor Certification who has received special engineering and technical training. Also, the STE has a Bachelor of Science degree in engineering or a related technical field. The SRS catalog contains approximately 2500 knowledge and ability (K/A) statements for SROs and SUPs at heavy water moderated production reactors. Each K/A statement has been rated for its importance to the safe operation of the plant in a manner ensuring the health and safety of the public. The SRS K/A catalog is presently organized into five major sections: Plant Systems grouped by Safety Function, Plant Wide Generic K/As, Emergency Plant Evolutions, Theory and Components (to be developed).

  11. Experimental Evaluation of the Thermal Performance of a Water Shield for a Surface Power Reactor

    SciTech Connect

    Pearson, J. Boise; Stewart, Eric T.; Reid, Robert S.

    2007-01-30

    Water based reactor shielding is being investigated for use on initial lunar surface power systems. A water shield may lower overall cost (as compared to development cost for other materials) and simplify operations in the setup and handling. The thermal hydraulic performance of the shield is of significant interest. The mechanism for transferring heat through the shield is natural convection. Natural convection in a 100 kWt lunar surface reactor shield design is evaluated with 2 kW power input to the water in the Water Shield Testbed (WST) at the NASA Marshall Space Flight Center. The experimental data from the WST is used to validate a CFD model. Performance of the water shield on the lunar surface is then predicted with a CFD model anchored to test data. The experiment had a maximum water temperature of 75 deg. C. The CFD model with 1/6-g predicts a maximum water temperature of 88 deg. C with the same heat load and external boundary conditions. This difference in maximum temperature does not greatly affect the structural design of the shield, and demonstrates that it may be possible to use water for a lunar reactor shield.

  12. EXPERIMENTAL EVALUATION OF THE THERMAL PERFORMANCE OF A WATER SHIELD FOR A SURFACE POWER REACTOR

    SciTech Connect

    REID, ROBERT S.; PEARSON, J. BOSIE; STEWART, ERIC T.

    2007-01-16

    Water based reactor shielding is being investigated for use on initial lunar surface power systems. A water shield may lower overall cost (as compared to development cost for other materials) and simplify operations in the setup and handling. The thermal hydraulic performance of the shield is of significant interest. The mechanism for transferring heat through the shield is natural convection. Natural convection in a 100 kWt lunar surface reactor shield design is evaluated with 2 kW power input to the water in the Water Shield Testbed (WST) at the NASA Marshall Space Flight Center. The experimental data from the WST is used to validate a CFD model. Performance of the water shield on the lunar surface is then predicted with a CFD model anchored to test data. The experiment had a maximum water temperature of 75 C. The CFD model with 1/6-g predicts a maximum water temperature of 88 C with the same heat load and external boundary conditions. This difference in maximum temperature does not greatly affect the structural design of the shield, and demonstrates that it may be possible to use water for a lunar reactor shield.

  13. Neural Network with Local Memory for Nuclear Reactor Power Level Control

    SciTech Connect

    Uluyol, Oender; Ragheb, Magdi; Tsoukalas, Lefteri

    2001-02-15

    A methodology is introduced for a neural network with local memory called a multilayered local output gamma feedback (LOGF) neural network within the paradigm of locally-recurrent globally-feedforward neural networks. It appears to be well-suited for the identification, prediction, and control tasks in highly dynamic systems; it allows for the presentation of different timescales through incorporation of a gamma memory. A learning algorithm based on the backpropagation-through-time approach is derived. The spatial and temporal weights of the network are iteratively optimized for a given problem using the derived learning algorithm. As a demonstration of the methodology, it is applied to the task of power level control of a nuclear reactor at different fuel cycle conditions. The results demonstrate that the LOGF neural network controller outperforms the classical as well as the state feedback-assisted classical controllers for reactor power level control by showing a better tracking of the demand power, improving the fuel and exit temperature responses, and by performing robustly in different fuel cycle and power level conditions.

  14. Nuclear reactor power for a space-based radar. SP-100 project

    NASA Technical Reports Server (NTRS)

    Bloomfield, Harvey; Heller, Jack; Jaffe, Leonard; Beatty, Richard; Bhandari, Pradeep; Chow, Edwin; Deininger, William; Ewell, Richard; Fujita, Toshio; Grossman, Merlin

    1986-01-01

    A space-based radar mission and spacecraft, using a 300 kWe nuclear reactor power system, has been examined, with emphasis on aspects affecting the power system. The radar antenna is a horizontal planar array, 32 X 64 m. The orbit is at 61 deg, 1088 km. The mass of the antenna with support structure is 42,000 kg; of the nuclear reactor power system, 8,300 kg; of the whole spacecraft about 51,000 kg, necessitating multiple launches and orbital assembly. The assembly orbit is at 57 deg, 400 km, high enough to provide the orbital lifetime needed for orbital assembly. The selected scenario uses six Shuttle launches to bring the spacecraft and a Centaur G upper-stage vehicle to assembly orbit. After assembly, the Centaur places the spacecraft in operational orbit, where it is deployed on radio command, the power system started, and the spacecraft becomes operational. Electric propulsion is an alternative and allows deployment in assembly orbit, but introduces a question of nuclear safety.

  15. Research on pressure control of pressurizer in pressurized water reactor nuclear power plant

    NASA Astrophysics Data System (ADS)

    Dai, Ling; Yang, Xuhong; Liu, Gang; Ye, Jianhua; Qian, Hong; Xue, Yang

    2010-07-01

    Pressurizer is one of the most important components in the nuclear reactor system. Its function is to keep the pressure of the primary circuit. It can prevent shutdown of the system from the reactor accident under the normal transient state while keeping the setting value in the normal run-time. This paper is mainly research on the pressure system which is running in the Daya Bay Nuclear Power Plant. A conventional PID controller and a fuzzy controller are designed through analyzing the dynamic characteristics and calculating the transfer function. Then a fuzzy PID controller is designed by analyzing the results of two controllers. The fuzzy PID controller achieves the optimal control system finally.

  16. Design of a Low Power, Fast-Spectrum, Liquid-Metal Cooled Surface Reactor System

    SciTech Connect

    Marcille, T. F.; Poston, D. I.; Kapernick, R. J.; Dixon, D. D.; Fischer, G. A.; Doherty, S. P.

    2006-01-20

    In the current 2005 US budget environment, competition for fiscal resources make funding for comprehensive space reactor development programs difficult to justify and accommodate. Simultaneously, the need to develop these systems to provide planetary and deep space-enabling power systems is increasing. Given that environment, designs intended to satisfy reasonable near-term surface missions, using affordable technology-ready materials and processes warrant serious consideration. An initial lunar application design incorporating a stainless structure, 880 K pumped NaK coolant system and a stainless/UO2 fuel system can be designed, fabricated and tested for a fraction of the cost of recent high-profile reactor programs (JIMO, SP-100). Along with the cost reductions associated with the use of qualified materials and processes, this design offers a low-risk, high-reliability implementation associated with mission specific low temperature, low burnup, five year operating lifetime requirements.

  17. Space reactor/Stirling cycle systems for high power lunar application

    NASA Technical Reports Server (NTRS)

    Schmitz, Paul C.; Mason, Lee S.

    1991-01-01

    An analysis is performed to mathematically model a 550 kWe lunar base power supply which uses a SP-100 reactor coupled with Stirling converters. The reactor is placed in an excavation to keep activated coolant in the hole and to allow maintenance of the components outside the hole. Two technology levels are considered. They are 1050 and 1300 K heater head Stirling converts. It is found that for a 1050 K converter the total mass which provided 1000 volts DC at 250 m is 14,366 kg while the 1300 K system mass is 12,104 kg. The radiation area of the 1050 and 1300 K systems are 641 and 356 sq m respectively. Comparisons are made with Brayton and thermionic systems with both near term and advanced technology considered.

  18. Investigation of applications for high-power, self-critical fissioning uranium plasma reactors

    NASA Technical Reports Server (NTRS)

    Rodgers, R. J.; Latham, T. S.; Krascella, N. L.

    1976-01-01

    Analytical studies were conducted to investigate potentially attractive applications for gaseous nuclear cavity reactors fueled by uranium hexafluoride and its decomposition products at temperatures of 2000 to 6000 K and total pressures of a few hundred atmospheres. Approximate operating conditions and performance levels for a class of nuclear reactors in which fission energy removal is accomplished principally by radiant heat transfer from the high temperature gaseous nuclear fuel to surrounding absorbing media were determined. The results show the radiant energy deposited in the absorbing media may be efficiently utilized in energy conversion system applications which include (1) a primary energy source for high thrust, high specific impulse space propulsion, (2) an energy source for highly efficient generation of electricity, and (3) a source of high intensity photon flux for heating working fluid gases for hydrogen production or MHD power extraction.

  19. Space reactor/Stirling cycle systems for high power lunar applications

    NASA Technical Reports Server (NTRS)

    Schmitz, Paul C.; Mason, Lee S.

    1991-01-01

    An analysis is performed to mathematically model a 550 kWe lunar base power supply which uses a SP-100 reactor coupled with Stirling converters. The reactor is placed in an excavation to keep activated coolant in the hole and to allow maintance of the components outside the hole. Two technology levels are considered. They are 1050 and 1300 K heater head Stirling converts. It is found that for a 1050 K converter the total mass which provided 1000 volts dc at 250 m is 14,366 kg while the 1300 K system mass is 12,104 kg. The radiation area of the 1050 and 1300 K systems are 641 and 356 sq m respectively. Comparisons are made with Brayton and thermionic systems with both near term and advanced technology considered.

  20. Reactor

    DOEpatents

    Evans, Robert M.

    1976-10-05

    1. A neutronic reactor having a moderator, coolant tubes traversing the moderator from an inlet end to an outlet end, bodies of material fissionable by neutrons of thermal energy disposed within the coolant tubes, and means for circulating water through said coolant tubes characterized by the improved construction wherein the coolant tubes are constructed of aluminum having an outer diameter of 1.729 inches and a wall thickness of 0.059 inch, and the means for circulating a liquid coolant through the tubes includes a source of water at a pressure of approximately 350 pounds per square inch connected to the inlet end of the tubes, and said construction including a pressure reducing orifice disposed at the inlet ends of the tubes reducing the pressure of the water by approximately 150 pounds per square inch.

  1. Detailed axial power profiles in a MOX fuel experiment in the Advanced Test Reactor

    SciTech Connect

    Chang, G.; Ryskamp, J.M.

    1998-12-31

    The US Department of Energy (DOE) has chosen two options to dispose of surplus weapons-grade (WG) plutonium (WGPu). One option is to burn the WGPu in mixed-oxide (MOX) fuel in light water reactors. An average power test (<10 kW/ft) of WG-MOX fuel was inserted in the Advanced Test Reactor (ATR) in February 1998. A high power test (<15 kW/ft) of WG-MOX fuel in ATR will follow the average-power test. The ability to accurately predict fuel power is essential in the high-power WG-MOX fuel capsule design for the test in ATR. Detailed fission power and temperature profiles may influence gallium migration in WG-MOX fuel pins. Most of the fission heat generated in the MOX fuel capsules will transfer radially to the water coolant. However, because of the short length (15.24 cm) of the MOX fuel pellet stacks, some of the fission heat will transfer through the end pellets axially. Compounded with peak fission power local-to-average ratios (LTAR) at the ends of MOX fuel stacks, the hot spot created may exceed the design limit. Therefore, the prediction of the axial fission power profiles over the MOX fuel stacks at the beginning of life (BOL) and end of life (EOL) are important for MOX fuel performance analysis and capsule design for testing in ATR. Continuous-energy MCNP linked with ORIGEN2 can generate the burnup-dependent cross sections and fission power distribution for fuel burnup analysis while accurately including the effects of self-shielding. This approach is very important for the prediction of plutonium content and LTAR in MOX fuel pellet stacks with HfO{sub 2} ends. MCWO can accurately determine fuel pin power distributions in the ATR experiment when the MOX fuel and HfO{sub 2} are depleted simultaneously. This is significant because the authors quickly provided the customer with the required detailed power distributions within the MOX pins using the new approach. The MOX fuel pin with HfO{sub 2} can flatten the axial power profiles from BOL to EOL and meet the MOX

  2. Optimization of power-cycle arrangements for Supercritical Water cooled Reactors (SCWRs)

    NASA Astrophysics Data System (ADS)

    Lizon-A-Lugrin, Laure

    The world energy demand is continuously rising due to the increase of both the world population and the standard of life quality. Further, to assure both a healthy world economy as well as adequate social standards, in a relatively short term, new energy-conversion technologies are mandatory. Within this framework, a Generation IV International Forum (GIF) was established by the participation of 10 countries to collaborate for developing nuclear power reactors that will replace the present technology by 2030. The main goals of these nuclear-power reactors are: economic competitiveness, sustainability, safety, reliability and resistance to proliferation. As a member of the GIF, Canada has decided to orient its efforts towards the design of a CANDU-type Super Critical Water-cooled Reactor (SCWR). Such a system must run at a coolant outlet temperature of about 625°C and at a pressure of 25 MPa. It is obvious that at such conditions the overall efficiency of this kind of Nuclear Power Plant (NPP) will compete with actual supercritical water-power boilers. In addition, from a heat-transfer viewpoint, the use of a supercritical fluid allows the limitation imposed by Critical Heat Flux (CHF) conditions, which characterize actual technologies, to be removed. Furthermore, it will be also possible to use direct thermodynamic cycles where the supercritical fluid expands right away in a turbine without the necessity of using intermediate steam generators and/or separators. This work presents several thermodynamic cycles that could be appropriate to run SCWR power plants. Improving both thermal efficiency and mechanical power constitutes a multi-objective optimization problem and requires specific tools. To this aim, an efficient and robust evolutionary algorithm, based on genetic algorithm, is used and coupled to an appropriate power plant thermodynamic simulation model. The results provide numerous combinations to achieve a thermal efficiency higher than 50% with a

  3. Feasibility study of the University of Utah TRIGA reactor power upgrade in respect to control rod system

    NASA Astrophysics Data System (ADS)

    Cutic, Avdo

    The objectives of this thesis are twofold: to determine the highest achievable power levels of the current University of Utah TRIG Reactor (UUTR) core configuration with the existing three control rods, and to design the core for higher reactor power by optimizing the control rod worth. For the current core configuration, the maximum reactor power, eigenvalue keff, shutdown margin, and excess reactivity have been measured and calculated. These calculated estimates resulted from thermal power calibrations, and the control rod worth measurements at various power levels. The results were then used as a benchmark to verify the MCNP5 core simulations for the current core and then to design a core for higher reactor power. This study showed that the maximum achievable power with the current core configuration and control rod system is 150kW, which is 50kW higher than the licensed power of the UUTR. The maximum achievable UUTR core power with the existing fuel is determined by optimizing the core configuration and control rod worth, showing that a power upgrade of 500 kW is achievable. However, it requires a new control rod system consisting of a total of four control rods. The cost of such an upgrade is $115,000.

  4. Post 9-11 Security Issues for Non-Power Reactor Facilities

    SciTech Connect

    Zaffuts, P. J.

    2003-02-25

    This paper addresses the legal and practical issues arising out of the design and implementation of a security-enhancement program for non power reactor nuclear facilities. The security enhancements discussed are derived from the commercial nuclear power industry's approach to security. The nuclear power industry's long and successful experience with protecting highly sensitive assets provides a wealth of information and lessons that should be examined by other industries contemplating security improvements, including, but not limited to facilities using or disposing of nuclear materials. This paper describes the nuclear industry's approach to security, the advantages and disadvantages of its constituent elements, and the legal issues that facilities will need to address when adopting some or all of these elements in the absence of statutory or regulatory requirements to do so.

  5. High-power-density approaches to magnetic fusion energy: Problems and promise of compact Reversed-Field Pinch Reactors (CRFPR)

    NASA Astrophysics Data System (ADS)

    Hagenson, R. L.; Krakowski, R. A.; Dreicer, H.

    If the cost assumptions upon which the positive assessment of conventional large superconducting fusion reactors are based proves optimistic, approaches that promise considerably increased system power density and reduced mass utilization are required. These more compact reactor embodiments generally must operate with reduced shield thickness and resistive magnets. Because of the unique magnetic topology associated with the Reversed Field Pinch (RFP), the compact reactor embodiment of this approach is particularly attractive from the view point of low field resistive coils operating with Ohmic losses that are small relative to the fusion power. The RFP, one example of a high power density (HPD) approach to magnetic fusion energy. A comprehensive system model is described and applied to select a unique, cost optimized design point that is used for a subsequent conceptual engineering design of the Compact RFP Reactor.

  6. Performance Analyses of 38 kWe Turbo-Machine Unit for Space Reactor Power Systems

    NASA Astrophysics Data System (ADS)

    Gallo, Bruno M.; El-Genk, Mohamed S.

    2008-01-01

    This paper developed a design and investigated the performance of 38 kWe turbo-machine unit for space nuclear reactor power systems with Closed Brayton Cycle (CBC) energy conversion. The compressor and turbine of this unit are scaled versions of the NASA's BRU developed in the sixties and seventies. The performance results of turbo-machine unit are calculated for rotational speed up to 45 krpm, variable reactor thermal power and system pressure, and fixed turbine and compressor inlet temperatures of 1144 K and 400 K. The analyses used a detailed turbo-machine model developed at the University of New Mexico that accounts for the various energy losses in the compressor and turbine and the effect of compressibility of the He-Xe (40 mole/g) working fluid with increased flow rate. The model also accounts for the changes in the physical and transport properties of the working fluid with temperature and pressure. Results show that a unit efficiency of 24.5% is achievable at rotation speed of 45 krpm and system pressure of 0.75 MPa, assuming shaft and electrical generator efficiencies of 86.7% and 90%. The corresponding net electric power output of the unit is 38.5 kWe, the flow rate of the working fluid is 1.667 kg/s, the pressure ratio and polytropic efficiency for the compressor are 1.60 and 83.1%, and 1.51 and 88.3% for the turbine.

  7. A fission matrix based validation protocol for computed power distributions in the advanced test reactor

    SciTech Connect

    Nielsen, J. W.; Nigg, D. W.; LaPorta, A. W.

    2013-07-01

    The Idaho National Laboratory (INL) has been engaged in a significant multi year effort to modernize the computational reactor physics tools and validation procedures used to support operations of the Advanced Test Reactor (ATR) and its companion critical facility (ATRC). Several new protocols for validation of computed neutron flux distributions and spectra as well as for validation of computed fission power distributions, based on new experiments and well-recognized least-squares statistical analysis techniques, have been under development. In the case of power distributions, estimates of the a priori ATR-specific fuel element-to-element fission power correlation and covariance matrices are required for validation analysis. A practical method for generating these matrices using the element-to-element fission matrix is presented, along with a high-order scheme for estimating the underlying fission matrix itself. The proposed methodology is illustrated using the MCNP5 neutron transport code for the required neutronics calculations. The general approach is readily adaptable for implementation using any multidimensional stochastic or deterministic transport code that offers the required level of spatial, angular, and energy resolution in the computed solution for the neutron flux and fission source. (authors)

  8. Performance Analyses of 38 kWe Turbo-Machine Unit for Space Reactor Power Systems

    SciTech Connect

    Gallo, Bruno M.; El-Genk, Mohamed S.

