Powered by Deep Web Technologies
Note: This page contains sample records for the topic "nuclear reactor design" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


1

Computer simulations help design new nuclear reactors  

NLE Websites -- All DOE Office Websites (Extended Search)

Computer simulations help design new nuclear reactors Director's Welcome Organization Achievements Highlights Fact Sheets, Brochures & Other Documents Multimedia Library About Nuclear Energy Nuclear Reactors Designed by Argonne Argonne's Nuclear Science and Technology Legacy Opportunities within NE Division Visit Argonne Work with Argonne Contact us For Employees Site Map Help Join us on Facebook Follow us on Twitter NE on Flickr Celebrating the 70th Anniversary of Chicago Pile 1 (CP-1) Argonne OutLoud on Nuclear Energy Argonne Energy Showcase 2012 Highlights Bookmark and Share Reprinted from "Argonne Now" - Spring 2008 Physicist Won-Sik Yang and computer scientist Andrew Siegel hold a fuel rod assembly in front of a model of the Experimental Breeder Reactor-II

2

Neutron transport analysis for nuclear reactor design  

DOE Patents (OSTI)

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

Vujic, J.L.

1993-11-30T23:59:59.000Z

3

Neutron transport analysis for nuclear reactor design  

DOE Patents (OSTI)

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

Vujic, Jasmina L. (Lisle, IL)

1993-01-01T23:59:59.000Z

4

Achievements: Nuclear Reactors designed/built by Argonne National  

NLE Websites -- All DOE Office Websites (Extended Search)

Achievements > Achievements > Argonne National Laboratory Reactors About Director's Welcome Organization Achievements Highlights Fact Sheets, Brochures & Other Documents Multimedia Library Visit Argonne Work with Argonne Contact us Nuclear Energy Why Nuclear Energy? Why are some people afraid of Nuclear Energy? How do nuclear reactors work? Cheaper & Safer Nuclear Energy Helping to Solve the Nuclear Waste Problem Nuclear Reactors Nuclear Reactors Early Exploration Training Reactors Basic and Applied Science Research LWR Technology Development BORAX-III lighting Arco, Idaho (Press Release) Heavy Water and Graphite Reactors Fast Reactor Technology Integral Fast Reactor Argonne Reactor Tree CP-1 70th Anniversary CP-1 70th Anniversary Argonne's Nuclear Science and Technology Legacy

5

Why Nuclear Energy? - Reactors designed/built by Argonne National  

NLE Websites -- All DOE Office Websites (Extended Search)

Nuclear Energy: Nuclear Energy: Why Nuclear Energy? About Director's Welcome Organization Achievements Highlights Fact Sheets, Brochures & Other Documents Multimedia Library Visit Argonne Work with Argonne Contact us Nuclear Energy Why Nuclear Energy? Why are some people afraid of Nuclear Energy? How do nuclear reactors work? Cheaper & Safer Nuclear Energy Helping to Solve the Nuclear Waste Problem Nuclear Reactors Nuclear Reactors Early Exploration Training Reactors Basic and Applied Science Research LWR Technology Development BORAX-III lighting Arco, Idaho (Press Release) Heavy Water and Graphite Reactors Fast Reactor Technology Integral Fast Reactor Argonne Reactor Tree CP-1 70th Anniversary CP-1 70th Anniversary Argonne's Nuclear Science and Technology Legacy Argonne's Nuclear Science and Technology Legacy

6

NUCLEAR REACTOR  

DOE Patents (OSTI)

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

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

1962-10-23T23:59:59.000Z

7

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

Office of Science (SC) Website

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

8

The role of integral experiments and nuclear cross section evaluations in space nuclear reactor design  

SciTech Connect

The importance of the nuclear and neutronic properties of candidate space reactor materials to the design process has been acknowledged as has been the use of benchmark reactor physics experiments to verify and qualify analytical tools used in design, safety, and performance evaluation. Since June 1966, the Cross Section Evaluation Working Group (CSEWG) has acted as an interagency forum for the assessment and evaluation of nuclear reaction data used in the nuclear design process. CSEWG data testing has involved the specification and calculation of benchmark experiments which are used widely for commercial reactor design and safety analysis. These benchmark experiments preceded the issuance oflthe industry standards for acceptance, but the benchmarks exceed the minimum acceptance criteria for such data. Thus, a starting place has been provided in assuring the accuracy and uncertainty of nuclear data important to space reactor applications. (FI)

Moses, D.L.; McKnight, R.D.

1987-01-01T23:59:59.000Z

9

Nuclear Reactors  

NLE Websites -- All DOE Office Websites (Extended Search)

Reactors Nuclear reactors created not only large amounts of plutonium needed for the weapons programs, but a variety of other interesting and useful radioisotopes. They produced...

10

Nuclear Design of the HOMER-15 Mars Surface Fission Reactor  

Science Conference Proceedings (OSTI)

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

Poston, David I. [Nuclear Systems Design Group, Decision Applications Division, Los Alamos National Laboratory, Los Alamos, New Mexico, 87545 (United States)

2002-07-01T23:59:59.000Z

11

NUCLEAR REACTOR  

DOE Patents (OSTI)

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

Grebe, J.J.

1959-07-14T23:59:59.000Z

12

Design Concept and Application of Small Nuclear Power Reactor  

Science Conference Proceedings (OSTI)

The outline of the recent design concepts and those features of the small nuclear power rector are described, including specifications, present design status, application and so on.

Minato, Akio [CRIEPI, Central Research Institute of Electric Power Industry, Tokyo (Japan); Sekimoto, Hiroshi [Center for Research into Innovative Nuclear Energy Systems (CRINES) Tokyo Institute of Technology 2-12-1, Ookayama, Meguro-ku, Tokyo, 152-8550 (Japan)

2009-03-31T23:59:59.000Z

13

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

Office of Science (SC) Website

Improved Design of Nuclear Reactor Improved Design of Nuclear Reactor Control System Nuclear Physics (NP) NP Home About Research Facilities Science Highlights Benefits of NP Spinoff Applications Spinoff Archives SBIR/STTR Applications of Nuclear Science and Technology Funding Opportunities Nuclear Science Advisory Committee (NSAC) News & Resources Contact Information Nuclear Physics U.S. Department of Energy SC-26/Germantown Building 1000 Independence Ave., SW Washington, DC 20585 P: (301) 903-3613 F: (301) 903-3833 E: sc.np@science.doe.gov More Information » Spinoff Archives Improved Design of Nuclear Reactor Control System Print Text Size: A A A RSS Feeds FeedbackShare Page Application/instrumentation: Improved Design of Nuclear Reactor Control System Developed at: Oak Ridge National Laboratory, Holifield Radioactive Ion Beam Facility (HRIBF)

14

NUCLEAR REACTOR  

DOE Patents (OSTI)

A boiling-water nuclear reactor is described wherein control is effected by varying the moderator-to-fuel ratio in the reactor core. This is accomplished by providing control tubes containing a liquid control moderator in the reactor core and providing means for varying the amount of control moderatcr within the control tubes.

Treshow, M.

1961-09-01T23:59:59.000Z

15

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

E-Print Network (OSTI)

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

Galvez, Cristhian

2011-01-01T23:59:59.000Z

16

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

E-Print Network (OSTI)

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

Gibbs, Jonathan Paul

2008-01-01T23:59:59.000Z

17

Advanced Reactor Development and Technology - Nuclear Engineering...  

NLE Websites -- All DOE Office Websites (Extended Search)

Capabilities Nuclear Systems Modeling and Design Analysis Reactor Physics and Fuel Cycle Analysis Nuclear Data Program Advanced Reactor Development Overview Advanced Fast Reactor...

18

NUCLEAR REACTOR  

DOE Patents (OSTI)

A nuclear reactor is described that includes spaced vertical fuel elements centrally disposed in a pressure vessel, a mass of graphite particles in the pressure vessel, means for fluidizing the graphite particles, and coolant tubes in the pressure vessel laterally spaced from the fuel elements. (AEC)

Post, R.G.

1963-05-01T23:59:59.000Z

19

NUCLEAR REACTOR  

DOE Patents (OSTI)

This patent relates to a combination useful in a nuclear reactor and is comprised of a casing, a mass of graphite irapregnated with U compounds in the casing, and at least one coolant tube extending through the casing. The coolant tube is spaced from the mass, and He is irtroduced irto the space between the mass and the coolant tube. (AEC)

Starr, C.

1963-01-01T23:59:59.000Z

20

NUCLEAR REACTOR  

DOE Patents (OSTI)

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

Young, G.

1963-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "nuclear reactor design" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


21

NUCLEAR REACTOR  

DOE Patents (OSTI)

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

Christy, R.F.

1958-07-15T23:59:59.000Z

22

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

Science Conference Proceedings (OSTI)

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.

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

2012-12-01T23:59:59.000Z

23

Spring design for use in the core of a nuclear reactor  

DOE Patents (OSTI)

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

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

1993-01-01T23:59:59.000Z

24

THE CNSG II--A CONCEPTUAL MERCHANT SHIP NUCLEAR REACTOR DESIGN  

SciTech Connect

The Consolidated Nuclear Steam Generator H consists of a pressurized water reactor, a steam generator, and a pressurizer combined in a sirgle pressure vessel. The design of the 66000 shaft horsepower system is presented, together with basic plant irformation for designs of 15000, 22000, and 30000 shaft horsepower. The economics, safety characteristics, and operational procedures of the plant are also discussed. (D.C.W.)

1963-09-01T23:59:59.000Z

25

Nuclear reactor  

DOE Patents (OSTI)

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

Pennell, William E. (Greensburg, PA); Rowan, William J. (Monroeville, PA)

1977-01-01T23:59:59.000Z

26

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

DOE Green Energy (OSTI)

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

Bruce G. Schnitzler

2012-01-01T23:59:59.000Z

27

Conceptual design for a re-entrant type fuel channel for supercritical water-cooled nuclear reactors.  

E-Print Network (OSTI)

??Current CANDU-type nuclear reactors use a once-through fuel-channel with an annulus gas insulating it from the moderator. The current reference design for a CANDU-type SuperCritical… (more)

Samuel, Jeffrey

2011-01-01T23:59:59.000Z

28

NUCLEAR REACTOR  

DOE Patents (OSTI)

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

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

1958-12-01T23:59:59.000Z

29

Nuclear Reactor Accidents  

NLE Websites -- All DOE Office Websites (Extended Search)

Reactor Accidents The accidents at the Three Mile Island (TMI) and Chernobyl nuclear reactors have triggered particularly intense concern about radiation hazards. The TMI accident,...

30

NUCLEAR REACTOR  

DOE Patents (OSTI)

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

Treshow, M.

1958-08-19T23:59:59.000Z

31

NUCLEAR REACTORS  

DOE Patents (OSTI)

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

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

1961-12-01T23:59:59.000Z

32

Guidebook to nuclear reactors  

SciTech Connect

A general introduction to reactor physics and theory is followed by descriptions of commercial nuclear reactor types. Future directions for nuclear power are also discussed. The technical level of the material is suitable for laymen.

Nero, A.V. Jr.

1976-05-01T23:59:59.000Z

33

FUNDAMENTALS IN THE OPERATION OF NUCLEAR TEST REACTORS. VOLUME 2. MATERIALS TESTING REACTOR DESIGN AND OPERATION  

SciTech Connect

The reactor components, building, control system and circuitry, and experimental and handling facilities are described and discussed, together with operation, shutdown, tank work and supplemental facilities. Training questions and answers are included. (D.C.W.)

1963-10-01T23:59:59.000Z

34

NUCLEAR REACTOR CONTROL SYSTEM  

DOE Patents (OSTI)

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

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

1959-11-01T23:59:59.000Z

35

THERMAL NUCLEAR REACTOR  

DOE Patents (OSTI)

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

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

1957-09-24T23:59:59.000Z

36

Small Fast Spectrum Reactor Designs Suitable for Direct Nuclear Thermal Propulsion  

SciTech Connect

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

Bruce G. Schnitzler; Stanley K. Borowski

2012-07-01T23:59:59.000Z

37

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

Office of Science (SC) Website

energy released during the neutron-induced fission of nuclear fuels is used for energy production in power reactors. The process of beta-n emission from fission products...

38

Light Water Reactors Technology Development - Nuclear Reactors  

NLE Websites -- All DOE Office Websites (Extended Search)

Light Water Reactors Light Water Reactors About Director's Welcome Organization Achievements Highlights Fact Sheets, Brochures & Other Documents Multimedia Library Visit Argonne Work with Argonne Contact us Nuclear Energy Why Nuclear Energy? Why are some people afraid of Nuclear Energy? How do nuclear reactors work? Cheaper & Safer Nuclear Energy Helping to Solve the Nuclear Waste Problem Nuclear Reactors Nuclear Reactors Early Exploration Training Reactors Basic and Applied Science Research LWR Technology Development BORAX-III lighting Arco, Idaho (Press Release) Heavy Water and Graphite Reactors Fast Reactor Technology Integral Fast Reactor Argonne Reactor Tree CP-1 70th Anniversary CP-1 70th Anniversary Argonne's Nuclear Science and Technology Legacy Argonne's Nuclear Science and Technology Legacy

39

Thermal Design of an Ultrahigh Temperature Vapor Core Reactor Combined Cycle Nuclear Power Plant  

SciTech Connect

Current work modeling high temperature compact heat exchangers may demonstrate the design feasibility of a Vapor Core Reactor (VCR) driven combined cycle power plant. For solid nuclear fuel designs, the cycle efficiency is typically limited by a metallurgical temperature limit which is dictated by fuel and structural melting points. In a vapor core, the gas/vapor phase nuclear fuel is uniformly mixed with the topping cycle working fluid. Heat is generated homogeneously throughout the working fluid thus extending the metallurgical temperature limit. Because of the high temperature, magnetohydrodynamic (MHD) generation is employed for topping cycle power extraction. MHD rejected heat is transported via compact heat exchanger to a conventional Brayton gas turbine bottoming cycle. High bottoming cycle mass flow rates are required to remove the waste heat because of low heat capacities for the bottoming cycle gas. High mass flow is also necessary to balance the high Uranium Tetrafluoride (UF{sub 4}) mass flow rate in the topping cycle. Heat exchanger design is critical due to the high temperatures and corrosive influence of fluoride compounds and fission products existing in VCR/MHD exhaust. Working fluid compositions for the topping cycle include variations of Uranium Tetrafluoride, Helium and various electrical conductivity seeds for the MHD. Bottoming cycle working fluid compositions include variations of Helium and Xenon. Some thought has been given to include liquid metal vapor in the bottoming cycle for a Cheng or evaporative cooled design enhancement. The NASA Glenn Lewis Research Center code Chemical Equilibrium with Applications (CEA) is utilized for evaluating chemical species existing in the gas stream. Work being conducted demonstrates the compact heat exchanger design, utilization of the CEA code, and assessment of different topping and bottoming working fluid compositions. (authors)

Bays, Samuel E.; Anghaie, Samim; Smith, Blair; Knight, Travis [Innovative Space Power and Propulsion Institute, University of Florida, 202 Nuclear Science Building, Gainesville, FL 32611 (United States)

2004-07-01T23:59:59.000Z

40

Designing Reactors to Facilitate Decommissioning  

SciTech Connect

Critics of nuclear power often cite issues with tail-end-of-the-fuel-cycle activities as reasons to oppose the building of new reactors. In fact, waste disposal and the decommissioning of large nuclear reactors have proven more challenging than anticipated. In the early days of the nuclear power industry the design and operation of various reactor systems was given a great deal of attention. Little effort, however, was expended on end-of-the-cycle activities, such as decommissioning and disposal of wastes. As early power and test reactors have been decommissioned difficulties with end-of-the-fuel-cycle activities have become evident. Even the small test reactors common at the INEEL were not designed to facilitate their eventual decontamination, decommissioning, and dismantlement. The results are that decommissioning of these facilities is expensive, time consuming, relatively hazardous, and generates large volumes of waste. This situation clearly supports critics concerns about building a new generation of power reactors.

Richard H. Meservey

2006-06-01T23:59:59.000Z

Note: This page contains sample records for the topic "nuclear reactor design" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


41

6 Nuclear Fuel Designs  

NLE Websites -- All DOE Office Websites (Extended Search)

Message from the Director Message from the Director 2 Nuclear Power & Researrh Reactors 3 Discovery of Promethium 4 Nuclear Isotopes 4 Nuclear Medicine 5 Nuclear Fuel Processes & Software 6 Nuclear Fuel Designs 6 Nuclear Safety 7 Nuclear Desalination 7 Nuclear Nonproliferation 8 Neutron Scattering 9 Semiconductors & Superconductors 10 lon-Implanted Joints 10 Environmental Impact Analyses 11 Environmental Quality 12 Space Exploration 12 Graphite & Carbon Products 13 Advanced Materials: Alloys 14 Advanced Materials: Ceramics 15 Biological Systems 16 Biological Systems 17 Computational Biology 18 Biomedical Technologies 19 Intelligent Machines 20 Health Physics & Radiation Dosimetry 21 Radiation Shielding 21 Information Centers 22 Energy Efficiency: Cooling & Heating

42

Preliminary neutronics design of china lead-alloy cooled demonstration reactor (CLEAR-III) for nuclear waste transmutation  

Science Conference Proceedings (OSTI)

China Lead-Alloy cooled Demonstration Reactor (CLEAR-III), which is the concept of lead-bismuth cooled accelerator driven sub-critical reactor for nuclear waste transmutation, was proposed and designed by FDS team in China. In this study, preliminary neutronics design studies have primarily focused on three important performance parameters including Transmutation Support Ratio (TSR), effective multiplication factor and blanket thermal power. The constraint parameters, such as power peaking factor and initial TRU loading, were also considered. In the specific design, uranium-free metallic dispersion fuel of (TRU-Zr)-Zr was used as one of the CLEAR-III fuel types and the ratio between MA and Pu was adjusted to maximize transmutation ratio. In addition, three different fuel zones differing in the TRU fraction of the fuel were respectively employed for this subcritical reactor, and the zone sizes and TRU fractions were determined such that the linear powers of these zones were close to each other. The neutronics calculations and analyses were performed by using Multi-Functional 4D Neutronics Simulation System named VisualBUS and nuclear data library HENDL (Hybrid Evaluated Nuclear Data Library). In the preliminary design, the maximum TSRLLMA was {approx}11 and the blanket thermal power was {approx}1000 MW when the effective multiplication factor was 0.98. The results showed that good performance of transmutation could be achieved based on the subcritical reactor loaded with uranium-free fuel. (authors)

Chen, Z. [Inst. of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); Southwest Science and Technology Univ., No.350 Shushanhu Road, Shushan District, Hefei, Anhui, 230031 (China); Chen, Y.; Bai, Y.; Wang, W.; Chen, Z.; Hu, L.; Long, P. [Inst. of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); School of Nuclear Science and Technology, Univ. of Science and Technology of China, Hefei, Anhui, 230031 (China)

2012-07-01T23:59:59.000Z

43

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

Science Conference Proceedings (OSTI)

Technical Paper / Advances in Nuclear Fuel Management - Increased Enrichment/High Burnup and Light Water Reactor Fuel Cycle Optimization

B. T. Rearden; W. J. Anderson; G. A. Harms

44

Nuclear reactor overflow line  

DOE Patents (OSTI)

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

Severson, Wayne J. (Pittsburgh, PA)

1976-01-01T23:59:59.000Z

45

Advanced Nuclear Research Reactor  

SciTech Connect

This report describes technical modifications implemented by INVAP to improve the safety of the Research Reactors the company designs and builds.

Lolich, J.V.

2004-10-06T23:59:59.000Z

46

Advanced Intermediate Heat Transport Loop Design Configurations for Hydrogen Production Using High Temperature Nuclear Reactors  

DOE Green Energy (OSTI)

The US Department of Energy is investigating the use of high-temperature nuclear reactors to produce hydrogen using either thermochemical cycles or high-temperature electrolysis. Although the hydrogen production processes are in an early stage of development, coupling either of these processes to the high-temperature reactor requires both efficient heat transfer and adequate separation of the facilities to assure that off-normal events in the production facility do not impact the nuclear power plant. An intermediate heat transport loop will be required to separate the operations and safety functions of the nuclear and hydrogen plants. A next generation high-temperature reactor could be envisioned as a single-purpose facility that produces hydrogen or a dual-purpose facility that produces hydrogen and electricity. Early plants, such as the proposed Next Generation Nuclear Plant (NGNP), may be dual-purpose facilities that demonstrate both hydrogen and efficient electrical generation. Later plants could be single-purpose facilities. At this stage of development, both single- and dual-purpose facilities need to be understood. A number of possible configurations for a system that transfers heat between the nuclear reactor and the hydrogen and/or electrical generation plants were identified. These configurations included both direct and indirect cycles for the production of electricity. Both helium and liquid salts were considered as the working fluid in the intermediate heat transport loop. Methods were developed to perform thermal-hydraulic evaluations and cycle-efficiency evaluations of the different configurations and coolants. The thermal-hydraulic evaluations estimated the sizes of various components in the intermediate heat transport loop for the different configurations. The relative sizes of components provide a relative indication of the capital cost associated with the various configurations. Estimates of the overall cycle efficiency of the various configurations were also determined. The evaluations determined which configurations and coolants are the most promising from thermal-hydraulic and efficiency points of view.

Chang Oh; Cliff Davis; Rober Barner; Paul Pickard

2005-11-01T23:59:59.000Z

47

Nuclear reactor apparatus  

DOE Patents (OSTI)

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

Wade, Elman E. (Ruffs Dale, PA)

1978-01-01T23:59:59.000Z

48

NUCLEAR REACTORS AND EARTHQUAKES  

SciTech Connect

A book is presented which supplies pertinent seismological information to engineers in the nuclear reactor field. Data are presented on the occurrence, intensity, and wave shapes. Techniques are described for evaluating the response of structures to such events. Certain reactor types and their modes of operation are described briefly. Various protection systems are considered. Earthquake experience in industrial and reactor plants is described. (D.L.C.)

1961-01-01T23:59:59.000Z

49

Design of the Next Generation Nuclear Plant Graphite Creep Experiments for Irradiation in the Advanced Test Reactor  

SciTech Connect

The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Program will be irradiating six gas reactor graphite creep experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the new United States Department of Energy’s lead laboratory for nuclear energy development. The ATR is one of the world’s premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These graphite irradiations are being accomplished to support development of the next generation reactors in the United States. The graphite experiments will be irradiated over the next six to eight years to support development of a graphite irradiation performance data base on the new nuclear grade graphites now available for use in high temperature gas reactors. The goals of the irradiation experiments are to obtain irradiation performance data at different temperatures and loading conditions to support design of the Next Generation Nuclear Plant Very High Temperature Gas Reactor, as well as other future gas reactors. The experiments will each consist of a single capsule that will contain seven separate stacks of graphite specimens. Six of the specimen stacks will have half of their graphite specimens under a compressive load, while the other half of the specimens will not be subjected to a compressive load during irradiation. The six stacks will be organized into pairs with a different compressive load being applied to the top half of each pair of specimen stacks. The seventh stack will not have a compressive load on the graphite specimens during irradiation. The specimens will be irradiated in an inert sweep gas atmosphere with on-line temperature and compressive load monitoring and control. There will also be the capability of sampling the sweep gas effluent to determine if any oxidation or off-gassing of the specimens occurs during initial start-up of the experiment. The final design phase for the first experiment was completed in September 2008, and the fabrication and assembly of the experiment test train as well as installation and testing of the control and support systems that will monitor and control the experiment during irradiation are being completed in early calendar 2009. The first experiment is scheduled to be ready for insertion in the ATR by April 30, 2009. This paper will discuss the design of the experiment including the test train and the temperature and compressive load monitoring, control, and data collection systems.

S. Blaine Grover

2009-05-01T23:59:59.000Z

50

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

DOE Patents (OSTI)

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

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

1999-01-01T23:59:59.000Z

51

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

DOE Patents (OSTI)

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

Crawford, Douglas C.; Porter, Douglas L.; Hayes, Steven L.; Hill, Robert N.

1997-12-01T23:59:59.000Z

52

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

DOE Patents (OSTI)

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

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

1999-03-23T23:59:59.000Z

53

GAS COOLED NUCLEAR REACTORS  

DOE Patents (OSTI)

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

Long, E.; Rodwell, W.

1958-06-10T23:59:59.000Z

54

HOMOGENEOUS NUCLEAR POWER REACTOR  

DOE Patents (OSTI)

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.

King, L.D.P.

1959-09-01T23:59:59.000Z

55

RADIATION FACILITY FOR NUCLEAR REACTORS  

DOE Patents (OSTI)

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

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

1961-12-12T23:59:59.000Z

56

Nuclear reactor reflector  

DOE Patents (OSTI)

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

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

1994-01-01T23:59:59.000Z

57

Nuclear reactor reflector  

DOE Patents (OSTI)

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

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

1994-06-07T23:59:59.000Z

58

Nuclear reactor control column  

DOE Patents (OSTI)

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

Bachovchin, Dennis M. (Plum Borough, PA)

1982-01-01T23:59:59.000Z

59

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

SciTech Connect

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

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

2005-08-15T23:59:59.000Z

60

Designing decommissioning into new reactor designs  

SciTech Connect

One of the lessons learned from decommissioning of existing reactors has been that decommissioning was not given much thought when these reactors were designed some three or four decades ago. Recently, the nuclear power has seen a worldwide resurgence and many new advanced reactor designs are either on the market or nearing design completion. Most of these designs are evolutionary in nature and build on the existing and proven technologies. They also incorporate many improvements and take advantage of the substantial operating experience. Nevertheless, by and large, the main factors driving the design of new reactors are the safety features, safeguards considerations, and the economic factors. With a large decommissioning experience that already exists in the nuclear industry, and with average decommissioning costs at around six hundred million dollars for each reactor in today's dollars, it is necessary that decommissioning factors also be considered as a part of the early design effort. Even though decommissioning may be sixty years down the road from the time they go on line, it is only prudent that new designs be optimized for eventual decommissioning, along with the other major considerations. (authors)

Devgun, J.S.; CHMM, Ph.D. [Nuclear Power Technologies, Sargent and Lundy LLC, Chicago, IL (United States)

2007-07-01T23:59:59.000Z

Note: This page contains sample records for the topic "nuclear reactor design" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


61

Design guide for category V reactors transient reactors  

SciTech Connect

The purpose of this Design Guide is to provide additional guidance to aid the DOE facility contractor in meeting the requirement that the siting, design, construction, modification, operation, maintenance, and decommissioning of DOE-owned reactors be in accordance with generally uniform standards, guides, and codes which are comparable to those applied to similar reactors licensed by the Nuclear Regulatory Commission (NRC). This Design Guide deals principally with the design and functional requirements of Category V reactor structures, components, and systems.