    2008-01-21

    This paper developed a design and investigated the performance of 38 kWe turbo-machine unit for space nuclear reactor power systems with Closed Brayton Cycle (CBC) energy conversion. The compressor and turbine of this unit are scaled versions of the NASA's BRU developed in the sixties and seventies. The performance results of turbo-machine unit are calculated for rotational speed up to 45 krpm, variable reactor thermal power and system pressure, and fixed turbine and compressor inlet temperatures of 1144 K and 400 K. The analyses used a detailed turbo-machine model developed at University of New Mexico that accounts for the various energy losses in the compressor and turbine and the effect of compressibility of the He-Xe (40 mole/g) working fluid with increased flow rate. The model also accounts for the changes in the physical and transport properties of the working fluid with temperature and pressure. Results show that a unit efficiency of 24.5% is achievable at rotation speed of 45 krpm and system pressure of 0.75 MPa, assuming shaft and electrical generator efficiencies of 86.7% and 90%. The corresponding net electric power output of the unit is 38.5 kWe, the flow rate of the working fluid is 1.667 kg/s, the pressure ratio and polytropic efficiency for the compressor are 1.60 and 83.1%, and 1.51 and 88.3% for the turbine.

  9. Collaborative investigations of in-service irradiated material from the Japan Power Demonstration Reactor pressure vessel

    SciTech Connect

    Corwin, W.R.; Broadhead, B.L.; Suzuki, M.; Kohsaka, A.

    1997-02-01

    There is a need to validate the results of irradiation effects research by the examination of material taken directly from the wall of a pressure vessel that has been irradiated during normal service. Just such an evaluation is currently being conducted on material from the wall of the pressure vessel from the Japan Power Demonstration Reactor (JPDR). The research is being jointly performed at the Tokai Research Establishment of the Japan Atomic Energy Research Institute (JAERI) and by the Nuclear Regulatory Commission (NRC)-funded Heavy-Section Steel Irradiation Program at the Oak Ridge National Laboratory (ORNL).

  10. Compatibility tests of materials for a lithium-cooled space power reactor concept

    NASA Technical Reports Server (NTRS)

    Sinclair, J. H.

    1973-01-01

    Materials for a lithium-cooled space power reactor concept must be chemically compatible for up to 50,000 hr at high temperature. Capsule tests at 1040 C (1900 F) were made of material combinations of prime interest: T-111 in direct contact with uranium mononitride (UN), Un in vacuum separated from T-111 by tungsten wire, UN with various oxygen impurity levels enclosed in tungsten wire lithium-filled T-111 capsules, and TZM and lithium together in T-111 capsules. All combinations were compatible for over 2800 hr except for T-111 in direct contact with UN.

  11. Startup thaw concept for the SP-100 space reactor power system

    NASA Technical Reports Server (NTRS)

    Kirpich, A.; Das, A.; Choe, H.; Mcnamara, E.; Switick, D.; Bhandari, P.

    1990-01-01

    A thaw concept for a space reactor power system which employs lithium as a circulant for both the heat-transport and the heat-rejection fluid loops is presented. An exemplary thermal analysis for a 100-kWe (i.e., SP-100) system is performed. It is shown that the design of the thaw system requires a thorough knowledge of the various physical states of the circulant throughout the system, both spatially and temporally, and that the design has to provide adequate margins for the system to avoid a structural or thermally induced damage.

  12. Source-term reevaluation for US commercial nuclear power reactors: a status report

    SciTech Connect

    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.

  13. Assessments of Longevity of Equipment Metal of Nuclear Power Plants equipped with Reactors VVER-1000

    SciTech Connect

    Gorbatykh, V.P.; Al Kassem, S.N.

    2004-07-01

    Characteristics of damage processes of metal of coffer-dams of steam generators collectors at nuclear power plants (NPPs) equipped with reactors VVER-1000 have been mentioned; principles of construction of longevity function has been cited and new approach has been shown while solving the problem of the longevity of the metal resource by substantiating the technological actions with new mode characteristics, performed with the help of specially developed equations and formulae, where practically all damage processes and all influencing factors can be accounted. (authors)

  14. Vibration behavior of fuel-element vibration suppressors for the advanced power reactor

    NASA Technical Reports Server (NTRS)

    Adams, D. W.; Fiero, I. B.

    1973-01-01

    Preliminary shock and vibration tests were performed on vibration suppressors for the advanced power reactor for space application. These suppressors position the fuel pellets in a pin type fuel element. The test determined the effect of varying axial clearance on the behavior of the suppressors when subjected to shock and vibratory loading. The full-size suppressor was tested in a mockup model of fuel and clad which required scaling of test conditions. The test data were correlated with theoretical predictions for suppressor failure. Good agreement was obtained. The maximum difference with damping neglected was about 30 percent. Neglecting damping would result in a conservative design.

  15. The scalability of OTR (out-of-core thermionic reactor) space nuclear power systems

    SciTech Connect

    Gallup, D.R.

    1990-03-01

    In this document, masses of the STAR-C power system and an optimized out-of-core thermionic reactor (OTR) power system versus power level are investigated. The impacts of key system parameters on system performance are also addressed. The STAR-C is mass competitive below about 15 kWe, but at higher power levels the scalability is relatively poor. An optimized OR is the least massive space nuclear power system below 25 kWe, and scales well to 50 kWe. The system parameters that have a significant impact on the scalability of the STAR-C are core thermal flux, thermionic converter efficiency, and core length to diameter ratio. The emissivity of the core surface is shown to be a relatively unimportant parameter. For an optimized OR power system, the most significant system parameter is the maximum allowable fuel temperature. It is also shown that if advanced radiation-hardened electronics are used in the satellite payload, a very large mass savings is realized. 10 refs., 23 figs., 7 tabs.

  16. EOIL power scaling in a 1-5 kW supersonic discharge-flow reactor

    NASA Astrophysics Data System (ADS)

    Davis, Steven J.; Lee, Seonkyung; Oakes, David B.; Haney, Julie; Magill, John C.; Paulsen, Dwane A.; Cataldi, Paul; Galbally-Kinney, Kristin L.; Vu, Danthu; Polex, Jan; Kessler, William J.; Rawlins, Wilson T.

    2008-02-01

    Scaling of EOIL systems to higher powers requires extension of electric discharge powers into the kW range and beyond with high efficiency and singlet oxygen yield. We have previously demonstrated a high-power microwave discharge approach capable of generating singlet oxygen yields of ~25% at ~50 torr pressure and 1 kW power. This paper describes the implementation of this method in a supersonic flow reactor designed for systematic investigations of the scaling of gain and lasing with power and flow conditions. The 2450 MHz microwave discharge, 1 to 5 kW, is confined near the flow axis by a swirl flow. The discharge effluent, containing active species including O II(a1Δ g, b1Σ g +), O( 3P), and O 3, passes through a 2-D flow duct equipped with a supersonic nozzle and cavity. I2 is injected upstream of the supersonic nozzle. The apparatus is water-cooled, and is modular to permit a variety of inlet, nozzle, and optical configurations. A comprehensive suite of optical emission and absorption diagnostics is used to monitor the absolute concentrations of O II(a), O II(b), O( 3P), O 3, I II, I(2P 3/2), I(2P 1/2), small-signal gain, and temperature in both the subsonic and supersonic flow streams. We discuss initial measurements of singlet oxygen and I* excitation kinetics at 1 kW power.

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

    SciTech Connect

    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.

  18. Assessment of the use of extended burnup fuel in light water power reactors

    SciTech Connect

    Baker, D.A.; Bailey, W.J.; Beyer, C.E.; Bold, F.C.; Tawil, J.J.

    1988-02-01

    This study has been conducted by Pacific Northwest Laboratory for the US Nuclear Regulatory Commission to review the environmental and economic impacts associated with the use of extended burnup nuclear fuel in light water power reactors. It has been proposed that current batch average burnup levels of 33 GWd/t uranium be increased to above 50 GWd/t. The environmental effects of extending fuel burnup during normal operations and during accident events and the economic effects of cost changes on the fuel cycle are discussed in this report. The physical effects of extended burnup on the fuel and the fuel assembly are also presented as a basis for the environmental and economic assessments. Environmentally, this burnup increase would have no significant impact over that of normal burnup. Economically, the increased burnup would have favorable effects, consisting primarily of a reduction: (1) total fuel requirements; (2) reactor downtime for fuel replacement; (3) the number of fuel shipments to and from reactor sites; and (4) repository storage requirements. 61 refs., 4 figs., 27 tabs.

  19. PR-EDB: Power Reactor Embrittlement Data Base, Version 2. Revision 2, Program description

    SciTech Connect

    Stallmann, F.W.; Wang, J.A.; Kam, F.B.K.; Taylor, B.J.

    1994-01-01

    Investigations of regulatory issues such as vessel integrity over plant life, vessel failure, and sufficiency of current codes Standard Review Plans (SRP`s) and Guides for license renewal can be greatly expedited by the use of a well-designed computerized data base. Also, such a data base is essential for the validation of embrittlement prediction models by researchers. The Power Reactor Embrittlement Data Base (PR-EDB) is such a comprehensive collection of data for US commercial nuclear reactors. The current version of the PR-EDB contains the Charpy test data that were irradiated in 252 capsules of 96 reactors and consists of 207 data points for heat-affected-zone (HAZ) materials (98 different HAZ), 227 data points for weld materials (105 different welds), 524 data points for base materials (136 different base materials), including 297 plate data points (85 different plates), 119 forging data points (31) different forging), and 108 correlation monitor materials data points (3 different plates). The data files are given in dBASE format and can be accessed with any computer using the DOS operating system. ``User-friendly`` utility programs are used to retrieve and select specific data, manipulate data, display data to the screen or printer, and to fit and plot Charpy impact data. The results of several studies investigated are presented in Appendix D.

  20. Dynamic parameters test of Haiyang Nuclear Power Engineering in reactor areas, China

    NASA Astrophysics Data System (ADS)

    Zhou, N.; Zhao, S.; Sun, L.

    2012-12-01

    Haiyang Nuclear Power Project is located in Haiyang city, China. It consists of 6×1000MW AP1000 Nuclear Power generator sets. The dynamic parameters of the rockmass are essential for the design of the nuclear power plant. No.1 and No.2 reactor area are taken as research target in this paper. Sonic logging, single hole and cross-hole wave velocity are carried out respectively on the site. There are four types of rock lithology within the measured depth. They are siltstone, fine sandstone, shale and allgovite. The total depth of sonic logging is 409.8m and 2049 test points. The sound wave velocity of the rocks are respectively 5521 m/s, 5576m/s, 5318 m/s and 5576 m/s. Accroding to the statistic data, among medium weathered fine sandstone, fairly broken is majority, broken and relatively integrity are second, part of integrity. Medium weathered siltstone, relatively integrity is mojority, fairly broken is second. Medium weathered shale, fairly broken is majority, broken and relatively integrity for the next and part of integrity. Slight weathered fine sandstone, siltstone, shale and allgovite, integrity is the mojority, relatively integrity for the next, part of fairly broken.The single hole wave velocity tests are set in two boreholesin No.1 reactor area and No.2 reactor area respectively. The test depths of two holes are 2-24m, and the others are 2-40m. The wave velocity data are calculated at different depth in each holes and dynamic parameters. According to the test statistic data, the wave velocity and the dynamic parameter values of rockmass are distinctly influenced by the weathering degree. The test results are list in table 1. 3 groups of cross hole wave velocity tests are set for No.1 and 2 reactor area, No.1 reactor area: B16, B16-1, B20(Direction:175°, depth: 100m); B10, B10-1, B11(269°, 40m); B21, B21-1, B17(154°, 40m); with HB16, HB10, HB21 as trigger holes; No.2 reactor area: B47, B47-1, HB51(176°, 100m); B40, B40-1, B41(272°, 40m); B42, B42-1, B

  1. 75 FR 5632 - Office of New Reactors; Interim Staff Guidance on the Review of Nuclear Power Plant Designs Using...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-02-03

    ... emergency alternating current power system. This ISG document provides guidance on the implementation of... EGTG systems that are air cooled and diesel oil fueled are considered in this interim guidance. DATES... COMMISSION Office of New Reactors; Interim Staff Guidance on the Review of Nuclear Power Plant Designs...

  2. 76 FR 8383 - Office of New Reactors; Interim Staff Guidance on Impacts of Construction of New Nuclear Power...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-02-14

    ... COMMISSION Office of New Reactors; Interim Staff Guidance on Impacts of Construction of New Nuclear Power... Staff Guidance (ISG) COL-ISG-022 entitled ``Impacts of Construction of ] New Nuclear Power Plants on Operating Units at Multi-Unit Sites'' (Agencywide Documents Access and Management System (ADAMS)...

  3. Radiotoxicity and decay heat power of spent nuclear fuel of VVER type reactors at long-term storage.

    PubMed

    Bergelson, B R; Gerasimov, A S; Tikhomirov, G V

    2005-01-01

    Radiotoxicity and decay heat power of the spent nuclear fuel of VVER-1000 type reactors are calculated during storage time up to 300,000 y. Decay heat power of radioactive waste (radwaste) determines parameters of the heat removal system for the safe storage of spent nuclear fuel. Radiotoxicity determines the radiological hazard of radwaste after its leakage and penetration into the environment.

  4. A Burst Mode, Ultrahigh Temperature UF4 Vapor Core Reactor Rankine Cycle Space Power System Concept

    NASA Technical Reports Server (NTRS)

    Dugan, E. T.; Kahook, S. D.; Diaz, N. J.

    1996-01-01

    Static and dynamic neutronic analyses have been performed on an innovative burst mode (100's of MW output for a few thousand seconds) Ulvahigh Temperature Vapor Core Reactor (UTVR) space nuclear power system. The NVTR employs multiple, neutronically-coupled fissioning cores and operates on a direct, closed Rankine cycle using a disk Magnetohydrodynamic (MHD) generater for energy conversion. The UTVR includes two types of fissioning core regions: (1) the central Ultrahigh Temperature Vapor Core (UTVC) which contains a vapor mixture of highly enriched UF4 fuel and a metal fluoride working fluid and (2) the UF4 boiler column cores located in the BeO moderator/reflector region. The gaseous nature of the fuel the fact that the fuel is circulating, the multiple coupled fissioning cores, and the use of a two phase fissioning fuel lead to unique static and dynamic neutronic characteristics. Static neutronic analysis was conducted using two-dimensional S sub n, transport theory calculations and three-dimensional Monte Carlo transport theory calculations. Circulating-fuel, coupled-core point reactor kinetics equations were used for analyzing the dynamic behavior of the UTVR. In addition to including reactivity feedback phenomena associated with the individual fissioning cores, the effects of core-to-core neutronic and mass flow coupling between the UTVC and the surrounding boiler cores were also included in the dynamic model The dynamic analysis of the UTVR reveals the existence of some very effectlve inherent reactivity feedback effects that are capable of quickly stabilizing this system, within a few seconds, even when large positive reactivity insertions are imposed. If the UTVC vapor fuel density feedback is suppressed, the UTVR is still inherently stable because of the boiler core liquid-fuel volume feedback; in contrast, suppression of the vapor fuel density feedback in 'conventional" gas core cavity reactors causes them to become inherently unstable. Due to the

  5. Estimation of Specific Mass for Multimegawatt NEP Systems Based on Vapor Core Reactors with MHD Power Conversion

    NASA Astrophysics Data System (ADS)

    Knight, Travis; Anghaie, Samim

    2004-02-01

    Very low specific-mass power generation in space is possible using Vapor Core Reactors with Magnetohydrodynamic (VCR/MHD) generator. These advanced reactors at the conceptual design level have potential for the generation of tens to hundreds of megawatts of power in space with specific mass of about 1 kg/kWe. Power for nuclear electric propulsion (NEP) is possible with almost direct power conditioning and coupling of the VCR/MHD power output to the VASIMR engine, MPD, and a whole host of electric thrusters. The VCR/MHD based NEP system is designed to power space transportation systems that dramatically reduce the mission time for human exploration of the entire solar system or for aggressive long-term robotic missions. There are more than 40 years of experience in the evaluation of the scientific and technical feasibility of gas and vapor core reactor concepts. The proposed VCR is based on the concept of a cavity reactor made critical through the use of a reflector such as beryllium or beryllium oxide. Vapor fueled cavity reactors that are considered for NEP applications operate at maximum core center and wall temperatures of 4000 K and 1500K, respectively. A recent investigation has resulted in the conceptual design of a uranium tetrafluoride fueled vapor core reactor coupled to a MHD generator. Detailed neutronic design and cycle analyses have been performed to establish the operating design parameters for 10 to 200 MWe NEP systems. An integral system engineering-simulation code is developed to perform parametric analysis and design optimization studies for the VCR/MHD power system. Total system weight and size calculated based on existing technology has proven the feasibility of achieving exceptionally low specific mass (α ~1 kg/kWe) with a VCR/MHD powered system.

  6. Toward Global Standards on Peaceful uses of Space Nuclear Reactor Power Systems

    NASA Astrophysics Data System (ADS)

    El-Genk, M. S.

    Space reactor power systems are enabling to future space exploration and outposts on the Moon, Mars and other celestial bodies. They could provide for electrical propulsion, thermal propulsion and bimodal operation of thermal propulsion and electrical power generation. The much higher specific impulse of nuclear thermal and electric propulsion, compared to chemical rockets, could significantly shorten the travel time to Mars and farthest destinations, reducing the exposure of equipment and astronauts to the harmful space radiation. This article addresses safety issues relevant to the design, operation and end-of-life storage of these systems in an attempt to stimulate a discussion of the need to establish specific and globally agreed upon safety standards and guidelines for future peaceful uses.

  7. SP-100, the US Space Nuclear Reactor Power Program. Technical information report

    SciTech Connect

    Truscello, V.C.

    1983-11-01

    DARPA, in conjunction with DOE`s Office of Nuclear Energy, and NASA`s Office of Aeronautics and Space Technology are jointly sponsoring a space nuclear reactor power system program known as the Space Power-100 (SP-100) Development Project. The program is presently in the critical technology phase. This phase, better known as technology assessment and advancement, includes mission requirements definition, system conceptual designs, and critical technology development. A ground test phase decision is scheduled for July 1985. If the decision is positive, the next phase would begin in fiscal year 1986. An overriding concern in conducting this program is to ensure that nuclear safety is being properly addressed even in these early stages.

  8. Office of Analysis and Evaluation of Operational Data 1989 annual report, Power reactors

    SciTech Connect

    1990-07-01

    The annual report of the US Nuclear Regulatory Commission's Office for Analysis and Evaluation of Operational Data (AEOD) is devoted to the activities performed during 1989. The report is published in two separate parts. This document, NUREG-1272, Vol. 4, No. 1, covers power reactors and presents an overview of the operating experience of the nuclear power industry from the NRC perspective, including comments about the trends of some key performance measures. The report also includes the principal findings and issues identified in AEOD studies over the past year and summarizes information from such sources as licensee event reports, diagnostic evaluations, and reports to the NRC's Operations Center. This report also compiles the status of staff actions resulting from previous Incident Investigation Team (IIT) reports. 16 figs., 9 tabs.

  9. An assessment and validation study of nuclear reactors for low power space applications

    NASA Technical Reports Server (NTRS)

    Klein, A. C.; Gedeon, S. R.; Morey, D. C.

    1987-01-01

    The feasibility and safety of six conceptual small, low power nuclear reactor designs was evaluated. Feasibility evaluations included the determination of sufficient reactivity margins for seven years of full power operation and safe shutdown as well as handling during pre-launch assembly phases. Safety evaluations were concerned with the potential for maintaining subcritical conditions in the event of launch or transportation accidents. These included water immersion accident scenarios both with and without water flooding the core. Results show that most of the concepts can potentially meet the feasibility and safety requirements; however, due to the preliminary nature of the designs considered, more detailed designs will be necessary to enable these concepts to fully meet the safety requirements.

  10. Experimental Evaluation of a Water Shield for a Surface Power Reactor

    NASA Technical Reports Server (NTRS)

    Pearson, J. B.; Reid, R.; Sadasivan, P.; Stewart, E.

    2007-01-01

    A water based shielding system is being investigated for use on initial lunar surface power systems. The use of water may lower overall cost (as compared to development cost for other materials) and simplify operations in the setup and handling. The thermal hydraulic performance of the shield is of significant interest. The mechanism for transferring heat through the shield is natural convection. A representative lunar surface reactor design is evaluated at various power levels in the Water Shield Testbed (WST) at the NASA Marshall Space Flight Center. The evaluation compares the experimental data from the WST to CFD models. Performance of a water shield on the lunar surface is predicted by CFD models anchored to test data, and by matching relevant dimensionless parameters.

  11. Test Results From a Direct Drive Gas Reactor Simulator Coupled to a Brayton Power Conversion Unit

    NASA Technical Reports Server (NTRS)

    Hervol, David S.; Briggs, Maxwell H.; Owen, Albert K.; Bragg-Sitton, Shannon M.

    2009-01-01

    The Brayton Power Conversion Unit (BPCU) located at NASA Glenn Research Center (GRC) in Cleveland, OH is a closed cycle system incorporating a turboaltemator, recuperator, and gas cooler connected by gas ducts to an external gas heater. For this series of tests, the BPCU was modified by replacing the gas heater with the Direct Drive Gas heater or DOG. The DOG uses electric resistance heaters to simulate a fast spectrum nuclear reactor similar to those proposed for space power applications. The combined system thermal transient behavior was the focus of these tests. The BPCU was operated at various steady state points. At each point it was subjected to transient changes involving shaft rotational speed or DOG electrical input. This paper outlines the changes made to the test unit and describes the testing that took place along with the test results.

  12. A feasibility assessment of nuclear reactor power system concepts for the NASA Growth Space Station

    NASA Technical Reports Server (NTRS)

    Bloomfield, H. S.; Heller, J. A.

    1986-01-01

    A preliminary feasibility assessment of the integration of reactor power system concepts with a projected growth Space Station architecture was conducted to address a variety of installation, operational, disposition and safety issues. A previous NASA sponsored study, which showed the advantages of Space Station - attached concepts, served as the basis for this study. A study methodology was defined and implemented to assess compatible combinations of reactor power installation concepts, disposal destinations, and propulsion methods. Three installation concepts that met a set of integration criteria were characterized from a configuration and operational viewpoint, with end-of-life disposal mass identified. Disposal destinations that met current aerospace nuclear safety criteria were identified and characterized from an operational and energy requirements viewpoint, with delta-V energy requirement as a key parameter. Chemical propulsion methods that met current and near-term application criteria were identified and payload mass and delta-V capabilities were characterized. These capabilities were matched against concept disposal mass and destination delta-V requirements to provide a feasibility of each combination.

  13. Technical evaluation of major candidate blanket systems for fusion power reactor

    NASA Astrophysics Data System (ADS)

    Tone, Tatsuzo; Seki, Masahiro; Minato, Akio

    1987-03-01

    The key functions required for tritium breeding blankets for a fusion power reactor are ; (1) self-sufficient tritium breeding, (2) in-situ tritium recovery and low tritium inventory, (3) high temperature cooling giving a high efficiency of electricity generation and (4) thermo-mechanical reliability and simplified remote maintenance to obtain high plant availability. Blanket performance is substantially governed by materials selection. Major options of structure/breeder/coolant/neutron multiplier materials considered for the present design study are PCA/Li/sub 2/O/H/sub 2/O/Be, Mo-alloy/Li/sub 2/O/He/Be, Mo-alloy/LiAlO/sub 2//He/Be, V-alloy/Li/Li/none, and Mo-alloy/Li/He/none. In addition, remote maintenance of blankets, tritium recovery system, heat transport and energy conversion have been investigated. In this report, technological problems and critical R and D issues for power reactor blanket development are identified and a comparison of major candidate blanket concepts is discussed in terms of the present materials data base, economic performance, prospects for future improvements, and engineering feasibility and difficulties based on the results obtained from individual design studies. improvements, and engineering feasibility and difficulties based on the results obtained from individual design studies.

  14. 14C release from a Soviet-designed pressurized water reactor nuclear power plant.

    PubMed

    Uchrin, G; Csaba, E; Hertelendi, E; Ormai, P; Barnabas, I

    1992-12-01

    The Paks Nuclear Power Plant in Hungary runs with four pressurized water reactors, each of 440-MWe capacity. Sampling systems have been developed and used to determine the 14C of various chemical forms (14CO2, 14CO, 14CnHm) in the airborne releases. The average normalized yearly discharge rates for the time period 1988-1991 are equal to 0.77 TBq GWe-1 y-1 for hydrocarbons and 0.05 TBq GWe-1 y-1 for CO2. The contribution of 14CO was less than 0.5% of the total emission. The 14C discharge rate is estimated to be four times higher than the corresponding mean data of Western European pressurized water reactors. The calculated effective dose equivalent to individuals living in the vicinity of the power plant, due to 14C release, was 0.64 microSv in 1991 while the effective dose equivalent due to the natural 14C level was 15 microSv y-1. The long-term global impact of the 14C release in the operational period of the plant (1982-1991) was 1,270 man-Sv. The 14C excess in the environmental air has been measured since 1989 by taking biweekly samples at a distance of 1.7 km from the nuclear power plant. The long-term average of radiocarbon excess coming from the power plant was 2 mBq m-3. The local 14C deposition was followed by tree ring analysis, too. No 14C increase higher than the uncertainty of the measurement (four per thousand = 0.17 mBq m-3) was observed.

  15. High-power-density approaches to magnetic fusion energy: Problems and promise of compact reversed-field pinch reactors (CRFPR)

    NASA Astrophysics Data System (ADS)

    Hagenson, Randy L.; Krakowski, Robert A.; Dreicer, Harry

    1983-03-01

    If the costing assumptions upon which the positive assessment of conventional large superconducting fusion reactors are based proves overly optimistic, approaches that promise considerably increased system power density and reduced mass utilization will be required. These more compact reactor embodiments generally must operate with reduced shield thickness and resistive magnets. Because of the unique magnetic topology associated with the Reversed-Field Pinch (RFP), the compact reactor embodiment for this approach is particularly attractive from the viewpoint of low-field resistive coils operating with ohmic losses that can be made small relative to the fusion power. The RFP, therefore, is used as one example of a high-power-density (HPD) approach to magnetic fusion energy. A comprehensive system model is described and applied to select a unique, cost-optimized design point that will be used for a subsequent conceptual engineering design of the compact RFP Reactor (CRFPR). This cost-optimized CRFPR design serves as an example of a HPD fusion reactor that would operate with system power densities and mass utilizations that are comparable to fission power plants, these measures of system performance being an order of magnitude more favorable than the conventional approaches to magnetic fusion energy (MFE).

  16. Combined on-board hydride slurry storage and reactor system and process for hydrogen-powered vehicles and devices

    SciTech Connect

    Brooks, Kriston P; Holladay, Jamelyn D; Simmons, Kevin L; Herling, Darrell R

    2014-11-18

    An on-board hydride storage system and process are described. The system includes a slurry storage system that includes a slurry reactor and a variable concentration slurry. In one preferred configuration, the storage system stores a slurry containing a hydride storage material in a carrier fluid at a first concentration of hydride solids. The slurry reactor receives the slurry containing a second concentration of the hydride storage material and releases hydrogen as a fuel to hydrogen-power devices and vehicles.

  17. Off-design temperature effects on nuclear fuel pins for an advanced space-power-reactor concept

    NASA Technical Reports Server (NTRS)

    Bowles, K. J.

    1974-01-01

    An exploratory out-of-reactor investigation was made of the effects of short-time temperature excursions above the nominal operating temperature of 990 C on the compatibility of advanced nuclear space-power reactor fuel pin materials. This information is required for formulating a reliable reactor safety analysis and designing an emergency core cooling system. Simulated uranium mononitride (UN) fuel pins, clad with tungsten-lined T-111 (Ta-8W-2Hf) showed no compatibility problems after heating for 8 hours at 2400 C. At 2520 C and above, reactions occurred in 1 hour or less. Under these conditions free uranium formed, redistributed, and attacked the cladding.

  18. Study on neutronic of very small Pb - Bi cooled no-onsite refueling nuclear power reactor (VSPINNOR)

    SciTech Connect

    Arianto, Fajar; Su'ud, Zaki; Zuhair

    2014-09-30

    A conceptual design study on Very Small Pb-Bi No-Onsite Refueling Cooled Nuclear Reactor (VSPINNOR) with Uranium nitride fuel using MCNPX program has been performed. In this design the reactor core is divided into three regions with different enrichment. At the center of the core is laid fuel without enrichment (internal blanket). While for the outer region using fuel enrichment variations. VSPINNOR fast reactor was operated for 10 years without refueling. Neutronic analysis shows optimized result of VSPINNOR has a core of 50 cm radius and 100 cm height with 300 MWth thermal power output at 60% fuel fraction that can be operated 18 years without refueling or fuel shuffling.

  19. Heat pipe space nuclear reactor design assessment. Volume 2: Feasibility study of upgrading the SP-100 heat pipe space nuclear power system

    NASA Astrophysics Data System (ADS)

    El-Genk, M. S.; Seo, J. T.

    1985-08-01

    This report investigated the feasibility of upgrading the power of the Heat Pipe Space Nuclear Reactor (HPSNR) system design. The report has also discussed the four primary methods for power upgrading: Increasing the thermal power output to the reactor core, pulse-mode operation, improving the heat rejection, and improving the thermal-to-electric energy conversion.

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

    SciTech Connect

    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.

  1. POWER CYCLE AND STRESS ANALYSES FOR HIGH TEMPERATURE GAS-COOLED REACTOR

    SciTech Connect

    Oh, Chang H; Davis, Cliff; Hawkes, Brian D; Sherman, Steven R

    2007-05-01

    The Department of Energy and the Idaho National Laboratory are developing a Next Generation Nuclear Plant (NGNP) to serve as a demonstration of state-of-the-art nuclear technology. The purpose of the demonstration is two fold 1) efficient low cost energy generation and 2) hydrogen production. Although a next generation plant could be developed as a single-purpose facility, early designs are expected to be dual-purpose. While hydrogen production and advanced energy cycles are still in its early stages of development, research towards coupling a high temperature reactor, electrical generation and hydrogen production is under way. Many aspects of the NGNP must be researched and developed in order to make recommendations on the final design of the plant. Parameters such as working conditions, cycle components, working fluids, and power conversion unit configurations must be understood. Three configurations of the power conversion unit were demonstrated in this study. A three-shaft design with three turbines and four compressors, a combined cycle with a Brayton top cycle and a Rankine bottoming cycle, and a reheated cycle with three stages of reheat were investigated. An intermediate heat transport loop for transporting process heat to a High Temperature Steam Electrolysis (HTSE) hydrogen production plant was used. Helium, CO2, and a 80% nitrogen, 20% helium mixture (by weight) were studied to determine the best working fluid in terms cycle efficiency and development cost. In each of these configurations the relative component size were estimated for the different working fluids. The relative size of the turbomachinery was measured by comparing the power input/output of the component. For heat exchangers the volume was computed and compared. Parametric studies away from the baseline values of the three-shaft and combined cycles were performed to determine the effect of varying conditions in the cycle. This gives some insight into the sensitivity of these cycles to

  2. Innovative self-powered submersible microbial electrolysis cell (SMEC) for biohydrogen production from anaerobic reactors.

    PubMed

    Zhang, Yifeng; Angelidaki, Irini

    2012-05-15

    A self-powered submersible microbial electrolysis cell (SMEC), in which a specially designed anode chamber and external electricity supply were not needed, was developed for in situ biohydrogen production from anaerobic reactors. In batch experiments, the hydrogen production rate reached 17.8 mL/L/d at the initial acetate concentration of 410 mg/L (5 mM), while the cathodic hydrogen recovery ( [Formula: see text] ) and overall systemic coulombic efficiency (CE(os)) were 93% and 28%, respectively, and the systemic hydrogen yield ( [Formula: see text] ) peaked at 1.27 mol-H(2)/mol-acetate. The hydrogen production increased along with acetate and buffer concentration. The highest hydrogen production rate of 32.2 mL/L/d and [Formula: see text] of 1.43 mol-H(2)/mol-acetate were achieved at 1640 mg/L (20 mM) acetate and 100 mM phosphate buffer. Further evaluation of the reactor under single electricity-generating or hydrogen-producing mode indicated that further improvement of voltage output and reduction of electron losses were essential for efficient hydrogen generation. In addition, alternate exchanging the electricity-assisting and hydrogen-producing function between the two cell units of the SMEC was found to be an effective approach to inhibit methanogens. Furthermore, 16S rRNA genes analysis showed that this special operation strategy resulted same microbial community structures in the anodic biofilms of the two cell units. The simple, compact and in situ applicable SMEC offers new opportunities for reactor design for a microbial electricity-assisted biohydrogen production system.

  3. A physical description of fission product behavior fuels for advanced power reactors.

    SciTech Connect

    Kaganas, G.; Rest, J.; Nuclear Engineering Division; Florida International Univ.

    2007-10-18

    The Global Nuclear Energy Partnership (GNEP) is considering a list of reactors and nuclear fuels as part of its chartered initiative. Because many of the candidate materials have not been explored experimentally under the conditions of interest, and in order to economize on program costs, analytical support in the form of combined first principle and mechanistic modeling is highly desirable. The present work is a compilation of mechanistic models developed in order to describe the fission product behavior of irradiated nuclear fuel. The mechanistic nature of the model development allows for the possibility of describing a range of nuclear fuels under varying operating conditions. Key sources include the FASTGRASS code with an application to UO{sub 2} power reactor fuel and the Dispersion Analysis Research Tool (DART ) with an application to uranium-silicide and uranium-molybdenum research reactor fuel. Described behavior mechanisms are divided into subdivisions treating fundamental materials processes under normal operation as well as the effect of transient heating conditions on these processes. Model topics discussed include intra- and intergranular gas-atom and bubble diffusion, bubble nucleation and growth, gas-atom re-solution, fuel swelling and ?scion gas release. In addition, the effect of an evolving microstructure on these processes (e.g., irradiation-induced recrystallization) is considered. The uranium-alloy fuel, U-xPu-Zr, is investigated and behavior mechanisms are proposed for swelling in the {alpha}-, intermediate- and {gamma}-uranium zones of this fuel. The work reviews the FASTGRASS kinetic/mechanistic description of volatile ?scion products and, separately, the basis for the DART calculation of bubble behavior in amorphous fuels. Development areas and applications for physical nuclear fuel models are identified.

  4. Lunar Nuclear Power Plant With Solid Core Reactor, Heatpipes and Thermoelectric Conversion

    NASA Astrophysics Data System (ADS)

    Sayre, Edwin D.; Ring, Peter J.; Brown, Neil; Elsner, Norbert B.; Bass, John C.

    2008-01-01

    This is a lunar nuclear power plant with the advantages of minimum mass, with no moving parts, no pumped liquid coolant, a solid metal rugged core, with no single point of failure. The electrical output is 100 kilowatts with a 500 kilowatt thermal reactor. The thermoelectric converters surround the potassium heatpipes from the core and water heatpipes surround the converter and connect to the radiator. The solid core reactor is made from HT9 alloy. The fuel is uranium oxide with 90% enrichment. The thermoelectric converter is bonded to the outside of the 1.10 inch ID heat pipe and is 30 inches long. The thermoelectric couple is Si/SiGe-Si/SiC Quantum Well with over 20% efficiency with an 890 K hot side and a 490 K cold side and produces 625 Watts. 176 converters produce 110 kWe. With less than 10% loss in controls this yields 100 kWe for use. The cylindrical thermoelectric converter is designed and fabricated by HIPing to keep brittle materials in compression and to ensure conductivity. The solid core is fabricated by machining the heatpipe tubes with 6 grooves that are diffusion bonded together by HIPing to form the fuel tubes. The maximum temperature of the heat pipes is 940 K and the return flow temperature is 890 K. The reactor core is hexagonal shaped, 61 cm. wide and 76.2 cm high with 12 rotating control drums surrounding it. There is shielding to protect components and human habitation. The radiator is daisy shaped at 45 degrees with each petal 5.5 meters long. The design life is ten years.

  5. Lunar Nuclear Power Plant With Solid Core Reactor, Heatpipes and Thermoelectric Conversion

    SciTech Connect

    Sayre, Edwin D.; Ring, Peter J.; Brown, Neil; Elsner, Norbert B.; Bass, John C.

    2008-01-21

    This is a lunar nuclear power plant with the advantages of minimum mass, with no moving parts, no pumped liquid coolant, a solid metal rugged core, with no single point of failure. The electrical output is 100 kilowatts with a 500 kilowatt thermal reactor. The thermoelectric converters surround the potassium heatpipes from the core and water heatpipes surround the converter and connect to the radiator. The solid core reactor is made from HT9 alloy. The fuel is uranium oxide with 90% enrichment. The thermoelectric converter is bonded to the outside of the 1.10 inch ID heat pipe and is 30 inches long. The thermoelectric couple is Si/SiGe-Si/SiC Quantum Well with over 20% efficiency with an 890 K hot side and a 490 K cold side and produces 625 Watts. 176 converters produce 110 kWe. With less than 10% loss in controls this yields 100 kWe for use. The cylindrical thermoelectric converter is designed and fabricated by HIPing to keep brittle materials in compression and to ensure conductivity. The solid core is fabricated by machining the heatpipe tubes with 6 grooves that are diffusion bonded together by HIPing to form the fuel tubes. The maximum temperature of the heat pipes is 940 K and the return flow temperature is 890 K. The reactor core is hexagonal shaped, 61 cm. wide and 76.2 cm high with 12 rotating control drums surrounding it. There is shielding to protect components and human habitation. The radiator is daisy shaped at 45 degrees with each petal 5.5 meters long. The design life is ten years.

  6. Innovative self-powered submersible microbial electrolysis cell (SMEC) for biohydrogen production from anaerobic reactors.