Brynda, W J; Karol, R C; Lobner, P R; Powell, R W; Straker, E A

1979-03-01T23:59:59.000Z

62

Analysis of Reference Design for Nuclear-Assisted Hydrogen Production at 750°C Reactor Outlet Temperature  

SciTech Connect

The use of High Temperature Electrolysis (HTE) for the efficient production of hydrogen without the greenhouse gas emissions associated with conventional fossil-fuel hydrogen production techniques has been under investigation at the Idaho National Engineering Laboratory (INL) for the last several years. The activities at the INL have included the development, testing and analysis of large numbers of solid oxide electrolysis cells, and the analyses of potential plant designs for large scale production of hydrogen using a high-temperature gas-cooled reactor (HTGR) to provide the process heat and electricity to drive the electrolysis process. The results of this research led to the selection in 2009 of HTE as the preferred concept in the U.S. Department of Energy (DOE) hydrogen technology down-selection process. However, the down-selection process, along with continued technical assessments at the INL, has resulted in a number of proposed modifications and refinements to improve the original INL reference HTE design. These modifications include changes in plant configuration, operating conditions and individual component designs. This report describes the resulting new INL reference design coupled to two alternative HTGR power conversion systems, a Steam Rankine Cycle and a Combined Cycle (a Helium Brayton Cycle with a Steam Rankine Bottoming Cycle). Results of system analyses performed to optimize the design and to determine required plant performance and operating conditions when coupled to the two different power cycles are also presented. A 600 MWt high temperature gas reactor coupled with a Rankine steam power cycle at a thermal efficiency of 44.4% can produce 1.85 kg/s of hydrogen and 14.6 kg/s of oxygen. The same capacity reactor coupled with a combined cycle at a thermal efficiency of 42.5% can produce 1.78 kg/s of hydrogen and 14.0 kg/s of oxygen.

Michael G. McKellar; Edwin A. Harvego

2010-05-01T23:59:59.000Z

63

Reactor & Nuclear Systems Publications | ORNL  

NLE Websites -- All DOE Office Websites (Extended Search)

Publications and Reports NSED Monthly Reports Reactor and Nuclear Systems Publications 2013 Publications 2012 Publications 2011 Publications 2010 and Older Publications Nuclear...

64

Nuclear Reactors and Technology  

SciTech Connect

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

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

1992-01-01T23:59:59.000Z

65

Role and use of nuclear theories and models in practical evaluation of neutron nuclear data needed for fission and fusion reactor design and other nuclear applications  

SciTech Connect

A review of the various nuclear models used in the evaluation of neutron nuclear data for fission and fusion reactors is presented. Computer codes embodying the principles of the relevant nuclear models are compared with each other and with experimental data. The regions of validity and limitations of the conceptual formalisms are also included, along with the effects of the numerical procedures used in the codes themselves. Conclusions and recommendations for future demands are outlined.15 tables, 15 figures, 90 references. (auth)

Prince, A.

1975-01-01T23:59:59.000Z

66

AETR NUCLEAR MOCKUP DESIGN  

SciTech Connect

The Advanced Engineering Test Reactor (AETR) nuclear mockup is designed to be a flexible, inexpensive developmental facility which duplicates the reactor portion of the AETR and which would be used to verify the operation of reactor controls. The mockup would also furnish measurements of temperature and void coefficients, flux shapes, and critical mass, and facilitate a reliable AETR design in a minimum time, and with minimum development cost. For even greater usefulness, the mockup is also designed for use in conjunction with an operating AETR to check the reactivity of experiments and fuel assemblies, and for other annular core reactor development purposes. In these respects, the mockup design embodies the concepts for a very low power nuclear auxiliary outlined in an earlier report. Detail design was limited to the reactor assembly and control systems and it is assumed that a simple building with supporting facilities may be readily supplied by private industry or the AEC. Results of negotiations with potential vendors and fabricators, a description of the facility components, and design drawings suitable for contractor use are presented. Preliminary quotations from fabricators and suppliers indicate that the facility can be in operation within an eight month period at a total cost not exceeding

Leonard, B.H.; Bertelson, P.C.; Kornfeld, M.J.; Wade, E.E.

1959-10-01T23:59:59.000Z

67

NUCLEAR REACTOR FUEL ELEMENT  

DOE Patents (OSTI)

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

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

1963-06-11T23:59:59.000Z

68

ME 361E Nuclear Reactor Engineering ABET EC2000 syllabus  

E-Print Network (OSTI)

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

Ben-Yakar, Adela

69

FUEL ELEMENT FOR NUCLEAR REACTORS  

DOE Patents (OSTI)

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

Bassett, C.H.

1961-05-16T23:59:59.000Z

70

Nuclear reactor safety device  

DOE Patents (OSTI)

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

Hutter, Ernest (Wilmette, IL)

1986-01-01T23:59:59.000Z

71

Heat dissipating nuclear reactor  

DOE Patents (OSTI)

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

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

1987-01-01T23:59:59.000Z

72

Heat dissipating nuclear reactor  

DOE Patents (OSTI)

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

Hunsbedt, A.; Lazarus, J.D.

1985-11-21T23:59:59.000Z

73

Nuclear Reactor Technologies | Department of Energy  

NLE Websites -- All DOE Office Websites (Extended Search)

Reactor Technologies Nuclear Reactor Technologies TVA Watts Bar Nuclear Power Plant | Photo courtesy of Tennessee Valley Authority TVA Watts Bar Nuclear Power Plant | Photo...

74

Licensed reactor nuclear safety criteria applicable to DOE reactors  

SciTech Connect

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

Not Available

1991-04-01T23:59:59.000Z

75

The Argonaut Reactor - Reactors designed/built by Argonne National  

NLE Websites -- All DOE Office Websites (Extended Search)

Achievements > Achievements > Argonne Reactors > Training Reactors About Director's Welcome Organization Achievements Highlights Fact Sheets, Brochures & Other Documents Multimedia Library Visit Argonne Work with Argonne Contact us Nuclear Energy Why Nuclear Energy? Why are some people afraid of Nuclear Energy? How do nuclear reactors work? Cheaper & Safer Nuclear Energy Helping to Solve the Nuclear Waste Problem Nuclear Reactors Nuclear Reactors Early Exploration Training Reactors Basic and Applied Science Research LWR Technology Development BORAX-III lighting Arco, Idaho (Press Release) Heavy Water and Graphite Reactors Fast Reactor Technology Integral Fast Reactor Argonne Reactor Tree CP-1 70th Anniversary CP-1 70th Anniversary Argonne's Nuclear Science and Technology Legacy

76

Slurry reactor design studies  

SciTech Connect

The objective of these studies was to perform a realistic evaluation of the relative costs of tublar-fixed-bed and slurry reactors for methanol, mixed alcohols and Fischer-Tropsch syntheses under conditions where they would realistically be expected to operate. The slurry Fischer-Tropsch reactor was, therefore, operated at low H{sub 2}/CO ratio on gas directly from a Shell gasifier. The fixed-bed reactor was operated on 2.0 H{sub 2}/CO ratio gas after adjustment by shift and CO{sub 2} removal. Every attempt was made to give each reactor the benefit of its optimum design condition and correlations were developed to extend the models beyond the range of the experimental pilot plant data. For the methanol design, comparisons were made for a recycle plant with high methanol yield, this being the standard design condition. It is recognized that this is not necessarily the optimum application for the slurry reactor, which is being proposed for a once-through operation, coproducing methanol and power. Consideration is also given to the applicability of the slurry reactor to mixed alcohols, based on conditions provided by Lurgi for an Octamix{trademark} plant using their standard tubular-fixed reactor technology. 7 figs., 26 tabs.

Fox, J.M.; Degen, B.D.; Cady, G.; Deslate, F.D.; Summers, R.L. (Bechtel Group, Inc., San Francisco, CA (USA)); Akgerman, A. (Texas A and M Univ., College Station, TX (USA)); Smith, J.M. (California Univ., Davis, CA (USA))

1990-06-01T23:59:59.000Z

77

Nuclear reactor building  

DOE Patents (OSTI)

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

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

1994-04-05T23:59:59.000Z

78

Nuclear reactor building  

DOE Patents (OSTI)

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

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

1994-01-01T23:59:59.000Z

79

Nuclear reactor safety device  

DOE Patents (OSTI)

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

Hutter, E.

1983-08-15T23:59:59.000Z

80

HOMOGENEOUS NUCLEAR REACTOR  

DOE Patents (OSTI)

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

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

1959-02-17T23:59:59.000Z

Note: This page contains sample records for the topic "nuclear reactor design" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


81

Nuclear reactor control apparatus  

DOE Patents (OSTI)

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

Sridhar, Bettadapur N. (Cupertino, CA)

1983-11-01T23:59:59.000Z

82

Thermal and flow design of helium-cooled reactors  

Science Conference Proceedings (OSTI)

This book continues the American Nuclear Society's series of monographs on nuclear science and technology. Chapters of the book include information on the first-generation gas-cooled reactors; HTGR reactor developments; reactor core heat transfer; mechanical problems related to the primary coolant circuit; HTGR design bases; core thermal design; gas turbines; process heat HTGR reactors; GCFR reactor thermal hydraulics; and gas cooling of fusion reactors.

Melese, G.; Katz, R.

1984-01-01T23:59:59.000Z

83

Nuclear Reactor Severe Accident Experiments  

NLE Websites -- All DOE Office Websites (Extended Search)

Nuclear Reactor Severe Accident Experiments Nuclear Reactor Severe Accident Experiments Capabilities Engineering Experimentation Reactor Safety Testing and Analysis Overview Nuclear Reactor Severe Accident Experiments MAX NSTF SNAKE Aerosol Experiments System Components Laser Applications Robots Applications Other Facilities Other Capabilities Work with Argonne Contact us For Employees Site Map Help Join us on Facebook Follow us on Twitter NE on Flickr Nuclear Reactor Severe Accident Experiments 1 2 3 4 5 6 7 We perform experiments simulating reactor core melt phenomena in which molten core debris ("corium") erodes the concrete floor of a containment building. This occurred during the Fukushima nuclear power plant accident though the extent of concrete damage is yet unknown. This video shows the top view of a churning molten pool of uranium oxide at 2000°C (3600°F) seen during an experiment at Argonne. Corium behaves much like lava.

84

Advanced Burner Test Reactor - Preconceptual Design Report  

NLE Websites -- All DOE Office Websites (Extended Search)

Burner Test Reactor Preconceptual Design Report ANL-ABR-1 (ANL-AFCI-173) Nuclear Engineering Division Disclaimer This report was prepared as an account of work sponsored by an...

85

Overview of Reactor and Nuclear  

E-Print Network (OSTI)

and Safety Gary Mays Nuclear Data and Criticality Safety Mike Dunn Nuclear Security Modeling Tim Valentine - Office of Environmental Management - Office of Intelligence · National Nuclear Security AdministrationOverview of Reactor and Nuclear Systems Division Cecil Parks RNS Division Director parkscv

86

FUEL ELEMENT FOR NUCLEAR REACTORS  

DOE Patents (OSTI)

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

Bassett, C.H.

1961-05-01T23:59:59.000Z

87

EMERGENCY SHUTDOWN FOR NUCLEAR REACTORS  

DOE Patents (OSTI)

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

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

1963-12-24T23:59:59.000Z

88

Nuclear reactor control apparatus  

DOE Patents (OSTI)

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

Sridhar, Bettadapur N. (Cupertino, CA)

1983-10-25T23:59:59.000Z

89

Nuclear reactor control  

DOE Patents (OSTI)

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

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

1982-01-01T23:59:59.000Z

90

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

E-Print Network (OSTI)

H. G. MacPherson The molten salt adventure Nuclear Scienceand P.F. Peterson, Molten-Salt-Cooled Advanced High-Clarno Assessment of candidate molten salt coolants for the

Galvez, Cristhian

2011-01-01T23:59:59.000Z

91

Heavy Liquid Metal Reactor Development - Nuclear Engineering Division  

NLE Websites -- All DOE Office Websites (Extended Search)

> Heavy Liquid Metal Reactor Development > Heavy Liquid Metal Reactor Development Capabilities Nuclear Systems Modeling and Design Analysis Reactor Physics and Fuel Cycle Analysis Nuclear Data Program Advanced Reactor Development Overview Advanced Fast Reactor (AFR) Heavy Liquid Metal Reactor Development Generation IV Nuclear Waste Form and Repository Performance Modeling Nuclear Energy Systems Design and Development Other Capabilities Work with Argonne Contact us For Employees Site Map Help Advanced Reactor Development and Technology Heavy Liquid Metal Reactor Development Bookmark and Share STAR-LM: Simplified, Modular, Small Reactor Featuring Flow-thru Fuel Cartridge STAR-LM: Simplified, Modular, Small Reactor Featuring Flow-thru Fuel Cartridge. Click on image to view larger image. Argonne has traditionally been the foremost institute in the US for

92

ESBWR... An Evolutionary Reactor Design  

Science Conference Proceedings (OSTI)

GE's latest evolution of the Boiling Water Reactor, the ESBWR, combines improvements in safety with design simplification and component standardization to produce a safer, more reliable nuclear power plant, with lower projected construction costs than plants in operation today. The ESBWR program started in the early 1990's when GE was developing the Simplified Boiling Water Reactor (SBWR). GE stopped this program because the power output of the SBWR was too small to generate the right economics for a new build project. The program was a success however, because the design proved many of the passive safety technology developments that are being utilized in the ESBWR. By harnessing these design concepts and testing results from the original SBWR and construction and operating experience from the Advanced Boiling Water Reactor (ABWR), the ESBWR design team has produced a simplified reactor with a standardized design and first-rate economics. Significant simplification of plant systems is achieved in the ESBWR. As a result, operating and maintenance staff requirements are reduced; low-level waste generation is reduced; dose rates are reduced; operational reliability is improved; and plant safety and security are improved. Each of these improvements provide distinct and unique advantages to the ESBWR design. First, fewer active components (in particular, active safety systems) reduce the maintenance and online surveillance requirements, thereby reducing operational exposure and dose rates. Second, fewer demands on plant operators and safety systems reduce plant operating staff while still providing direct improvements in accident and transient response. Finally, reductions in building volumes and required manufactured components shorten the length of time needed for ESBWR construction, resulting in improved financial returns for plant owners. The ESBWR is designed to meet the needs of nuclear power plant owners today and into the future, with a 60-year design life. Through design simplification and standardization, ESBWR offers improved safety, increased reliability, and ease of operation. Yet compared to current nuclear power plants, the ESBWR requires only a fraction of traditional plant operating and maintenance staff, offers faster construction and lower costs of construction, while also reducing operational costs. (authors)

Gamble, Robert E. [GE Energy - Nuclear, 1989 Little Orchard St., San Jose, CA 95125 (United States); Hinds, David H.; Hucik, Steven A.; Maslak, Chris E. [GE Energy - Nuclear, 3901 Castle Hayne Road A-30, Wilmington, NC 28402 (United States)

2006-07-01T23:59:59.000Z

93

Design modification for the modular helium reactor for higher temperature operation and reliability studies for nuclear hydrogen production processes  

E-Print Network (OSTI)

Design options have been evaluated for the Modular Helium Reactor (MHR) for higher temperature operation. An alternative configuration for the MHR coolant inlet flow path is developed to reduce the peak vessel temperature (PVT). The coolant inlet path is shifted from the annular path between reactor core barrel and vessel wall through the permanent side reflector (PSR). The number and dimensions of coolant holes are varied to optimize the pressure drop, the inlet velocity, and the percentage of graphite removed from the PSR to create this inlet path. With the removal of ~10% of the graphite from PSR the PVT is reduced from 541 0C to 421 0C. A new design for the graphite block core has been evaluated and optimized to reduce the inlet coolant temperature with the aim of further reduction of PVT. The dimensions and number of fuel rods and coolant holes, and the triangular pitch have been changed and optimized. Different packing fractions for the new core design have been used to conserve the number of fuel particles. Thermal properties for the fuel elements are calculated and incorporated into these analyses. The inlet temperature, mass flow and bypass flow are optimized to limit the peak fuel temperature (PFT) within an acceptable range. Using both of these modifications together, the PVT is reduced to ~350 0C while keeping the outlet temperature at 950 0C and maintaining the PFT within acceptable limits. The vessel and fuel temperatures during low pressure conduction cooldown and high pressure conduction cooldown transients are found to be well below the design limits. The reliability and availability studies for coupled nuclear hydrogen production processes based on the sulfur iodine thermochemical process and high temperature electrolysis process have been accomplished. The fault tree models for both these processes are developed. Using information obtained on system configuration, component failure probability, component repair time and system operating modes and conditions, the system reliability and availability are assessed. Required redundancies are made to improve system reliability and to optimize the plant design for economic performance. The failure rates and outage factors of both processes are found to be well below the maximum acceptable range.

Reza, S.M. Mohsin

2007-05-01T23:59:59.000Z

94

Designing a component-based architecture for the modeling and simulation of nuclear fuels and reactors  

Science Conference Proceedings (OSTI)

Concerns over the environment and energy security have recently prompted renewed interest in the U. S. in nuclear energy. Recognizing this, the U. S. Dept. of Energy has launched an initiative to revamp and modernize the role that modeling and simulation ...

Jay J. Billings; Wael R. Elwasif; Lee M. Hively; David E. Bernholdt; John M. Hetrick, III; Tim Bohn

2009-11-01T23:59:59.000Z

95

Russian RBMK reactor design information  

Science Conference Proceedings (OSTI)

This document concerns the systems, design, and operations of the graphite-moderated, boiling, water-cooled, channel-type (RBMK) reactors located in the former Soviet Union (FSU). The Russian Academy of Sciences Nuclear Safety Institute (NSI) in Moscow, Russia, researched specific technical questions that were formulated by the Pacific Northwest Laboratory (PNL) and provided detailed technical answers to those questions. The Russian response was prepared in English by NSI in a question-and-answer format. This report presents the results of that technical exchange in the context they were received from the NSI organization. Pacific Northwest Laboratory is generating this document to support the US Department of Energy (DOE) community in responding to requests from FSU states, which are seeking Western technological and financial assistance to improve the safety systems of the Russian-designed reactors. This report expands upon information that was previously available to the United States through bilateral information exchanges, international nuclear society meetings, International Atomic Energy Agency (IAEA) reactor safety programs, and Research and Development Institute of Power Engineering (RDIPE) reports. The response to the PNL questions have not been edited or reviewed for technical consistency or accuracy by PNL staff or other US organizations, but are provided for use by the DOE community in the form they were received.

Not Available

1993-11-01T23:59:59.000Z

96

Fast Reactor Curriculum Workshop - Nuclear Engineering Division...  

NLE Websites -- All DOE Office Websites (Extended Search)

Fast Reactor Curriculum Workshop Director's Welcome Organization Achievements Highlights Fact Sheets, Brochures & Other Documents Multimedia Library About Nuclear Energy Nuclear...

97

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

E-Print Network (OSTI)

A. L. London, Compact Heat Exchangers 3 rd Edition McGraw-A. L. London, Compact Heat Exchangers 3 rd Edition McGraw-that short and compact NDHX heat exchanger designs can be

Galvez, Cristhian

2011-01-01T23:59:59.000Z

98

Incorporation of Hydride Nuclear Fuels in Commercial Light Water Reactors  

E-Print Network (OSTI)

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

Terrani, Kurt Amir

2010-01-01T23:59:59.000Z

99

Nuclear Reactors and Technology; (USA)  

SciTech Connect

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

Cason, D.L.; Hicks, S.C. (eds.)

1991-01-01T23:59:59.000Z

100

Nuclear reactor I  

DOE Patents (OSTI)

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

Ference, Edward W. (Central City, PA); Houtman, John L. (Acme, PA); Waldby, Robert N. (New Stanton, PA)

1977-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "nuclear reactor design" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


101

Advanced Nuclear Reactors | Department of Energy  

NLE Websites -- All DOE Office Websites (Extended Search)

in Sodium Reactor Subassemblies Sodium-cooled Fast Reactor (SFR) Coolant Flow At the heart of a nuclear power plant is the reactor. The fuel assembly is placed inside a reactor...

102

Reactor Physics and Fuel Cycle Analysis - Nuclear Engineering Division  

NLE Websites -- All DOE Office Websites (Extended Search)

Analysis Analysis Capabilities Nuclear Systems Modeling and Design Analysis Reactor Physics and Fuel Cycle Analysis Overview Current Projects Software Nuclear Plant Dynamics and Safety Nuclear Data Program Advanced Reactor Development Nuclear Waste Form and Repository Performance Modeling Nuclear Energy Systems Design and Development Other Capabilities Work with Argonne Contact us For Employees Site Map Help Join us on Facebook Follow us on Twitter NE on Flickr Reactor Physics and Fuel Cycle Analysis Bookmark and Share Reactor physics and fuel cycle analysis is a core competency of the Nuclear Engineering (NE) Division. The Division has played a major role in the design and analysis of advanced reactors, particularly liquid-metal-cooled reactors. NE researchers have concentrated on developing computer codes for

103

Proliferation Resistant Nuclear Reactor Fuel  

Science Conference Proceedings (OSTI)

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

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

2011-02-18T23:59:59.000Z

104

NUCLEAR REACTOR FUEL SYSTEMS  

DOE Patents (OSTI)

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

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

1959-09-15T23:59:59.000Z

105

Export possibilities for small nuclear reactors  

Science Conference Proceedings (OSTI)

The worldwide deployment of peaceful nuclear technology is predicated on conformance with the Nuclear Non-Proliferation Treaty of 1972. Under this international treaty, countries have traded away pursuit of nuclear weapons in exchange for access to commercial nuclear technology that could help them grow economically. Realistically, however, most nuclear technology has been beyond the capacity of the NPT developing countries to afford. Even if the capital cost of the plant is managed, the costs of the infrastructure and the operational complexity of most nuclear technology have taken it out of the hands of the nations who need it the most. Now, a new class of small sodium cooled reactors has been specifically designed to meet the electrical power, water, hydrogen and heat needs of small and remote users. These reactors feature small size, long refueling interval, no onsite fuel storage, and simplified operations. Sized in the 10 MW(e) to 50 MW(e) range these reactors are modularized for factory production and for rapid site assembly. The fuel would be <20% U-235 uranium fuel with a 30-year core life. This new reactor type more appropriately fills the needs of countries for lower power distributed systems that can fill the gap between large developed infrastructure and primitive distributed energy systems. Looking at UN Resolution 1540 and the impact of other agreements, there is a need to address the issues of nuclear security, fuel, waste, and economic/legal/political-stakeholder concerns. This paper describes the design features of this new reactor type that specifically address these issues in a manner that increases the availability of commercial nuclear technology to the developing nations of the world. (authors)

Campagna, M.S.; Hess, C.; Moor, P. [Burns and Roe Enterprises, Inc., Oradell, NJ (United States); Sawruk, W. [ABSG Consulting, Inc., Shillington, PA (United States)

2007-07-01T23:59:59.000Z

106

Uncertainties in the Anti-neutrino Production at Nuclear Reactors  

E-Print Network (OSTI)

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

Djurcic, Zelimir

2009-01-01T23:59:59.000Z

107

THE HGCR-1, A DESIGN STUDY OF A NUCLEAR POWER STATION EMPLOYING A HIGH- TEMPERATURE GAS-COOLED REACTOR WITH GRAPHITE-UO$sub 2$ FUEL ELEMENTS  

SciTech Connect

The preliminary design of a 3095-Mw(thermal), helium-cooled, graphite- moderated reactor employing sign conditions, 1500 deg F reactor outlet gas would be circulated to eight steam generators to produce 1050 deg F, 1450-psi steam which would be converted to electrical power in eight 157-Mw(electrical) turbine- generators. The over-all efficiency of this nuclear power station is 36.5%. The significant activities released from the unclad graphite-UO/sub 2/ fuel appear to be less than 0.2% of those produced and would be equivalent to 0.002 curie/ cm/ sup 3/ in the primary helium circuit. The maintenance problems associated with this contamination level are discussed. A cost analysis indicates that the capital cost of this nuclear station per electrical kilowatt would be around 0, and that the production cost of electrical power would be 7.8 mills/kwhr. (auth)

Cottrell, W.B.; Copenhaver, C.M.; Culver, H.N.; Fontana, M.H.; Kelleghan, V.J.; Samuels, G.

1959-07-28T23:59:59.000Z

108

Heat pipe nuclear reactor for space power  

SciTech Connect

A heat-pipe cooled nuclear reactor has been designed to provide 3.2 MW(t) 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/sup 0/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 UO/sub 2/ 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.

Koenig, D.R.

1976-01-01T23:59:59.000Z

109

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

SciTech Connect

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

D. Kokkinos

2005-04-28T23:59:59.000Z

110

Nuclear Systems Modeling and Design Analysis - Nuclear Engineering Division  

NLE Websites -- All DOE Office Websites (Extended Search)

Nuclear Systems Nuclear Systems Modeling and Design Analysis CAPABILITIES Overview Nuclear Systems Modeling and Design Analysis Nuclear Systems Technologies Risk and Safety Assessments Nonproliferation and National Security Materials Testing Engineering Computation & Design Engineering Experimentation Work with Argonne Contact us For Employees Site Map Help Join us on Facebook Follow us on Twitter NE on Flickr Celebrating the 70th Anniversary of Chicago Pile 1 (CP-1) Argonne OutLoud on Nuclear Energy Argonne Energy Showcase 2012 Capabilities Nuclear Systems Modeling and Design Analysis Bookmark and Share Reactor Physics and Fuel Cycle Analysis Reactor Physics and Fuel Cycle Analysis We have played a major role in the design and analysis of most existing and past reactor types and of many

111

Innovative design of uranium startup fast reactors  

E-Print Network (OSTI)

Sodium Fast Reactors are one of the three candidates of GEN-IV fast reactors. Fast reactors play an important role in saving uranium resources and reducing nuclear wastes. Conventional fast reactors rely on transuranic ...

Fei, Tingzhou

2012-01-01T23:59:59.000Z

112

HOMOGENEOUS NUCLEAR REACTOR  

SciTech Connect

This homogeneous reactor comprises a core occupied by a solution of a fissile material in a moderator liquid and a breeder region enclosing the core and having a suspension of fertile material in the same moderator liquid. There is communication between the core and breeder to allow mass transfer and pressure equalization between the regions. The zones each have a separate circuit for removing heat by a mixer chamber situated inside the reactor vessel. The effluents coming from the two regions are mixed and led to a common device for separation into a clear solution and suspension, which are each led back to its corresponding circuit. To control the relative concentration of the two regions, an evaporator is provided separating a part of the moderator liquid from the solution occupying the core, the condensed separated moderator liquid being led into the breeder region. (NPO)

1960-07-11T23:59:59.000Z

113

Nuclear Reactors and Technology; (USA)  

SciTech Connect

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

Cason, D.L.; Hicks, S.C. (eds.)

1991-01-01T23:59:59.000Z

114

METHOD OF OPERATING NUCLEAR REACTORS  

DOE Patents (OSTI)

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

Untermyer, S.

1958-10-14T23:59:59.000Z

115

Generic small modular reactor plant design.  

SciTech Connect

This report gives an overview of expected design characteristics, concepts, and procedures for small modular reactors. The purpose of this report is to provide those who are interested in reducing the cost and improving the safety of advanced nuclear power plants with a generic design that possesses enough detail in a non-sensitive manner to give merit to their conclusions. The report is focused on light water reactor technology, but does add details on what could be different in a more advanced design (see Appendix). Numerous reactor and facility concepts were used for inspiration (documented in the bibliography). The final design described here is conceptual and does not reflect any proposed concept or sub-systems, thus any details given here are only relevant within this report. This report does not include any design or engineering calculations.

Lewis, Tom Goslee,; Cipiti, Benjamin B.; Jordan, Sabina Erteza; Baum, Gregory A.

2012-12-01T23:59:59.000Z

116

Advanced Nuclear Reactors | Department of Energy  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Advanced Nuclear Advanced Nuclear Reactors Advanced Nuclear Reactors Turbulent Flow of Coolant in an Advanced Nuclear Reactor Visualizing Coolant Flow in Sodium Reactor Subassemblies Sodium-cooled Fast Reactor (SFR) Coolant Flow At the heart of a nuclear power plant is the reactor. The fuel assembly is placed inside a reactor vessel where all the nuclear reactions occur to produce the heat and steam used for power generation. Nonetheless, an entire power plant consists of many other support components and key structures like coolant pipes; pumps and tanks including their surrounding steel framing; and concrete containment and support structures. The Reactors Product Line within NEAMS is concerned with modeling the reactor vessel as well as those components of a complete power plant that

117

Economic analysis of nuclear reactors  

SciTech Connect

The report presents several methods for estimating the power costs of nuclear reactors. When based on a consistent set of economic assumptions, total power costs may be useful in comparing reactor alternatives. The principal items contributing to the total power costs of a nuclear power plant are: (1) capital costs, (2) fuel cycle costs, (3) operation and maintenance costs, and (4) income taxes and fixed charges. There is a large variation in capital costs and fuel expenses among different reactor types. For example, the standard once-through LWR has relatively low capital costs; however, the fuel costs may be very high if U/sub 3/O/sub 8/ is expensive. In contrast, the FBR has relatively high capital costs but low fuel expenses. Thus, the distribution of expenses varies significantly between these two reactors. In order to compare power costs, expenses and revenues associated with each reactor may be spread over the lifetime of the plant. A single annual cost, often called a levelized cost, may be obtained by the methods described. Levelized power costs may then be used as a basis for economic comparisons. The paper discusses each of the power cost components. An exact expression for total levelized power costs is derived. Approximate techniques of estimating power costs will be presented.

Owen, P.S.; Parker, M.B.; Omberg, R.P.

1979-05-01T23:59:59.000Z

118

Software: Reactor Physics and Fuel Cycle Analysis - Nuclear Engineering  

NLE Websites -- All DOE Office Websites (Extended Search)

Analysis > Analysis > Software Capabilities Nuclear Systems Modeling and Design Analysis Reactor Physics and Fuel Cycle Analysis Overview Current Projects Software Nuclear Plant Dynamics and Safety Nuclear Data Program Advanced Reactor Development Nuclear Waste Form and Repository Performance Modeling Nuclear Energy Systems Design and Development Other Capabilities Work with Argonne Contact us For Employees Site Map Help Join us on Facebook Follow us on Twitter NE on Flickr Reactor Physics and Fuel Cycle Analysis Software Bookmark and Share An extensive powerful suite of computer codes developed and validated by the NE Division and its predecessor divisions at Argonne supports the development of fast reactors; many of these codes are also applicable to other reactor types. A brief description of these codes follows. Contact

119

Reactor and Nuclear Systems Division (RNSD)  

NLE Websites -- All DOE Office Websites (Extended Search)

RNSD Home RNSD Home Research Groups Advanced Reactor Systems & Safety Nuclear Data & Criticality Safety Nuclear Security Modeling Radiation Safety Information Computational Center Radiation Transport Reactor Physics Thermal Hydraulics & Irradiation Engineering Used Fuel Systems Staff Details (CV/Bios) Publications Org Chart Contact Us ORNL Staff Only Research Groups Advanced Reactor Systems & Safety Nuclear Data & Criticality Safety Nuclear Security Modeling Radiation Safety Information Computational Center Radiation Transport Reactor Physics Thermal Hydraulics & Irradiation Engineering Used Fuel Systems Reactor and Nuclear Systems Division News Highlights U.S. Rep. Fleischmann touts ORNL as national energy treasure Martin Peng wins Fusion Power Associates Leadership Award

120

Propellant actuated nuclear reactor steam depressurization valve  

DOE Patents (OSTI)

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

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

1992-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "nuclear reactor design" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


121

Reactors for nuclear electric propulsion  

SciTech Connect

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

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

1981-01-01T23:59:59.000Z

122

New Reactor Designs  

U.S. Energy Information Administration (EIA)

LWRs generate power through steam turbines similar to those used for most power generated by burning coal or fuel ... combined with British Nuclear Fuels Limited to ...

123

Flow duct for nuclear reactors  

DOE Patents (OSTI)

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

Straalsund, Jerry L. (Richland, WA)

1978-01-01T23:59:59.000Z

124

NUCLEAR REACTOR COMPENENT CLADDING MATERIAL  

DOE Patents (OSTI)

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

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

1959-01-27T23:59:59.000Z

125

Secretary Chu Statement on AP1000 Reactor Design Certification | Department  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Secretary Chu Statement on AP1000 Reactor Design Certification Secretary Chu Statement on AP1000 Reactor Design Certification Secretary Chu Statement on AP1000 Reactor Design Certification December 22, 2011 - 3:25pm Addthis Washington, D.C. - U.S. Energy Secretary Steven Chu issued the following statement today in support of the Nuclear Regulatory Commission's (NRC) decision to certify Westinghouse Electric's AP1000 nuclear reactor design, a significant step towards constructing a new generation of U.S. nuclear reactors. In February 2010, the Obama Administration announced the offer of a conditional commitment for a $8.33 billion loan guarantee for the construction and operation of two AP1000 reactors at Alvin W. Vogtle Electric Generation Plant in Burke, Georgia. "The Administration and the Energy Department are committed to restarting

126

NUCLEAR ENERGY RESEARCH INITIATIVE (NERI) PROGRAM GRANT NUMBER DE-FG03-00SF22168 TECHNICAL PROGRESS REPORT (Nov. 15, 2001 - Feb. 15,2002) ''Design and Layout Concepts for Compact, Factory-Produced, Transportable, Generation IV Reactor Systems''  

SciTech Connect

The objectives of this project are to develop and evaluate nuclear power plant designs and layout concepts to maximize the benefits of compact modular Generation IV reactor concepts including factory fabrication and packaging for optimal transportation and siting. Three nuclear power plant concepts are being studied representing water, helium and lead-bismuth coolants. This is the sixth quarterly progress report.

Fred R. Mynatt; Andy Kadak; Marc Berte; Larry Miller; Mohammed Khan; Joe McConn; Lawrence Townsend; Wesley Williams; Martin Williamson

2002-03-15T23:59:59.000Z

127

Solid0Core Heat-Pipe Nuclear Batterly Type Reactor  

DOE Green Energy (OSTI)

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

Ehud Greenspan

2008-09-30T23:59:59.000Z

128

Nuclear safety as applied to space power reactor systems  

SciTech Connect

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

Cummings, G.E.

1987-01-01T23:59:59.000Z

129

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

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

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

130

Fuel Reformation: Microchannel Reactor Design  

DOE Green Energy (OSTI)

Fuel processing is used to extract hydrogen from conventional vehicle fuel and allow fuel cell powered vehicles to use the existing petroleum fuel infrastructure. Kilowatt scale micro-channel steam reforming, water-gas shift and preferential oxida-tion reactors have been developed capable of achieving DOE required system performance metrics. Use of a microchannel design effectively supplies heat to the highly endothermic steam reforming reactor to maintain high conversions, controls the temperature profile for the exothermic water gas shift reactor, which optimizes the overall reaction conversion, and removes heat to prevent the unwanted hydrogen oxidation in the prefer-ential oxidation reactor. The reactors combined with micro-channel heat exchangers, when scaled to a full sized 50 kWe automotive system, will be less than 21 L in volume and 52 kg in weight.

Brooks, Kriston P.; Davis, James M.; Fischer, Christopher M.; King, David L.; Pederson, Larry R.; Rawlings, Gregg C.; Stenkamp, Victoria S.; TeGrotenhuis, Ward E.; Wegeng, Robert S.; Whyatt, Greg A.

2005-09-01T23:59:59.000Z

131

Reactivity control assembly for nuclear reactor  

DOE Patents (OSTI)

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

Bollinger, Lawrence R. (Schenectady, NY)

1984-01-01T23:59:59.000Z

132

Electric Power Produced from Nuclear Reactor | National Nuclear Security  

NLE Websites -- All DOE Office Websites (Extended Search)

Electric Power Produced from Nuclear Reactor | National Nuclear Security Electric Power Produced from Nuclear Reactor | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure Continuing Management Reform Countering Nuclear Terrorism About Us Our Programs Our History Who We Are Our Leadership Our Locations Budget Our Operations Media Room Congressional Testimony Fact Sheets Newsletters Press Releases Speeches Events Social Media Video Gallery Photo Gallery NNSA Archive Federal Employment Apply for Our Jobs Our Jobs Working at NNSA Blog Home > About Us > Our History > NNSA Timeline > Electric Power Produced from Nuclear Reactor Electric Power Produced from Nuclear Reactor December 20, 1951 Arco, ID Electric Power Produced from Nuclear Reactor

133

Nuclear Facility Design  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Design Design FUNCTIONAL AREA GOAL: Headquarters and Field organizations and their contractors ensure that nuclear facilities are designed to assure adequate protection for the public, workers, and the environment from nuclear hazards. REQUIREMENTS: ï‚· 10 CFR 830.120 ï‚· 10 CFR 830 subpart B ï‚· DOE O 413.3 ï‚· DOE O 420.1B ï‚· DOE O 414.1C ï‚· DOE O 226.1 ï‚· DOE M 426.1 ï‚· DEAR 970-5404-2 Guidance: ï‚· DOE G 420.1-1 ï‚· Project Management Practices, Integrated Quality ( Rev E, June 2003) ï‚· DOE Implementation Plan for DNSB Recommendation 2004-2 Performance Objective 1: Contractor Program Documentation Contracts between and the contractors who operate nuclear facilities contain adequate requirements concerning the conduct of nuclear facility safety design for nuclear facility capital projects and major modifications and the

134

Security of Nuclear Reactors and Special Nuclear Materials This revisiono  

E-Print Network (OSTI)

Provides requirements for the recovery of lost, seized, or stolen special nuclear material (para 2-1b). o Prescribes that unclassified information pertaining to plans, procedures, and equipment for the physical protection of nuclear reactors and special nuclear material will be safeguarded as DoD Unclassified Controlled Nuclear Information (para 2-1f). o Requires the conduct of a vulnerability assessment at each facility where special nuclear material is used or stored (para 2-2a). o Provides that Headquarters, U. S. Army Materiel Command will develop the postulated threat as the basis for the vulnerability assessment (para 2-2b), as well as the standardized format for documenting the results of the assessment and for the after action reports (para 2-2h). o Designates special nuclear material as inherently dangerous to others for use of force purposes (para 2-4a). o Prescribes minimum storage standards for special nuclear material (para 3-1). o Provides for the protection of vital equipment (para 3-3). o Explains the concept of the required security system for nuclear reactors and special nuclear material (para 4-2). o Establishes specific physical security standards for the protection of nuclear reactors and special nuclear material (para 4-4), to include required access controls (para 4-5). o Prohibits the locksmith from being designated as the key control officer or lock custodian (para 4-5g(25)). o Provides guidance on meeting requirement to continuously man two alarm monitoring facilities (para 4-6b). o Allows continued use of monitoring console systems installed prior to publication of this regulation that do not meet the map or video display requirement (para 4-6g(1)). o Provides guidance for testing the perimeter intrusion detection system (para 4-6n(2)). o Requires appropriate security personnel be trained to manually start the standby generator if the automatic starter fails to function properly (para 4-9b(4)). o Provides that the size, composition, and response time of the response force will be set by the major subordinate commander and approved by the Commanding

unknown authors

1993-01-01T23:59:59.000Z

135

Reactors: Modern-Day Alchemy - Argonne's Nuclear Science and Technology  

NLE Websites -- All DOE Office Websites (Extended Search)

Achievements > Achievements > Legacy > Reactors: Modern-Day Alchemy About Director's Welcome Organization Achievements Highlights Fact Sheets, Brochures & Other Documents Multimedia Library Visit Argonne Work with Argonne Contact us Nuclear Energy Why Nuclear Energy? Why are some people afraid of Nuclear Energy? How do nuclear reactors work? Cheaper & Safer Nuclear Energy Helping to Solve the Nuclear Waste Problem Nuclear Reactors Nuclear Reactors Early Exploration Training Reactors Basic and Applied Science Research LWR Technology Development BORAX-III lighting Arco, Idaho (Press Release) Heavy Water and Graphite Reactors Fast Reactor Technology Integral Fast Reactor Argonne Reactor Tree CP-1 70th Anniversary CP-1 70th Anniversary Argonne's Nuclear Science and Technology Legacy

136

Innovative fuel designs for high power density pressurized water reactor  

E-Print Network (OSTI)

One of the ways to lower the cost of nuclear energy is to increase the power density of the reactor core. Features of fuel design that enhance the potential for high power density are derived based on characteristics of ...

Feng, Dandong, Ph. D. Massachusetts Institute of Technology

2006-01-01T23:59:59.000Z

137

Simulated nuclear reactor fuel assembly  

DOE Patents (OSTI)

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

Berta, Victor T. (Idaho Falls, ID)

1993-01-01T23:59:59.000Z

138

Advanced burner test reactor preconceptual design report.  

Science Conference Proceedings (OSTI)

The goals of the Global Nuclear Energy Partnership (GNEP) are to expand the use of nuclear energy to meet increasing global energy demand, to address nuclear waste management concerns and to promote non-proliferation. Implementation of the GNEP requires development and demonstration of three major technologies: (1) Light water reactor (LWR) spent fuel separations technologies that will recover transuranics to be recycled for fuel but not separate plutonium from other transuranics, thereby providing proliferation-resistance; (2) Advanced Burner Reactors (ABRs) based on a fast spectrum that transmute the recycled transuranics to produce energy while also reducing the long term radiotoxicity and decay heat loading in the repository; and (3) Fast reactor fuel recycling technologies to recover and refabricate the transuranics for repeated recycling in the fast reactor system. The primary mission of the ABR Program is to demonstrate the transmutation of transuranics recovered from the LWR spent fuel, and hence the benefits of the fuel cycle closure to nuclear waste management. The transmutation, or burning of the transuranics is accomplished by fissioning and this is most effectively done in a fast spectrum. In the thermal spectrum of commercial LWRs, some transuranics capture neutrons and become even heavier transuranics rather than being fissioned. Even with repeated recycling, only about 30% can be transmuted, which is an intrinsic limitation of all thermal spectrum reactors. Only in a fast spectrum can all transuranics be effectively fissioned to eliminate their long-term radiotoxicity and decay heat. The Advanced Burner Test Reactor (ABTR) is the first step in demonstrating the transmutation technologies. It directly supports development of a prototype full-scale Advanced Burner Reactor, which would be followed by commercial deployment of ABRs. The primary objectives of the ABTR are: (1) To demonstrate reactor-based transmutation of transuranics as part of an advanced fuel cycle; (2) To qualify the transuranics-containing fuels and advanced structural materials needed for a full-scale ABR; and (3) To support the research, development and demonstration required for certification of an ABR standard design by the U.S. Nuclear Regulatory Commission. The ABTR should also address the following additional objectives: (1) To incorporate and demonstrate innovative design concepts and features that may lead to significant improvements in cost, safety, efficiency, reliability, or other favorable characteristics that could promote public acceptance and future private sector investment in ABRs; (2) To demonstrate improved technologies for safeguards and security; and (3) To support development of the U.S. infrastructure for design, fabrication and construction, testing and deployment of systems, structures and components for the ABRs. Based on these objectives, a pre-conceptual design of a 250 MWt ABTR has been developed; it is documented in this report. In addition to meeting the primary and additional objectives listed above, the lessons learned from fast reactor programs in the U.S. and worldwide and the operating experience of more than a dozen fast reactors around the world, in particular the Experimental Breeder Reactor-II have been incorporated into the design of the ABTR to the extent possible.

Chang, Y. I.; Finck, P. J.; Grandy, C.; Cahalan, J.; Deitrich, L.; Dunn, F.; Fallin, D.; Farmer, M.; Fanning, T.; Kim, T.; Krajtl, L.; Lomperski, S.; Moisseytsev, A.; Momozaki, Y.; Sienicki, J.; Park, Y.; Tang, Y.; Reed, C.; Tzanos, C; Wiedmeyer, S.; Yang, W.; Chikazawa, Y.; JAEA

2008-12-16T23:59:59.000Z

139

Nuclear reactor characteristics and operational history  

Gasoline and Diesel Fuel Update (EIA)

Nuclear > U.S. reactor operation status tables Nuclear > U.S. reactor operation status tables Nuclear Reactor Operational Status Tables Release date: November 22, 2011 Next release date: November 2012 See also: Table 1. Capacity and Generation, Table 2. Ownership Data Table 3. Nuclear Reactor Characteristics and Operational History PDF XLS Plant Name Generator ID Type Reactor Supplier and Model Construction Start Grid Connection Original Expiration Date License Renewal Application License Renewal Issued Extended Expiration Arkansas Nuclear One 1 PWR Babcock&Wilcox, Lower Loop 10/1/1968 8/17/1974 5/20/2014 2/1/2000 6/20/2001 5/20/2034 Arkansas Nuclear One 2 PWR Combustion Eng. 7/1/1971 12/26/1978 7/17/2018 10/15/2003 6/30/2005 7/17/2038

140

Nuclear reactor characteristics and operational history  

U.S. Energy Information Administration (EIA) Indexed Site

Nuclear > U.S. reactor operation status tables Nuclear > U.S. reactor operation status tables Nuclear Reactor Operational Status Tables Release date: November 22, 2011 Next release date: November 2012 See also: Table 2. Ownership Data, Table 3. Characteristics and Operational History Table 1. Nuclear Reactor, State, Type, Net Capacity, Generation, and Capacity Factor PDF XLS Plant/Reactor Name Generator ID State Type 2009 Summer Capacity Net MW(e)1 2010 Annual Generation Net MWh2 Capacity Factor Percent3 Arkansas Nuclear One 1 AR PWR 842 6,607,090 90 Arkansas Nuclear One 2 AR PWR 993 8,415,588 97 Beaver Valley 1 PA PWR 892 7,119,413 91 Beaver Valley 2 PA PWR 885 7,874,151 102 Braidwood Generation Station 1 IL PWR 1,178 9,196,689 89

Note: This page contains sample records for the topic "nuclear reactor design" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


141

CERNA WORKING PAPER SERIES What drives innovation in nuclear reactors technologies?  

E-Print Network (OSTI)

, rapidly shifted toward the development of nuclear reactor design technologies especially as NPPs designs evolved toward more standardized technologies (e.g., Light Water Reactors (LWRs)) by the late 1960s (OECD organizations is especially strong for nuclear reactors technology development (OECD/NEA, 2007). 19 Forward

Paris-Sud XI, Université de

142

Licensed reactor nuclear safety criteria applicable to DOE reactors  

SciTech Connect

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

Not Available

1993-11-01T23:59:59.000Z

143

Operational safety enhancement of Soviet-designed nuclear reactors via development of nuclear power plant simulators and transfer of related technology  

SciTech Connect

The US Department of Energy (DOE), under the US government`s International Nuclear Safety Program (INSP), is implementing a program of developing and providing simulators for many of the Russian and Ukrainian Nuclear Power Plants (NPPs). Pacific Northwest National Laboratory (PNNL) and Brookhaven National Laboratory (BNL) manage and provide technical oversight of the various INSP simulator projects for DOE. The program also includes a simulator technology transfer process to simulator design organizations in Russia and Ukraine. Training programs, installation of new simulators, and enhancements in existing simulators are viewed as providing a relatively fast and cost-effective technology transfer that will result in measurable improvement in the safety culture and operation of NPPs. A review of this program, its present status, and its accomplishments are provided in this paper.

Kohut, P.; Epel, L.G.; Tutu, N.K. [and others

1998-08-01T23:59:59.000Z

144

The design of a functionally graded composite for service in high temperature lead and lead-bismuth cooled nuclear reactors  

E-Print Network (OSTI)

A material that resists lead-bismuth eutectic (LBE) attack and retains its strength at 700°C would be an enabling technology for LBE-cooled reactors. No single alloy currently exists that can economically meet the required ...

Short, Michael Philip

2010-01-01T23:59:59.000Z

145

Nuclear Reactor Technologies | Department of Energy  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Reactor Technologies Reactor Technologies Nuclear Reactor Technologies TVA Watts Bar Nuclear Power Plant | Photo courtesy of Tennessee Valley Authority TVA Watts Bar Nuclear Power Plant | Photo courtesy of Tennessee Valley Authority Nuclear power has reliably and economically contributed almost 20% of electrical generation in the United States over the past two decades. It remains the single largest contributor (more than 70%) of non-greenhouse-gas-emitting electric power generation in the United States. Small Modular Reactor Technologies Small modular reactors can also be made in factories and transported to sites where they would be ready to "plug and play" upon arrival, reducing both capital costs and construction times. The smaller size also makes these reactors ideal for small electric grids and for locations that

146

Small Modular Nuclear Reactors | Department of Energy  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Reactor Technologies » Small Modular Reactor Technologies » Small Modular Nuclear Reactors Small Modular Nuclear Reactors Cutaway of 2-Unit Generation mPower SMR Installation. | © 2012 Generation mPower LLC. All Rights Reserved. Reprinted with permission. Cutaway of 2-Unit Generation mPower SMR Installation. | © 2012 Generation mPower LLC. All Rights Reserved. Reprinted with permission. The development of clean, affordable nuclear power options is a key element of the Department of Energy's Office of Nuclear Energy (DOE-NE) Nuclear Energy Research and Development Roadmap. As a part of this strategy, a high priority of the Department has been to help accelerate the timelines for the commercialization and deployment of small modular reactor (SMR) technologies through the SMR Licensing Technical Support program. Begun

147

Development of Applicable Benchmark Experiments for (Th,Pu)O2 Power Reactor Designs Using TSUNAMI Analysis.  

E-Print Network (OSTI)

?? When simulating reactor physics experiments, uncertainties in nuclear data result in a bias between simulated and experimental values. For new reactor designs or for… (more)

Langton, Stephanie E

2013-01-01T23:59:59.000Z

148

Nuclear reactor characteristics and operational history  

Gasoline and Diesel Fuel Update (EIA)

1. Capacity and Generation, Table 3. Characteristics and Operational History 1. Capacity and Generation, Table 3. Characteristics and Operational History Table 2. U.S. Nuclear Reactor Ownership Data PDF XLS Plant/Reactor Name Generator ID Utility Name - Operator Owner Name % Owned Arkansas Nuclear One 1 Entergy Arkansas Inc Entergy Arkansas Inc 100 Arkansas Nuclear One 2 Entergy Arkansas Inc Entergy Arkansas Inc 100 Beaver Valley 1 FirstEnergy Nuclear Operating Company FirstEnergy Nuclear Generation Corp 100 Beaver Valley 2 FirstEnergy Nuclear Operating Company FirstEnergy Nuclear Generation Corp 100 Braidwood Generation Station 1 Exelon Nuclear Exelon Nuclear 100 Braidwood Generation Station 2 Exelon Nuclear Exelon Nuclear 100 Browns Ferry 1 Tennessee Valley Authority Tennessee Valley Authority 100

149

NUCLEAR REACTOR FUEL ELEMENT ASSEMBLY  

DOE Patents (OSTI)

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

Stengel, F.G.

1963-12-24T23:59:59.000Z

150

Nuclear reactor control apparatus. [FBR  

DOE Patents (OSTI)

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

Sridhar, B.N.

1981-04-16T23:59:59.000Z

151

Nuclear reactor composite fuel assembly  

DOE Patents (OSTI)

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

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

1980-01-01T23:59:59.000Z

152

ALPR PRELIMINARY DESIGN STUDY (ARGONNE LOW POWER REACTOR). PHASE I  

SciTech Connect

A preliminary design study, Phase I of the ALPR . project, has been made in accordance with the Army Reactors Branch specifications for a nuclear ''package'' power plant with a 200 to 260 kw electric and 400 kw heating capacity..The plant is to be installed at the Idaho Reactor Testing Station as a prototype for remote arctic installations. The ''conventiornl'' power plant as well as the exterior reactor components are described, and cost estimates are given. ''Nuclear'' components of the reactor are described. (auth)

Treshow, M.; Hamer, E.; Pearlman, H.; Rossin, D.; Shaftman, D.

1956-04-20T23:59:59.000Z

153

Digital computer operation of a nuclear reactor  

DOE Patents (OSTI)

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

Colley, R.W.

1982-06-29T23:59:59.000Z

154

Liquid metal cooled nuclear reactor plant system  

DOE Patents (OSTI)

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

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

1993-01-01T23:59:59.000Z

155

Digital computer operation of a nuclear reactor  

DOE Patents (OSTI)

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

Colley, Robert W. (Richland, WA)

1984-01-01T23:59:59.000Z

156

Safer nuclear reactors could result from Los Alamos research  

NLE Websites -- All DOE Office Websites (Extended Search)

Calendar Video Newsroom News Releases News Releases - 2010 March Safer nuclear reactors could result from research Safer nuclear reactors could result from Los...

157

Early Exploration - Reactors designed/built by Argonne National Laboratory  

NLE Websites -- All DOE Office Websites (Extended Search)

Early Exploration Early Exploration About Director's Welcome Organization Achievements Highlights Fact Sheets, Brochures & Other Documents Multimedia Library Visit Argonne Work with Argonne Contact us Nuclear Energy Why Nuclear Energy? Why are some people afraid of Nuclear Energy? How do nuclear reactors work? Cheaper & Safer Nuclear Energy Helping to Solve the Nuclear Waste Problem Nuclear Reactors Nuclear Reactors Early Exploration Training Reactors Basic and Applied Science Research LWR Technology Development BORAX-III lighting Arco, Idaho (Press Release) Heavy Water and Graphite Reactors Fast Reactor Technology Integral Fast Reactor Argonne Reactor Tree CP-1 70th Anniversary CP-1 70th Anniversary Argonne's Nuclear Science and Technology Legacy Argonne's Nuclear Science and Technology Legacy

158

Reactor Technology | Nuclear Science | ORNL  

NLE Websites -- All DOE Office Websites (Extended Search)

Research Areas Fuel Cycle Science & Technology Fusion Nuclear Science Isotope Development and Production Nuclear Security Science & Technology Nuclear Systems Modeling, Simulation...