    PubMed

    Zhang, Yifeng; Angelidaki, Irini

    2012-05-15

    A self-powered submersible microbial electrolysis cell (SMEC), in which a specially designed anode chamber and external electricity supply were not needed, was developed for in situ biohydrogen production from anaerobic reactors. In batch experiments, the hydrogen production rate reached 17.8 mL/L/d at the initial acetate concentration of 410 mg/L (5 mM), while the cathodic hydrogen recovery ( [Formula: see text] ) and overall systemic coulombic efficiency (CE(os)) were 93% and 28%, respectively, and the systemic hydrogen yield ( [Formula: see text] ) peaked at 1.27 mol-H(2)/mol-acetate. The hydrogen production increased along with acetate and buffer concentration. The highest hydrogen production rate of 32.2 mL/L/d and [Formula: see text] of 1.43 mol-H(2)/mol-acetate were achieved at 1640 mg/L (20 mM) acetate and 100 mM phosphate buffer. Further evaluation of the reactor under single electricity-generating or hydrogen-producing mode indicated that further improvement of voltage output and reduction of electron losses were essential for efficient hydrogen generation. In addition, alternate exchanging the electricity-assisting and hydrogen-producing function between the two cell units of the SMEC was found to be an effective approach to inhibit methanogens. Furthermore, 16S rRNA genes analysis showed that this special operation strategy resulted same microbial community structures in the anodic biofilms of the two cell units. The simple, compact and in situ applicable SMEC offers new opportunities for reactor design for a microbial electricity-assisted biohydrogen production system. PMID:22402271

  7. NRC review of Electric Power Research Institute's Advanced Light Reactor Utility Requirements Document - Program summary, Project No. 669

    SciTech Connect

    Not Available

    1992-08-01

    The staff of the US Nuclear Regulatory Commission has prepared Volume 1 of a safety evaluation report (SER), NRC Review of Electric Power Research Institute's Advanced Light Water Reactor Utility Requirements Document -- Program Summary,'' to document the results of its review of the Electric Power Research Institute's Advanced Light Water Reactor Utility Requirements Document.'' This SER provides a discussion of the overall purpose and scope of the Requirements Document, the background of the staff's review, the review approach used by the staff, and a summary of the policy and technical issues raised by the staff during its review.

  8. Process Model of A Fusion Fuel Recovery System for a Direct Drive IFE Power Reactor

    NASA Astrophysics Data System (ADS)

    Natta, Saswathi; Aristova, Maria; Gentile, Charles

    2008-11-01

    A task has been initiated to develop a detailed representative model for the fuel recovery system (FRS) in the prospective direct drive inertial fusion energy (IFE) reactor. As part of the conceptual design phase of the project, a chemical process model is developed in order to observe the interaction of system components. This process model is developed using FEMLAB Multiphysics software with the corresponding chemical engineering module (CEM). Initially, the reactants, system structure, and processes are defined using known chemical species of the target chamber exhaust. Each step within the Fuel recovery system is modeled compartmentally and then merged to form the closed loop fuel recovery system. The output, which includes physical properties and chemical content of the products, is analyzed after each step of the system to determine the most efficient and productive system parameters. This will serve to attenuate possible bottlenecks in the system. This modeling evaluation is instrumental in optimizing and closing the fusion fuel cycle in a direct drive IFE power reactor. The results of the modeling are presented in this paper.

  9. Worldwide assessment of steam-generator problems in pressurized-water-reactor nuclear power plants

    SciTech Connect

    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.

  10. Preliminary analysis of hot spot factors in an advanced reactor for space electric power systems

    NASA Technical Reports Server (NTRS)

    Lustig, P. H.; Holms, A. G.; Davison, H. W.

    1973-01-01

    The maximum fuel pin temperature for nominal operation in an advanced power reactor is 1370 K. Because of possible nitrogen embrittlement of the clad, the fuel temperature was limited to 1622 K. Assuming simultaneous occurrence of the most adverse conditions a deterministic analysis gave a maximum fuel temperature of 1610 K. A statistical analysis, using a synthesized estimate of the standard deviation for the highest fuel pin temperature, showed probabilities of 0.015 of that pin exceeding the temperature limit by the distribution free Chebyshev inequality and virtually nil assuming a normal distribution. The latter assumption gives a 1463 K maximum temperature at 3 standard deviations, the usually assumed cutoff. Further, the distribution and standard deviation of the fuel-clad gap are the most significant contributions to the uncertainty in the fuel temperature.

  11. Geomagnetic origin for transient particle events from nuclear reactor-powered satellites.

    PubMed

    Share, G H; Kurfess, J D; Marlow, K W; Messina, D C

    1989-04-28

    Transient events observed since 1980 by the Gamma-Ray Spectrometer experiment on the Solar Maximum Mission satellite (SMM) have been identified with radiation emitted from 18 different Soviet nuclear reactor-powered satellites. Most of these satellites are similar to Cosmos 954 and 1402 which reentered the atmosphere. Gamma radiation from these satellites was detected when they passed within about 400 to 500 kilometers of SMM. Positron annihilation line radiation (511 kiloelectron volts) and charged-particle events were detected when SMM encountered clouds of positrons and electrons emitted by these satellites and stored up to tens of minutes in the geomagnetic field. The rate of these events varied from about 1 in 5 days to over 30 per day and was strongly dependent on the operating altitudes of the Cosmos satellites and density of the upper atmosphere.

  12. LWR (Light Water Reactor) power plant simulations using the AD10 and AD100 systems

    SciTech Connect

    Wulff, W.; Cheng, H.S.; Chien, C.J.; Jang, J.Y.; Lin, H.C.; Mallen, A.N.; Wang, S.J.; Institute of Nuclear Energy Research, Lung-Tan; Tawian Power Co., Taipei; Brookhaven National Lab., Upton, NY; Institute of Nuclear Energy Research, Lung-Tan )

    1989-01-01

    Boiling (BWR) and Pressurized (PWR) Water Reactor Power Plants are being simulated at BNL with the AD10 and AD100 Peripheral Processor Systems. The AD10 system has been used for BWR simulations since 1984 for safety analyses, emergency training and optimization studies. BWR simulation capabilities have been implemented recently on the AD100 system and PWR simulation capabilities are currently being developed under the auspices of international cooperation. Modeling and simulation methods are presented with emphasis on the simulation of the Nuclear Steam Supply System. Results are presented for BWR simulation and performance characteristics are compared of the AD10 and AD100 systems. It will be shown that the AD100 simulates two times faster than two AD10 processors operating in parallel and that the computing capacity of one AD100 (with FMU processor) is twice as large as that of two AD10 processors. 9 refs., 5 figs., 1 tab.

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

    SciTech Connect

    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.

  14. Passive and Active Radiation Measurements Capability at the INL Zero Power Physics Reactor (ZPPR) Facility

    SciTech Connect

    Robert Neibert; John Zabriskie; Collin Knight; James L. Jones

    2010-12-01

    The Zero Power Physics Reactor (ZPPR) facility is a Department of Energy facility located in the Idaho National Laboratory’s (INL) Materials and Fuels Complex. It contains various nuclear and non-nuclear materials that are available to support many radiation measurement assessments. User-selected, single material, nuclear and non-nuclear materials can be readily utilized with ZPPR clamshell containers with almost no criticality concerns. If custom, multi-material configurations are desired, the ZPPR clamshell or an approved aluminum Inspection Object (IO) Box container may be utilized, yet each specific material configuration will require a criticality assessment. As an example of the specialized material configurations possible, the National Nuclear Security Agency’s Office of Nuclear Verification (NNSA/NA 243) has sponsored the assembly of six material configurations. These are shown in the Appendixes and have been designated for semi-permanent storage that can be available to support various radiation measurement applications.

  15. Monte Carlo simulation of a research reactor with nominal power of 7 MW to design new control safety rods

    NASA Astrophysics Data System (ADS)

    Shoushtari, M. K.; Kakavand, T.; Sadat Kiai, S. M.; Ghaforian, H.

    2010-03-01

    The Monte Carlo simulation has been established for a research reactor with nominal power of 7 MW. A detailed model of the reactor core was employed including standard and control fuel elements, reflectors, irradiation channels, control rods, reactor pool and thermal column. The following physical parameters of reactor core were calculated for the present LEU core: core reactivity ( ρ), control rod (CR) worth, thermal and epithermal neutron flux distributions, shutdown margin and delayed neutron fraction. Reduction of unfavorable effects of blockage probability of control safety rod (CSR)s in their interiors because of not enough space in their sites, and lack of suitable capabilities to fabricate very thin plates for CSR cladding, is the main aim of the present study. Making the absorber rod thinner and CSR cladding thicker by introducing a better blackness absorbing material and a new stainless steel alloy, respectively, are two studied ways to reduce the effects of mentioned problems.

  16. An investigation of temperature measurement methods in nuclear power plant reactor pressure vessel annealing

    SciTech Connect

    Acton, R.U.; Gill, W.; Sais, D.J.; Schulze, D.H.; Nakos, J.T.

    1996-05-01

    The objective of this project was to provide an assessment of several methods by which the temperature of a commercial nuclear power plant reactor pressure vessel (RPV) could be measured during an annealing process. This project was a coordinated effort between DOE`s Office of Nuclear Energy, Science and Technology; DOE`s Light Water Reactor Technology Center at Sandia National Laboratories; and the Electric Power Research Institute`s Non- Destructive Evaluation Center. Ball- thermocouple probes similar to those described in NUREG/CR-5760, spring-loaded, metal- sheathed thermocouple probes, and 1778 air- suspended thermocouples were investigated in experiments that heated a section of an RPV wall to simulate a thermal annealing treatment. A parametric study of ball material, emissivity, thermal conductivity, and thermocouple function locations was conducted. Also investigated was a sheathed thermocouple failure mode known as shunting (electrical breakdown of insulation separating the thermocouple wires). Large errors were found between the temperature as measured by the probes and the true RPV wall temperature during heat-up and cool-down. At the annealing soak temperature, in this case 454{degrees}C [850`F], all sensors measured the same temperature within about {plus_minus}5% (23.6{degrees}C [42.5{degrees}F]). Because of these errors, actual RPV wall heating and cooling rates differed from those prescribed (by up to 29%). Shunting does not appear to be a problem under these conditions. The large temperature measurement errors led to the development of a thermal model that predicts the RPV wall temperature from the temperature of a ball- probe. Comparisons between the model and the experimental data for ball-probes indicate that the model could be a useful tool in predicting the actual RPV temperature based on the indicated ball- probe temperature. The model does not predict the temperature as well for the spring-loaded and air suspended probes.

  17. Technology Perspectives on the Management of Spent-Resin Wastes Generated From Nuclear Power Reactor Operations

    SciTech Connect

    Shiv Vijayan; Makoto Kikuchi; Akihiro Komatsu

    2002-07-01

    Organic-resin wastes (spent resins) are generated by different purification systems employed in all types of nuclear power reactors during routine and non-routine operations. The quantities of such resin wastes, and their inventories of contaminants vary depend on the operational goals of the individual power plant. Depending on the regulatory target in the particular jurisdiction where the reactor is located, the type and amounts of radionuclides, metals and other chemical contaminants in the resin waste determine the extent of treatment required for interim storage or final disposal of the waste. Resin-waste treatment comprises different operations such as pretreatment, conditioning/stabilization and containerisation that produce a waste package suitable for handling, transport, storage and disposal. One aspect of the contaminants that has significant impact on waste conditioning and the overall cost of managing such wastes are the concentrations of short half-life (arbitrarily less than approximately 30 years) radionuclides, and long half-life radionuclides, in particular carbon-14, and toxic metals present in the waste. A spectrum of resin-waste conditioning methods is available. Some methods have been applied to specific situations while others are being developed for future applications to meet the need for reducing worker dose, environmental releases, and waste-storage and disposal costs. This paper describes waste treatment options for low-level radioactive resin wastes and potential options of resin wastes containing appreciable amounts of carbon-14. Indications are that drying of the resin waste containing long half-life radionuclides such as carbon-14 and compaction or pelletizing can be favourable to allow interim dry-storage of the waste and to provide sufficient flexibility in the preparation of a suitable waste form to meet applicable waste acceptance criteria for the eventual disposal of such wastes. (authors)

  18. Neutronic design studies of a conceptual DCLL fusion reactor for a DEMO and a commercial power plant

    NASA Astrophysics Data System (ADS)

    Palermo, I.; Veredas, G.; Gómez-Ros, J. M.; Sanz, J.; Ibarra, A.

    2016-01-01

    Neutronic analyses or, more widely, nuclear analyses have been performed for the development of a dual-coolant He/LiPb (DCLL) conceptual design reactor. A detailed three-dimensional (3D) model has been examined and optimized. The design is based on the plasma parameters and functional materials of the power plant conceptual studies (PPCS) model C. The initial radial-build for the detailed model has been determined according to the dimensions established in a previous work on an equivalent simplified homogenized reactor model. For optimization purposes, the initial specifications established over the simplified model have been refined on the detailed 3D design, modifying material and dimension of breeding blanket, shield and vacuum vessel in order to fulfil the priority requirements of a fusion reactor in terms of the fundamental neutronic responses. Tritium breeding ratio, energy multiplication factor, radiation limits in the TF coils, helium production and displacements per atom (dpa) have been calculated in order to demonstrate the functionality and viability of the reactor design in guaranteeing tritium self-sufficiency, power efficiency, plasma confinement, and re-weldability and structural integrity of the components. The paper describes the neutronic design improvements of the DCLL reactor, obtaining results for both DEMO and power plant operational scenarios.

  19. Closed Brayton Cycle power system with a high temperature pellet bed reactor heat source for NEP applications

    NASA Astrophysics Data System (ADS)

    Juhasz, Albert J.; El-Genk, Mohamed S.; Harper, William B., Jr.

    1992-10-01

    Capitalizing on past and future development of high temperature gas reactor (HTGR) technology, a low mass 15 MWe closed gas turbine cycle power system using a pellet bed reactor heating helium working fluid is proposed for Nuclear Electric Propulsion (NEP) applications. Although the design of this directly coupled system architecture, comprising the reactor/power system/space radiator subsystems, is presented in conceptual form, sufficient detail is included to permit an assessment of overall system performance and mass. Furthermore, an attempt is made to show how tailoring of the main subsystem design characteristics can be utilized to achieve synergistic system level advantages that can lead to improved reliability and enhanced system life while reducing the number of parasitic load driven peripheral subsystems.

  20. Closed Brayton Cycle power system with a high temperature pellet bed reactor heat source for NEP applications

    NASA Technical Reports Server (NTRS)

    Juhasz, Albert J.; El-Genk, Mohamed S.; Harper, William B., Jr.

    1992-01-01

    Capitalizing on past and future development of high temperature gas reactor (HTGR) technology, a low mass 15 MWe closed gas turbine cycle power system using a pellet bed reactor heating helium working fluid is proposed for Nuclear Electric Propulsion (NEP) applications. Although the design of this directly coupled system architecture, comprising the reactor/power system/space radiator subsystems, is presented in conceptual form, sufficient detail is included to permit an assessment of overall system performance and mass. Furthermore, an attempt is made to show how tailoring of the main subsystem design characteristics can be utilized to achieve synergistic system level advantages that can lead to improved reliability and enhanced system life while reducing the number of parasitic load driven peripheral subsystems.

  1. Impact of the use of low or medium enriched uranium on the masses of space nuclear reactor power systems

    SciTech Connect

    1994-12-01

    The design process for determining the mass increase for the substitution of low-enriched uranium (LEU) for high-enriched uranium (HEU) in space nuclear reactor systems is an optimization process which must simultaneously consider several variables. This process becomes more complex whenever the reactor core operates on an in-core thermionic power conversion, in which the fissioning of the nuclear fuel is used to directly heat thermionic emitters, with the subsequent elimination of external power conversion equipment. The increased complexity of the optimization process for this type of system is reflected in the work reported herein, where considerably more information has been developed for the moderated in-core thermionic reactors.

  2. 10 CFR 73.55 - Requirements for physical protection of licensed activities in nuclear power reactors against...

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... power reactor licensee, licensed under 10 CFR part 50, shall implement the requirements of this section... applicants for an operating license under 10 CFR part 50, or combined license under 10 CFR part 52 who have... integration of systems, technologies, programs, equipment, supporting processes, and implementing...

  3. 10 CFR 73.55 - Requirements for physical protection of licensed activities in nuclear power reactors against...

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... power reactor licensee, licensed under 10 CFR part 50, shall implement the requirements of this section... applicants for an operating license under 10 CFR part 50, or combined license under 10 CFR part 52 who have... integration of systems, technologies, programs, equipment, supporting processes, and implementing...

  4. 10 CFR 73.55 - Requirements for physical protection of licensed activities in nuclear power reactors against...

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... power reactor licensee, licensed under 10 CFR part 50, shall implement the requirements of this section... applicants for an operating license under 10 CFR part 50, or combined license under 10 CFR part 52 who have... integration of systems, technologies, programs, equipment, supporting processes, and implementing...

  5. 10 CFR 73.55 - Requirements for physical protection of licensed activities in nuclear power reactors against...

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... power reactor licensee, licensed under 10 CFR part 50, shall implement the requirements of this section... applicants for an operating license under 10 CFR part 50, or combined license under 10 CFR part 52 who have... access into the protected area, the licensee shall: (i) Confirm the identity of individuals. (ii)...

  6. 10 CFR 73.55 - Requirements for physical protection of licensed activities in nuclear power reactors against...

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... power reactor licensee, licensed under 10 CFR part 50, shall implement the requirements of this section... applicants for an operating license under 10 CFR part 50, or combined license under 10 CFR part 52 who have... access into the protected area, the licensee shall: (i) Confirm the identity of individuals. (ii)...

  7. 10 CFR 50.64 - Limitations on the use of highly enriched uranium (HEU) in domestic non-power reactors.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... 10 Energy 1 2012-01-01 2012-01-01 false Limitations on the use of highly enriched uranium (HEU) in... Permits § 50.64 Limitations on the use of highly enriched uranium (HEU) in domestic non-power reactors. (a... applicant proposes to use highly enriched uranium (HEU) fuel, unless the applicant demonstrates that...

  8. 10 CFR 50.64 - Limitations on the use of highly enriched uranium (HEU) in domestic non-power reactors.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... 10 Energy 1 2013-01-01 2013-01-01 false Limitations on the use of highly enriched uranium (HEU) in... Permits § 50.64 Limitations on the use of highly enriched uranium (HEU) in domestic non-power reactors. (a... applicant proposes to use highly enriched uranium (HEU) fuel, unless the applicant demonstrates that...

  9. 10 CFR 50.64 - Limitations on the use of highly enriched uranium (HEU) in domestic non-power reactors.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 10 Energy 1 2011-01-01 2011-01-01 false Limitations on the use of highly enriched uranium (HEU) in... Permits § 50.64 Limitations on the use of highly enriched uranium (HEU) in domestic non-power reactors. (a... applicant proposes to use highly enriched uranium (HEU) fuel, unless the applicant demonstrates that...

  10. 10 CFR 50.64 - Limitations on the use of highly enriched uranium (HEU) in domestic non-power reactors.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... 10 Energy 1 2014-01-01 2014-01-01 false Limitations on the use of highly enriched uranium (HEU) in... Permits § 50.64 Limitations on the use of highly enriched uranium (HEU) in domestic non-power reactors. (a... applicant proposes to use highly enriched uranium (HEU) fuel, unless the applicant demonstrates that...

  11. 10 CFR 50.64 - Limitations on the use of highly enriched uranium (HEU) in domestic non-power reactors.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 10 Energy 1 2010-01-01 2010-01-01 false Limitations on the use of highly enriched uranium (HEU) in... Permits § 50.64 Limitations on the use of highly enriched uranium (HEU) in domestic non-power reactors. (a... applicant proposes to use highly enriched uranium (HEU) fuel, unless the applicant demonstrates that...