159

Gas-cooled nuclear reactor  

DOE Patents (OSTI)

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

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

1985-01-01T23:59:59.000Z

160

FUEL ELEMENT FOR NUCLEAR REACTORS  

DOE Patents (OSTI)

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

Bassett, C.H.

1961-11-21T23:59:59.000Z

Note: This page contains sample records for the topic "nuclear reactor design" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


161

Fast-acting nuclear reactor control device  

DOE Patents (OSTI)

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

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

1993-01-01T23:59:59.000Z

162

Shutdown system for a nuclear reactor  

DOE Patents (OSTI)

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

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

1984-06-05T23:59:59.000Z

163

Shutdown system for a nuclear reactor  

DOE Patents (OSTI)

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

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

1984-01-01T23:59:59.000Z

164

Shutdown system for a nuclear reactor  

DOE Patents (OSTI)

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

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

1982-01-20T23:59:59.000Z

165

LANL | Nuclear Design and Risk Analysis, D-5 | Home Page  

NLE Websites -- All DOE Office Websites (Extended Search)

Nuclear Design and Risk Analysis, D-5 Nuclear Design and Risk Analysis, D-5 Home About Us CAPABILITIES Computational Fluid Dynamics International Nuclear Risk Analysis Nuclear Facility Safety Nuclear Reactor Safety/ Risk Analysis Nuclear Weapons Safety Programmatic Risk Analysis Radiation Transport Modeling (MCNPX) Risk Based Decision Support Seismic Risk Analysis Small Reactor Design CONTACTS Group Leader Pratap Sadasivan (505) 665-5853 Deputy Group Leader Jay Elson Office Administrator Amanda Braithwaite Innovative design and analysis of nuclear systems The Nuclear Design and Risk Analysis Group (D-5) is a multidisciplinary team of scientists and engineers. We provide modeling and analysis capabilities to design and evaluate the potential risks of complex systems, with a focus on nuclear systems. D-5 goes beyond just providing an answer: we provide answers in the context of the overall decision process. We ensure that decision makers have all of the knowledge available to make an informed regulatory, design, or risk decision.

166

Nuclear propulsion apparatus with alternate reactor segments  

DOE Patents (OSTI)

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

Szekely, Thomas (Santa Monica, CA)

1979-04-03T23:59:59.000Z

167

Large Scale Weather Control Using Nuclear Reactors  

E-Print Network (OSTI)

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

Moninder Singh Modgil

2002-10-02T23:59:59.000Z

168

Cost-Shared Development of Innovative Small Modular Reactor Designs |  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Cost-Shared Development of Innovative Small Modular Reactor Designs Cost-Shared Development of Innovative Small Modular Reactor Designs Cost-Shared Development of Innovative Small Modular Reactor Designs The Small Modular Reactor (SMR) Licensing Technical Support (LTS) program, sponsored by the U.S. Department of Energy (DOE) Office of Nuclear Energy (NE), through this Funding Opportunity Announcement (FOA) seeks to facilitate the development of innovative SMR designs that have the potential to address the nation's economic, environmental and energy security goals. Specifically, the Department is soliciting applications for SMR designs that offer unique and innovative solutions for achieving the objectives of enhanced safety, operations, and performance relative to currently certified designs. This FOA focuses on design development and

169

Cost-Shared Development of Innovative Small Modular Reactor Designs |  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Cost-Shared Development of Innovative Small Modular Reactor Designs Cost-Shared Development of Innovative Small Modular Reactor Designs Cost-Shared Development of Innovative Small Modular Reactor Designs The Small Modular Reactor (SMR) Licensing Technical Support (LTS) program, sponsored by the U.S. Department of Energy (DOE) Office of Nuclear Energy (NE), through this Funding Opportunity Announcement (FOA) seeks to facilitate the development of innovative SMR designs that have the potential to address the nation's economic, environmental and energy security goals. Specifically, the Department is soliciting applications for SMR designs that offer unique and innovative solutions for achieving the objectives of enhanced safety, operations, and performance relative to currently certified designs. This FOA focuses on design development and

170

Nuclear reactor shield including magnesium oxide  

DOE Patents (OSTI)

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

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

1981-01-01T23:59:59.000Z

171

Fuel handling apparatus for a nuclear reactor  

DOE Patents (OSTI)

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

Hawke, Basil C. (Solana Beach, CA)

1987-01-01T23:59:59.000Z

172

Design guide for Category III reactors: pool type reactors. [US DOE  

SciTech Connect

The Department of Energy (DOE) in the ERDA Manual requires that all DOE-owned reactors be sited, designed, constructed, modified, operated, maintained, and decommissioned in a manner that gives adequate consideration to health and safety factors. Specific guidance pertinent to the safety of DOE-owned reactors is found in Chapter 0540 of the ERDA Manual. The purpose of this Design Guide is to provide additional guidance to aid the DOE facility contractor in meeting the requirement that the siting, design, construction, modification, operation, maintenance, and decommissioning of DOE-owned reactors be in accordance with generally uniform standards, guides, and codes which are comparable to those applied to similar reactors licensed by the Nuclear Regulatory Commission (NRC). This Design Guide deals principally with the design and functional requirement of Category III reactor structures, components, and systems.

Brynda, W J; Lobner, P R; Powell, R W; Straker, E A

1978-11-01T23:59:59.000Z

173

Today and Future Neutrino Experiments at Krasnoyarsk Nuclear Reactor  

E-Print Network (OSTI)

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

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

1999-12-21T23:59:59.000Z

174

Dynamic detection of nuclear reactor core incident  

Science Conference Proceedings (OSTI)

Surveillance, safety and security of evolving systems are a challenge to prevent accident. The dynamic detection of a hypothetical and theoretical blockage incident in the Phenix nuclear reactor is investigated. Such an incident is characterized by abnormal ... Keywords: Contrast, Dynamic detection of perturbations, Evolving system, Fast-neutron reactor, Neighbourhood, Noise

Laurent Hartert; Danielle Nuzillard; Jean-Philippe Jeannot

2013-02-01T23:59:59.000Z

175

MODERATOR ELEMENTS FOR UNIFORM POWER NUCLEAR REACTOR  

DOE Patents (OSTI)

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)

Balent, R.

1963-03-12T23:59:59.000Z

176

Spent nuclear fuel discharges from US reactors 1993  

SciTech Connect

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

Not Available

1995-02-01T23:59:59.000Z

177

Early Argonne reactor lit the way for worldwide nuclear industry -  

NLE Websites -- All DOE Office Websites (Extended Search)

Early Argonne reactor lit the way for worldwide Early Argonne reactor lit the way for worldwide nuclear industry About Director's Welcome Organization Achievements Highlights Fact Sheets, Brochures & Other Documents Multimedia Library Visit Argonne Work with Argonne Contact us Nuclear Energy Why Nuclear Energy? Why are some people afraid of Nuclear Energy? How do nuclear reactors work? Cheaper & Safer Nuclear Energy Helping to Solve the Nuclear Waste Problem Nuclear Reactors Nuclear Reactors Early Exploration Training Reactors Basic and Applied Science Research LWR Technology Development BORAX-III lighting Arco, Idaho (Press Release) Heavy Water and Graphite Reactors Fast Reactor Technology Integral Fast Reactor Argonne Reactor Tree CP-1 70th Anniversary CP-1 70th Anniversary Argonne's Nuclear Science and Technology Legacy

178

Nuclear reactor vessel fuel thermal insulating barrier  

DOE Patents (OSTI)

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

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

2013-03-19T23:59:59.000Z

179

Design guide for category VI reactors: air-cooled graphite reactors  

SciTech Connect

The purpose of this Design Guide is to provide additional guidance to aid the DOE facility contractor in meeting the requirement that the siting, design, construction, modification, operation, maintenance, and decommissioning of DOE-owned air-cooled graphite reactors be in accordance with generally uniform standards, guides, and codes which are comparable to those applied to similar reactors licensed by the Nuclear Regulatory Commission (NRC).

Brynda, W.J.; Karol, R.; Powell, R.W.

1979-02-01T23:59:59.000Z

180

High-Fidelity Light Water Reactor Analysis with the Numerical Nuclear Reactor  

Science Conference Proceedings (OSTI)

Technical Paper / Mathematics and Computation, Supercomputing, Reactor Physics and Nuclear and Biological Applications

David P. Weber; Tanju Sofu; Won Sik Yang; Thomas J. Downar; Justin W. Thomas; Zhaopeng Zhong; Jin Young Cho; Kang Seog Kim; Tae Hyun Chun; Han Gyu Joo; Chang Hyo Kim

Note: This page contains sample records for the topic "nuclear reactor design" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


181

Heat dissipating nuclear reactor with metal liner  

DOE Patents (OSTI)

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

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

1987-01-01T23:59:59.000Z

182

Heat dissipating nuclear reactor with metal liner  

DOE Patents (OSTI)

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

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

1985-11-21T23:59:59.000Z

183

Reactivity control assembly for nuclear reactor. [LMFBR  

DOE Patents (OSTI)

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

Bollinger, L.R.

1982-03-17T23:59:59.000Z

184

Nuclear reactor construction with bottom supported reactor vessel  

DOE Patents (OSTI)

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

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

1987-01-01T23:59:59.000Z

185

IMPROVEMENTS RELATING TO NUCLEAR REACTORS  

SciTech Connect

In order to reduce the pumping power for the coolant in a steam-cooled reactor, in which the steam being passed through successive sections of the reactor core and being superheated there, the sections are connected in series with one another, while a plurality of de-superheaters is provided such that steam flowing from one section to the next passes through a de-superheater. The condensed steam returning to the reactor from the means utilizing the steam heat content is divided into a number of separate streams. The first stream going to the first section in the reactor core is raised at least to saturated steam outside the reactor, while the remaining streams of condensed steam are conveyed to the de-superheaters to be mixed with steam passing therethrough between successive sections of the reactor, cooling in this manner said steam and being themselves converted into steam. Increasing amounts of condensate are added in successive de-superheaters until the steam returning to the reactor from the final desuperheater is equivalent to the full mass flow of steam circulating to the heat utilizing means. (NPO)

1960-08-01T23:59:59.000Z

186

Reactor closure design for a pool-type fast reactor  

SciTech Connect

The reactor closure is the topmost structural part of a reactor module. For a pool-type fast reactor it is an especially important structure because it provides the interface between the primary coolant system and the main access area above the closure. The reactor closure comprises a stationary deck, a rotatable plug, the boundary elements of primary system and containment penetrations for equipment and auxiliary systems. This paper evaluates two different reactor closure design concepts, referred to as ''warm'' deck and ''hot'' deck, for a pool-type fast reactor with respect to their design features, technical merits, and economic benefits. The evaluation also includes functional, structural, and thermal analyses of the two deck design concepts. Issues related to their fabrication and shipping to the plant site are also addressed. The warm deck is a thick solid steel plate with under-the-deck insulation consisting of many layers of steel plates. The hot deck is a box-type structure consisting of a bottom plate reinforced with vertical ribs and cylinders. For insulation and radiation shielding, the region of the hot deck above the bottom plate is filled with steel balls. Conventional insulation is added on the top to further reduce heat loss into area above the deck. The design choice of the closure deck is strongly dependent on design features of the reactor; especially on the reactor module support. While the warm deck is preferable with the top support, the hot deck is better suited for the bottom support design of the module.

Chung, H.; Seidensticker, R.W.; Kann, W.J.; Bump, T.R.; Schatmeier, C.

1986-01-01T23:59:59.000Z

187

STEAM GENERATOR FOR NUCLEAR REACTOR  

DOE Patents (OSTI)

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

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

1963-07-16T23:59:59.000Z

188

MOLTEN FLUORIDE NUCLEAR REACTOR FUEL  

DOE Patents (OSTI)

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

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

1960-01-01T23:59:59.000Z

189

22.39 Integration of Reactor Design, Operations, and Safety, Fall 2005  

E-Print Network (OSTI)

This course integrates studies of reactor physics and engineering sciences into nuclear power plant design. Topics include materials issues in plant design and operations, aspects of thermal design, fuel depletion and ...

Todreas, Neil E.

190

Fusion-breeder-reactor design studies  

SciTech Connect

Studies of the technical and economic feasibility of producing fissile fuel in tandem mirrors and in tokamaks for use in fission reactors are presented. Fission-suppressed fusion breeders promise unusually good safety features and can provide make-up fuel for 11 to 18 LWRs of equal nuclear power depending on the fuel cycle. The increased revenues from sales of both electricity and fissile material might allow the commercial application of fusion technology significantly earlier than would be possible with electricity production from fusion alone. Fast-fission designs might allow a fusion reactor with a smaller fusion power and lower Q value to be economical and thus make this application of fusion even earlier. A demonstration reactor with a fusion power of 400 MW could produce 600 kg of fissile material per year at a capacity factor of 50%. The critical issues, for which small scale experiments are either being carried out or planned, are: (1) material compatibility, (2) beryllium feasibility, (3) MHD effects, and (4) pyrochemical reprocessing.

Moir, R.W.; Lee, J.D.; Coops, M.S.

1983-04-05T23:59:59.000Z

191

Optimally moderated nuclear fission reactor and fuel source therefor  

DOE Patents (OSTI)

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

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

2008-07-22T23:59:59.000Z

192

Performance and safety design of the advanced liquid metal reactor  

SciTech Connect

The Advanced Liquid Metal Reactor (ALMR) program led by General Electric is developing, under U.S. Department of Energy sponsorship, a conceptual design for an advanced sodium-cooled liquid metal reactor plant. This design is intended to improve the already excellent level of plant safety achieved by the nuclear power industry while at the same time providing significant reductions in plant construction and operating costs. In this paper, the plant design and performance are reviewed, with emphasis on the ALMR's unique passive design safety features and its capability to utilize as fuel the actinides in LWR spent fuel.

Berglund, R.C.; Magee, P.M.; Boardman, C.E.; Gyorey, G.L. (General Electric Co., San Jose, CA (United States). Advanced Nuclear Technology)

1991-01-01T23:59:59.000Z

193

Structural mechanics of fast spectrum nuclear reactor cores  

NLE Websites -- All DOE Office Websites (Extended Search)

mechanics of fast spectrum nuclear reactor cores A fast reactor core is composed of a closely packed hexagonal arrangement of fuel, control, blanket , and shielding assemblies....

194

Shielding considerations for advanced space nuclear reactor systems  

SciTech Connect

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

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

1982-01-01T23:59:59.000Z

195

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

E-Print Network (OSTI)

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

2004-01-01T23:59:59.000Z

196

Industry Participation Sought for Design of Next Generation Nuclear Plant |  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Industry Participation Sought for Design of Next Generation Nuclear Industry Participation Sought for Design of Next Generation Nuclear Plant Industry Participation Sought for Design of Next Generation Nuclear Plant June 29, 2006 - 2:41pm Addthis Gen IV Reactor Capable of Producing Electricity and/or Hydrogen WASHINGTON, DC - The U.S. Department of Energy (DOE) is seeking expressions of interest from prospective industry teams interested in participating in the development and conceptual design for the Next Generation Nuclear Plant (NGNP), a very high temperature gas-cooled nuclear reactor prototype with the capability to produce process heat, electricity and/or hydrogen. The very high temperature reactor is based on research and development activities supported by DOE's Generation IV nuclear energy systems initiative.

197

Cooling system for a nuclear reactor  

DOE Patents (OSTI)

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

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

1982-01-01T23:59:59.000Z

198

Fast-acting nuclear reactor control device  

DOE Patents (OSTI)

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

Kotlyar, O.M.; West, P.B.

1992-12-31T23:59:59.000Z

199

Nuclear reactor fissile isotopes antineutrino spectra  

E-Print Network (OSTI)

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

Sinev, V

2012-01-01T23:59:59.000Z

200

Nuclear reactor fissile isotopes antineutrino spectra  

E-Print Network (OSTI)

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

V. Sinev

2012-07-30T23:59:59.000Z

Note: This page contains sample records for the topic "nuclear reactor design" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


201

Nuclear power plant design analysis  

SciTech Connect

Information concerning the engineering aspects of the design of commercial nuclear power plants is presented. Topics discussed include: electric utility economics; nuclear plant cconomics; thermal-transport systems and core design; nuclear analysis methods; safcty requirements; fuel-system analysis; dcsign considerations; and optimization approaches. (DCC)

Sesonske, A.

1973-01-01T23:59:59.000Z

202

The Integral Fast Reactor (IFR) - Reactors designed/built by Argonne  

NLE Websites -- All DOE Office Websites (Extended Search)

Integral Fast Reactor Integral Fast Reactor About Director's Welcome Organization Achievements Highlights Fact Sheets, Brochures & Other Documents Multimedia Library Visit Argonne Work with Argonne Contact us Nuclear Energy Why Nuclear Energy? Why are some people afraid of Nuclear Energy? How do nuclear reactors work? Cheaper & Safer Nuclear Energy Helping to Solve the Nuclear Waste Problem Nuclear Reactors Nuclear Reactors Early Exploration Training Reactors Basic and Applied Science Research LWR Technology Development BORAX-III lighting Arco, Idaho (Press Release) Heavy Water and Graphite Reactors Fast Reactor Technology Integral Fast Reactor Argonne Reactor Tree CP-1 70th Anniversary CP-1 70th Anniversary Argonne's Nuclear Science and Technology Legacy Argonne's Nuclear Science and Technology Legacy

203

Theta 13 Determination with Nuclear Reactors  

E-Print Network (OSTI)

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

F. Dalnoki-Veress

2004-06-24T23:59:59.000Z

204

Integral Fast Reactor: A future source of nuclear energy  

SciTech Connect

Argonne National Laboratory is developing a reactor concept that would be an important part of the worlds energy future. This report discusses the Integral Fast Reactor (IFR) concept which provides significant improvements over current generation reactors in reactor safety, plant complexity, nuclear proliferation, and waste generation. Two major facilities, a reactor and a fuel cycle facility, make up the IFR concept. The reactor uses fast neutrons and metal fuel in a sodium coolant at atmospheric pressure that relies on laws of physics to keep it safe. The fuel cycle facility is a hot cell using remote handling techniques for fabricating reactor fuel. The fuel feed stock includes spent fuel from the reactor, and potentially, spent light water reactor fuel and plutonium from weapons. This paper discusses the unique features of the IFR concept and the differences the quality assurance program has from current commercial practices. The IFR concept provides an opportunity to design a quality assurance program that makes use of the best contemporary ideas on management and quality.

Southon, R.

1993-09-01T23:59:59.000Z

205

Reactor physics design of supercritical CO?-cooled fast reactors  

E-Print Network (OSTI)

Gas-Cooled Fast Reactors (GFRs) are among the GEN-IV designs proposed for future deployment. Driven by anticipated plant cost reduction, the use of supercritical CO? (S-CO?) as a Brayton cycle working fluid in a direct ...

Pope, Michael A. (Michael Alexander)

2004-01-01T23:59:59.000Z

206

Reactor protection system design alternatives for sodium fast reactors  

E-Print Network (OSTI)

Historically, unprotected transients have been viewed as design basis events that can significantly challenge sodium-cooled fast reactors. The perceived potential consequences of a severe unprotected transient in a ...

DeWitte, Jacob D. (Jacob Dominic)

2011-01-01T23:59:59.000Z

207

Design of a sodium-cooled epithermal long-term exploration nuclear engine  

E-Print Network (OSTI)

To facilitate the mission to Mars initiative, the current work has focused on conceptual designs for transformational and enabling space nuclear reactor technologies. A matrix of design alternatives for both the reactor ...

Yarsky, Peter

2004-01-01T23:59:59.000Z

208

Fusion reactor design studies. [ARIES Tokamak  

SciTech Connect

This report discusses the following topics on the ARIES tokamak: systems; plasma power balance; impurity control and fusion ash removal; fusion product ripple loss; energy conversion; reactor fueling; first wall design; shield design; reactor safety; and fuel cost and resources. (LSP)

Emmert, G.A.; Kulcinski, G.L.; Santarius, J.F.

1990-10-12T23:59:59.000Z

209

Office of Nuclear Reactor Regulation  

E-Print Network (OSTI)

The U.S. Nuclear Regulatory Commission (NRC) is considering renewal of the operating licenses for the Edwin I. Hatch Nuclear Plant, Units 1 and 2 (HNP) for a period of an additional 20 years. The purpose of this assessment is to provide information to the U.S. National Marine Fisheries Service concerning the impacts of continued operation of the HNP on the shortnose sturgeon, Acipenser brevirostrum. The

unknown authors

2000-01-01T23:59:59.000Z

210

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

SciTech Connect

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.

Radulescu, Laura ['Horia Hulubei' National Institute of Nuclear Physics and Engineering, PO BOX MG-6, Bucharest 077125 (Romania); Pavelescu, Margarit [Academy of Romanian Scientists, Bucharest (Romania)

2010-01-21T23:59:59.000Z

211

22.312 Engineering of Nuclear Reactors, Fall 2004  

E-Print Network (OSTI)

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

Buongiorno, Jacopo, 1971-

212

22.312 Engineering of Nuclear Reactors, Fall 2002  

E-Print Network (OSTI)

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

Todreas, Neil E.

213

Nuclear safety criteria and specifications for space nuclear reactors  

SciTech Connect

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

1982-08-01T23:59:59.000Z

214

Investigation of bond graphs for nuclear reactor simulations  

E-Print Network (OSTI)

This work proposes a simple and effective approach to modeling multiphysics nuclear reactor problems using bond graphs. The conventional method of modeling the coupled multiphysics transients in nuclear reactors is operator ...

Sosnovsky, Eugeny

2010-01-01T23:59:59.000Z

215

More About NNSA's Naval Reactors Office | National Nuclear Security...  

National Nuclear Security Administration (NNSA)

Naval Reactors Office The Naval Nuclear Propulsion Program provides militarily effective nuclear propulsion plants and ensures their safe, reliable and long-lived operation. This...

216

Domestic U.S. Reactor Conversions: Fact Sheet | National Nuclear...  

NLE Websites -- All DOE Office Websites (Extended Search)

U.S. Reactor Conversions: Fact Sheet | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency...

217

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

Science Conference Proceedings (OSTI)

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

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

2009-10-01T23:59:59.000Z

218

Damper mechanism for nuclear reactor control elements  

DOE Patents (OSTI)

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

Taft, William Elwood (Los Gatos, CA)

1976-01-01T23:59:59.000Z

219

NUCLEAR REACTOR FUEL-BREEDER FUEL ELEMENT  

DOE Patents (OSTI)

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

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

1962-08-14T23:59:59.000Z

220

Current Abstracts Nuclear Reactors and Technology  

SciTech Connect

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

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

1993-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "nuclear reactor design" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


221

Nuclear reactor shutdown control rod assembly  

DOE Patents (OSTI)

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

Bilibin, Konstantin (North Hollywood, CA)

1988-01-01T23:59:59.000Z

222

Characterization of Nuclear Reactor Materials and Components with ...  

Science Conference Proceedings (OSTI)

About this Symposium. Meeting, 2013 TMS Annual Meeting & Exhibition. Symposium, Characterization of Nuclear Reactor Materials and Components with  ...

223

Characterization of Nuclear Reactor Materials and Components with ...  

Science Conference Proceedings (OSTI)

About this Symposium. Meeting, 2011 TMS Annual Meeting & Exhibition. Symposium, Characterization of Nuclear Reactor Materials and Components with  ...

224

Passive heat transfer means for nuclear reactors  

DOE Patents (OSTI)

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

Burelbach, James P. (Glen Ellyn, IL)

1984-01-01T23:59:59.000Z

225

Safety and core design of large liquid-metal cooled fast breeder reactors  

E-Print Network (OSTI)

and A. SESONSKE. Nuclear Reactor Engineering: Third Edition.E. LEWIS. Fundamentals of Nuclear Reactor Physics. Elseviervan DAM. “Physics of nuclear reactor safety”. In: Reports on

Qvist, Staffan Alexander

2013-01-01T23:59:59.000Z

226

MA50177: Scientific Computing Nuclear Reactor Simulation Generalised Eigenvalue Problems  

E-Print Network (OSTI)

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

Scheichl, Robert

227

1 INTRODUCTION Modern nuclear reactor concepts make use of pas-  

E-Print Network (OSTI)

1 INTRODUCTION Modern nuclear reactor concepts make use of pas- sive safety features (Fong et al systems in advanced nuclear reactors; in (Cardoso et al. 2008), Artificial Neural Networks (ANNs: Special Issue "Natural Circulation in Nuclear Reactor Systems", Hindawi Publishing Corpo- ration, Paper

228

Polynomial regression with derivative information in nuclear reactor uncertainty quantification*  

E-Print Network (OSTI)

1 Polynomial regression with derivative information in nuclear reactor uncertainty quantification in the outputs. The usual difficulties in modeling the work of the nuclear reactor models include the large size, applying the existing AD tools to nuclear reactor models still takes considerable development effort

Anitescu, Mihai

229

LIMITED POWER BURSTS IN DISTRIBUTED MODELS OF NUCLEAR REACTORS  

E-Print Network (OSTI)

of a nuclear reactor with feedback," in: Applied Problems in the Theory of Oscillations [in RussianLIMITED POWER BURSTS IN DISTRIBUTED MODELS OF NUCLEAR REACTORS M. V. Bazhenov and E. F. Sabaev UDC of Nuclear Reactors [in Russian], l~nergoatomizdat, Moscow (1990). F. R. Gantmakher and V. A. Yakubovich

Bazhenov, Maxim

230

FUEL ELEMENT FOR NUCLEAR REACTOR  

DOE Patents (OSTI)

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

Carney, K.G. Jr.

1959-07-14T23:59:59.000Z

231

DECOMMISSIONING OF NUCLEAR POWER REACTORS  

E-Print Network (OSTI)

Decommissioning means permanently removing a nuclear facility from service and reducing radioactive material on the licensed site to levels that would permit termination of the NRC license. On June 27, 1988, the NRC issued general requirements on decommissioning that contained technical and financial criteria and dealt with planning needs, timing, funding mechanisms, and environmental review

unknown authors

2000-01-01T23:59:59.000Z

232

Requirements for advanced simulation of nuclear reactor and chemicalseparation plants.  

SciTech Connect

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

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

2006-12-11T23:59:59.000Z

233

Foundational development of an advanced nuclear reactor integrated safety code.  

SciTech Connect

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

Clarno, Kevin (Oak Ridge National Laboratory, Oak Ridge, TN); Lorber, Alfred Abraham; Pryor, Richard J.; Spotz, William F.; Schmidt, Rodney Cannon; Belcourt, Kenneth (Ktech Corporation, Albuquerque, NM); Hooper, Russell Warren; Humphries, Larry LaRon

2010-02-01T23:59:59.000Z

234

Conceptual Nuclear Decommissioning Knowledge Management System Design  

Science Conference Proceedings (OSTI)

Taiwan Research Reactor (TRR) was shut down in January 1988, along with the termination of a few nuclear facilities within Institute of Nuclear Energy Research (INER). In the past few years, INER has continually dismantled its related nuclear facilities. ...

I-Hsin Chou; Chin-Feng Fan; Yen-Chang Tzeng

2005-07-01T23:59:59.000Z

235

Nuclear reactor control room construction  

DOE Patents (OSTI)

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

Lamuro, R.C.; Orr, R.

1993-11-16T23:59:59.000Z

236

FUEL ELEMENT FOR NUCLEAR REACTORS  

DOE Patents (OSTI)

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

Dickson, J.J.

1963-09-24T23:59:59.000Z

237

CRC handbook of nuclear reactors calculations. Vol. II  

Science Conference Proceedings (OSTI)

This handbook breaks down the complex field of nuclear reactor calculations into major steps. Each step presents a detailed analysis of the problems to be solved, the parameters involved, and the elaborate computer programs developed to perform the calculations. This book bridges the gap between nuclear reactor theory and the implementation of that theory, including the problems to be encountered and the level of confidence that should be given to the methods described. Volume II: Monte Carlo Calculations for Nuclear Reactors. In-Core Management of Four Reactor Types. In-Core Management in CANDU-PHW Reactors. Reactor Dynamics. The Theory of Neutron Leakage in Reactor Lattices. Index.

Ronen, Y.

1986-01-01T23:59:59.000Z

238

PEBBLE-BED NUCLEAR REACTOR SYSTEM PHYSICS AND FUEL UTILIZATION  

E-Print Network (OSTI)

The Generation IV Pebble Bed Modular Reactor (PMBR) design may be used for electricity production, co-generation applications (industrial heat, hydrogen production, desalination, etc.), and could potentially eliminate some high level nuclear wastes. Because of these advantages, as well as the ability to build cost-effective small-to-medium sized reactors, this design is currently being considered for construction in many countries, from Japan, where test reactors are being analyzed, to China. The use of TRISO-coated micro-particles as a fuel in these reactors leads to multi-heterogeneity physics features that must be properly treated and accounted for. Inherent interrelationships of neutron interactions, temperature effects, and structural effects, further challenge computational evaluations of High Temperature Reactors (HTRs). The developed models and computational techniques have to be validated in code-to-code and, most importantly, code-to-experiment benchmark studies. This report quantifies the relative accuracy of various multi-heterogeneity treatments in whole-core 3D models for parametric studies of Generation IV Pebble Bed Modular Reactors as well as provide preliminary results of the PBMR performance analysis. Data is gathered from two different models, one based upon a benchmark for the African PBMR-400 design, and another based on the PROTEUS criticality experiment, since the African design is a more realistic power reactor, but the PROTEUS experiment model can be used for calculations that cannot be performed on the more complex model. Early data was used to refine final models, and the resulting final models were used to conduct parametric studies on composition and geometry optimization based on pebble bed reactor physics in order to improve fuel utilization.

Kelly, Ryan 1989-

2011-05-01T23:59:59.000Z

239

Rodded shutdown system for a nuclear reactor  

DOE Patents (OSTI)

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

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

1978-01-01T23:59:59.000Z

240

GAS COOLED NUCLEAR REACTOR STUDY. Final Report  

SciTech Connect

An investigntion was made of the performance of a gas-cooled reactor, designed to provide a source of high temperature heat to a stream of helium. This reactor, in turn, is used as a source of heat for the air stream in a gas- turbine power plant. The reactor design was predicted primarily on the requirement for transferring a large amount of heat to the helium stream with a pressure drop low enough that it will not represent a major loss of power in the power plant. The mass of uranium e uired far criticality under various circumstances was investigated by multigroup calculations, both on desk calculators and on an IBM-704 machine. The gasturbine power plant perfarmance was studied based on a Studebaker-Packard-designed gas-turbine power plant for the propulsion of destroyer-escort vessels. A small experimental program was carried out to study some effects of helium on graphite and on structural steels. (auth)

Thompson, A.S.

1956-07-31T23:59:59.000Z

Note: This page contains sample records for the topic "nuclear reactor design" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


241

Accelerator Laboratory AGN-201M Nuclear Reactor Laboratory  

E-Print Network (OSTI)

Laboratory Nuclear Power Institute (NPI) Nuclear Science Center (1MW Triga Reactor) (NSC) Nuclear SecurityAccelerator Laboratory AGN-201M Nuclear Reactor Laboratory Center for Large-scale Scientific Simulations (CLASS) Fuel Cycle and Materials Laboratory (FCML) Institute for National Security, Education

242

Fuel rod retention device for a nuclear reactor  

DOE Patents (OSTI)

A device is described for supporting a nuclear fuel rod in a fuel rod assembly which allows the rod to be removed without disturbing other rods in the assembly. A fuel rod cap connects the rod to a bolt which is supported in the assembly end fitting by means of a locking assembly. The device is designed so that the bolt is held securely during normal reactor operation yet may be easily disengaged and the fuel rod removed when desired.

Hylton, Charles L. (Madison Heights, VA)

1984-01-01T23:59:59.000Z

243

Multiple microprocessor based nuclear reactor power monitor  

SciTech Connect

The reactor power monitor is a portable multiple-microprocessor controlled data acquisition device being built for the International Atomic Energy Association. Its function is to measure and record the hourly integrated operating thermal power level of a nuclear reactor for the purpose of detecting unannounced plutonium production. The monitor consists of a /sup 3/He proportional neutron detector, a write-only cassette tape drive and control electronics based on two INTEL 8748 microprocessors. The reactor power monitor operates from house power supplied by the plant operator, but has eight hours of battery backup to cover power interruptions. Both the hourly power levels and any line power interruptions are recorded on tape and in memory. Intermediate dumps from the memory to a data terminal or strip chart recorder can be performed without interrupting data collection.

Lewis, P.S.; Ethridge, C.D.

1979-01-01T23:59:59.000Z

244

Conceptual design study of spheromak reactors  

SciTech Connect

Preliminary design studies are carried out for a spheromak fusion reactor. Simplified circuit theory is applied to obtain characteristic relations among various parameters of the spheromak configuration for an aspect ratio A greater than or equal to 1.6. These relations are used to calculate the parameters for the conceptual designs of three types of fusion reactor: (1) DT two-component, (2) DT ignited, and, (3) catalyzed DD ignited reactors. With a total wall loading of approx. 4 MWm/sup -2/, it is found that edge magnetic fields of only approx. 4T (DT) and approx. 9T (cat. DD) are required for ignited reactors of one-meter plasma (minor) radius with output powers in the gigawatt range. Assessment of various methods of generating reactor-grade spheromak plasmas is discussed briefly.

Katsurai, M.; Yamada, M.

1980-07-01T23:59:59.000Z

245

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

Science Conference Proceedings (OSTI)

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

Not Available

1993-01-01T23:59:59.000Z

246

DOE fundamentals handbook: Nuclear physics and reactor theory  

SciTech Connect

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

Not Available

1993-01-01T23:59:59.000Z

247

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

SciTech Connect

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

Not Available

1993-01-01T23:59:59.000Z

248

FUEL ELEMENT FOR A NUCLEAR REACTOR  

DOE Patents (OSTI)

A fuel element structure particularly useful in high temperature nuclear reactors is presented. Basically, the structure comprises two coaxial graphite sleeves integrally joined together by radial fins. Due to the high structural strength of graphite at high temperatures and the rigidity of this structure, nuclear fuel encased within the inner sleeve in contiguous relation therewith is supported and prevented from expanding radially at high temperatures. Thus, the necessity of relying on the usual cladding materials with relatively low temperature limitations for structural strength is removed. (AEC)

Davidson, J.K.

1963-11-19T23:59:59.000Z

249

NUCLEAR REACTOR AND THERMIONIC FUEL ELEMENT THEREFOR  

DOE Patents (OSTI)

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

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

1963-12-01T23:59:59.000Z

250

CRC handbook of nuclear reactors calculations. Vol. III  

Science Conference Proceedings (OSTI)

This handbook breaks down the complex field of nuclear reactor calculations into major steps. Each step presents a detailed analysis of the problems to be solved, the parameters involved, and the elaborate computer programs developed to perform the calculations. This book bridges the gap between nuclear reactor theory and the implementation of that theory, including the problems to be encountered and the level of confidence that should be given to the methods described. Volume III: Control Rods and Burnable Absorber Calculations. Perturbation Theory for Nuclear Reactor Analysis. Thermal Reactors Calculations. Fast Reactor Calculations. Seed-Blanket Reactors. Index.

Ronen, Y.

1986-01-01T23:59:59.000Z

251

Mirror Advanced Reactor Study interim design report  

DOE Green Energy (OSTI)

The status of the design of a tenth-of-a-kind commercial tandem-mirror fusion reactor is described at the midpoint of a two-year study. When completed, the design is to serve as a strategic goal for the mirror fusion program. The main objectives of the Mirror Advanced Reactor Study (MARS) are: (1) to design an attractive tandem-mirror fusion reactor producing electricity and synfuels (in alternate versions), (2) to identify key development and technology needs, and (3) to exploit the potential of fusion for safety, low activation, and simple disposal of radioactive waste. In the first year we have emphasized physics and engineering of the central cell and physics of the end cell. Design optimization and trade studies are continuing, and we expect additional modifications in the end cells to further improve the performance of the final design.

Not Available

1983-04-01T23:59:59.000Z

252

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

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Safety Culture in the US Nuclear Regulatory Commission's Reactor Safety Culture in the US Nuclear Regulatory Commission's Reactor Oversight Process Safety Culture in the US Nuclear Regulatory Commission's Reactor Oversight Process September 19, 2012 Presenter: Undine Shoop, Chief, Health Physics and Human Performance Branch, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission Topics covered: Purpose of the Reactor Oversight Process (ROP) ROP Framework Safety Culture within the ROP Safety Culture Assessments Safety Culture in the US Nuclear Regulatory Commission's Reactor Oversight Process More Documents & Publications A Commissioner's Perspective on USNRC Actions in Response to the Fukushima Nuclear Accident Comparison of Integrated Safety Analysis (ISA) and Probabilistic Risk Assessment (PRA) for Fuel Cycle Facilities, 2/17/11

253

Advanced nuclear reactor public opinion project  

SciTech Connect

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

Benson, B.

1991-07-25T23:59:59.000Z

254

Uncertainties in the Anti-neutrino Production at Nuclear Reactors  

E-Print Network (OSTI)

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

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

2008-08-06T23:59:59.000Z

255

Uncertainties in the Anti-neutrino Production at Nuclear Reactors  

E-Print Network (OSTI)

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

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

2008-01-01T23:59:59.000Z

256

NUCLEAR ENERGY RESEARCH INITIATIVE (NERI) PROGRAM GRANT NUMBER DE-FG03-00SF22168 TECHNICAL PROGRESS REPORT (Nov. 15, 2001 - Feb. 15,2002) ''Design and Layout Concepts for Compact, Factory-Produced, Transportable, Generation IV Reactor Systems''  

SciTech Connect

The objectives of this project are to develop and evaluate nuclear power plant designs and layout concepts to maximize the benefits of compact modular Generation IV reactor concepts including factory fabrication and packaging for optimal transportation and siting. Three nuclear power plant concepts are being studied representing water, helium and lead-bismuth coolants. This is the sixth quarterly progress report.

Fred R. Mynatt; Andy Kadak; Marc Berte; Larry Miller; Mohammed Khan; Joe McConn; Lawrence Townsend; Wesley Williams; Martin Williamson

2002-03-15T23:59:59.000Z

257

Nuclear reactor pressure vessel support system  

DOE Patents (OSTI)

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

Sepelak, George R. (McMurray, PA)

1978-01-01T23:59:59.000Z

258

CONFERENCES AND SYMPOSIA FUSION REACTOR DESIGN IV  

E-Print Network (OSTI)

step and present large scale tokamak design groups should be encouraged. 3. The next reactor design (8th April 1984). These devices are the so called 'Large Tokamaks', because their parameters (Table 1 AND SYMPOSIA TABLE 1-1. CURRENT DATA BASE FOR LARGE TOKAMAKS Parameter/data TFTR JET JT-6O T-15 Major radius (m

Abdou, Mohamed

259

FAST FUEL TEST REACTOR-FFTR CONCEPTUAL DESIGN STUDY  

SciTech Connect

The Fast Fuel Test Reactor (FFTR) is a nuclear facility for the purpose of irradiating samples of fuels and structural components for use in fast reactors. The core consisis of a plate type element in a square configuration. Beryllium metal between the fuel elements is used to obtain a neutron energy spectrum in the hard intermediate region. Cooling of the core and test specimens is accomplished by means of liquid sodium. The design concept was carried through in sufficient degree in the following areas of preliminary concern: number and size of irradiation facilities, sample power requirements, plant layout to evaluate site requirements, plant and nuclear design parameters to evaluate essential equipment requirements. plant-capital-cost estimate, annual- operating-cost estimate, and estimate of construction time schedule. (W.D.M.)

Brubaker, R.; Hummel, H.H.; McArthy, A.; Smaardyk, A.; Kittel, J.H.

1960-08-01T23:59:59.000Z

260

Spatial multi-taper spectrum estimation for nuclear reactor modelling  

Science Conference Proceedings (OSTI)

Multi-taper univariate and cross-spectral analysis is used to investigate the structure of spatial variation in the temperatures measured across the surface of a nuclear reactor. The construction of the spatial tapers over the approximate circular reactor ...

C. J. Scarrott; G. Tunnicliffe Wilson

2009-10-01T23:59:59.000Z

Note: This page contains sample records for the topic "nuclear reactor design" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


261

Design options for a bunsen reactor.  

SciTech Connect

This work is being performed for Matt Channon Consulting as part of the Sandia National Laboratories New Mexico Small Business Assistance Program (NMSBA). Matt Channon Consulting has requested Sandia's assistance in the design of a chemical Bunsen reactor for the reaction of SO2, I2 and H2O to produce H2SO4 and HI with a SO2 feed rate to the reactor of 50 kg/hour. Based on this value, an assumed reactor efficiency of 33%, and kinetic data from the literature, a plug flow reactor approximately 1%E2%80%9D diameter and and 12 inches long would be needed to meet the specification of the project. Because the Bunsen reaction is exothermic, heat in the amount of approximately 128,000 kJ/hr would need to be removed using a cooling jacket placed around the tubular reactor. The available literature information on Bunsen reactor design and operation, certain support equipment needed for process operation and a design that meet the specification of Matt Channon Consulting are presented.

Moore, Robert Charles

2013-10-01T23:59:59.000Z

262

Fuel handling system for a nuclear reactor  

DOE Patents (OSTI)

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

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

1986-01-01T23:59:59.000Z

263

Closure head for a nuclear reactor  

DOE Patents (OSTI)

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

Wade, Elman E. (South Huntingdon, PA)

1980-01-01T23:59:59.000Z

264

Nuclear reactor insulation and preheat system  

DOE Patents (OSTI)

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

Wampole, Nevin C. (Latrobe, PA)

1978-01-01T23:59:59.000Z

265

Nuclear reactor flow control method and apparatus  

DOE Patents (OSTI)

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

Church, J.P.

1993-03-30T23:59:59.000Z

266

Nuclear reactor flow control method and apparatus  

DOE Patents (OSTI)

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

Church, J.P.

1991-04-23T23:59:59.000Z

267

Neutrino Oscillation Experiments at Nuclear Reactors  

E-Print Network (OSTI)

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

Giorgio Gratta

1999-05-06T23:59:59.000Z

268

Nuclear Thermal Rockets: The Physics of the Fission Reactor  

E-Print Network (OSTI)

Nuclear Thermal Rockets: The Physics of the Fission Reactor Shane D. Ross Control and Dynamical combustion are those powered by nuclear fission. Comparison of Chemical and Nuclear Rockets. Most existent.g., hydrogen and oxygen). In a nuclear rocket, or more precisely, a nuclear thermal rocket, the propellant

Ross, Shane

269

Ground test facility for nuclear testing of space reactor subsystems  

SciTech Connect

Two major reactor facilities at the INEL have been identified as easily adaptable for supporting the nuclear testing of the SP-100 reactor subsystem. They are the Engineering Test Reactor (ETR) and the Loss of Fluid Test Reactor (LOFT). In addition, there are machine shops, analytical laboratories, hot cells, and the supporting services (fire protection, safety, security, medical, waste management, etc.) necessary to conducting a nuclear test program. This paper presents the conceptual approach for modifying these reactor facilities for the ground engineering test facility for the SP-100 nuclear subsystem. 4 figs.

Quapp, W.J.; Watts, K.D.

1985-01-01T23:59:59.000Z

270

Liquid metal pump for nuclear reactors  

DOE Patents (OSTI)

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

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

1975-10-01T23:59:59.000Z

271

Measuring Neutrino Oscillations with Nuclear Reactors  

SciTech Connect

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

McKeown, R. D. [W. K. Kellogg Radiation Laboratory, California Institute of Technology, Pasadena, CA (United States)

2007-10-26T23:59:59.000Z

272

Nuclear reactor containment spray testing system. [PWR  

SciTech Connect

Disclosed is a method for periodic testing of a spray system in a nuclear reactor containment. The method includes injecting a gas into the spray system such that a temperature differential exists between the gas and the containment atmosphere. Scanning the gas jet discharged from the spray nozzles with infrared apparatus then provides a real-time thermal image on a monitor, such as a cathode ray tube, and detects any partially or completely blocked nozzles in the spray system. The scanning may be performed from the containment operating deck. 1 claim, 4 figures.

Rubin, K.

1978-01-10T23:59:59.000Z

273

Incorporation of Hydride Nuclear Fuels in Commercial Light Water Reactors  

E-Print Network (OSTI)

experience in the nuclear fuels field. I am also extremelyreactor core components, nuclear fuel-element design hasreactors, commercial nuclear fuel still consists of uranium

Terrani, Kurt Amir

2010-01-01T23:59:59.000Z

274

More About NNSA's Naval Reactors Office | National Nuclear Security...  

National Nuclear Security Administration (NNSA)

to skip to the main content Facebook Flickr RSS Twitter YouTube More About NNSA's Naval Reactors Office | National Nuclear Security Administration Our Mission Managing the...

275

TABLE 1. Nuclear Reactor, State, Type, Net Capacity ...  

U.S. Energy Information Administration (EIA)

Nuclear Reactor, State, Type, Net Capacity, ... Quad Cities Generating Station River Bend San Onofre Seabrook Sequoyah South Texas Project St Lucie ...

276

Characterization of Nuclear Reactor Materials and Components with ...  

Science Conference Proceedings (OSTI)

Mar 6, 2013 ... Characterization of Nuclear Reactor Materials and Components with ... Results are discussed in terms of existing theoretical models for hydride ...

277

Light Water Reactor Materials for Commercial Nuclear Power ...  

Science Conference Proceedings (OSTI)

Presentation Title, Light Water Reactor Materials for Commercial Nuclear ... First- Principles Theory of Magnetism, Crystal Field and Phonon Spectrum of UO2.

278

Nuclear Reactor Materials at the Atomic Scale - Programmaster.org  

Science Conference Proceedings (OSTI)

Presentation Title, Nuclear Reactor Materials at the Atomic Scale ... Study of the Interaction of Solutes with Interfaces in Iron Using Density-Functional Theory.

279

Fast Reactor Technology - Reactors designed/built by Argonne...  

NLE Websites -- All DOE Office Websites (Extended Search)

Nuclear Safety Materials Disposition Decontamination & Decommissioning Nuclear Criticality Safety Nuclear Data Program Nuclear Waste Form Modeling Departments Engineering...

280

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

E-Print Network (OSTI)

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

Danon, Yaron

Note: This page contains sample records for the topic "nuclear reactor design" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


281

Fluid sampling system for a nuclear reactor  

DOE Patents (OSTI)

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

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

1994-11-22T23:59:59.000Z

282

Fuel performance comparison between Savannah River reactors and the US commercial nuclear reactors  

SciTech Connect

This document provides a review of fuel/target performance of the Savannah River Reactors which was made to compare their in-core performance with that of the commercial nuclear reactors in the US.

Paik, I.K.; Ellison, P.G.

1989-01-01T23:59:59.000Z

283

DOE/NNSA perspective safeguard by design: GEN III/III+ light water reactors and beyond  

SciTech Connect

An overview of key issues relevant to safeguards by design (SBD) for GEN III/IV nuclear reactors is provided. Lessons learned from construction of typical GEN III+ water reactors with respect to SBD are highlighted. Details of SBD for safeguards guidance development for GEN III/III+ light water reactors are developed and reported. This paper also identifies technical challenges to extend SBD including proliferation resistance methodologies to other GEN III/III+ reactors (except HWRs) and GEN IV reactors because of their immaturity in designs.

Pan, Paul Y [Los Alamos National Laboratory

2010-12-10T23:59:59.000Z

284

Basic and Applied Science Research Reactors - Reactors designed...  

NLE Websites -- All DOE Office Websites (Extended Search)

BORAX-III lighting Arco, Idaho (Press Release) Heavy Water and Graphite Reactors Fast Reactor Technology Integral Fast Reactor Argonne Reactor Tree CP-1 70th Anniversary CP-1 70th...

285

Minimizing or eliminating refueling of nuclear reactor  

DOE Patents (OSTI)

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

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

1989-01-01T23:59:59.000Z

286

Electrochemistry of Water-Cooled Nuclear Reactors  

DOE Green Energy (OSTI)

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

Dgiby Macdonald; Mirna Urquidi-Macdonald; John Mahaffy, Amit Jain, Han Sang Kim, Vishisht Gupta; Jonathan Pitt

2006-08-08T23:59:59.000Z

287

Nuclear Archeology for CANDU Power Reactors  

SciTech Connect

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

Broadhead, Bryan L [ORNL

2011-01-01T23:59:59.000Z

288

DOI Designates B Reactor at DOE's Hanford Site as a National Historic  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

DOI Designates B Reactor at DOE's Hanford Site as a National DOI Designates B Reactor at DOE's Hanford Site as a National Historic Landmark DOI Designates B Reactor at DOE's Hanford Site as a National Historic Landmark August 25, 2008 - 3:20pm Addthis DOE to offer regular public tours in 2009 WASHINGTON, DC - U.S. Department of the Interior (DOI) Deputy Secretary Lynn Scarlett and U.S. Department of Energy (DOE) Acting Deputy Secretary Jeffrey F. Kupfer today announced the designation of DOE's B Reactor as a National Historic Landmark and unveiled DOE's plan for a new public access program to enable American citizens to visit B Reactor during the 2009 tourist season. The B Reactor at DOE's Hanford Site in southeast Washington State was the world's first industrial-scale nuclear reactor and produced plutonium for the atomic weapon that was dropped on Nagasaki,

289

DOI Designates B Reactor at DOE's Hanford Site as a National Historic  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

DOI Designates B Reactor at DOE's Hanford Site as a National DOI Designates B Reactor at DOE's Hanford Site as a National Historic Landmark DOI Designates B Reactor at DOE's Hanford Site as a National Historic Landmark August 25, 2008 - 3:20pm Addthis DOE to offer regular public tours in 2009 WASHINGTON, DC - U.S. Department of the Interior (DOI) Deputy Secretary Lynn Scarlett and U.S. Department of Energy (DOE) Acting Deputy Secretary Jeffrey F. Kupfer today announced the designation of DOE's B Reactor as a National Historic Landmark and unveiled DOE's plan for a new public access program to enable American citizens to visit B Reactor during the 2009 tourist season. The B Reactor at DOE's Hanford Site in southeast Washington State was the world's first industrial-scale nuclear reactor and produced plutonium for the atomic weapon that was dropped on Nagasaki,

290

Nuclear reactor core and fuel element therefor  

SciTech Connect

This patent describes a nuclear reactor core. This core consists of vertical columns of disengageable fuel elements stacked one atop another. These columns are arranged in side-by-side relationship to form a substantially continuous horizontal array. Each of the fuel elements include a block of refractory material having relatively good thermal conductivity and neutron moderating characteristics. The block has a pair of parallel flat top and bottom end faces and sides which are substantially prependicular to the end faces. The sides of each block is aligned vertically within a vertical column, with the sides of vertically adjacent blocks. Each of the blocks contains fuel chambers, including outer rows containing only fuel chambers along the sides of the block have nuclear fuel material disposed in them. The blocks also contain vertical coolant holes which are located inside the fuel chambers in the outer rows and the fuel chambers which are not located in the outer rows with the fuel chambers and which extend axially completely through from end face to end face and form continuous vertical intracolumn coolant passageways in the reactor core. The blocks have vertical grooves extending along the sides of the blocks form interblock channels which align in groups to form continuous vertical intercolumn coolant passsageways in the reactor core. The blocks are in the form of a regular hexagonal prism with each side of the block having vertical gooves defining one half of one of the coolant interblock channels, six corner edges on the blocks have vertical groves defining one-third of an interblock channel, the vertical sides of the blocks defining planar vertical surfaces.

Fortescue, P.

1986-02-11T23:59:59.000Z

291

Office of Nuclear Facility Basis & Facility Design  

NLE Websites -- All DOE Office Websites (Extended Search)

Office of Nuclear Safety Basis & Facility Design(HS-31) Reports to the Office of Nuclear Safety About Us The Office of Nuclear Safety Basis & Facility Design establishes safety...

292

Nuclear safety as applied to space power reactor systems  

SciTech Connect

To develop a strategy for incorporating and demonstrating safety, it is necessary to enumerate the unique aspects of space power reactor systems from a safety standpoint. These features must be differentiated from terrestrial nuclear power plants so that our experience can be applied properly. Some ideas can then be developed on how safe designs can be achieved so that they are safe and perceived to be safe by the public. These ideas include operating only after achieving a stable orbit, developing an inherently safe design, ''designing'' in safety from the start and managing the system development (design) so that it is perceived safe. These and other ideas are explored further in this paper.

Cummings, G.E.

1987-01-01T23:59:59.000Z

293

A New Nuclear Reactor Neutrino Experiment to Measure theta 13  

E-Print Network (OSTI)

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

K. Anderson

2004-02-26T23:59:59.000Z

294

Determination of parameters of a nuclear reactor through noise measurements  

DOE Patents (OSTI)

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

Cohn, C.E.