  12. Bimodal space nuclear power system with fast reactor and Topaz II-type single-cell TFE

    SciTech Connect

    Ponomarev-Stepnoi, N.N.; Usov, V.A.; Ogloblin, B.G.; Shalaev, A.I.; Klimov, A.V.; Kirillov, E.Y.; Shumov, D.P.; Radchenko, I.S.; Nicolaev, Y.V.

    1996-03-01

    The paper deals with characteristics and conceptual studies of a bimodal space thermionic system with a fast reactor and single-cell TFEs which is designed to operate in two modes: rated power mode providing power supply to space vehicle-mounted systems with energy consumption level of 10{endash}80 kW(e) and forced thermal propulsion mode with thrust of 2200 N. {copyright} {ital 1996 American Institute of Physics.}

  13. Technology, safety and costs of decommissioning a Reference Boiling Water Reactor Power Station. Main report. Volume 1

    SciTech Connect

    Oak, H.D.; Holter, G.M.; Kennedy, W.E. Jr.; Konzek, G.J.

    1980-06-01

    Technology, safety and cost information is given for the conceptual decommissioning of a large (1100MWe) boiling water reactor (BWR) power station. Three approaches to decommissioning, immediate dismantlement, safe storage with deferred dismantlement and entombment, were studied to obtain comparisons between costs, occupational radiation doses, potential dose to the public and other safety impacts. It also shows the sensitivity of decommissioning safety and costs to the power rating of a BWR in the range of 200 to 1100 MWe.

  14. Radiotoxicity and decay heat power of spent nuclear fuel of VVER type reactors at long-term storage.

    PubMed

    Bergelson, B R; Gerasimov, A S; Tikhomirov, G V

    2005-01-01

    Radiotoxicity and decay heat power of the spent nuclear fuel of VVER-1000 type reactors are calculated during storage time up to 300,000 y. Decay heat power of radioactive waste (radwaste) determines parameters of the heat removal system for the safe storage of spent nuclear fuel. Radiotoxicity determines the radiological hazard of radwaste after its leakage and penetration into the environment. PMID:16381764

  15. Self-powered denitration of landfill leachate through ammonia/nitrate coupled redox fuel cell reactor.

    PubMed

    Zhang, Huimin; Xu, Wei; Feng, Daolun; Liu, Zhanmeng; Wu, Zucheng

    2016-03-01

    In order to explore the feasibility of energy-free denitrifying N-rich wastewater, a self-powered device was uniquely assembled, in which ammonia/nitrate coupled redox fuel cell (CRFC) reactor was served as removing nitrogen and harvesting electric energy simultaneously. Ammonia is oxidized at anodic compartment and nitrate is reduced at cathodic compartment spontaneously by electrocatalysis. In 7.14 mM ammonia+0.2M KOH anolyte and 4.29 mM KNO3+0.1M H2SO4 catholyte, the nitrate removal efficiency was 46.9% after 18 h. Meanwhile, a maximum power density of 170 mW m(-2) was achieved when applying Pd/C cathode. When NH4Cl/nitrate and ammonia/nitrite CRFCs were tested, 26.2% N-NH4Cl and 91.4% N-NO2(-) were removed respectively. Nitrogen removal efficiency for real leachate at the same initial NH3-N concentration is 22.9% and nitrification of ammonia in leachate can be used as nitrate source. This work demonstrated a new way for N-rich wastewater remediation with electricity generation. PMID:26720140

  16. Self-powered denitration of landfill leachate through ammonia/nitrate coupled redox fuel cell reactor.

    PubMed

    Zhang, Huimin; Xu, Wei; Feng, Daolun; Liu, Zhanmeng; Wu, Zucheng

    2016-03-01

    In order to explore the feasibility of energy-free denitrifying N-rich wastewater, a self-powered device was uniquely assembled, in which ammonia/nitrate coupled redox fuel cell (CRFC) reactor was served as removing nitrogen and harvesting electric energy simultaneously. Ammonia is oxidized at anodic compartment and nitrate is reduced at cathodic compartment spontaneously by electrocatalysis. In 7.14 mM ammonia+0.2M KOH anolyte and 4.29 mM KNO3+0.1M H2SO4 catholyte, the nitrate removal efficiency was 46.9% after 18 h. Meanwhile, a maximum power density of 170 mW m(-2) was achieved when applying Pd/C cathode. When NH4Cl/nitrate and ammonia/nitrite CRFCs were tested, 26.2% N-NH4Cl and 91.4% N-NO2(-) were removed respectively. Nitrogen removal efficiency for real leachate at the same initial NH3-N concentration is 22.9% and nitrification of ammonia in leachate can be used as nitrate source. This work demonstrated a new way for N-rich wastewater remediation with electricity generation.

  17. Plant Modernization with Digital Reactor Protection System Safety System Upgrades at US Nuclear Power Stations

    SciTech Connect

    Heckle, Wm. Lloyd; Bolian, Tricia W.

    2006-07-01

    As the current fleet of nuclear power plants in the US reaches 25+ years of operation, obsolescence is driving many utilities to implement upgrades to both their safety and non-safety-related Instrumentation and Control (I and C) Systems. Digital technology is the predominant replacement technology for these upgrades. Within the last 15 years, digital control systems have been deployed in non-safety- related control applications at many utilities. In addition, a few utilities have replaced small safety-related systems utilizing digital technology. These systems have shown digital technology to be robust, reliable and simpler to maintain. Based upon this success, acceptance of digital technology has gained momentum with both utilities and regulatory agencies. Today, in an effort to extend the operating lives of their nuclear stations and resolve obsolescence of critical components, utilities are now pursuing digital technology for replacement of their primary safety systems. AREVA is leading this effort in the United States with the first significant digital upgrade of a major safety system. AREVA has previously completed upgrades to safety-related control systems emergency diesel engine controls and governor control systems for a hydro station which serves as the emergency power source for a nuclear station. Currently, AREVA is implementing the replacement of both the Reactor Protection System (RPS) and the Engineered Safety Features Actuation System (ESFAS) on all three units at a US PWR site. (authors)

  18. Experimental Evaluation of the Thermal Performance of a Water Shield for a Surface Power Reactor

    NASA Technical Reports Server (NTRS)

    Pearson, J. Boise; Stewart, Eric T.; Reid, Robert S.

    2007-01-01

    A water based shielding system is being investigated for use on initial lunar surface power systems. The use of water may lower overall cost (as compared to development cost for other materials) and simplify operations in the setup and handling. The thermal hydraulic performance of the shield is of significant interest. The mechanism for transferring heat through the shield is natural convection. Natural convection in a representative lunar surface reactor shield design is evaluated at various power levels in the Water Shield Testbed (WST) at the NASA Marshall Space Flight Center. The experimental data from the WST is used to anchor a CFD model. Performance of a water shield on the lunar surface is then predicted by CFD models anchored to test data. The accompanying viewgraph presentation includes the following topics: 1) Testbed Configuration; 2) Core Heater Placement and Instrumentation; 3) Thermocouple Placement; 4) Core Thermocouple Placement; 5) Outer Tank Thermocouple Placement; 6) Integrated Testbed; 7) Methodology; 8) Experimental Results: Core Temperatures; 9) Experimental Results; Outer Tank Temperatures; 10) CFD Modeling; 11) CFD Model: Anchored to Experimental Results (1-g); 12) CFD MOdel: Prediction for 1/6-g; and 13) CFD Model: Comparison of 1-g to 1/6-g.

  19. Induced Radioactivity and Waste Classification of Reactor Zone Components of the Chernobyl Nuclear Power Plant Unit 1 After Final Shutdown

    SciTech Connect

    Bylkin, Boris K.; Davydova, Galina B.; Zverkov, Yuri A.; Krayushkin, Alexander V.; Neretin, Yuri A.; Nosovsky, Anatoly V.; Seyda, Valery A.; Short, Steven M.

    2001-10-15

    The dismantlement of the reactor core materials and surrounding structural components is a major technical concern for those planning closure and decontamination and decommissioning of the Chernobyl Nuclear Power Plant (NPP). Specific issues include when and how dismantlement should be accomplished and what the radwaste classification of the dismantled system would be at the time it is disassembled. Whereas radiation levels and residual radiological characteristics of the majority of the plant systems are directly measured using standard radiation survey and radiochemical analysis techniques, actual measurements of reactor zone materials are not practical due to high radiation levels and inaccessibility. For these reasons, neutron transport analysis was used to estimate induced radioactivity and radiation levels in the Chernobyl NPP Unit 1 reactor core materials and structures.Analysis results suggest that the optimum period of safe storage is 90 to 100 yr for the Unit 1 reactor. For all of the reactor components except the fuel channel pipes (or pressure tubes), this will provide sufficient decay time to allow unlimited worker access during dismantlement, minimize the need for expensive remote dismantlement, and allow for the dismantled reactor components to be classified as low- or medium-level radioactive waste. The fuel channel pipes will remain classified as high-activity waste requiring remote dismantlement for hundreds of years due to the high concentration of induced {sup 63}Ni in the Zircaloy pipes.

  20. Defining the "proven technology" technical criterion in the reactor technology assessment for Malaysia's nuclear power program

    NASA Astrophysics Data System (ADS)

    Anuar, Nuraslinda; Kahar, Wan Shakirah Wan Abdul; Manan, Jamal Abdul Nasir Abd

    2015-04-01

    Developing countries that are considering the deployment of nuclear power plants (NPPs) in the near future need to perform reactor technology assessment (RTA) in order to select the most suitable reactor design. The International Atomic Energy Agency (IAEA) reported in the Common User Considerations (CUC) document that "proven technology" is one of the most important technical criteria for newcomer countries in performing the RTA. The qualitative description of five desired features for "proven technology" is relatively broad and only provides a general guideline to its characterization. This paper proposes a methodology to define the "proven technology" term according to a specific country's requirements using a three-stage evaluation process. The first evaluation stage screens the available technologies in the market against a predefined minimum Technology Readiness Level (TRL) derived as a condition based on national needs and policy objectives. The result is a list of technology options, which are then assessed in the second evaluation stage against quantitative definitions of CUC desired features for proven technology. The potential technology candidates produced from this evaluation is further narrowed down to obtain a list of proven technology candidates by assessing them against selected risk criteria and the established maximum allowable total score using a scoring matrix. The outcome of this methodology is the proven technology candidates selected using an accurate definition of "proven technology" that fulfills the policy objectives, national needs and risk, and country-specific CUC desired features of the country that performs this assessment. A simplified assessment for Malaysia is carried out to demonstrate and suggest the use of the proposed methodology. In this exercise, ABWR, AP1000, APR1400 and EPR designs assumed the top-ranks of proven technology candidates according to Malaysia's definition of "proven technology".

  1. Flow tests of a single fuel element coolant channel for a compact fast reactor for space power

    NASA Technical Reports Server (NTRS)

    Springborn, R. H.

    1971-01-01

    Water flow tests were conducted on a single-fuel-element cooling channel for a nuclear concept to be used for space power. The tests established a method for measuring coolant flow rate which is applicable to water flow testing of a complete mockup of the reference reactor. The inlet plenum-to-outlet plenum pressure drop, which approximates the overall core pressure drop, was measured and correlated with flow rate. This information can be used for reactor coolant flow and heat transfer calculations. An analytical study of the flow characteristics was also conducted.

  2. Pebble Bed Reactor Power Systems for Lunar Outposts: Long Operation Life and End-of Life Storage

    NASA Astrophysics Data System (ADS)

    El-Genk, Mohamed S.; Schriener, Timothy M.

    2010-09-01

    The Pellet Bed Reactor(PeBR) and power system for supporting future lunar outposts offer many desirable design, operation and safety features and address post operation storage of spent nuclear fuel. In addition to its long, full power operation life of 66 year, the PeBR is launched without fuel and loaded after placement below grade on the lunar surface with spherical fuel pellets, designed to fully contain fission products. The fuel pellets(~1.0 cm dia.) are launched separately in subcritical canisters. The post-operation PeBR is stored below grade for > 300 year to allow the radioactivity in the spent fuel to decay to a sufficiently low level. The PeBR power system, designed for avoidance of single point failures in reactor cooling and energy conversion, nominally generates ~100 kWe at a thermal efficiency of ~ 21%. In addition to the sectored reactor core, it uses three Closed Brayton Cycle loops with centrifugal flow turbo-machines for energy conversion and He-Xe(40 g/mol) binary gas mixture working fluid and reactor coolant.

  3. The Dynomak: An advanced spheromak reactor system with imposed-dynamo current drive and next-generation nuclear power technologies

    NASA Astrophysics Data System (ADS)

    Sutherland, D. A.; Jarboe, T. R.; Marklin, G.; Morgan, K. D.; Nelson, B. A.

    2013-10-01

    A high-beta spheromak reactor system has been designed with an overnight capital cost that is competitive with conventional power sources. This reactor system utilizes recently discovered imposed-dynamo current drive (IDCD) and a molten salt blanket system for first wall cooling, neutron moderation and tritium breeding. Currently available materials and ITER developed cryogenic pumping systems were implemented in this design on the basis of technological feasibility. A tritium breeding ratio of greater than 1.1 has been calculated using a Monte Carlo N-Particle (MCNP5) neutron transport simulation. High-temperature superconducting tapes (YBCO) were used for the equilibrium coil set, substantially reducing the recirculating power fraction when compared to previous spheromak reactor studies. Using zirconium hydride for neutron shielding, a limiting equilibrium coil lifetime of at least thirty full-power years has been achieved. The primary FLiBe loop was coupled to a supercritical carbon dioxide Brayton cycle due to attractive economics and high thermal efficiencies. With these advancements, an electrical output of 1000 MW from a thermal output of 2486 MW was achieved, yielding an overall plant efficiency of approximately 40%. A paper concerning the Dynomak reactor design is currently being reviewed for publication.

  4. LOS ALAMOS NEUTRON SCIENCE CENTER CONTRIBUTIONS TO THE DEVELOPMENT OF FUTURE POWER REACTORS

    SciTech Connect

    GAVRON, VICTOR I.; HILL, TONY S.; PITCHER, ERIC J.; TOVESSON, FREDERIK K.

    2007-01-09

    measurements in progress include {sup 240}Pu and {sup 242}Pu. The United States recently announced the Global Nuclear Energy Partnership (GNEP), with the goal of closing the commercial nuclear fuel cycle while minimizing proliferation risk. GNEP achieves these goals using fast-spectrum nuclear reactors powered by new transmutation fuels that contain significant quantities of minor actinides. The proposed Materials Test Station (MTS) will provide the GNEP with a cost-effective means of obtaining domestic fast-spectrum irradiations of advanced transmutation fuel forms and structural materials, which is an important step in the fuels qualification process. The MTS will be located at the LANSCE, and will be driven by a 1.08-MW proton beam. Th epeak neutron flux in the irradiation region is 1.67 x 10{sup 15} n/cm{sup 2}/s, and the energy spectrum is similar to that of a fast reactor, with the addition of a high-energy tail. The facility is expected to operate at least 4,400 hours per year. Fuel burnup rates will exceed 4% per year, and the radiation damage rate in iron will be 18 dpa (displacements per atom) per year. The construction cost is estimated to be $73M (including 25% contingency), with annual operating costs in the range of $6M to $10M. Appropriately funded, the MTS could begin operation in 2010.

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

    SciTech Connect

    Zdarek, J.; Pecinka, L.

    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.

  6. Measurement instruments for automatically monitoring the water chemistry of reactor coolant at nuclear power stations equipped with VVER reactors. Selection of measurement instruments and experience gained from their operation at Russian and foreign NPSs

    NASA Astrophysics Data System (ADS)

    Ivanov, Yu. A.

    2007-12-01

    An analytical review is given of Russian and foreign measurement instruments employed in a system for automatically monitoring the water chemistry of the reactor coolant circuit and used in the development of projects of nuclear power stations equipped with VVER-1000 reactors and the nuclear station project AES 2006. The results of experience gained from the use of such measurement instruments at nuclear power stations operating in Russia and abroad are presented.

  7. Kalkar nuclear power plant (SNR-300) - A sodium-cooled fast breeder reactor prototype

    SciTech Connect

    Morgenstern, F.H.

    1987-09-01

    The status of the Kalkar nuclear power plant in early summer 1986 is that, apart from later alterations to the workshop building, the assembly and non-nuclear commissioning work has practically been completed. From a technical point of view, nuclear commissioning of the plant can begin, but vital factors for this are the necessary nuclear licenses. The most important licensing prerequisites have been fulfilled;all essential appraisals have been available since January/February 1986. At the beginning of April 1986, the Reactor Safety Commission and the Radiation Protection Commission cast a positive vote for initial fuel loading. Before the accident in Chernobyl, but particularly since then, the issuing of the licenses has come under the political pressure of the commencing election campaign phase for the federal elections in January 1987. The initial project definition phase, the organizational boundary conditions, and the major requirements for the construction of the plant are summarized in chronological form. To provide the total picture, references dealing with general and technical aspects of the project are listed.

  8. Independent Safety Assessment of the TOPAZ-II space nuclear reactor power system (Revised)

    SciTech Connect

    1993-09-01

    The Independent Safety Assessment described in this study report was performed to assess the safety of the design and launch plans anticipated by the U.S. Department of Defense (DOD) in 1993 for a Russian-built, U.S.-modified, TOPAZ-II space nuclear reactor power system. Its conclusions, and the bases for them, were intended to provide guidance for the U.S. Department of Energy (DOE) management in the event that the DOD requested authorization under section 91b. of the Atomic Energy Act of 1954, as amended, for possession and use (including ground testing and launch) of a nuclear-fueled, modified TOPAZ-II. The scientists and engineers who were engaged to perform this assessment are nationally-known nuclear safety experts in various disciplines. They met with participants in the TOPAZ-II program during the spring and summer of 1993 and produced a report based on their analysis of the proposed TOPAZ-II mission. Their conclusions were confined to the potential impact on public safety and did not include budgetary, reliability, or risk-benefit analyses.

  9. Planetary Surface Power and Interstellar Propulsion Using Fission Fragment Magnetic Collimator Reactor

    SciTech Connect

    Tsvetkov, Pavel V.; Hart, Ron R.; King, Don B.; Rochau, Gary E.

    2006-01-20

    Fission energy can be used directly if the kinetic energy of fission fragments is converted to electricity and/or thrust before turning into heat. The completed US DOE NERI Direct Energy Conversion (DEC) Power Production project indicates that viable DEC systems are possible. The US DOE NERI DEC Proof of Principle project began in October of 2002 with the goal to demonstrate performance principles of DEC systems. One of the emerging DEC concepts is represented by fission fragment magnetic collimator reactors (FFMCR). Safety, simplicity, and high conversion efficiency are the unique advantages offered by these systems. In the FFMCR, the basic energy source is the kinetic energy of fission fragments. Following escape from thin fuel layers, they are captured on magnetic field lines and are directed out of the core and through magnetic collimators to produce electricity and thrust. The exiting flow of energetic fission fragments has a very high specific impulse that allows efficient planetary surface power and interstellar propulsion without carrying any conventional propellant onboard. The objective of this work was to determine technological feasibility of the concept. This objective was accomplished by producing the FFMCR design and by analysis of its performance characteristics. The paper presents the FFMCR concept, describes its development to a technologically feasible level and discusses obtained results. Performed studies offer efficiencies up to 90% and velocities approaching speed of light as potentially achievable. The unmanned 10-tons probe with 1000 MW FFMCR propulsion unit would attain mission velocity of about 2% of the speed of light. If the unit is designed for 4000 MW, then in 10 years the unmanned 10-tons probe would attain mission velocity of about 10% of the speed of light.