1975-07-15T23:59:59.000Z

295

PRISM; The plant design concept for the U. S. advanced liquid metal reactor program  

SciTech Connect

The US program for development of an advanced liquid metal reactor (ALMR) is proceeding into a new phase of focused design development. This new phase started at the beginning of 1989; its objective is to complete the conceptual design of the US ALMR, with supporting key feature tests, sufficiently to enter a more detailed design phase and subsequent construction of a prototype reactor plant. A project goal is to demonstrate by actual performance of the reactor its passive, inherent safety features and thereby provide the technical basis for certification of the design by the Nuclear Regulatory Commission (NRC). This paper reports on the PRISM (power reactor inherently safe module) reactor concept which in combination with the IFR (integral fast reactor) metal fuel cycle being developed by Argonne National Laboratory, was selected by DOE in 1988 as the reference design for the US ALMR program.

Berglund, R.C.; Tippets, F.E. (GE Nuclear Energy, Advance Nuclear Technology, San Jose, CA (US))

1989-01-01T23:59:59.000Z

296

University Research Reactor Task Force to the Nuclear Energy Research  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

University Research Reactor Task Force to the Nuclear Energy University Research Reactor Task Force to the Nuclear Energy Research Advisory Committee University Research Reactor Task Force to the Nuclear Energy Research Advisory Committee In mid-February, 2001 The University Research Reactor (URR) Task Force (TF), a sub-group of the Department of Energy (DOE) Nuclear Energy Research Advisory Committee (NERAC), was asked to: * Analyze information collected by DOE, the NERAC "Blue Ribbon Panel," universities, and other sources pertaining to university reactors including their research and training capabilities, costs to operate, and operating data, and * Provide DOE with clear, near-term recommendations as to actions that should be taken by the Federal Government and a long-term strategy to assure the continued operation of vital university reactor facilities in

297

Design studies of mirror machine reactors  

SciTech Connect

An overview is presented of a mirror fusion reactor design study. The general methodology used in the study is discussed, the reactor is described, and some design alternatives to the present approach are enumerated. The system chosen for this design study is a mirror machine with direct conversion using D- T fuel. The nominal power output is 200 MW. The coil geometry is the Yin Yang, minimum B with a vacuum mirror ratio of 3. The coil is of particular utility because of its simple conductor shapes and because the two separate conductors, by proper B-field biasing, allow the charged particles to escape preferentially through one mirror only and through a relatively small window'' of that mirror. This is necessary for direct converter economy. (auth)

Werner, R.W.; Carlson, G.A.; Hovingh, J.; Lee, J.D.; Peterson, M.A.

1973-12-01T23:59:59.000Z

298

Modeling and Simulation for Nuclear Reactors Hub | Department of Energy  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Modeling and Simulation for Nuclear Reactors Hub Modeling and Simulation for Nuclear Reactors Hub Modeling and Simulation for Nuclear Reactors Hub August 1, 2010 - 4:20pm Addthis Scientists and engineers are working to help the nuclear industry make reactors more efficient through computer modeling and simulation. Scientists and engineers are working to help the nuclear industry make reactors more efficient through computer modeling and simulation. The Department's Energy Innovation Hubs are helping to advance promising areas of energy science and engineering from the earliest stages of research to the point of commercialization where technologies can move to the private sector by bringing together leadings scientists to collaborate on critical energy challenges. The Energy Innovation Hubs aim to develop innovation through a unique

299

Spent nuclear fuel discharges from U.S. reactors 1994  

Science Conference Proceedings (OSTI)

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

NONE

1996-02-01T23:59:59.000Z

300

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

DOE Green Energy (OSTI)

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.

Buden, D.

1993-08-11T23:59:59.000Z

Note: This page contains sample records for the topic "nuclear reactor design" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


301

Statement on Defense Nuclear Nonproliferation and Naval Reactors Activities  

National Nuclear Security Administration (NNSA)

Defense Nuclear Nonproliferation and Naval Reactors Activities Defense Nuclear Nonproliferation and Naval Reactors Activities before the House Committee on Appropriations Subcommittee on Energy & Water Development | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure Continuing Management Reform Countering Nuclear Terrorism About Us Our Programs Our History Who We Are Our Leadership Our Locations Budget Our Operations Media Room Congressional Testimony Fact Sheets Newsletters Press Releases Speeches Events Social Media Video Gallery Photo Gallery NNSA Archive Federal Employment Apply for Our Jobs Our Jobs Working at NNSA Blog Home > Media Room > Congressional Testimony > Statement on Defense Nuclear

302

Statement on Defense Nuclear Nonproliferation and Naval Reactors Activities  

NLE Websites -- All DOE Office Websites (Extended Search)

Defense Nuclear Nonproliferation and Naval Reactors Activities Defense Nuclear Nonproliferation and Naval Reactors Activities before the House Committee on Appropriations Subcommittee on Energy & Water Development | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure Continuing Management Reform Countering Nuclear Terrorism About Us Our Programs Our History Who We Are Our Leadership Our Locations Budget Our Operations Media Room Congressional Testimony Fact Sheets Newsletters Press Releases Speeches Events Social Media Video Gallery Photo Gallery NNSA Archive Federal Employment Apply for Our Jobs Our Jobs Working at NNSA Blog Home > Media Room > Congressional Testimony > Statement on Defense Nuclear

303

Nuclear reactors built, being built, or planned 1992  

Science Conference Proceedings (OSTI)

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

Not Available

1993-07-01T23:59:59.000Z

304

Control rod for a nuclear reactor  

DOE Patents (OSTI)

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

Roman, Walter G. (Pittsburgh, PA); Sutton, Jr., Harry G. (Pittsburgh, PA)

1979-01-01T23:59:59.000Z

305

FUEL ELEMENT FOR A NUCLEAR REACTOR  

DOE Patents (OSTI)

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

Duffy, J.G. Jr.

1961-05-30T23:59:59.000Z

306

FUEL ELEMENT FOR A NUCLEAR REACTOR  

DOE Patents (OSTI)

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

Duffy, J.G. Jr.

1961-05-30T23:59:59.000Z

307

Nuclear reactor fuel rod attachment system  

DOE Patents (OSTI)

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

Christiansen, David W. (Kennewick, WA)

1982-01-01T23:59:59.000Z

308

Nuclear reactor cooling system decontamination reagent regeneration  

DOE Patents (OSTI)

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

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

1985-01-01T23:59:59.000Z

309

Uncertainties in the Anti-neutrino Production at Nuclear Reactors  

Science Conference Proceedings (OSTI)

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

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

2008-08-06T23:59:59.000Z

310

SPRING DRIVEN ACTUATING MECHANISM FOR NUCLEAR REACTOR CONTROL  

DOE Patents (OSTI)

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

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

1962-01-23T23:59:59.000Z

311

Weld monitor and failure detector for nuclear reactor system  

DOE Patents (OSTI)

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

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

1987-01-01T23:59:59.000Z

312

Energy Secretary to Visit Georgia Nuclear Reactor Site and Tennessee  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Secretary to Visit Georgia Nuclear Reactor Site and Secretary to Visit Georgia Nuclear Reactor Site and Tennessee Laboratory to Highlight Administration Support for Nuclear Energy Energy Secretary to Visit Georgia Nuclear Reactor Site and Tennessee Laboratory to Highlight Administration Support for Nuclear Energy February 13, 2012 - 6:16pm Addthis WASHINGTON, D.C. - U.S. Secretary of Energy Secretary Steven Chu will visit the Vogtle nuclear power plant in Waynesboro, Georgia, and Oak Ridge National Laboratory on Wednesday, February 15 to highlight steps the Obama Administration is taking to restart America's nuclear energy industry. In Waynesboro, Secretary Chu will join Southern Company CEO Thomas A. Fanning, Georgia Power CEO W. Paul Bowers, and local leaders for a tour of Vogtle units 3 and 4 -- the site of the first two new nuclear power units

313

Independent Confirmatory Survey Report for the University of Arizona Nuclear Reactor Laboratory, Tucson, Arizona  

SciTech Connect

The University of Arizona (University) research reactor is a TRIGA swimming pool type reactor designed by General Atomics and constructed at the University in 1958. The reactor first went into operation in December of 1958 under U.S. Nuclear Regulatory Commission (NRC) license R-52 until final shut down on May 18, 2010. Initial site characterization activities were conducted in February 2009 during ongoing reactor operations to assess the radiological status of the Nuclear Reactor Laboratory (NRL) excluding the reactor tank, associated components, and operating systems. Additional post-shutdown characterization activities were performed to complete characterization activities as well as verify assumptions made in the Decommissioning Plan (DP) that were based on a separate activation analysis (ESI 2009 and WMG 2009). Final status survey (FSS) activities began shortly after the issuance of the FSS plan in May 2011. The contractor completed measurement and sampling activities during the week of August 29, 2011.

Nick A. Altic

2011-11-11T23:59:59.000Z

314

Overview of Sandia National Laboratories pulse nuclear reactors  

SciTech Connect

Sandia National Laboratories has designed, constructed and operated bare metal Godiva-type and pool-type pulse reactors since 1961. The reactor facilities were designed to support a wide spectrum of research, development, and testing activities associated with weapon and reactor systems.

Schmidt, T.R. [Sandia National Labs., Albuquerque, NM (United States); Reuscher, J.A. [Texas A& M Univ., College Station, TX (United States)

1994-10-01T23:59:59.000Z

315

FUNDAMENTALS IN THE OPERATION OF NUCLEAR TEST REACTORS. VOLUME 1. REACTOR SCIENCE AND TECHNOLOGY  

SciTech Connect

A resume of nuclear physics basic to reactor operation precedes discussion of aspects of reactor physics, engineering, chemistry, metallurgy, instrumentation, control, kinetics, and safety. The object is to provide an approach to and understanding of problems in irradiation test programs in the Materials Testing and Engineering Test Reactors. (D.C.W.)

1963-06-01T23:59:59.000Z

316

SELF-REGULATING BOILING-WATER NUCLEAR REACTORS  

DOE Patents (OSTI)

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

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

1960-08-16T23:59:59.000Z

317

Development of NERVA reactor for space nuclear propulsion  

Science Conference Proceedings (OSTI)

The general technology development and demonstration of a Nuclear Engine for Rocket Vehicle Application (NERVA), a joint AEC-NASA program, was undertaken successfully in the 1960's and terminated in 1971 for lack of a specific mission. Detailed flight engine specifications were defined and several candidate designs which would satisfy these specifications were completed just prior to termination of these efforts. However, the technology interest continued and efforts were extended during the early 1970's to consider space power applications including a manned Mars mission and dual mode (propulsion power and electrical power) operation. Subsequent efforts have continued in developing electric power applications. Light-weight solid core reactor nuclear power sources have been conceptually studied based upon this technology. This paper provides a short summary of the technology that evolved in this very complex and frequently changing program with some specific references to the Mars mission propulsion application as it evolved from the NERVA development program.

Holman, R.R.; Pierce, B.L.

1986-01-01T23:59:59.000Z

318

Maria Research Reactor loaded with LEU - Otwock, Poland | National Nuclear  

National Nuclear Security Administration (NNSA)

Maria Research Reactor loaded with LEU - Otwock, Poland | National Nuclear Maria Research Reactor loaded with LEU - Otwock, Poland | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure Continuing Management Reform Countering Nuclear Terrorism About Us Our Programs Our History Who We Are Our Leadership Our Locations Budget Our Operations Media Room Congressional Testimony Fact Sheets Newsletters Press Releases Speeches Events Social Media Video Gallery Photo Gallery NNSA Archive Federal Employment Apply for Our Jobs Our Jobs Working at NNSA Blog Home > Media Room > Video Gallery > Maria Research Reactor loaded with LEU - ... Maria Research Reactor loaded with LEU - Otwock, Poland Maria Research Reactor loaded with LEU - Otwock, Poland

319

An optimality condition for the assembly distribution in a nuclear reactor Laurent Thevenot  

E-Print Network (OSTI)

We give here an optimality condition for the optimization problem of the assembly distribution in a nuclear reactor, by using the homogenization method. In this paper the reactivity of the reactor core is measured by the critical eigenvalue for both continuous and multigroup neutron transport modelings. In particular, we extend the spectral theory of the critical eigenvalue and prove the dierentiability of this latter with respect to the design parameter, the con guration of the fuels. MS Classi cation: 49K20, 49J50, 35P05, 35F15. Key words: Neutron Transport, Nuclear Reactor, Optimization, Optimality Condition, Homogenization, Critical Eigenvalue, Eigenvalue Derivative.

Departement Of Applied; Laurent Thevenot

2003-01-01T23:59:59.000Z

320

CONSTRUCTION OF WEB-ACCESSIBLE MATERIALS HANDBOOK FORGENERATION IV NUCLEAR REACTORS  

Science Conference Proceedings (OSTI)

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

Ren, Weiju [ORNL

2005-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "nuclear reactor design" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


321

Nuclear reactors built, being built, or planned, 1991  

Science Conference Proceedings (OSTI)

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

Simpson, B.

1992-07-01T23:59:59.000Z

322

System aspects of a Space Nuclear Reactor Power System  

Science Conference Proceedings (OSTI)

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.

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

1988-01-01T23:59:59.000Z

323

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

Science Conference Proceedings (OSTI)

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.

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

1990-12-01T23:59:59.000Z

324

BOILING NUCLEAR SUPERHEATER (BONUS) POWER STATION. Final Summary Design Report  

SciTech Connect

The design and construction of the Boiling Nuclear Superheater (BONUS) Power Station at Punta Higuera on the seacoast at the westernmost tip of Puerto Rico are described. The reactor has an output of 17.5 Mw(e). This report will serve as a source of information for personnel engaged in management, evaluation, and training. (N.W.R.)

1962-05-01T23:59:59.000Z

325

First Step to Spur U.S. Manufacturing of Small Modular Nuclear Reactors |  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

First Step to Spur U.S. Manufacturing of Small Modular Nuclear First Step to Spur U.S. Manufacturing of Small Modular Nuclear Reactors First Step to Spur U.S. Manufacturing of Small Modular Nuclear Reactors January 25, 2012 - 5:06pm Addthis Brenda DeGraffenreid The Energy Department recently announced the first step toward manufacturing small modular nuclear reactors (SMRs) in the United States, demonstrating the Administration's commitment to advancing U.S. manufacturing leadership in low-carbon, next generation energy technologies and restarting the nation's nuclear industry. The release of a draft Funding Opportunity Announcement (FOA) last week presents supply-chain procurement opportunities for our nation's small businesses down the line, as industry provides input in advance of a full FOA on engineering, design certification, and licensing through a

326

How Brazil spun the atom [nuclear power reactors  

Science Conference Proceedings (OSTI)

This paper describes the Resende nuclear complex in Brazil which will house hundreds of uranium centrifuges to produce enriched uranium that will fuel its nuclear power reactors. By consistently fulfilling its obligations as a party to the Nuclear Non-Proliferation ...

E. Guizzo

2006-03-01T23:59:59.000Z

327

100-MW NUCLEAR POWER PLANT UTILIZING A SODIUM COOLED, GRAPHITE MODERATED REACTOR  

SciTech Connect

The conceptual design of a 100 Mw(e) nuclear power plant is described. The plant utilized a sodium-cooled graphite-moderated reactor with stainless- steel clad. slightiy enriched UO/sub 2/ fuel. The reactor is provided with three main coolant circuits, and the steam cycle has three stages of regenerative heating. The plant control system allows automatic operation over the range of 20 to 100% load, or manual operation at all loads. The site, reactor, sodium systems, reactor auxiliaries, fuel handling, instrumentation, turbine-generator, buildings. and safety measures are described. Engineering drawings are included. (W.D.M.)

1958-02-28T23:59:59.000Z

328

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

Science Conference Proceedings (OSTI)

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

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

2012-09-01T23:59:59.000Z

329

DESIGN OF A TOKAMAK FUSION REACTOR FIRST WALL ARMOR AGAINST NEUTRAL BEAM IMPINGEMENT  

E-Print Network (OSTI)

of Niobium," BNES Nuclear Fusion Reactor Conference, CulhamWall Erosion in Fusion Reactors," Nuclear Fusion. g. 31.and Reactors," Fifth Conference Pro- ceedings on Plasma Physics and Controlled Nuclear Fusion

Myers, Richard Allen

2011-01-01T23:59:59.000Z

330

EEE 562 Nuclear Reactor Theory and Design (3 hrs) Catalog Description: Principles of neutron chain reacting systems. Neutron diffusion and moderation.  

E-Print Network (OSTI)

important part of reactor physics. As the neutrons are produced in neutron induced fission of a fissile physical processes like neutron capture, elastic and inelastic scattering, upscattering, etc. Neutron technique is a scattering of free neutrons by matter. It is used in biophysics, physics, chemistry

331

Application of Nuclear Energy for Seawater Desalination: Design Concepts of Nuclear Desalination Plants  

SciTech Connect

Nuclear energy is playing an important role in electricity generation, producing 16% of the world's electricity. However, most of the world's energy consumption is in the form of heat, in which case nuclear energy could also play an important role. In particular, process heat for seawater desalination using nuclear energy has been of growing interest to some Member States of the International Atomic Energy Agency over the past two decades. This growing interest stems from increasingly acute freshwater shortages in many arid and semi-arid zones around the world. Indeed, several national and international nuclear desalination demonstration programs are already under way or being planned. Of particular interest are projects for seawater nuclear desalination plants in coastal regions, where saline feed water can serve the dual purpose of cooling water for the nuclear reactor and as feed water for the desalination plant. In principle any nuclear reactor can provide energy (low-grade heat and/or electricity), as required by desalination processes. However, there are some additional requirements to be met under specific conditions in order to introduce nuclear desalination. Technical issues include meeting more stringent safety requirements (nuclear reactors themselves and nuclear-desalination integrated complexes in particular), and performance improvement of the integrated systems. Economic competitiveness is another important factor to be considered for a broader deployment of nuclear desalination. For technical robustness and economic competitiveness a number of design variants of coupling configurations of nuclear desalination integrated plant concepts are being evaluated. This paper identifies and discusses various factors, which support the attractiveness of nuclear desalination. It further summarizes some of the key approaches recommended for nuclear desalination complex design and gives an overview of various design concepts of nuclear desalination plants, which are experienced, being implemented or evaluated by several IAEA Member States. Operating experience with nuclear desalination using a liquid-metal reactor BN-350 in Kazakhstan and several Pressurized Water Reactor units in Japan is discussed. Results of economic analyses of nuclear desalination are also presented in order to illustrate the competitiveness of this technology with other conventional desalination operations. (authors)

Faibish, R.S.; Konishi, T.; Gasparini, M. [International Atomic Energy Agency - IAEA, P.O. Box 100, Wagramer Strasse 5, A-1400 Vienna (Austria)

2002-07-01T23:59:59.000Z

332

Current Projects for Reactor Physics and Fuel Cycle Analysis...  

NLE Websites -- All DOE Office Websites (Extended Search)

Nuclear Systems Modeling and Design Analysis > Reactor Physics and Fuel Cycle Analysis > Current Projects Capabilities Nuclear Systems Modeling and Design Analysis Reactor Physics...

333

Research Reactor Conversion | National Nuclear Security Administration  

National Nuclear Security Administration (NNSA)

Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure...

334

Passive cooling safety system for liquid metal cooled nuclear reactors  

DOE Patents (OSTI)

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

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

1991-01-01T23:59:59.000Z

335

Indirect passive cooling system for liquid metal cooled nuclear reactors  

DOE Patents (OSTI)

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

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

1990-01-01T23:59:59.000Z

336

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

SciTech Connect

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

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

1994-01-01T23:59:59.000Z

337

Safety and core design of large liquid-metal cooled fast breeder reactors  

E-Print Network (OSTI)

well known from basic reactor theory, the flux distributionof a fast reactor using the perturbation theory”. In: Atomicbeam theory and are not specific to a nuclear reactor core.

Qvist, Staffan Alexander

2013-01-01T23:59:59.000Z

338

Idaho Site Obtains Patent for Nuclear Reactor Sodium Cleanup Treatment |  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Obtains Patent for Nuclear Reactor Sodium Cleanup Obtains Patent for Nuclear Reactor Sodium Cleanup Treatment Idaho Site Obtains Patent for Nuclear Reactor Sodium Cleanup Treatment March 28, 2013 - 12:00pm Addthis CWI engineers Jeff Jones, David Tolman, right, and Kirk Dooley (seated) developed a treatment to safely dissolve a bicarbonate crust and treat and remove the sodium in the Experimental Breeder Reactor-II at the Idaho site. CWI engineers Jeff Jones, David Tolman, right, and Kirk Dooley (seated) developed a treatment to safely dissolve a bicarbonate crust and treat and remove the sodium in the Experimental Breeder Reactor-II at the Idaho site. Piping in the east boiler basement of the sodium processing building was color coded for easy identification. Orange indicates sodium and green identifies cooling water.

339

Nuclear reactors built, being built, or planned 1996  

Science Conference Proceedings (OSTI)

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

NONE

1997-08-01T23:59:59.000Z

340

Idaho Site Obtains Patent for Nuclear Reactor Sodium Cleanup Treatment |  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Idaho Site Obtains Patent for Nuclear Reactor Sodium Cleanup Idaho Site Obtains Patent for Nuclear Reactor Sodium Cleanup Treatment Idaho Site Obtains Patent for Nuclear Reactor Sodium Cleanup Treatment March 28, 2013 - 12:00pm Addthis CWI engineers Jeff Jones, David Tolman, right, and Kirk Dooley (seated) developed a treatment to safely dissolve a bicarbonate crust and treat and remove the sodium in the Experimental Breeder Reactor-II at the Idaho site. CWI engineers Jeff Jones, David Tolman, right, and Kirk Dooley (seated) developed a treatment to safely dissolve a bicarbonate crust and treat and remove the sodium in the Experimental Breeder Reactor-II at the Idaho site. Piping in the east boiler basement of the sodium processing building was color coded for easy identification. Orange indicates sodium and green identifies cooling water.

Note: This page contains sample records for the topic "nuclear reactor design" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


341

ENGINEERING STUDY ON UNDERGROUND CONSTRUCTION OF NUCLEAR POWER REACTORS  

SciTech Connect

The advantages, disadvantages, and cost of constructing a auclear power reactor underground are outlinedData on underground construction of hydroelectric plants, other structures, and underground reactor projects in Norway and Sweden are reviewed. A hypothetical underground Experimental Boiling Water Reactor design and sketch are given with cost estimates(T.R.H.)

Beck, C.

1958-04-15T23:59:59.000Z

342

Incorporation of Hydride Nuclear Fuels in Commercial Light Water Reactors  

E-Print Network (OSTI)

52] J.H. Rust. Nuclear Power Plant Engineering. Buchanan,the economics of nuclear power plants. In addition, the longin commercial nuclear power plants. The fuel designs and

Terrani, Kurt Amir

2010-01-01T23:59:59.000Z

343

Fuel assembly transfer basket for pool type nuclear reactor vessels  

DOE Patents (OSTI)

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

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

1991-01-01T23:59:59.000Z

344

Observer-based fault detection for nuclear reactors  

E-Print Network (OSTI)

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

Li, Qing, 1972-

2001-01-01T23:59:59.000Z

345

Nuclear Engineering and Design 242 (2012) 1925 Contents lists available at SciVerse ScienceDirect  

E-Print Network (OSTI)

2011 Accepted 14 September 2011 a b s t r a c t Graphite is a widely used material in nuclear reactors of creep data in reactor design. In: 1st International Nuclear Graphite Specialists Meeting (INGSM), Oak.J.E., 1974. Theory of irradiation creep in reactor graphite--dislocation pinning­unpinning model. Carbon 12

Martin, Ralph R.

346

Liquid metal cooled nuclear reactors with passive cooling system  

SciTech Connect

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

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

1991-01-01T23:59:59.000Z

347

Nuclear reactors built, being built, or planned: 1995  

Science Conference Proceedings (OSTI)

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

NONE

1996-08-01T23:59:59.000Z

348

Nuclear reactors built, being built, or planned, 1994  

Science Conference Proceedings (OSTI)

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

NONE

1995-07-01T23:59:59.000Z

349

CRC handbook of nuclear reactors calculations. Vol. I  

Science Conference Proceedings (OSTI)

This handbook breaks down the complex field of nuclear reactor calculations into major steps. Each step presents a detailed analysis of the problems to be solved, the parameters involved, and the elaborate computer programs developed to perform the calculations. This book bridges the gap between nuclear reactor theory and the implementation of that theory, including the problems to be encountered and the level of confidence that should be given to the methods described.

Ronen, Y.

1986-01-01T23:59:59.000Z

350

NUCLEAR TRAINING REACTOR. Preliminary Report and Hazards Analysis  

SciTech Connect

Complete descriptions of the reactor, building, and site are given. Reactor operation, bazards associated with the reactor, and the maximum credible accident are discussed. The reactor is a BSR swimming pool type with a design power level of 10 kw and a maximum thermal flux in the fuel region of 2.23 x 10/ sup 11/ n/cm/sup 2/-sec. (W.D.M.)

1959-09-01T23:59:59.000Z

351

Radionuclides in United States commercial nuclear power reactors  

SciTech Connect

In the next ten to twenty years, many of the commercial nuclear power reactors in the United States will be reaching their projected lifetime of forty years. As these power plants are decommissioned, it seems prudent to consider the recycling of structural materials such as stainless steel. Some of these materials and components have become radioactive through either nuclear activation of the elements within the components or surface contamination with radioactivity form the operational activities. In order to understand the problems associated with recycling stainless steel from decommissioned nuclear power reactors, it is necessary to have information on the radionuclides expected on or in the contaminated materials. A study has been conducted of radionuclide contamination information that is available for commercial nuclear power reactors in the United States. There are two types of nuclear power reactors in commercial use in the United States, pressurized water reactors (PWRs) and boiling water reactors (BWRs). Before presenting radionuclide activities information, a brief discussion is given on the major components and operational differences for the PWRs and BWRs. Radionuclide contamination information is presented from 11 PWRs and over 8 BWRs. These data include both the radionuclides within the circulating reactor coolant water as well as radionuclide contamination on and within component parts.