  10. Occupational radiation exposure at commercial nuclear power reactors and other facilities 1994. Twenty-seventh annual report

    SciTech Connect

    Thomas, M.L.; Hagemeyer, D.

    1996-01-01

    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). Annual reports for 1994 were received from a total of 303 NRC licensees, of which 109 were operators of nuclear power reactors in commercial operation. Compilations of the reports submitted by the 303 licensees indicated that 152,028 individuals were monitored, 79,780 of whom received a measurable dose. The collective dose incurred by these individuals was 24,740 person-cSv (person-rem){sup 2} which represents a 15% decrease from the 1993 value. The number of workers receiving a measurable dose also decreased, resulting in the average measurable dose of 0.31 cSv (rem) for 1994. The average measurable dose is defined to be the total collective dose (TEDE) divided by the number of workers receiving a measurable dose. These figures have been adjusted to account for transient reactor workers. In 1994, the annual collective dose per reactor for light water reactor licensees (LWRs) was 198 person-cSv (person-rem). This represents a 18% decrease from the 1993 value of 242 person-cSv (person-rem). The annual collective dose per reactor for boiling water reactors (BWRs) was 327 person-cSv (person-rem) and, for pressurized water reactors (PWRs), it was 131 person-cSv (person-rem). Analyses of transient worker data indicate that 18,178 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 workers by multiple licensees. In 1994, the average measurable dose calculated from reported data was 0.28 cSv (rem). The corrected dose distribution resulted in an average measurable dose of 0.31 cSv (rem).

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

    SciTech Connect

    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.

  12. Long Term Storage with Surveillance of Canadian Prototype Nuclear Power Reactors

    SciTech Connect

    Janzen, Rick

    2008-01-15

    Atomic Energy of Canada (AECL) was originally formed by the government of Canada in 1952 to perform research in radiation and nuclear areas. In the mid 1950's Canada decided to limit itself to peaceful uses of nuclear energy and AECL embarked on several research and development programs, one of them being the development of nuclear power plants. This led to the development of the CANDU{sup TM} design of heavy water power reactors, of which there are now 29 operating around the world. This presentation discusses the present state of the first two CANDU{sup TM} prototype reactors and a prototype boiling light water reactor and lessons learnt after being in a long-term storage with surveillance state for more than 20 years. AECL facilities undergo decommissioning by either a prompt or a deferred removal approach. Both approaches are initiated after an operating facility has been declared redundant and gone through final operational shutdown. For the deferred approach, initial decommissioning activities are performed to put the facility into a sustainable, safe, shutdown state to minimize the hazards and costs of the ensuing extended storage with surveillance (SWS) or Safestor phase. At the appropriate time, the facility is dismantled and removed, or put into a suitable condition for re-use. AECL has a number of facilities that were built during its history, and some of these are now redundant or will become redundant in the near future. The deferred removal approach is part of AECL's decommissioning strategy for several reasons: 1. Reduction in radiation doses to workers during the final dismantling, 2. No facilities are available yet in Canada for the management of quantity of wastes arising from decommissioning, 3. Financial constraints presented by the number of facilities that will undergo decommissioning, compared to the availability of funds to carry out the work. This has led to the development of a comprehensive decommissioning plan that includes all of AECL

  13. CONVECTION REACTOR

    DOEpatents

    Hammond, R.P.; King, L.D.P.

    1960-03-22

    An homogeneous nuclear power reactor utilizing convection circulation of the liquid fuel is proposed. The reactor has an internal heat exchanger looated in the same pressure vessel as the critical assembly, thereby eliminating necessity for handling the hot liquid fuel outside the reactor pressure vessel during normal operation. The liquid fuel used in this reactor eliminates the necessity for extensive radiolytic gas rocombination apparatus, and the reactor is resiliently pressurized and, without any movable mechanical apparatus, automatically regulates itself to the condition of criticality during moderate variations in temperature snd pressure and shuts itself down as the pressure exceeds a predetermined safe operating value.

  14. Technology, safety and costs of decommissioning a reference boiling water reactor power station. Classification of decommissioning wastes. Addendum 2

    SciTech Connect

    Murphy, E.S.

    1984-09-01

    The radioactive wastes expected to result from decommissioning of the reference boiling water reactor power station are reviewed and classified in accordance with 10 CFR 61. The 18,949 cubic meters of waste from DECON are classified as follows: Class A, 97.5%; Class B, 2.0%; Class C, 0.3%. About 0.2% (47 cubic meters) of the waste would be generally unacceptable for disposal using near-surface disposal methods.

  15. Design of a proteus lattice representative of a burnt and fresh fuel interface at power conditions in light water reactors

    SciTech Connect

    Hursin, M.; Perret, G.

    2012-07-01

    The research program LIFE (Large-scale Irradiated Fuel Experiment) between PSI and Swissnuclear has been started in 2006 to study the interaction between large sets of burnt and fresh fuel pins in conditions representative of power light water reactors. Reactor physics parameters such as flux ratios and reaction rate distributions ({sup 235}U and {sup 238}U fissions and {sup 238}U capture) are calculated to estimate an appropriate arrangement of burnt and fresh fuel pins within the central element of the test zone of the zero-power research reactor PROTEUS. The arrangement should minimize the number of burnt fuel pins to ease fuel handling and reduce costs, whilst guaranteeing that the neutron spectrum in both burnt and fresh fuel regions and at their interface is representative of a large uniform array of burnt and fresh pins in the same moderation conditions. First results are encouraging, showing that the burnt/fresh fuel interface is well represented with a 6 x 6 bundle of burnt pins. The second part of the project involves the use of TSUNAMI, CASMO-4E and DAKOTA to perform parametric and optimization studies on the PROTEUS lattice by varying its pitch (P) and fraction of D{sub 2}O in moderator (F{sub D2O}) to be as representative as possible of a power light water reactor core at hot full power conditions at beginning of cycle (BOC). The parameters P and F{sub D2O} that best represent a PWR at BOC are 1.36 cm and 5% respectively. (authors)

  16. Technology, safety and costs of decommissioning a reference pressurized water reactor power station. Classification of decommissioning wastes. Addendum 3

    SciTech Connect

    Murphy, E.S.

    1984-09-01

    The radioactive wastes expected to result from decommissioning of the reference pressurized water reactor power station are reviewed and classified in accordance with 10 CFR 61. The 17,885 cubic meters of waste from DECON are classified as follows: Class A, 98.0%; Class B, 1.2%; Class C, 0.1%. About 0.7% (133 cubic meters) of the waste would be generally unacceptable for disposal using near-surface disposal methods.

  17. Study of coolant activation and dose rates with flow rate and power perturbations in pool-type research reactors

    SciTech Connect

    Mirza, N.M.; Mirza, S.M.; Ahmad, N. )

    1991-12-01

    This paper reports on a computer code using the multigroup diffusion theory based LEOPARD and ODMUG programs that has been developed to calculate the activity in the coolant leaving the core of a pool-type research reactor. Using this code, the dose rates at various locations along the coolant path with varying coolant flow rate and reactor power perturbations are determined. A flow rate decrease from 1000 to 145 m{sup 3}/h is considered. The results indicate that a flow rate decrease leads to an increase in the coolant outlet temperature, which affects the neutron group constants and hence the group fluxes. The activity in the coolant leaving the core increases with flow rate decrease. However, at the inlet of the holdup tank, the total dose rate first increases, then passes through a maximum at {approximately} 500 m{sup 3}/h, and finally decreases with flow rate decrease. The activity at the outlet of the holdup tank is mainly due to {sup 24}Na and {sup 56}Mn, and it increases by {approximately} 2% when the flow rate decreases from 1000 to 145 m{sup 3}/h. In an accidental power rise at constant flow rate, the activity in the coolant increases, and the dose rates at all the points along the coolant path show a slight nonlinear rise as the reactor power density increases.

  18. EVALUATION OF ZERO-POWER, ELEVATED-TEMPERATURE MEASUREMENTS AT JAPAN’S HIGH TEMPERATURE ENGINEERING TEST REACTOR

    SciTech Connect

    John D. Bess; Nozomu Fujimoto; James W. Sterbentz; Luka Snoj; Atsushi Zukeran

    2011-03-01

    The High Temperature Engineering Test Reactor (HTTR) of the Japan Atomic Energy Agency (JAEA) is a 30 MWth, graphite-moderated, helium-cooled reactor that was constructed with the objectives to establish and upgrade the technological basis for advanced high-temperature gas-cooled reactors (HTGRs) as well as to conduct various irradiation tests for innovative high-temperature research. The core size of the HTTR represents about one-half of that of future HTGRs, and the high excess reactivity of the HTTR, necessary for compensation of temperature, xenon, and burnup effects during power operations, is similar to that of future HTGRs. During the start-up core physics tests of the HTTR, various annular cores were formed to provide experimental data for verification of design codes for future HTGRs. The experimental benchmark performed and currently evaluated in this report pertains to the data available for two zero-power, warm-critical measurements with the fully-loaded HTTR core. Six isothermal temperature coefficients for the fully-loaded core from approximately 340 to 740 K have also been evaluated. These experiments were performed as part of the power-up tests (References 1 and 2). Evaluation of the start-up core physics tests specific to the fully-loaded core (HTTR-GCR-RESR-001) and annular start-up core loadings (HTTR-GCR-RESR-002) have been previously evaluated.

  19. An experimental study of catalytic and non-catalytic reaction in heat recirculating reactors and applications to power generation

    NASA Astrophysics Data System (ADS)

    Ahn, Jeongmin

    An experimental study of the performance of a Swiss roll heat exchanger and reactor was conducted, with emphasis on the extinction limits and comparison of results with and without Pt catalyst. At Re<40, the catalyst was required to sustain reaction; with the catalyst self-sustaining reaction could be obtained at Re less than 1. Both lean and rich extinction limits were extended with the catalyst, though rich limits were extended much further. At low Re, the lean extinction limit was rich of stoichiometric and rich limit had equivalence ratios 80 in some cases. Non-catalytic reaction generally occurred in a flameless mode near the center of the reactor. With or without catalyst, for sufficiently robust conditions, a visible flame would propagate out of the center, but this flame could only be re-centered with catalyst. Gas chromatography indicated that at low Re, CO and non-C3 H8 hydrocarbons did not form. For higher Re, catalytic limits were slightly broader but had much lower limit temperatures. At sufficiently high Re, catalytic and gas-phase limits merged. Experiments with titanium Swiss rolls have demonstrated reducing wall thermal conductivity and thickness leads to lower heat losses and therefore increases operating temperatures and extends flammability limits. By use of Pt catalysts, reaction of propane-air mixtures at temperatures 54°C was sustained. Such low temperatures suggest that polymers may be employed as a reactor material. A polyimide reactor was built and survived prolonged testing at temperatures up to 500°C. Polymer reactors may prove more practical for microscale devices due to their lower thermal conductivity and ease of manufacturing. Since the ultimate goal of current efforts is to develop combustion driven power generation devices at MEMS like scales, a thermally self-sustaining miniature power generation device was developed utilizing a single-chamber solid-oxide-fuel-cell (SOFC) placed in a Swiss roll. With the single-chamber design

  20. Modeling of Coolant Flow in the Fuel Assembly of the Reactor of a Floating Nuclear Power Plant Using the Logos CFD Program

    NASA Astrophysics Data System (ADS)

    Dmitriev, S. M.; Dobrov, A. A.; Legchanov, M. A.; Khrobostov, A. E.

    2015-09-01

    Results of computer modeling of coolant flow in the fuel assembly of the reactor of a floating nuclear power plant using the LOGOS CFD programs have been given. The possibility of using the obtained results to improve models built into the engineering programs of thermohydraulic calculation of nuclear-reactor cores has been considered.

  1. Space nuclear power system based on thermionic reactor with single-cell TFEs and zirconium hydride moderator

    SciTech Connect

    Ponomarev-Stepnoi, N.N.; Usov, V.A. ); Nickitin, V.P.; Ogloblin, B.G.; Lutov, J.I.; Luppov, A.N.; Gabrusev, V.N.; Klimov, A.V. ); Nicolaev, J.V.; Kucherov, R.J.; Eremin, S.A. )

    1993-01-15

    The results of research and development work on the space nuclear power system with net electric power of 40 kW performed with cooperation of Russian research and engineering organizations in 1991--92 are presented in this paper. Development work was carried out on the basis of experience gained in the course of Topaz-2 SNPS creation. As the basis for SNPS the 127-channel intermediate reactor is used with in-core single-cell TFEs having improved parameters. The SNPS parameters, distinctive features of decision on design and layout, safety ensuring concepts, proposals on the future stages of SNPS development are also considered in the paper.

  2. Radiator Heat Pipes with Carbon-Carbon Fins and Armor for Space Nuclear Reactor Power Systems

    NASA Astrophysics Data System (ADS)

    Tournier, Jean-Michel; El-Genk, Mohamed

    2005-02-01

    Technologies for Space Reactor Power Systems are being developed to enable future NASA's missions early next decade to explore the farthest planets in the solar system. The choices of the energy conversion technology for these power systems require radiator temperatures that span a wide range, from 350 K to 800 K. Heat pipes with carbon-carbon fins and armor are the preferred choice for these radiators because of inherent redundancy and efficient spreading and rejection of waste heat into space at a relatively small mass penalty. The performance results and specific masses of radiator heat pipes with cesium, rubidium, and potassium working fluids are presented and compared in this paper. The heat pipes operate at 40% of the prevailing operation limit (a design margin of 60%), typically the sonic and/or capillary limit. The thickness of the carbon-carbon fins is 0.5 mm but the width is varied, and the evaporator and condenser sections are 0.15 and 1.35 m long, respectively. The 400-mesh wick and the heat pipe thin metal wall are titanium, and the carbon-carbon armor (~ 2 mm-thick) provides both structural strength and protection against meteoroids impacts. The cross-section area of the D-shaped radiator heat pipes is optimized for minimum mass. Because of the low vapor pressure of potassium and its very high Figure-Of-Merit (FOM), radiator potassium heat pipes are the best performers at temperatures above 800 K, where the sonic limit is no longer an issue. On the other hand, rubidium heat pipes are limited by the sonic limit below 762 K and by the capillary limit at higher temperature. The transition temperature between these two limits for the cesium heat pipes occurs at a lower temperature of 724 K, since cesium has lower FOM than rubidium. The present results show that with a design margin of 60%, the cesium heat pipes radiator is best at 680-720 K, the rubidium heat pipes radiator is best at 720-800 K, while the potassium heat pipes radiator is the best

  3. Radiator Heat Pipes with Carbon-Carbon Fins and Armor for Space Nuclear Reactor Power Systems

    SciTech Connect

    Tournier, Jean-Michel; El-Genk, Mohamed

    2005-02-06

    Technologies for Space Reactor Power Systems are being developed to enable future NASA's missions early next decade to explore the farthest planets in the solar system. The choices of the energy conversion technology for these power systems require radiator temperatures that span a wide range, from 350 K to 800 K. Heat pipes with carbon-carbon fins and armor are the preferred choice for these radiators because of inherent redundancy and efficient spreading and rejection of waste heat into space at a relatively small mass penalty. The performance results and specific masses of radiator heat pipes with cesium, rubidium, and potassium working fluids are presented and compared in this paper. The heat pipes operate at 40% of the prevailing operation limit (a design margin of 60%), typically the sonic and/or capillary limit. The thickness of the carbon-carbon fins is 0.5 mm but the width is varied, and the evaporator and condenser sections are 0.15 and 1.35 m long, respectively. The 400-mesh wick and the heat pipe thin metal wall are titanium, and the carbon-carbon armor ({approx} 2 mm-thick) provides both structural strength and protection against meteoroids impacts. The cross-section area of the D-shaped radiator heat pipes is optimized for minimum mass. Because of the low vapor pressure of potassium and its very high Figure-Of-Merit (FOM), radiator potassium heat pipes are the best performers at temperatures above 800 K, where the sonic limit is no longer an issue. On the other hand, rubidium heat pipes are limited by the sonic limit below 762 K and by the capillary limit at higher temperature. The transition temperature between these two limits for the cesium heat pipes occurs at a lower temperature of 724 K, since cesium has lower FOM than rubidium. The present results show that with a design margin of 60%, the cesium heat pipes radiator is best at 680-720 K, the rubidium heat pipes radiator is best at 720-800 K, while the potassium heat pipes radiator is the best

  4. Technical Bases to Aid in the Decision of Conducting Full Power Ground Nuclear Tests for Space Fission Reactors

    NASA Astrophysics Data System (ADS)

    Hixson, Laurie L.; Houts, Michael G.; Clement, Steven D.

    2004-02-01

    The extent to which, if any, full power ground nuclear testing of space reactors should be performed has been a point of discussion within the industry for decades. Do the benefits outweigh the risks? Are there equivalent alternatives? Can a test facility be constructed (or modified) in a reasonable amount of time? Is the test article an accurate representation of the flight system? Are the costs too restrictive? The obvious benefits of full power ground nuclear testing; obtaining systems integrated reliability data on a full-scale, complete end-to-end system; come at some programmatic risk. Safety related information is not obtained from a full-power ground nuclear test. This paper will discuss and assess these and other technical considerations essential in the decision to conduct full power ground nuclear-or alternative-tests.

  5. The use of supercritical parameters of a coolant—A promising path to development of nuclear power plant water-cooled reactors in the 21st century

    NASA Astrophysics Data System (ADS)

    Gabaraev, B. A.; Smolin, V. N.; Solov'ev, S. L.

    2006-09-01

    A modern concept of the development of water-cooled reactors of nuclear power plants (NPPs) is considered. Data on the design of NPPs with supercritical-parameters coolant and the results of experimental studies are presented.

  6. Megawatt Class Nuclear Space Power Systems (MCNSPS) conceptual design and evaluation report. Volume 2, technologies 1: Reactors, heat transport, integration issues

    NASA Technical Reports Server (NTRS)

    Wetch, J. R.

    1988-01-01

    The objectives of the Megawatt Class Nuclear Space Power System (MCNSPS) study are summarized and candidate systems and subsystems are described. Particular emphasis is given to the heat rejection system and the space reactor subsystem.

  7. Multigroup calculation of criticality and power distribution in a two-pass fast spectrum cermet-fueled reactor

    SciTech Connect

    Anghaie, S.; Feller, G.J. ); Peery, S.D.; Parsley, R.C. )

    1992-01-01

    The advanced propulsion group at Pratt Whitney has developed a nuclear thermal rocket concept, the XNR2000, for use on lunear, Mars, and deep-space planetary missions. The XNR2000 engine is powered by a fast spectrum cermet-fueled nuclear reactor that heats up hydrogen propellant to a maximum of 2850 K. An expander cycle is used to deliver 12 kg/s hydrogen to the core, producing 25,000 lb[sub f] thrust at 944 s of specific impulse. The reactor comprises a beryllium-reflected outer annulus core and an inner core with the hydrogen propellant entering from the bottom of the outer core and exiting from the bottom part of the inner core to the thrust chamber. Both the outer and inner cores are loaded with prismatic cermet fuel elements. The baseline XNR2000 reactor core consists of 90 fuel elements in the outer core and 61 in the inner core, arranged in the pattern. This paper focuses on the neutronic analysis of the baseline XNR2000 reactor.