Bechtold, T.E. [ed.] [Westinghouse Idaho Nuclear Co., Inc., Idaho Falls, ID (United States); Dyer, N.C. [Oregon Graduate Inst. of Science and Technology, Beaverton, OR (United States)

1994-01-01T23:59:59.000Z

352

Design and Nuclear-Safety Related Simulations of Bare-Pellet Test Irradiations for the Production of Pu-238 in the High Flux Isotope Reactor using COMSOL  

Science Conference Proceedings (OSTI)

The Oak Ridge National Laboratory (ORNL)is developing technology to produce plutonium-238 for the National Aeronautics and Space Administration (NASA) as a power source material for powering vehicles while in deep-space[1]. The High Flux Isotope Reactor (HFIR) of ORNL has been utilized to perform test irradiations of incapsulated neptunium oxide (NpO2) and aluminum powder bare pellets for purposes of understanding the performance of the pellets during irradiation[2]. Post irradiation examinations (PIE) are currently underway to assess the effect of temperature, thermal expansion, swelling due to gas production, fission products, and other phenomena

Freels, James D [ORNL; Jain, Prashant K [ORNL; Hobbs, Randy W [ORNL

2012-01-01T23:59:59.000Z

353

Predicted nuclear heating and temperatures in gas-cooled nuclear reactors for process heat applications  

SciTech Connect

The high-temperature gas-cooled nuclear reactor (HTGR) is an attractive potential source of primary energy for many industrial and chemical process applications. Significant modification of current HTGR core design will be required to achieve the required elevations in exit gas temperatures without exceeding the maximum allowable temperature limits for the fuel material. A preliminary evaluation of the effects of various proposed design modifications by predicting the resulting fuel and gas temperatures with computer calculational modeling techniques is reported. The design modifications evaluated are generally those proposed by the General Atomic Company (GAC), a manufacturer of HTGRs, and some developed at the LASL. The GAC modifications do result in predicted fuel and exit gas temperatures which meet the proposed design objectives. (auth)

Cort, G.E.; Vigil, J.C.; Jiacoletti, R.J.

1975-09-01T23:59:59.000Z

354

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

SciTech Connect

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.

Whitehead, Donnie Wayne; Varnado, G. Bruce

2008-09-01T23:59:59.000Z

355

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

Science Conference Proceedings (OSTI)

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

NONE

1997-10-01T23:59:59.000Z

356

Table 3. Nuclear Reactor Characteristics and Operational ...  

U.S. Energy Information Administration (EIA)

Point Beach Nuclear Plant Quad Cities Generating Station R.E. Ginna Nuclear Power Plant PSEG Salem Generating Station Harris South Texas Project PPL ...

357

Space Nuclear Power Plant Pre-Conceptual Design Report, For Information  

Science Conference Proceedings (OSTI)

This letter transmits, for information, the Project Prometheus Space Nuclear Power Plant (SNPP) Pre-Conceptual Design Report completed by the Naval Reactors Prime Contractor Team (NRPCT). This report documents the work pertaining to the Reactor Module, which includes integration of the space nuclear reactor with the reactor radiation shield, energy conversion, and instrumentation and control segments. This document also describes integration of the Reactor Module with the Heat Rejection segment, the Power Conditioning and Distribution subsystem (which comprise the SNPP), and the remainder of the Prometheus spaceship.

B. Levine

2006-01-27T23:59:59.000Z

358

Advanced High Temperature Reactor Neutronic Core Design  

Science Conference Proceedings (OSTI)

The AHTR is a 3400 MW(t) FHR class reactor design concept intended to serve as a central generating station type power plant. While significant technology development and demonstration remains, the basic design concept appears sound and tolerant of much of the remaining performance uncertainty. No fundamental impediments have been identified that would prevent widespread deployment of the concept. This paper focuses on the preliminary neutronic design studies performed at ORNL during the fiscal year 2011. After a brief presentation of the AHTR design concept, the paper summarizes several neutronic studies performed at ORNL during 2011. An optimization study for the AHTR core is first presented. The temperature and void coefficients of reactivity are then analyzed for a few configurations of interest. A discussion of the limiting factors due to the fast neutron fluence follows. The neutronic studies conclude with a discussion of the control and shutdown options. The studies presented confirm that sound neutronic alternatives exist for the design of the AHTR to maintain full passive safety features and reasonable operation conditions.

Ilas, Dan [ORNL; Holcomb, David Eugene [ORNL; Varma, Venugopal Koikal [ORNL

2012-01-01T23:59:59.000Z

359

Nuclear Energy Enabling Technologies (NEET) Reactor Materials  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Enabling Technologies (NEET) Reactor Materials Enabling Technologies (NEET) Reactor Materials Award Recipient Estimated Award Amount* Award Location Supporting Organizations Project Description University of Nebraska $979,978 Lincoln, NE Massachusetts Institute of Technology (Cambridge, MA), Texas A&M (College Station, TX) Project will explore the development of advanced metal/ceramic composites. These improvements could lead to more efficient production of electricity in advanced reactors. Oak Ridge National Laboratory $849,000 Oak Ridge, TN University of Wisconsin-Madison (Madison, WI) Project will develop novel high-temperature high-strength steels with the help of computational modeling, which could lead to increased efficiency in advanced reactors. Pacific Northwest National Laboratory

360

Materials Challenges in Next Generation Nuclear Reactors  

Science Conference Proceedings (OSTI)

Materials under active consideration for use in different reactor components ... A Theoretical Model of Corrosion Rate Distribution in Liquid LBE Flow Loop at ...

Note: This page contains sample records for the topic "nuclear reactor design" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


361

Heavy Water and Graphite Reactors - Reactors designed/built by...  

NLE Websites -- All DOE Office Websites (Extended Search)

experiments, necessary to achieve higher precision for the determination of reactor power distribution patterns, effect of non-uniform void distributions, kinetic behavior,...

362

Reactivity Control Schemes for Fast Spectrum Space Nuclear Reactors  

Science Conference Proceedings (OSTI)

Several different reactivity control schemes are considered for future space nuclear reactor power systems. Each of these control schemes uses a combination of boron carbide absorbers and/or beryllium oxide reflectors to achieve sufficient reactivity swing to keep the reactor subcritical during launch and to provide sufficient excess reactivity to operate the reactor over its expected 7–15 year lifetime. The size and shape of the control system directly impacts the size and mass of the space reactor's reflector and shadow shield

Aaron E. Craft; Jeffrey C. King

2008-01-01T23:59:59.000Z

363

Fuel leak detection apparatus for gas cooled nuclear reactors  

SciTech Connect

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

Burnette, Richard D. (San Diego, CA)

1977-01-01T23:59:59.000Z

364

Spent nuclear fuel discharges from US reactors 1992  

SciTech Connect

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

Not Available

1994-05-05T23:59:59.000Z

365

TRISO Fuel Performance: Modeling, Integration into Mainstream Design Studies, and Application to a Thorium-fueled Fusion-Fission Hybrid Blanket  

E-Print Network (OSTI)

Pebble-Bed Type Nuclear Reactor. ” Thesis. Delft UniversityFundamental Aspects of Nuclear Reactor Fuel Elements. U.S.technology for various nuclear reactor design projects that

Powers, Jeffrey

2011-01-01T23:59:59.000Z

366

Manhattan Project: Production Reactor (Pile) Design, Met Lab, 1942  

Office of Scientific and Technical Information (OSTI)

Schematic of the X-10 Graphite Reactor, Oak Ridge PRODUCTION REACTOR (PILE) DESIGN Schematic of the X-10 Graphite Reactor, Oak Ridge PRODUCTION REACTOR (PILE) DESIGN (Met Lab, 1942) Events > The Plutonium Path to the Bomb, 1942-1944 Production Reactor (Pile) Design, 1942 DuPont and Hanford, 1942 CP-1 Goes Critical, December 2, 1942 Seaborg and Plutonium Chemistry, 1942-1944 Final Reactor Design and X-10, 1942-1943 Hanford Becomes Operational, 1943-1944 By 1942, scientists had established that some of the uranium exposed to radioactivity in a reactor (pile) would eventually decay into plutonium, which could then be separated by chemical means from the uranium. Important theoretical research on this was ongoing, but the work was scattered at various universities from coast to coast. In early 1942, Arthur Compton arranged for all pile research to be moved to the Met Lab at the University of Chicago.

367

SNIF: A Futuristic Neutrino Probe for Undeclared Nuclear Fission Reactors  

E-Print Network (OSTI)

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

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

2010-01-01T23:59:59.000Z

368

Materials Solutions for the Nuclear Renaissance  

Science Conference Proceedings (OSTI)

Nuclear reactors present a complex, challenging environment where innovations in materials ... Materials design for fast burner reactors and fusion research

369

Nuclear Reactor Safeguards and Monitoring with Antineutrino Detectors  

E-Print Network (OSTI)

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

Adam Bernstein; Yifang Wang; Giorgio Gratta; Todd West

2001-08-01T23:59:59.000Z

370

Development of a Scatter Search Optimization Algorithm for Boiling Water Reactor Fuel Lattice Design  

Science Conference Proceedings (OSTI)

Technical Paper / Mathematics and Computation, Supercomputing, Reactor Physics and Nuclear and Biological Applications

Juan-Luis François; Cecilia Martín-del-Campo; Luis B. Morales; Miguel-Angel Palomera

371

Design Basis Threat | National Nuclear Security Administration  

National Nuclear Security Administration (NNSA)

Design Basis Threat | National Nuclear Security Administration Design Basis Threat | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure Continuing Management Reform Countering Nuclear Terrorism About Us Our Programs Our History Who We Are Our Leadership Our Locations Budget Our Operations Media Room Congressional Testimony Fact Sheets Newsletters Press Releases Speeches Events Social Media Video Gallery Photo Gallery NNSA Archive Federal Employment Apply for Our Jobs Our Jobs Working at NNSA Blog Design Basis Threat Home > About Us > Our Programs > Nuclear Security > Design Basis Threat Design Basis Threat NNSA has taken aggressive action to improve the security of its nuclear weapons material (often referred to as special nuclear material, or SNM)

372

HIGH FLUX ISOTOPE REACTOR PRELIMINARY DESIGN STUDY  

SciTech Connect

A comparison of possible types of research reactors for the production of transplutonium elements and other isotopes indicates that a flux-trap reactor consisting of a beryllium-reflecteds light-water-cooled annular fuel region surrounding a light-water island provides the required thermal neutron fluxes at minimum cost. The preliminary desigu of such a reactor was carried out on the basis of a parametric study of the effect of dimensions of the island and fuel regions heat removal rates, and fuel loading on the achievable thermal neutmn fluxes in the island and reflector. The results indicate that a 12- to 14-cm- diam. island provides the maximum flux for a given power density. This is in good agreement with the US8R critical experiments. Heat removal calculations indicate that average power densities up to 3.9 Mw/liter are achievable with H/ sub 2/O-cooled, platetype fuel elements if the system is pressurized to 650 psi to prevent surface boiling. On this basis, 100 Mw of heat can be removed from a 14-cm-ID x 36-cm-OD x 30.5-cm-long fuel regions resulting in a thermal neutron flux of 3 x 10/sup 15/ in the island after insertion of 100 g of Cm/sup 244/ or equivalent. The resulting production of Cf/sup 252/ amounts to 65 mg for a 1 1/2- year irradiation. Operation of the reactor at the more conservative level of 67 Mw, providing an irradiation flux of 2 x 10/sup 15/ in the islands will result in the production of 35 mg of Cf/sup 252/ per 18 months from 100 g of Cm/sup 244/. A development program is proposed to answer the question of the feasibility of the higher power operation. In addition to the central irradiation facility for heavyelement productions the HFIR contains ten hydraulic rabbit tubes passing through the beryllium reflector for isotope production and four beam holes for basic research, Preliminary estimates indicate that the cost of the facility, designed for an operating power level of 100 Mw, will be approximately 2 million. (auth)

Lane, J.A.; Cheverton, R.D.; Claiborne, G.C.; Cole, T.E.; Gambill, W.R.; Gill, J.P.; Hilvety, N.; McWherther, J.R.; Vroom, D.W.

1959-03-20T23:59:59.000Z

373

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

DOE Green Energy (OSTI)

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

M. J. Russell

2006-06-01T23:59:59.000Z

374

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

SciTech Connect

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

William Anderson; James Tulenko; Bradley Rearden; Gary Harms

2008-09-11T23:59:59.000Z

375

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

SciTech Connect

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

Wittenbrock, N. G.

1982-01-01T23:59:59.000Z

376

Updated Uranium Fuel Cycle Environmental Impacts for Advanced Reactor Designs  

Science Conference Proceedings (OSTI)

The purpose of this project was to update the environmental impacts from the uranium fuel cycle for select advanced (GEN III+) reactor designs.

Nitschke, R.

2004-10-03T23:59:59.000Z

377

Space Reactor Radiation Shield Design Summary, for Information  

SciTech Connect

The purpose of this letter is to provide a summary of the Prometheus space reactor radiation shield design status at the time of program restructuring.

EC Pheil

2006-02-17T23:59:59.000Z

378

Reactor Design for CO2 Capture Using Sorbents  

NLE Websites -- All DOE Office Websites (Extended Search)

Reactor Design for CO 2 Capture Using Sorbents Background Carbon Sequestration is rapidly becoming accepted as a viable option to reduce the amount of carbon dioxide (CO 2 )...

379

DESIGN AND LAYOUT CONCEPTS FOR COMPACT, FACTORY-PRODUCED, TRANSPORTABLE, GENERATION IV REACTOR SYSTEMS  

SciTech Connect

The purpose of this research project is to develop compact (100 to 400 MWe) Generation IV nuclear power plant design and layout concepts that maximize the benefits of factory-based fabrication and optimal packaging, transportation and siting. The reactor concepts selected were compact designs under development in the 2000 to 2001 period. This interdisciplinary project was comprised of three university-led nuclear engineering teams identified by reactor coolant type (water, gas, and liquid metal) and a fourth Industrial Engineering team. The reactors included a Modular Pebble Bed helium-cooled concept being developed at MIT, the IRIS water-cooled concept being developed by a team led by Westinghouse Electric Company, and a Lead-Bismuth-cooled concept developed by UT. In addition to the design and layout concepts this report includes a section on heat exchanger manufacturing simulations and a section on construction and cost impacts of proposed modular designs.

Mynatt Fred R.; Townsend, L.W.; Williamson, Martin; Williams, Wesley; Miller, Laurence W.; Khan, M. Khurram; McConn, Joe; Kadak, Andrew C.; Berte, Marc V.; Sawhney, Rapinder; Fife, Jacob; Sedler, Todd L.; Conway, Larry E.; Felde, Dave K.

2003-11-12T23:59:59.000Z

380

Advanced Nuclear Technology Advanced Light Water Reactor Utility Requirements Document, Revision 12  

Science Conference Proceedings (OSTI)

The utility requirement document (URD) is an industry-developed technical foundation for the design of advanced light water reactors (ALWRs). It was created with the objective of providing a comprehensive set of plant functional requirements that are considered important to utilities considering the construction of a nuclear plant and in ensuring successful deployment and operation of the plant. The scope of the URD is broad, addressing the entire plant (including the nuclear steam supply system, ...

2013-12-16T23:59:59.000Z

Note: This page contains sample records for the topic "nuclear reactor design" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


381

Economics and utilization of thorium in nuclear reactors. Technical annexes 1 and 2  

SciTech Connect

An assessment of the impact of utilizing the /sup 233/U/thorium fuel cycle in the U.S. nuclear economy is strongly dependent upon several decisions involving nuclear energy policy. These decisions include: (1) to recycle or not recycle fissile material; (2) if fissile material is recycled, to recycle plutonium, /sup 233/U, or both; and (3) to deploy or not to deploy advanced reactor designs such as Fast Breeder Reactors (FBR's), High Temperature Gas Reactors (HTGR's), and Canadian Deuterium Uranium Reactors (CANDU's). This report examines the role of thorium in the context of the above policy decisions while focusing special attention on economics and resource utilization.

1978-05-01T23:59:59.000Z

382

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

E-Print Network (OSTI)

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

Learned, John

383

Structural Design Challenges in Design Certification Applications for New Reactors  

SciTech Connect

The licensing framework established by the U.S. Nuclear Regulatory Commission under Title 10 of the Code of Federal Regulations (10 CFR) Part 52, “Licenses, Certifications, and Approvals for Nuclear Power Plants,” provides requirements for standard design certifications (DCs) and combined license (COL) applications. The intent of this process is the early reso- lution of safety issues at the DC application stage. Subsequent COL applications may incorporate a DC by reference. Thus, the COL review will not reconsider safety issues resolved during the DC process. However, a COL application that incorporates a DC by reference must demonstrate that relevant site-specific de- sign parameters are confined within the bounds postulated by the DC, and any departures from the DC need to be justified. This paper provides an overview of structural design chal- lenges encountered in recent DC applications under the 10 CFR Part 52 process, in which the authors have participated as part of the safety review effort.

Miranda, M.; Braverman, J.; Wei, X.; Hofmayer, C.; Xu, J.

2011-07-17T23:59:59.000Z

384

TerraPower Traveling Wave Reactor: Design and Development Status...  

NLE Websites -- All DOE Office Websites (Extended Search)

Aug 28 2013 09:00 AM - 10:00 AM Pat Schweiger, TerraPower, LLC, Bellevue, Washington Reactor and Nuclear Systems Division Seminar ORNL Conference Center (Bldg. 5200), TN Rm...

385

Design considerations in inertially-confined fusion reactors  

SciTech Connect

This paper discusses the effects of short time pulses of energetic particles and waves typical of inertially-confined thermonuclear reactions on the first wall, blanket and shield of conceptual reactors. Several reactor designs are presented which attempt to cope with the various problems from the microexplosion debris. Fusion-fission hybrid reactors are also discussed. Emphasis is placed on the first-wall problems of laser-initiated, inertially confined fusion reactors using the deuterium-tritium fuel cycle.

Hovingh, J.

1976-08-01T23:59:59.000Z

386

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

SciTech Connect

Technical nuclear forensics (TNF) refers to the collection, analysis and evaluation of pre- and post-detonation radiological or nuclear materials, devices, and/or debris. TNF is an integral component, complementing traditional forensics and investigative work, to help enable the attribution of discovered radiological or nuclear material. Research is needed to improve the capabilities of TNF. One research area of interest is determining the isotopic signatures of research reactors. Research reactors are a potential source of both radiological and nuclear material. Research reactors are often the least safeguarded type of reactor; they vary greatly in size, fuel type, enrichment, power, and burn-up. Many research reactors are fueled with highly-enriched uranium (HEU), up to {approx}93% {sup 235}U, which could potentially be used as weapons material. All of them have significant amounts of radiological material with which a radioactive dispersal device (RDD) could be built. Therefore, the ability to attribute if material originated from or was produced in a specific research reactor is an important tool in providing for the security of the United States. Currently there are approximately 237 operating research reactors worldwide, another 12 are in temporary shutdown and 224 research reactors are reported as shut down. Little is currently known about the isotopic signatures of spent research reactor fuel. An effort is underway at Savannah River National Laboratory (SRNL) to analyze spent research reactor fuel to determine these signatures. Computer models, using reactor physics codes, are being compared to the measured analytes in the spent fuel. This allows for improving the reactor physics codes in modeling research reactors for the purpose of nuclear forensics. Currently the Oak Ridge Research reactor (ORR) is being modeled and fuel samples are being analyzed for comparison. Samples of an ORR spent fuel assembly were taken by SRNL for analytical and radiochemical analysis. The fuel assembly was modeled using MONTEBURNS(MCNP5/ ORIGEN2.2) and MCNPX/CINDER90. The results from the models have been compared to each other and to the measured data.

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

2011-07-18T23:59:59.000Z

387

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

DOE Green Energy (OSTI)

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

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

2012-01-17T23:59:59.000Z

388

New Research Center to Increase Safety and Power Output of U.S. Nuclear Reactors  

Energy.gov (U.S. Department of Energy (DOE))

The Department of Energy dedicated the Consortium for Advanced Simulation of Light Water Reactors (CASL), an advanced research facility that will accelerate the advancement of nuclear reactor technology.

389

Dual annular rotating "windowed" nuclear reflector reactor control system  

DOE Patents (OSTI)

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

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

1994-01-01T23:59:59.000Z

390

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

E-Print Network (OSTI)

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

391

Strengthening the nuclear-reactor fuel cycle against proliferation  

SciTech Connect

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

Travelli, A.; Snelgrove, J.; Persiani, P. [Argonne National Lab., IL (United States). Arms Control and Nonproliferation Program

1992-12-31T23:59:59.000Z

392

Emergency heat removal system for a nuclear reactor  

DOE Patents (OSTI)

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

Dunckel, Thomas L. (Potomac, MD)

1976-01-01T23:59:59.000Z

393

Nuclear reactor safety. Progress report, January 1-March 31, 1982  

SciTech Connect

The work that is highlighted here represents accomplishments for the period January 1-March 31, 1982 by the groups at Los Alamos involved in reactor safety research for the Division of Accident Evaluation, Office of Nuclear Regulatory Research of the US Nuclear Regulatory Commission. Presented are brief overviews compiled by project, along with a bibliography of Technical Notes and publications written during this quarter. Information is presented concerning the TRAC code development; thermal-hydraulic analysis for PWR after ECCS operation; failure criteria for graphites used in HTGR type reactors; upper structure dynamics experiments; CRBR loss-of-flow accident analysis; and LWR severe accident analysis.

Stevenson, M.G. (comp.)

1982-08-01T23:59:59.000Z

394

Self-sustaining nuclear pumped laser-fusion reactor experiment  

DOE Green Energy (OSTI)

The features of a neutron feedback nuclear pumped (NFNP) laser-fusion reactor equipment were studied with the intention of establishing the feasibility of the concept. The NFNP laser-fusion concept is compared schematically to electrically pumped laser fusion. The study showed that, once a method of energy storage has been demonstrated, a self-sustaining fusion-fission hybrid reactor with a ''blanket multiplication'' of two would be feasible using nuclear pumped Xe F* excimer lasers having efficiencies of 1 to 2 percent and D-D-T pellets with gains of 50 to 100. (MHR)

Boody, F.P.; Choi, C.K.; Miley, G.H.

1977-01-01T23:59:59.000Z

395

Small Modular Fast Reactor Design Description Joint Effort  

NLE Websites -- All DOE Office Websites (Extended Search)

July 1, 2005 ANL-SMFR-1 July 1, 2005 ANL-SMFR-1 Small Modular Fast Reactor Design Description Joint Effort by Argonne National Laboratory (ANL) Commissariat a l'Energie Atomique (CEA) and Japan Nuclear Cycle Development Institute (JNC) Project Leaders Y. I. Chang and C. Grandy, ANL P. Lo Pinto, CEA M. Konomura, JNC Technical Contributors ANL: J. Cahalan, F. Dunn, M. Farmer, S. Kamal, L. Krajtl, A. Moisseytsev, Y. Momozaki, J. Sienicki, Y. Park, Y. Tang, C. Reed, C. Tzanos, S. Wiedmeyer, and W. Yang CEA: P. Allegre, J. Astegiano, F. Baque, L. Cachon, M. S. Chenaud, J-L Courouau, Ph. Dufour, J. C. Klein, C. Latge, C. Thevenot, and F. Varaine JNC: M. Ando, Y. Chikazawa, M. Nagamura, Y. Okano, Y. Sakamoto,

396

Laser Inertial Fusion-based Energy: Neutronic Design Aspects of a Hybrid Fusion-Fission Nuclear Energy System  

E-Print Network (OSTI)

for magnetic fusion reactors and IFMIF. Journal of NuclearFusion reactors blanket nucleonics. In Progress in NuclearFusion-Fission hybrid reactors. In Advances in Nuclear

Kramer, Kevin James

2010-01-01T23:59:59.000Z

397

Laser Inertial Fusion-based Energy: Neutronic Design Aspects of a Hybrid Fusion-Fission Nuclear Energy System  

E-Print Network (OSTI)

Louis J. Hamilton. Nuclear Reactor Analysis. John Wiley &the past and current fleet of nuclear reactors has generatedgenerate fuel for fast nuclear reactors, although Basov and

Kramer, Kevin James

2010-01-01T23:59:59.000Z

398

NUCLEAR REACTOR SLUG PROVIDED WITH THERMOCOUPLE  

DOE Patents (OSTI)

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

Kanne, W.R.

1958-10-14T23:59:59.000Z

399

CONTROL MEANS FOR A NUCLEAR REACTOR  

DOE Patents (OSTI)

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

Teitel, R.J.

1961-09-01T23:59:59.000Z

400

Revised design for the Tokamak experimental power reactor  

DOE Green Energy (OSTI)

A new, preliminary design has been identified for the tokamak experimental power reactor (EPR). The revised EPR design is simpler, more compact, less expensive and has somewhat better performance characteristics than the previous design, yet retains many of the previously developed design concepts. This report summarizes the principle features of the new EPR design, including performance and cost.

Stacey, W.M. Jr.; Abdou, M.A.; Brooks, J.N.

1977-03-01T23:59:59.000Z

Note: This page contains sample records for the topic "nuclear reactor design" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


401

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

E-Print Network (OSTI)

With funding provided by Nuclear Energy Research Initiative (NERI), design of Secure, Transportable, Autonomous Reactors (STAR) to aid countries with insufficient energy supplies is underway. The development of a new monitoring system that allows remote access to data from the reactor site is an important part of this project. The two goals of this monitoring system are to control the use of nuclear materials and to monitor the performance of the facility from a remote location. I have designed a prototype system for this NERI project that utilizes LabVIEW software and global network technologies to monitor the Nuclear Science Center (NSC) reactor at Texas A&M University. LabVIEW and its applications have all the needed features to build a monitoring system for many types of facilities, including STAR reactors. This system takes data from reactor cooling systems, power monitoring channels, fuel temperature indicators, control rod drives, security alarm sensors and stores it on local and remote hard drives, sends it through an output port to remote clients, and graphically displays these data in the reactor control room. Data from NSC TRIGA reactor is fed to a computer program that analyzes and predicts reactor performance in real time. To provide a remote observation of the working area and fissile material, this system uses cameras, triggered by alarm sensors and LabVIEW vision applications. Operators at the local and remote control stations may view and store all the images from these cameras. The system has been in operation for many months at the NSC with outstanding results and further development is continuing.

Jiltchenkov, Dmitri Victorovich

2002-01-01T23:59:59.000Z

402

DESIGN OF THE ARGONNE LOW POWER REACTOR (ALPR)  

SciTech Connect

A description is given of the design of a prototype "packaged" nuclear power plant. The purpose of the plant is to alleviate fuel oil logistics and storage problems posed by remote auxiliary DEW Line radar statibns north of the Arctic Circle. The ALPR (redesignated SL-1) is a 3 Mwt, heterogeneous, highly enriched uranium- fueled, naturalcirculation boiling water reactor, ccoled and moderated with light water. Steam at 300 psig, dry and saturated (421 deg F) is passed directly from the reactor to a conventional turbine-generator to produce electric power (300 kw nominal) and space-heating (400 kw) requirements consistent with rigid mechanical and structural specifications prescribed by the military, and dictated by the extreme geophysics prevailing at the ultimate site. The over all design criteria emphasize: simplicity and reliability of operation and maintenance, with minimum supervision; minimum on-site construction; maximum use of standard components; limited water supply; utilization of local gravel for biological shielding; transportability by air lift; and nominal 3-year fuel operating lifetime per core loading. The "packaged" concept is incorporated for the initial erection. The plant is not designed for relocation. The design criteria for the prototype necessitate special features. The fuel plates are clad with an alurninurn-nickel alloy (X8001). Burnable-poison (BIO) strips are mechancally attached to the fuel assemblies to compensate the excess reactivity required for a nominal 3-year core operating lifetime. The control rods are actuated by rackand-pinion drive extensions which incorporate rotary seals. Fuel exchange is accomplished without the removal of the pressure vessel head. The electrical power generated is used to operate plant auxiliaries; the "net electric power" is dissipated by resistors. The hot water for space heating is heated in a heat exchanger by 20-psig steam, use being made of the latent heat of vaporization, and all the heat is dissipated by a finned-tube, air-cooled heat exchanger. (auth)

Grant, N.R.; Hamer, E.E.; Hooker, H.H.; Jorgensen, G.L.; Kann, W.J.; Lipinski, W.C.; Milak, G.C.; Rossin, A.D.; Shaftman, D.H.; Smaardyk, A.; Treshow, M.