  8. A Computational Fluid Dynamic and Heat Transfer Model for Gaseous Core and Gas Cooled Space Power and Propulsion Reactors

    NASA Technical Reports Server (NTRS)

    Anghaie, S.; Chen, G.

    1996-01-01

    A computational model based on the axisymmetric, thin-layer Navier-Stokes equations is developed to predict the convective, radiation and conductive heat transfer in high temperature space nuclear reactors. An implicit-explicit, finite volume, MacCormack method in conjunction with the Gauss-Seidel line iteration procedure is utilized to solve the thermal and fluid governing equations. Simulation of coolant and propellant flows in these reactors involves the subsonic and supersonic flows of hydrogen, helium and uranium tetrafluoride under variable boundary conditions. An enthalpy-rebalancing scheme is developed and implemented to enhance and accelerate the rate of convergence when a wall heat flux boundary condition is used. The model also incorporated the Baldwin and Lomax two-layer algebraic turbulence scheme for the calculation of the turbulent kinetic energy and eddy diffusivity of energy. The Rosseland diffusion approximation is used to simulate the radiative energy transfer in the optically thick environment of gas core reactors. The computational model is benchmarked with experimental data on flow separation angle and drag force acting on a suspended sphere in a cylindrical tube. The heat transfer is validated by comparing the computed results with the standard heat transfer correlations predictions. The model is used to simulate flow and heat transfer under a variety of design conditions. The effect of internal heat generation on the heat transfer in the gas core reactors is examined for a variety of power densities, 100 W/cc, 500 W/cc and 1000 W/cc. The maximum temperature, corresponding with the heat generation rates, are 2150 K, 2750 K and 3550 K, respectively. This analysis shows that the maximum temperature is strongly dependent on the value of heat generation rate. It also indicates that a heat generation rate higher than 1000 W/cc is necessary to maintain the gas temperature at about 3500 K, which is typical design temperature required to achieve high

  9. Feasibility Study of Supercritical Light Water Cooled Reactors for Electric Power Production

    SciTech Connect

    Philip MacDonald; Jacopo Buongiorno; James Sterbentz; Cliff Davis; Robert Witt; Gary Was; J. McKinley; S. Teysseyre; Luca Oriani; Vefa Kucukboyaci; Lawrence Conway; N. Jonsson: Bin Liu

    2005-02-13

    The supercritical water reactor (SCWR) has been the object of interest throughout the nuclear Generation IV community because of its high potential: a simple, direct cycle, compact configuration; elimination of many traditional LWR components, operation at coolant temperatures much higher than traditional LWRs and thus high thermal efficiency. It could be said that the SWR was viewed as the water counterpart to the high temperature gas reactor.

  10. An analysis of thermionic space nuclear reactor power system: II. Merits of using safety drums for backup control

    NASA Astrophysics Data System (ADS)

    El-Genk, Mahomed S.; Xue, Huimin

    1993-01-01

    An analysis is performed to investigate the merits of using the TOPAZ-II safety drums for a backup control to prevent a reactivity excursion, stabilize the reactor, and achieve steady-state power operation, following a severe hypothetical reactivity initiated accident (RIA). Such an RIA is assumed to occur during the system start-up in orbit due to a malfunction of the drive mechanism of the control drums, causing the nine drums to accidentally rotate the full 180° outward. Results show that an immediate, inward rotation of the three safety drums to an angle of 80° will shutdown the reactor, however, a delay time of 10 s will not only prevents a reactivity excursion, but also enables operating the reactor at a steady-state thermal power of about 33.3 kW (0.9 kW per TFE). Conversely, when the immediate rotation of the safety drums is to a larger angle of 100°, a steady-state operation at about 37 kW can be achieved, but a delay of 10 s causes a reactivity excursion and overheating of the TFEs. It is therefor concluded that, should the drive mechanism be modified to enable rotating the safety drums for TOPAZ-II reactor at variable speeds of and below 22.5°/s, the three safety drums could be used successfully for a backup control, following an RIA. However, since the reactivity worth of the three safety drums is only 2.0, the maximum steady-state electric power achievable for the system is limited to approximately 0.25 kW, at which the fission power is about 37 kW and the emitter temperature is approximaely 1500 K. To alleviate such a limitation and enable operation at nominal design conditions (fission power of about 107 kW or a system's total electric power of 5.6 kW), the reactivity worth of the safety drums would have to be increased by at least 0.24. An additional increase in the safety drums' worth will also be necessary to maintain steady-state operation of the system at nominal conditions throughout the mission lifetime, with all nine control drums fully

  11. An analysis of thermionic space nuclear reactor power system: II. Merits of using safety drums for backup control

    SciTech Connect

    El-Genk, M.S.; Huimin Xue )

    1993-01-10

    An analysis is performed to investigate the merits of using the TOPAZ-II safety drums for a backup control to prevent a reactivity excursion, stabilize the reactor, and achieve steady-state power operation, following a severe hypothetical reactivity initiated accident (RIA). Such an RIA is assumed to occur during the system start-up in orbit due to a malfunction of the drive mechanism of the control drums, causing the nine drums to accidentally rotate the full 180[degree] outward. Results show that an immediate, inward rotation of the three safety drums to an angle of 80[degree] will shutdown the reactor, however, a delay time of 10 s will not only prevents a reactivity excursion, but also enables operating the reactor at a steady-state thermal power of about 33.3 kW (0.9 kW per TFE). Conversely, when the immediate rotation of the safety drums is to a larger angle of 100[degree], a steady-state operation at about 37 kW can be achieved, but a delay of 10 s causes a reactivity excursion and overheating of the TFEs. It is therefor concluded that, should the drive mechanism be modified to enable rotating the safety drums for TOPAZ-II reactor at variable speeds of and below 22.5[degree]/s, the three safety drums could be used successfully for a backup control, following an RIA. However, since the reactivity worth of the three safety drums is only $2.0, the maximum steady-state electric power achievable for the system is limited to approximately 0.25 kW, at which the fission power is about 37 kW and the emitter temperature is approximaely 1500 K. To alleviate such a limitation and enable operation at nominal design conditions (fission power of about 107 kW or a system's total electric power of 5.6 kW), the reactivity worth of the safety drums would have to be increased by at least $0.24.

  12. Determination of the relative power density distribution in a heterogeneous reactor from the results of measurements of the reactivity effects and the neutron importance function

    NASA Astrophysics Data System (ADS)

    Bobrov, A. A.; Glushkov, E. S.; Zimin, A. A.; Kapitonova, A. V.; Kompaniets, G. V.; Nosov, V. I.; Petrushenko, R. P.; Smirnov, O. N.

    2012-12-01

    A method for experimental determination of the relative power density distribution in a heterogeneous reactor based on measurements of fuel reactivity effects and importance of neutrons from a californium source is proposed. The method was perfected on two critical assembly configurations at the NARCISS facility of the Kurchatov Institute, which simulated a small-size heterogeneous nuclear reactor. The neutron importance measurements were performed on subcritical and critical assemblies. It is shown that, along with traditionally used activation methods, the developed method can be applied to experimental studies of special features of the power density distribution in critical assemblies and reactors.

  13. Determination of the relative power density distribution in a heterogeneous reactor from the results of measurements of the reactivity effects and the neutron importance function

    SciTech Connect

    Bobrov, A. A.; Glushkov, E. S.; Zimin, A. A.; Kapitonova, A. V.; Kompaniets, G. V.; Nosov, V. I. Petrushenko, R. P.; Smirnov, O. N.

    2012-12-15

    A method for experimental determination of the relative power density distribution in a heterogeneous reactor based on measurements of fuel reactivity effects and importance of neutrons from a californium source is proposed. The method was perfected on two critical assembly configurations at the NARCISS facility of the Kurchatov Institute, which simulated a small-size heterogeneous nuclear reactor. The neutron importance measurements were performed on subcritical and critical assemblies. It is shown that, along with traditionally used activation methods, the developed method can be applied to experimental studies of special features of the power density distribution in critical assemblies and reactors.

  14. Calculated performance of a mercury-compressor-jet powered airplane using a nuclear reactor as an energy source

    NASA Technical Reports Server (NTRS)

    Doyle, R B

    1951-01-01

    An analysis was made at a flight Mach number of 1.5, an altitude of 45,000 feet, a turbine-inlet temperature of 1460 degrees R, of a mercury compressor-jet powered airplane using a nuclear reactor as an energy source. The calculations covered a range of turbine-exhaust and turbine-inlet pressures and condenser-inlet Mach numbers. For a turbine--inlet pressure of 40 pounds per square inch absolute, a turbine-exhaust pressure of 14 pounds per square inch absolute, and a condenser-inlet Mach number of 0.23 the calculated airplane gross weight required to carry a 20,000 pound payload was 322000 pounds and the reactor heat release per unit volume was 8.9 kilowatts per cubic inch. These do not represent optimum operating conditions.

  15. In-vessel Retention Strategy for High Power Reactors - K-INERI Final Report (includes SBLB Test Results for Task 3 on External Reactor Vessel Cooling (ERVC) Boiling Data and CHF Enhancement Correlations)

    SciTech Connect

    F. B. Cheung; J. Yang; M. B. Dizon; J. Rempe

    2005-01-01

    In-vessel retention (IVR) of core melt is a key severe accident management strategy adopted by some operating nuclear power plants and proposed for some advanced light water reactors (ALWRs). If there were inadequate cooling during a reactor accident, a significant amount of core material could become molten and relocate to the lower head of the reactor vessel, as happened in the Three Mile Island Unit 2 (TMI-2) accident. If it is possible to ensure that the vessel head remains intact so that relocated core materials are retained within the vessel, the enhanced safety associated with these plants can reduce concerns about containment failure and associated risk. For example, the enhanced safety of the Westinghouse Advanced 600 MWe PWR (AP600), which relied upon External Reactor Vessel Cooling (ERVC) for IVR, resulted in the U.S. Nuclear Regulatory Commission (US NRC) approving the design without requiring certain conventional features common to existing LWRs. However, it is not clear that currently proposed external reactor vessel cooling (ERVC) without additional enhancements could provide sufficient heat removal for higher-power reactors (up to 1500 MWe). Hence, a collaborative, three-year, U.S. - Korean International Nuclear Energy Research Initiative (INERI) project was completed in which the Idaho National Engineering and Environmental Laboratory (INEEL), Seoul National University (SNU), Pennsylvania State University (PSU), and the Korea Atomic Energy Research Institute (KAERI) investigated the performance of ERVC and an in-vessel core catcher (IVCC) to determine if IVR is feasible for reactors up to 1500 MWe.

  16. Performance Analysis of Potassium Heat Pipes Radiator for HP-STMCs Space Reactor Power System

    SciTech Connect

    El-Genk, Mohamed S.; Tournier, Jean-Michel

    2004-02-04

    A detailed design and performance results of C-C finned, and armored potassium heat pipes radiator for a 110 kWe Heat Pipes-Segmented Thermoelectric Module Converters (HP-STMCs) Space Reactor Power system (SRPS) are presented. The radiator consists of two sections; each serves an equal number of STMCs and has 162 longitudinal potassium heat pipes with 0.508 mm thick C-C fins. The width of the C-C fins at the minor diameter of the radiator is almost zero, but increases with distance along the radiator to reach 3.7 cm at the radiator's major diameter. The radiator's heat pipes (OD = 2.42 cm in front and 3.03 cm in rear) have thin titanium (0.0762 mm thick) liners and wicks (0.20 mm thick with an effective pore radius of 12-16 {mu}m) and a 1.016 mm thick C-C wall. The wick is separated from the titanium liner by a 0.4 mm annulus filled with liquid potassium to increase the capillary limit. The outer surfaces of the heat pipes in the front and rear sections of the radiator are protected with a C-C armor that is 2.17 mm and 1.70 mm thick, respectively. The inside surface of the heat pipes in the front radiator is thermally insulated while the C-C finned condensers of the rear heat pipes are exposed, radiating into space through the rear opening of the radiator cavity. The heat pipes in both the front and the rear radiators have a 1.5 m long evaporator section and each dissipates 4.47 kW while operating at 43.6% of the prevailing sonic limit. The front and rear radiator sections are 5.29 m and 2.61 m long with outer surface area and mass of 47.1 m2 and 314.3 kg, and 39.9 m2 and 243.2 kg, respectively. The total radiator is 7.63 m long and has minor and major diameters of 1.48 m and 5.57 m, respectively, and a total surface area of 87 m2; however, the effective radiator area, after accounting for heat rejection through the rear of the radiator cavity, is 98.8 m2. The radiator's total mass including the C-C armor is 557.5 kg and the specific area and specific mass are 6

  17. Performance Analysis of Potassium Heat Pipes Radiator for HP-STMCs Space Reactor Power System

    NASA Astrophysics Data System (ADS)

    El-Genk, Mohamed S.; Tournier, Jean-Michel

    2004-02-01

    A detailed design and performance results of C-C finned, and armored potassium heat pipes radiator for a 110 kWe Heat Pipes-Segmented Thermoelectric Module Converters (HP-STMCs) Space Reactor Power system (SRPS) are presented. The radiator consists of two sections; each serves an equal number of STMCs and has 162 longitudinal potassium heat pipes with 0.508 mm thick C-C fins. The width of the C-C fins at the minor diameter of the radiator is almost zero, but increases with distance along the radiator to reach 3.7 cm at the radiator's major diameter. The radiator's heat pipes (OD = 2.42 cm in front and 3.03 cm in rear) have thin titanium (0.0762 mm thick) liners and wicks (0.20 mm thick with an effective pore radius of 12-16 μm) and a 1.016 mm thick C-C wall. The wick is separated from the titanium liner by a 0.4 mm annulus filled with liquid potassium to increase the capillary limit. The outer surfaces of the heat pipes in the front and rear sections of the radiator are protected with a C-C armor that is 2.17 mm and 1.70 mm thick, respectively. The inside surface of the heat pipes in the front radiator is thermally insulated while the C-C finned condensers of the rear heat pipes are exposed, radiating into space through the rear opening of the radiator cavity. The heat pipes in both the front and the rear radiators have a 1.5 m long evaporator section and each dissipates 4.47 kW while operating at 43.6% of the prevailing sonic limit. The front and rear radiator sections are 5.29 m and 2.61 m long with outer surface area and mass of 47.1 m2 and 314.3 kg, and 39.9 m2 and 243.2 kg, respectively. The total radiator is 7.63 m long and has minor and major diameters of 1.48 m and 5.57 m, respectively, and a total surface area of 87 m2; however, the effective radiator area, after accounting for heat rejection through the rear of the radiator cavity, is 98.8 m2. The radiator's total mass including the C-C armor is 557.5 kg and the specific area and specific mass are 6

  18. Gas Turbine Energy Conversion Systems for Nuclear Power Plants Applicable to LiFTR Liquid Fluoride Thorium Reactor Technology

    NASA Technical Reports Server (NTRS)

    Juhasz, Albert J.

    2014-01-01

    This panel plans to cover thermal energy and electric power production issues facing our nation and the world over the next decades, with relevant technologies ranging from near term to mid-and far term.Although the main focus will be on ground based plants to provide baseload electric power, energy conversion systems (ECS) for space are also included, with solar- or nuclear energy sources for output power levels ranging tens of Watts to kilo-Watts for unmanned spacecraft, and eventual mega-Watts for lunar outposts and planetary surface colonies. Implications of these technologies on future terrestrial energy systems, combined with advanced fracking, are touched upon.Thorium based reactors, and nuclear fusion along with suitable gas turbine energy conversion systems (ECS) will also be considered by the panelists. The characteristics of the above mentioned ECS will be described, both in terms of their overall energy utilization effectiveness and also with regard to climactic effects due to exhaust emissions.

  19. Technology Development Program for an Advanced Potassium Rankine Power Conversion System Compatible with Several Space Reactor Designs

    SciTech Connect

    Yoder, G.L.

    2005-10-03

    This report documents the work performed during the first phase of the National Aeronautics and Space Administration (NASA), National Research Announcement (NRA) Technology Development Program for an Advanced Potassium Rankine Power Conversion System Compatible with Several Space Reactor Designs. The document includes an optimization of both 100-kW{sub e} and 250-kW{sub e} (at the propulsion unit) Rankine cycle power conversion systems. In order to perform the mass optimization of these systems, several parametric evaluations of different design options were investigated. These options included feed and reheat, vapor superheat levels entering the turbine, three different material types, and multiple heat rejection system designs. The overall masses of these Nb-1%Zr systems are approximately 3100 kg and 6300 kg for the 100- kW{sub e} and 250-kW{sub e} systems, respectively, each with two totally redundant power conversion units, including the mass of the single reactor and shield. Initial conceptual designs for each of the components were developed in order to estimate component masses. In addition, an overall system concept was presented that was designed to fit within the launch envelope of a heavy lift vehicle. A technology development plan is presented in the report that describes the major efforts that are required to reach a technology readiness level of 6. A 10-year development plan was proposed.

  20. Using single-chamber microbial fuel cells as renewable power sources of electro-Fenton reactors for organic pollutant treatment.

    PubMed

    Zhu, Xiuping; Logan, Bruce E

    2013-05-15

    Electro-Fenton reactions can be very effective for organic pollutant degradation, but they typically require non-sustainable electrical power to produce hydrogen peroxide. Two-chamber microbial fuel cells (MFCs) have been proposed for pollutant treatment using Fenton-based reactions, but these types of MFCs have low power densities and require expensive membranes. Here, more efficient dual reactor systems were developed using a single-chamber MFC as a low-voltage power source to simultaneously accomplish H2O2 generation and Fe(2+) release for the Fenton reaction. In tests using phenol, 75 ± 2% of the total organic carbon (TOC) was removed in the electro-Fenton reactor in one cycle (22 h), and phenol was completely degraded to simple and readily biodegradable organic acids. Compared to previously developed systems based on two-chamber MFCs, the degradation efficiency of organic pollutants was substantially improved. These results demonstrate that this system is an energy-efficient and cost-effective approach for industrial wastewater treatment of certain pollutants.

  1. System Evaluation and Economic Analysis of a Nuclear Reactor Powered High-Temperature Electrolysis Hydrogen-Production Plant

    SciTech Connect

    E. A. Harvego; M. G. McKellar; M. S. Sohal; J. E. O'Brien; J. S. Herring

    2010-06-01

    A reference design for a commercial-scale high-temperature electrolysis (HTE) plant for hydrogen production was developed to provide a basis for comparing the HTE concept with other hydrogen production concepts. The reference plant design is driven by a high-temperature helium-cooled nuclear reactor coupled to a direct Brayton power cycle. The reference design reactor power is 600 MWt, with a primary system pressure of 7.0 MPa, and reactor inlet and outlet fluid temperatures of 540°C and 900°C, respectively. The electrolysis unit used to produce hydrogen includes 4,009,177 cells with a per-cell active area of 225 cm2. The optimized design for the reference hydrogen production plant operates at a system pressure of 5.0 MPa, and utilizes an air-sweep system to remove the excess oxygen that is evolved on the anode (oxygen) side of the electrolyzer. The inlet air for the air-sweep system is compressed to the system operating pressure of 5.0 MPa in a four-stage compressor with intercooling. The alternating current (AC) to direct current (DC) conversion efficiency is 96%. The overall system thermal-to-hydrogen production efficiency (based on the lower heating value of the produced hydrogen) is 47.1% at a hydrogen production rate of 2.356 kg/s. An economic analysis of this plant was performed using the standardized H2A Analysis Methodology developed by the Department of Energy (DOE) Hydrogen Program, and using realistic financial and cost estimating assumptions. The results of the economic analysis demonstrated that the HTE hydrogen production plant driven by a high-temperature helium-cooled nuclear power plant can deliver hydrogen at a competitive cost. A cost of $3.23/kg of hydrogen was calculated assuming an internal rate of return of 10%.

  2. Verification Survey of the Building 315 Zero Power Reactor-6 Facility, Argonne National Laboratory-East, Argonne, Illinois

    SciTech Connect

    W. C. Adams

    2007-05-25

    Oak Ridge Institute for Science and Education (ORISE) conducted independent verification radiological survey activities at Argonne National Laboratory’s Building 315, Zero Power Reactor-6 facility in Argonne, Illinois. Independent verification survey activities included document and data reviews, alpha plus beta and gamma surface scans, alpha and beta surface activity measurements, and instrumentation comparisons. An interim letter report and a draft report, documenting the verification survey findings, were submitted to the DOE on November 8, 2006 and February 22, 2007, respectively (ORISE 2006b and 2007).

  3. Comparison of solar photovoltaic and nuclear reactor power systems for a human-tended lunar observatory

    NASA Technical Reports Server (NTRS)

    Hickman, J. M.; Bloomfield, H. S.

    1989-01-01

    Photovoltaic and nuclear surface power systems were examined at the 20 to 100 kW power level range for use at a human-tended lunar astronomical observatory, andestimates of the power system masses were made. One system, consisting of an SP-100 thermoelectric nuclear power supply integrated with a lunar lander, is recommended for further study due to its low system mass, potential for modular growth, and applicability to other surface power missions, particularly in the Martian system.

  4. Comparison of solar photovoltaic and nuclear reactor power systems for a human-tended lunar observatory

    NASA Technical Reports Server (NTRS)

    Hickman, J. M.; Bloomfield, H. S.

    1989-01-01

    Photovoltaic and nuclear surface power systems were examined at the 20 to 100 kW power level range for use at a human-tended lunar astronomical observatory, and estimates of the power system masses were made. One system, consisting of an SP-100 thermoelectric nuclear power supply integrated with a lunar lander, is recommended for further study due to its low system mass, potential for modular growth, and applicability to other surface power missions, particularly in the Martian system.

  5. A Code to Produce Cell Averaged Cross Sections for Fast Critical Assemblies and Fast Power Reactors.

    1987-05-14

    Version 00 SLAROM solves the neutron integral transport equations to determine the flux distribution and spectra in a fast reactor lattice and calculates cell averaged effective cross sections. The code uses multigroup data of the type in DLC-111/JFS that use Bondarenko factors for resonance effects.

  6. Environmental degradation of materials in nuclear power systems-water reactors

    SciTech Connect

    Not Available

    1984-01-01

    This book presents the papers given at a conference which focused on corrosion damage in light water reactors. Topics considered include material degradation issues in key components, material degradation in service, microstructural and compositional effects, the effects of the environment, the effects of mechanical variables, and environment and material remedies.

  7. Reanalysis of the gas-cooled fast reactor experiments at the zero power facility proteus - Spectral indices

    SciTech Connect

    Perret, G.; Pattupara, R. M.; Girardin, G.; Chawla, R.

    2012-07-01

    The gas-cooled fast reactor (GCFR) concept was investigated experimentally in the PROTEUS zero power facility at the Paul Scherrer Inst. during the 1970's. The experimental program was aimed at neutronics studies specific to the GCFR and at the validation of nuclear data in fast spectra. A significant part of the program used thorium oxide and thorium metal fuel either distributed quasi-homogeneously in the reference PuO{sub 2}/UO{sub 2} lattice or introduced in the form of radial and axial blanket zones. Experimental results obtained at the time are still of high relevance in view of the current consideration of the Gas-cooled Fast Reactor (GFR) as a Generation-IV nuclear system, as also of the renewed interest in the thorium cycle. In this context, some of the experiments have been modeled with modern Monte Carlo codes to better account for the complex PROTEUS whole-reactor geometry and to allow validating recent continuous neutron cross-section libraries. As a first step, the MCNPX model was used to test the JEFF-3.1, JEFF-3.1.1, ENDF/B-VII.0 and JENDL-3.3 libraries against spectral indices, notably involving fission and capture of {sup 232}Th and {sup 237}Np, measured in GFR-like lattices. (authors)

  8. Dynamic Comparison of Three- and Four-Equation Reactor Core Models in a Full-Scope Power Plant Training Simulator

    SciTech Connect

    Espinosa-Paredes, Gilberto; Alvarez-Ramirez, Jose; Nunez-Carrera, Alejandro; Garcia-Gutierrez, Alfonso; Martinez-Mendez, Elizabeth Jeannette

    2004-02-15

    A comparative analysis of the dynamic behavior of a boiling water reactor in a full-scope power plant simulator for operator training is presented. Three- and four-equation reactor core models were used to examine three transients following tests described in acceptance test procedures: scram, loss of feedwater flow, and closure of main isolation valves. The three-equation model consists of water and steam mixture momentum, including mass and energy balances. The four-equation model is based on liquid and gas phase mass balances, together with a drift-flux approach for the analysis of phase separation. Analysis of the models showed that the scram transient was slightly different for three- and four-equation models. The drift-flux effects can explain such differences. Regarding the loss-of-feedwater transient, the predicted steam flow after scram is larger for the three-equation model. Finally, for the transient related to the closure of main steam isolation valves, the three-equation model provides slightly different results for the pressure change, which affects reactor level behavior.

  9. Conceptual Design for a 2 GW Inertial Fusion Energy (IFE) Direct-Drive Power Reactor Employing Magnetic Intervention

    NASA Astrophysics Data System (ADS)

    Tresemer, K. R.; Gentile, C. A.

    2007-11-01

    Presented is a conceptual design for a 2 GW IFE direct drive fusion power reactor. This design employs a cusp field to deflect IFE-generated ions away from the dry first wall of the target chamber and into specifically designed ion dumps. The reactor operates at 5 Hz, consuming ˜450,000 tritium targets/day, injected at >100 m/s into the target chamber and uniformly illuminated by laser light, stimulating detonation. The resulting fusion energy is collected by equatorial ion dumps equipped with heat exchangers. The reactor will breed and recycle its own fuel through the use of breeder blankets and a fuel recovery system. To minimize target-particle interference, the chamber will be kept at <0.5 mTorr through the use of magnetically levitated turbomolecular pumps (TMPs) and corresponding backing pumps. Under investigation are the principles of magnetohydrodynamics (MHD) which may be applied to attenuate and harness the energy residing in the post detonation ion fields.

  10. Subtask 7.4 - Power River Basin Subbituminous Coal-Biomass Cogasification Testing in a Transport Reactor

    SciTech Connect

    Michael Swanson; Daniel Laudal

    2009-03-01

    The U.S. Department of Energy (DOE) National Energy Technology Laboratory Office of Coal and Environmental Systems has as its mission to develop advanced gasification-based technologies for affordable, efficient, zero-emission power generation. These advanced power systems, which are expected to produce near-zero pollutants, are an integral part of DOE's Vision 21 Program. DOE has also been developing advanced gasification systems that lower the capital and operating costs of producing syngas for chemical production. A transport reactor has shown potential to be a low-cost syngas producer compared to other gasification systems since its high-throughput-per-unit cross-sectional area reduces capital costs. This work directly supports the Power Systems Development Facility utilizing the Kellogg Brown and Root transport reactor located at the Southern Company Services Wilsonville, Alabama, site. Over 3600 hours of operation on 17 different coals ranging from bituminous to lignite along with a petroleum coke has been completed to date in the pilot-scale transport reactor development unit (TRDU) at the Energy & Environmental Research Center (EERC). The EERC has established an extensive database on the operation of these various fuels in both air- and oxygen-blown modes utilizing a pilot-scale transport reactor gasifier. This database has been useful in determining the effectiveness of design changes on an advanced transport reactor gasifier and for determining the performance of various feedstocks in a transport reactor. The effects of different fuel types on both gasifier performance and the operation of the hot-gas filter system have been determined. It has been demonstrated that corrected fuel gas heating values ranging from 90 to 130 Btu/scf have been achieved in air-blown mode, while heating values up to 230 Btu/scf on a dry basis have been achieved in oxygen-blown mode. Carbon conversions up to 90% have also been obtained and are highly dependent on the oxygen

  11. 76 FR 39922 - Office of New Reactors; Proposed Revision 4 to Standard Review Plan Section 8.1 on Electric Power...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-07-07

    ... COMMISSION Office of New Reactors; Proposed Revision 4 to Standard Review Plan Section 8.1 on Electric Power... on May 31, 2011 (76 FR 31381), that announced the proposed Revision 4 to Standard Review Plan Section 8.1 on ``Electric Power-- Introduction.'' The notice period for this notice closes on June 30,...

  12. Fast Neutron Spectrum Potassium Worth for Space Power Reactor Design Validation

    SciTech Connect

    Bess, John D.; Marshall, Margaret A.; Briggs, J. Blair; Tsiboulia, Anatoli; Rozhikhin, Yevgeniy; Mihalczo, John T.

    2015-03-01

    A variety of critical experiments were constructed of enriched uranium metal (oralloy ) during the 1960s and 1970s at the Oak Ridge Critical Experiments Facility (ORCEF) in support of criticality safety operations at the Y-12 Plant. The purposes of these experiments included the evaluation of storage, casting, and handling limits for the Y-12 Plant and providing data for verification of calculation methods and cross-sections for nuclear criticality safety applications. These included solid cylinders of various diameters, annuli of various inner and outer diameters, two and three interacting cylinders of various diameters, and graphite and polyethylene reflected cylinders and annuli. Of the hundreds of delayed critical experiments, one was performed that consisted of uranium metal annuli surrounding a potassium-filled, stainless steel can. The outer diameter of the annuli was approximately 13 inches (33.02 cm) with an inner diameter of 7 inches (17.78 cm). The diameter of the stainless steel can was 7 inches (17.78 cm). The critical height of the configurations was approximately 5.6 inches (14.224 cm). The uranium annulus consisted of multiple stacked rings, each with radial thicknesses of 1 inch (2.54 cm) and varying heights. A companion measurement was performed using empty stainless steel cans; the primary purpose of these experiments was to test the fast neutron cross sections of potassium as it was a candidate for coolant in some early space power reactor designs.The experimental measurements were performed on July 11, 1963, by J. T. Mihalczo and M. S. Wyatt (Ref. 1) with additional information in its corresponding logbook. Unreflected and unmoderated experiments with the same set of highly enriched uranium metal parts were performed at the Oak Ridge Critical Experiments Facility in the 1960s and are evaluated in the International Handbook for Evaluated Criticality Safety Benchmark Experiments (ICSBEP Handbook) with the identifier HEU MET FAST 051. Thin

  13. Contamination of the transformer oil of power transformers and shunting reactors by metal-containing colloidal particles

    SciTech Connect

    L'vov, S. Yu.; Komarov, V. B.; Bondareva, V. N.; Seliverstov, A. F.; Lyut'ko, E. O.; L'vov, Yu. N.; Ershov, B. G.

    2011-05-15

    The results of a measurement of the contamination of the oil in 66 transformers by metal-containing colloidal particles, formed as a result of the interaction of the oil with the structural materials (the copper of the windings, the iron of the tank and core etc.), and also the results of measurements of the optical turbidity of the oil in 136 transformers when they were examined at the Power Engineering Research and Development Center Company are presented. Methods of determining the concentration of copper and iron in transformer oil are considered. The limiting values of the optical turbidity factors, the copper and iron content are determined. These can serve as a basis for taking decisions on whether to replace the silica gel of the filters for continuously purifying the oil of power transformers and the shunting reactors in addition to the standardized oil contamination factors, namely, the dielectric loss tangent and the acidity number of the oil.

  14. Report to the US Nuclear Regulatory Commission on analysis and evaluation of operational data - 1987: Power reactors

    SciTech Connect

    1988-10-01

    This annual report of the US Nuclear Regulatory Commission's Office for Analysis and Evaluation of Operational Data (AEOD) is devoted to the activities performed during 1987. The report is published in two volumes. NUREG-1272, Vol. 2, No. 1, covers Power Reactors and presents an overview of the operating experience of the nuclear power industry, with comments regarding the trends of some key performance measures. The report also includes the principal findings and issues identified in AEOD studies over the past year, and summarizes information from Licensee Event Reports, the NRC's Operations Center, and Diagnostic Evaluations. NUREG-1272, Vol. 2, No. 2, covers Nonreactors and presents a review of the nonreactors events and misadministration reports that were reported in 1987 and a brief synopsis of AEOD studies published in 1987. Each volume contains a list of the AEOD Reports issued for 1980-1987.

  15. Feasibility Study of Supercritical Light Water Cooled Fast Reactors for Actinide Burning and Electric Power Production Progress Report for Year 1, Quarter 2 (January - March 2002)

    SciTech Connect

    Mac Donald, Philip Elsworth; Buongiorno, Jacopo; Davis, Cliff Bybee; Weaver, Kevan Dean

    2002-03-01

    The use of light water at supercritical pressures as the coolant in a nuclear reactor offers the potential for considerable plant simplification and consequent capital and O&M cost reduction compared with current light water reactor (LWR) designs. Also, given the thermodynamic conditions of the coolant at the core outlet (i.e. temperature and pressure beyond the water critical point), very high thermal efficiencies of the power conversion cycle are possible (i.e. up to about 45%). Because no change of phase occurs in the core, the need for steam separators and dryers as well as for BWR-type re-circulation pumps is eliminated, which, for a given reactor power, results in a substantially shorter reactor vessel and smaller containment building than the current BWRs. Furthermore, in a direct cycle the steam generators are not needed.

  16. Dynamic compensation of the Silver self-powered neutron detector in the ramp program at the OSIRIS reactor

    SciTech Connect

    Moulin, D.J.

    1994-12-31

    The Silver self-powered neutron detector (SPND) is a common detector used in the ramp program at the OSIRIS reactor. The Silver SPND signal is a reference during steady states, but its response is too slow for monitoring transient tests. In order to compensate for the inherent time delay a mathematical processing method of the Silver SPND signal was developed. Based on a convolution-type resolution of the kinetics equations, a dynamic compensation algorithm can be used for transient conditions as well as steady state conditions. A computer program reconstructs, in real-time, the dynamic neutron flux sensed by the Silver detector from the current measured between the emitter and the collector of the SPND. Although this method decreases slightly the signal-to-noise ratio, it maintains the SPND`s characteristics and reduces the response time from about 10 minutes to less than 4 seconds for a step change in flux. This provides for prompt and accurate measurement of fuel rod power during ramp experiments in the OSIRIS reactor. This development makes the Silver SPND very suitable for many on-line monitoring applications.

  17. Helium heater design for the helium direct cycle component test facility. [for gas-cooled nuclear reactor power plant

    NASA Technical Reports Server (NTRS)

    Larson, V. R.; Gunn, S. V.; Lee, J. C.

    1975-01-01

    The paper describes a helium heater to be used to conduct non-nuclear demonstration tests of the complete power conversion loop for a direct-cycle gas-cooled nuclear reactor power plant. Requirements for the heater include: heating the helium to a 1500 F temperature, operating at a 1000 psia helium pressure, providing a thermal response capability and helium volume similar to that of the nuclear reactor, and a total heater system helium pressure drop of not more than 15 psi. The unique compact heater system design proposed consists of 18 heater modules; air preheaters, compressors, and compressor drive systems; an integral control system; piping; and auxiliary equipment. The heater modules incorporate the dual-concentric-tube 'Variflux' heat exchanger design which provides a controlled heat flux along the entire length of the tube element. The heater design as proposed will meet all system requirements. The heater uses pressurized combustion (50 psia) to provide intensive heat transfer, and to minimize furnace volume and heat storage mass.

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

    SciTech Connect

    Rouelle, P.; Roy, F.

    1998-12-31

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

  19. Fluoride-Salt-Cooled High-Temperature Reactor (FHR) for Power and Process Heat

    SciTech Connect

    Forsberg, Charles; Hu, Lin-wen; Peterson, Per; Sridharan, Kumar

    2015-01-21

    In 2011 the U.S. Department of Energy through its Nuclear Energy University Program (NEUP) awarded a 3- year integrated research project (IRP) to the Massachusetts Institute of Technology (MIT) and its partners at the University of California at Berkeley (UCB) and the University of Wisconsin at Madison (UW). The IRP included Westinghouse Electric Company and an advisory panel chaired by Regis Matzie that provided advice as the project progressed. The first sentence of the proposal stated the goals: The objective of this Integrated Research Project (IRP) is to develop a path forward to a commercially viable salt-cooled solid-fuel high-temperature reactor with superior economic, safety, waste, nonproliferation, and physical security characteristics compared to light-water reactors. This report summarizes major results of this research.

  20. Direct-Drive Gas-Cooled Reactor Power System: Concept and Preliminary Testing

    NASA Astrophysics Data System (ADS)

    Wright, S. A.; Lipinski, R. J.; Godfroy, T. J.; Bragg-Sitton, S. M.; van Dyke, M. K.

    2003-01-01

    This paper describes the concept and preliminary component testing of a gas-cooled. UN-fueled, pin-type reactor which uses He/Xe gas that goes directly into a recuperated Brayton system to produce electricity for nuclear electric propulsion. This Direct-Drive Gas-Cooled Reactor (DDG) is designed to be subcritical under water or wet-sand immersion in case of a launch accident. Because the gas-cooled reactor can directly drive the Brayton turbomachinery, it is possible to configure the system such that there are no external surfaces or pressure boundaries that are refractory metal, even though the gas delivered to the turbine is 1144 K. The He/Xe gas mixture is a good heat transport medium when flowing, and a good insulator when stagnant. Judicious use of stagnant cavities as insulating regions allows transport of the 1144-K gas while keeping all external surfaces below 900 K. At this temperature super-alloys (Hastelloy or Inconel) can be used instead of refractory metals. Super-alloys reduce the technology risk because they are easier to fabricate than refractory metals, we have a much more extensive knowledge base on their characteristics, and, because they have a greater resistance to oxidation, system testing is eased. The system is also relatively simple in its design: no additional coolant pumps, heat exchanger, or freeze-thaw systems are required. Key to success of this concept is a good knowledge of the heat transfer between the fuel pins and the gas, as well as the pressure drop through the system. This paper describes preliminary testing to obtain this key information, as well as experience in demonstrating electrically heated testing of simulated reactor components.