1961-05-01T23:59:59.000Z

403

Operating strategy generators for nuclear reactors  

Science Conference Proceedings (OSTI)

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

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

2011-12-15T23:59:59.000Z

404

Regulatory Process for Decommissioning Nuclear Power Reactors  

Science Conference Proceedings (OSTI)

The NRC revised decommissioning rule 10 CFR 50.82 in 1996 to make significant changes in the regulatory process for nuclear power plant licensees. This report provides a summary of ongoing federal agency and industry activities. It also describes the regulatory requirements applicable, or no longer applicable, to nuclear power plants at the time of permanent shutdown through the early decommissioning stage. The report describes the major components of a typical decommissioning plan, and provides industry...

1998-03-26T23:59:59.000Z

405

Production capabilities in US nuclear reactors for medical radioisotopes  

SciTech Connect

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

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

1992-11-01T23:59:59.000Z

406

Summary of the MARS tandem-mirror reactor design  

SciTech Connect

A recently completed two-year study of a commercial tandem-mirror reactor design (Mirror Advanced Reactor Study (MARS)) is briefly reviewed. The end plugs are designed for trapped-particle stability, MHD ballooning, balanced geodesic curvature, and small radial electric fields in the central cell. New technologies such as lithium-lead blankets, 24 T hybrid coils, gridless direct converters and plasma halo vacuum pumps are highlighted. General characteristics of the MARS tandem mirror and STARFIRE tokamak reactor design are compared. A design of an upgrade of MFTF-B incorporating many of the MARS features is discussed.

Logan, B.G.

1983-09-01T23:59:59.000Z

407

N reactor individual risk comparison to quantitative nuclear safety goals  

Science Conference Proceedings (OSTI)

A full-scope level III probabilistic risk assessment (PRA) has been completed for N reactor, a US Department of Energy (DOE) production reactor located on the Hanford Reservation in the state of Washington. Sandia National Laboratories (SNL) provided the technical leadership for this work, using the state-of-the-art NUREG-1150 methodology developed for the US Nuclear Regulatory Commission (NRC). The main objectives of this effort were to assess the risks to the public and to the on-site workers posed by the operation of N reactor, to identify changes to the plant that could reduce the overall risk, and to compare those risks to the proposed NRC and DOE quantitative safety goals. This paper presents the methodology adopted by Westinghouse Hanford Company (WHC) and SNL for individual health risk evaluation, its results, and a comparison to the NRC safety objectives and the DOE nuclear safety guidelines. The N reactor results, are also compared with the five NUREG-1150 nuclear plants. Only internal events are compared here because external events are not yet reported in the current draft NUREG-1150. This is the first full-scope level III PRA study with a detailed quantitative safety goal comparison performed for DOE production reactors.

Wang, O.S.; Rainey, T.E.; Zentner, M.D.

1990-01-01T23:59:59.000Z

408

Advanced reactors, passive safety, and acceptance of nuclear energy  

SciTech Connect

If nuclear power is to make a serious impact on CO{sub 2} emission, the industry will have to be very large. A 1000-MWe coal-fired power plant releases about 1.4 {times} 10{sup {minus}3} gigatons of carbon per year in the form of CO{sub 2}. The total of 6 GTC/yr of carbon released by human use of 300 quads/yr of energy worldwide then corresponds to the equivalent of about 4000 one-gigawatt power plants. By the middle of the next century, the world's energy demand might grow to about 500 quads/yr. One might halve the implied 10 GTC/yr by deploying 3500 1000-megawatt large reactors. Now the median core melt probability of today's fleet of reactors is according to Rasmussen 5 {times} 10{sup {minus}5} per reactor year which corresponds to a core melt frequency in such a large nuclear system of 0.18/yr - one accident equivalent to that at Three Mile Island Unit 2 every six years. This is almost surely unacceptable. Thus one concludes that a necessary condition for deployment of nuclear reactors on a scale sufficient to contribute significantly to mitigation of the greenhouse effect is reduction of the core melt probability considerably below Rasmussen's fiducial figure. In this paper, the authors summarize developments, both institutional and technical, since 1985 in the development of safer, if not inherently safe, reactors.

Forsberg, C.W. (Chemical Technology Div., Oak Ridge National Lab., Oak Ridge, TN (US)); Weinberg, A.M. (Oak Ridge Associated Univ., Oak Ridge, TN (US))

1990-01-01T23:59:59.000Z

409

Passive cooling system for top entry liquid metal cooled nuclear reactors  

SciTech Connect

This patent describes a passive cooling system for liquid metal cooled, top entry loop nuclear fission reactors. It comprises: a liquid metal cooled nuclear reactor plant; a passive cooling system; and a secondary passive cooling system.

Boardman, C.E.; Hunsbedt, A.; Hui, M.M.

1992-10-27T23:59:59.000Z

410

Chu Visits Site of America's First New Nuclear Reactor in Three...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Chu Visits Site of America's First New Nuclear Reactor in Three Decades Chu Visits Site of America's First New Nuclear Reactor in Three Decades February 15, 2012 - 2:12pm Addthis...

411

Chu Visits Site of America's First New Nuclear Reactor in Three...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Chu Visits Site of America's First New Nuclear Reactor in Three Decades Chu Visits Site of America's First New Nuclear Reactor in Three Decades February 15, 2012 - 12:40pm Addthis...

412

Design Considerations for Economically Competitive Sodium Cooled Fast Reactors  

SciTech Connect

The technological viability of sodium cooled fast reactors (SFR) has been established by various experimental and prototype (demonstration) reactors such as EBR-II, FFTF, Phénix, JOYO, BN-600 etc. However, the economic competitiveness of SFR has not been proven yet. The perceived high cost premium of SFRs over LWRs has been the primary impediment to the commercial expansion of SFR technologies. In this paper, cost reduction options are discussed for advanced SFR designs. These include a hybrid loop-pool design to optimize the primary system, multiple reheat and intercooling helium Brayton cycle for the power conversion system and the potential for suppression of intermediate heat transport system. The design options for the fully passive decay heat removal systems are also thoroughly examined. These include direct reactor auxiliary cooling system (DRACS), reactor vessel auxiliary cooling system (RVACS) and the newly proposed pool reactor auxiliary cooling system (PRACS) in the context of the hybrid loop-pool design.

Hongbin Zhang; Haihua Zhao

2009-05-01T23:59:59.000Z

413

Chu Visits Site of America?s First New Nuclear Reactor in Three...  

NLE Websites -- All DOE Office Websites (Extended Search)

5, 2012 Chu Visits Site of Americas First New Nuclear Reactor in Three Decades Energy Secretary Announces New Nuclear Energy Research Grants and Next Steps on Used Fuel...

414

System Engineering Design [Nuclear Waste Management using  

NLE Websites -- All DOE Office Websites (Extended Search)

System Engineering System Engineering Design Nuclear Fuel Cycle and Waste Management Technologies Overview Modeling and analysis Unit Process Modeling Mass Tracking System Software Waste Form Performance Modeling Safety Analysis, Hazard and Risk Evaluations Development, Design, Operation Overview Systems and Components Development Expertise System Engineering Design Other Major Programs Work with Argonne Contact us For Employees Site Map Help Join us on Facebook Follow us on Twitter NE Division on Flickr Nuclear Waste Management using Electrometallurgical Technology System Engineering Design Bookmark and Share Two major pieces of electrometallurgical process equipment are the Electrorefiner and the Cathode Processor. NE personnel have been involved in the conceptual design, final design, procurement, manufacture,

415

Material Control and Accounting Design Considerations for High-Temperature Gas Reactors  

Science Conference Proceedings (OSTI)

The subject of this report is domestic safeguards and security by design (2SBD) for high-temperature gas reactors, focusing on material control and accountability (MC&A). The motivation for the report is to provide 2SBD support to the Next Generation Nuclear Plant (NGNP) project, which was launched by Congress in 2005. This introductory section will provide some background on the NGNP project and an overview of the 2SBD concept. The remaining chapters focus specifically on design aspects of the candidate high-temperature gas reactors (HTGRs) relevant to MC&A, Nuclear Regulatory Commission (NRC) requirements, and proposed MC&A approaches for the two major HTGR reactor types: pebble bed and prismatic. Of the prismatic type, two candidates are under consideration: (1) GA's GT-MHR (Gas Turbine-Modular Helium Reactor), and (2) the Modular High-Temperature Reactor (M-HTR), a derivative of Areva's Antares reactor. The future of the pebble-bed modular reactor (PBMR) for NGNP is uncertain, as the PBMR consortium partners (Westinghouse, PBMR [Pty] and The Shaw Group) were unable to agree on the path forward for NGNP during 2010. However, during the technology assessment of the conceptual design phase (Phase 1) of the NGNP project, AREVA provided design information and technology assessment of their pebble bed fueled plant design called the HTR-Module concept. AREVA does not intend to pursue this design for NGNP, preferring instead a modular reactor based on the prismatic Antares concept. Since MC&A relevant design information is available for both pebble concepts, the pebble-bed HTGRs considered in this report are: (1) Westinghouse PBMR; and (2) AREVA HTR-Module. The DOE Office of Nuclear Energy (DOE-NE) sponsors the Fuel Cycle Research and Development program (FCR&D), which contains an element specifically focused on the domestic (or state) aspects of SBD. This Material Protection, Control and Accountancy Technology (MPACT) program supports the present work summarized in this report, namely the development of guidance to support the consideration of MC&A in the design of both pebble-bed and prismatic-fueled HTGRs. The objective is to identify and incorporate design features into the facility design that will cost effectively aid in making MC&A more effective and efficient, with minimum impact on operations. The theft of nuclear material is addressed through both MC&A and physical protection, while the threat of sabotage is addressed principally through physical protection.

Trond Bjornard; John Hockert

2011-08-01T23:59:59.000Z

416

Automatic coolant flow control device for a nuclear reactor assembly  

DOE Patents (OSTI)

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

Hutter, E.

1984-01-27T23:59:59.000Z

417

Adaptive nuclear reactor control for integral quadratic cost functions  

Science Conference Proceedings (OSTI)

The problem of optimally controlling the power level changes of a nuclear reactor is considered. The model of an existing power plant is used, which is a ninth-order nonlinear system, having time-varying parameters. A closed form solution of the optimal ...

George T. Bereznai; Naresh K. Sinha

1973-09-01T23:59:59.000Z

418

Natural circulating passive cooling system for nuclear reactor containment structure  

DOE Patents (OSTI)

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

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

1990-01-01T23:59:59.000Z

419

Method of controlling crystallite size in nuclear-reactor fuels  

DOE Patents (OSTI)

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

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

1985-01-01T23:59:59.000Z

420

Passive cooling system for nuclear reactor containment structure  

DOE Patents (OSTI)

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

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

1989-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "nuclear reactor design" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


421

Automatic coolant flow control device for a nuclear reactor assembly  

DOE Patents (OSTI)

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

Hutter, Ernest (Wilmette, IL)

1986-01-01T23:59:59.000Z

422

INSTRUMENT TRANSMITTERS FOR HIGH-PRESSURE, AQUEOUS, NUCLEAR REACTORS  

SciTech Connect

A review of the criteria involved in the selection of primary sensing elements for the measurement of process variables in high-pressure, aqueous, nuclear reactors is presented. Some acceptable types of sensing elements now in use at ORNL are described. (auth)

Moore, R.L.

1958-10-28T23:59:59.000Z

423

CONTROL ROD FOR A NUCLEAR REACTOR AND METHOD OF PREPARATION  

DOE Patents (OSTI)

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

Hausner, H.H.

1958-12-30T23:59:59.000Z

424

METHOD OF FORMING A FUEL ELEMENT FOR A NUCLEAR REACTOR  

DOE Patents (OSTI)

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

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

1962-01-23T23:59:59.000Z

425

Packed rod neutron shield for fast nuclear reactors  

DOE Patents (OSTI)

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

Eck, John E. (Hempfield Township, Westmoreland County, PA); Kasberg, Alvin H. (Murrysville, PA)

1978-01-01T23:59:59.000Z

426

POWER PLANT USING A STEAM-COOLED NUCLEAR REACTOR  

SciTech Connect

A method of providing efficient and economic means for obtaining reheat from nuclear heat is described. A steamcooled steam-moderated reactor produces high-pressure, high-temperature steam. A multi-stage steam turbine partially expands the high-pressure steam, which is then withdrawn and reheated, and then further expanded for producing useful power. The saturated steam is superheated by leading it through tubular passages provided in the fuel assemblies of a nuclear reactor, leading the useful part of the superheated steam into a steam turbine in which it expands to a predetermined intermediate pressure, leading the steam at that reduced pressure from the turbine back into the reactor where it is reheated by flowing through other tubular passages in the fuel assemblies, and returning the reheated steam to the turbine for further expansion. (M.C.G.)

Nettel, F.; Nakanishi, T.

1963-10-29T23:59:59.000Z

427

Thermal Hydraulics of Sodium-Cooled Fast Reactors: Key Design and Safety Issues and Highlights  

Science Conference Proceedings (OSTI)

Technical Paper / Special Issue on the 14th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-14) / Fission Reactors; Thermal Hydraulics

Hisashi Ninokata; Hideki Kamide

428

Transpiring wall supercritical water oxidation test reactor design report  

Science Conference Proceedings (OSTI)

Sandia National Laboratories is working with GenCorp, Aerojet and Foster Wheeler Development Corporation to develop a transpiring wall supercritical water oxidation reactor. The transpiring wall reactor promises to mitigate problems of salt deposition and corrosion by forming a protective boundary layer of pure supercritical water. A laboratory scale test reactor has been assembled to demonstrate the concept. A 1/4 scale transpiring wall reactor was designed and fabricated by Aerojet using their platelet technology. Sandia`s Engineering Evaluation Reactor serves as a test bed to supply, pressurize and heat the waste; collect, measure and analyze the effluent; and control operation of the system. This report describes the design, test capabilities, and operation of this versatile and unique test system with the transpiring wall reactor.

Haroldsen, B.L.; Ariizumi, D.Y.; Mills, B.E.; Brown, B.G. [Sandia National Labs., Livermore, CA (United States). Engineering for Transportation and Environment Dept.; Rousar, D.C. [GenCorp Aerojet, Sacramento, CA (United States)

1996-02-01T23:59:59.000Z

429

Future Directions [Irradiation Performance] - Nuclear Engineering...  

NLE Websites -- All DOE Office Websites (Extended Search)

Achievements Awards Patents Argonne Distinguished Fellows Professional Societies About Nuclear Energy Reactors Designed by Argonne Argonne's Nuclear Science and Technology...

430

Other Current Activities [Irradiation Performance] - Nuclear...  

NLE Websites -- All DOE Office Websites (Extended Search)

Achievements Awards Patents Argonne Distinguished Fellows Professional Societies About Nuclear Energy Reactors Designed by Argonne Argonne's Nuclear Science and Technology...

431

Passive and inherent safety technologies for light-water nuclear reactors  

SciTech Connect

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

Forsberg, C.W.

1990-07-01T23:59:59.000Z

432

Nuclear Power Plant Design Project  

E-Print Network (OSTI)

................................................. 22 5.1.16 Decommissioning: AP600, HTGR, ALWR ............................................................................................................... 35 7.3.4 Decommissioning Cost #12;9 decommissioning. The long delayed nuclear waste disposal facility at Yucca Mountain is becoming

433

Requirements for Advanced Simulation of Nuclear Reactor and Chemical Separation Plants  

E-Print Network (OSTI)

Requirements for Advanced Simulation of Nuclear Reactor and Chemical Separation Plants ANL-AFCI-168 of Nuclear Reactor and Chemical Separation Plants ANL-AFCI-168 by G. Palmiotti, J. Cahalan, P. Pfeiffer, T;2 ANL-AFCI-168 Requirements for Advanced Simulation of Nuclear Reactor and Chemical Separation Plants G

Anitescu, Mihai

434

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

E-Print Network (OSTI)

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

435

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

E-Print Network (OSTI)

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

Bazant, Martin Z.

436

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

E-Print Network (OSTI)

EU in push for support on nuclear fusion reactor September 26, 2004 Page Tool EU ministers have agreed to try to win broad international support for a plan to build a futuristic nuclear reactor to obtain power through nuclear fusion, a clean energy source. But views are split on where the ITER reactor

437

Design and construction of the Qinshan nuclear power plant  

Science Conference Proceedings (OSTI)

The construction work of China's first prototype nuclear power plant Qinshan NPP has already begun. This plant is located on the northern shore of Hangzhou Bay. It is designed to have a capacity of 300 MW electricity. The thermal power of the reactor is 966 MW. The primary circuit consists of 2 coolant loops and 16 auxiliary systems for normal operation. Safety is always the priority in design consideration. Preliminary safety analyses have been done and satisfactory results have been obtained. Excavation and clean out work has been finished.

Yu, O.

1985-01-01T23:59:59.000Z

438

Nuclear Reactor/Hydrogen Process Interface Including the HyPEP Model  

DOE Green Energy (OSTI)

The Nuclear Reactor/Hydrogen Plant interface is the intermediate heat transport loop that will connect a very high temperature gas-cooled nuclear reactor (VHTR) to a thermochemical, high-temperature electrolysis, or hybrid hydrogen production plant. A prototype plant called the Next Generation Nuclear Plant (NGNP) is planned for construction and operation at the Idaho National Laboratory in the 2018-2021 timeframe, and will involve a VHTR, a high-temperature interface, and a hydrogen production plant. The interface is responsible for transporting high-temperature thermal energy from the nuclear reactor to the hydrogen production plant while protecting the nuclear plant from operational disturbances at the hydrogen plant. Development of the interface is occurring under the DOE Nuclear Hydrogen Initiative (NHI) and involves the study, design, and development of high-temperature heat exchangers, heat transport systems, materials, safety, and integrated system models. Research and development work on the system interface began in 2004 and is expected to continue at least until the start of construction of an engineering-scale demonstration plant.

Steven R. Sherman

2007-05-01T23:59:59.000Z

439

Virtual environments for nuclear power plant design  

SciTech Connect

In the design and operation of nuclear power plants, the visualization process inherent in virtual environments (VE) allows for abstract design concepts to be made concrete and simulated without using a physical mock-up. This helps reduce the time and effort required to design and understand the system, thus providing the design team with a less complicated arrangement. Also, the outcome of human interactions with the components and system can be minimized through various testing of scenarios in real-time without the threat of injury to the user or damage to the equipment. If implemented, this will lead to a minimal total design and construction effort for nuclear power plants (NPP).

Brown-VanHoozer, S.A.; Singleterry, R.C. Jr.; King, R.W. [and others

1996-03-01T23:59:59.000Z

440

DESIGN AND FEASIBILITY STUDY OF A PEBBLE BED REACTOR-STEAM POWER PLANT  

SciTech Connect

Originally issued as S and P 1963A, Parts I and II. A design and feasibility study of a pebble bed reactorsteam power plant is presented, The reactor design which evolved from this study is a 125 Mwe heliumcooled two-region thermal breeder, operating on the uranium-thorium cycle, in which all core structural materials are graphite. Fuel is in the form of unclad spherical elements of graphite, containing fissile and fertile material. The primary loop consists of the reactor plus three steam generators and blowers in parallel. Nuclear characteristics, costs, etc., are given. (W.D.M.)

1958-05-01T23:59:59.000Z

Note: This page contains sample records for the topic "nuclear reactor design" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


441

Composite Materials under Extreme Radiation and Temperature Environments of the Next Generation Nuclear Reactors  

Science Conference Proceedings (OSTI)

In the nuclear energy renaissance, driven by fission reactor concepts utilizing very high temperatures and fast neutron spectra, materials with enhanced performance that exceeds are expected to play a central role. With the operating temperatures of the Generation III reactors bringing the classical reactor materials close to their performance limits there is an urgent need to develop and qualify new alloys and composites. Efforts have been focused on the intricate relations and the high demands placed on materials at the anticipated extreme states within the next generation fusion and fission reactors which combine high radiation fluxes, elevated temperatures and aggressive environments. While nuclear reactors have been in operation for several decades, the structural materials associated with the next generation options need to endure much higher temperatures (1200 C), higher neutron doses (tens of displacements per atom, dpa), and extremely corrosive environments, which are beyond the experience on materials accumulated to-date. The most important consideration is the performance and reliability of structural materials for both in-core and out-of-core functions. While there exists a great body of nuclear materials research and operating experience/performance from fission reactors where epithermal and thermal neutrons interact with materials and alter their physio-mechanical properties, a process that is well understood by now, there are no operating or even experimental facilities that will facilitate the extreme conditions of flux and temperature anticipated and thus provide insights into the behaviour of these well understood materials. Materials, however, still need to be developed and their interaction and damage potential or lifetime to be quantified for the next generation nuclear energy. Based on material development advances, composites, and in particular ceramic composites, seem to inherently possess properties suitable for key functions within the operating envelope of both fission and fusion reactors. In advanced fission reactors composite materials are being designed in an effort to extend the life and improve the reliability of fuel rod cladding as well as structural materials. Composites are being considered for use as core internals in the next generation of gas-cooled reactors. Further, next-generation plasma-fusion reactors, such as the International Thermonuclear Experimental Reactor (ITER) will rely on the capabilities of advanced composites to safely withstand extremely high neutron fluxes while providing superior thermal shock resistance.

Simos, N.

2011-05-01T23:59:59.000Z

442

An autonomous long-term fast reactor system and the principal design limitations of the concept  

E-Print Network (OSTI)

The objectives of this dissertation were to find a principal domain of promising and technologically feasible reactor physics characteristics for a multi-purpose, modular-sized, lead-cooled, fast neutron spectrum reactor fueled with an advanced uranium-transuranic-nitride fuel and to determine the principal limitations for the design of an autonomous long-term multi-purpose fast reactor (ALM-FR) within the principal reactor physics characteristic domain. The objectives were accomplished by producing a conceptual design for an ALM-FR and by analysis of the potential ALM-FR performance characteristics. The ALM-FR design developed in this dissertation is based on the concept of a secure transportable autonomous reactor for hydrogen production (STAR-H2) and represents further refinement of the STAR-H2 concept towards an economical, proliferation-resistant, sustainable, multi-purpose nuclear energy system. The development of the ALM-FR design has been performed considering this reactor within the frame of the concept of a self-consistent nuclear energy system (SCNES) that satisfies virtually all of the requirements for future nuclear energy systems: efficient energy production, safety, self-feeding, non-proliferation, and radionuclide burning. The analysis takes into consideration a wide range of reactor design aspects including selection of technologically feasible fuels and structural materials, core configuration optimization, dynamics and safety of long-term operation on one fuel loading, and nuclear material non-proliferation. Plutonium and higher actinides are considered as essential components of an advanced fuel that maintains long-term operation. Flexibility of the ALM-FR with respect to fuel compositions is demonstrated acknowledging the principal limitations of the long-term burning of plutonium and higher actinides. To ensure consistency and accuracy, the modeling has been performed using state-of-the-art computer codes developed at Argonne National Laboratory. As a result of the computational analysis performed in this work, the ALM-FR design provides for the possibility of continuous operation during about 40 years on one fuel loading containing mixture of depleted uranium with plutonium and higher actinides. All reactor physics characteristics of the ALM-FR are kept within technological limits ensuring safety of ultra-long autonomous operation. The results obtained provide for identification of physical features of the ALM-FR that significantly influence flexibility of the design and its applications. The special emphasis is given to existing limitations on the utilization of higher actinides as a fuel component.

Tsvetkova, Galina Valeryevna

2003-12-01T23:59:59.000Z

443

Nuclear reactor spacer grid and ductless core component  

DOE Patents (OSTI)

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

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

1989-01-01T23:59:59.000Z

444

SNIF: A Futuristic Neutrino Probe for Undeclared Nuclear Fission Reactors  

E-Print Network (OSTI)

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

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

2010-11-16T23:59:59.000Z

445

OPERATING EXPERIENCE WITH THE SODIUM REACTOR EXPERIMENT AND ITS APPLICATION TO THE HALLAM NUCLEAR POWER FACILITY  

SciTech Connect

The Sodium Reactor Experiment (SRE) was constructed to demonstrate the feasibility of sodium-oooled graphitemoderated reactors for central station power. The operating experience of SRE has provided valuable data for the design of the Hallam Nuclear Power Facillty (HNPF) now under construction. Some of the difficulties found in the SRE, which HNPF will be designed to avoid, are the sodium-sodium intermediate heat exchanger (horizontal position in SRE gave trouble; a vertical position will be used in HNPF), sodium pumps, handling of broken fuel elements, and excessive thermal stresses due to inadequate coolant flow. Other features of HNPF include the addition of an activity monitoring system for the core cover gas, elimination of tetralin for auxiliary cooling of plant equipment, instrumentation of fuel elements, and addition of carbon traps in the primary sodium system. SRE operation has demonstrated unusual reactor stability and capability for rapid power changes. (D.L.C.)

Beeley, R.J.; Mahlmeister, J.E.

1960-01-01T23:59:59.000Z

446

SUPPLEMENT ANALYSIS OF FOREIGN RESEARCH REACTOR srENT NUCLEAR FUEL  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

FOREIGN RESEARCH REACTOR srENT NUCLEAR FUEL FOREIGN RESEARCH REACTOR srENT NUCLEAR FUEL TRANSPORTATION ALONG OTHER THAN~. PRESENTATIVE ROUTE FROM CONCORD NAVAL WEAPO~~ STATION TO IDAHO NATIONAL ENGINEERING AND ENVIRONMENTAL LADORA TORY Introduction The Department of Energy is planning to transport foreign research reactor spent nuclear fuel by rail from the Concord Naval Weapons Station (CNWS), Concord, California, to the Idaho National Engineering and Environmental Laboratory (INEEL). The environmental analysis supporting the decision to transport, by rail or truck, foreign research reactor spent nuclear fuel from CNWS to the INEEL is contained in +he Final Environmental Impact Statement on a Proposed Nuclear Weapons Nonproliftration Policy Concerning Foreign Research Reactor

447

Heat barrier for use in a nuclear reactor facility  

DOE Patents (OSTI)

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

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

1988-01-01T23:59:59.000Z

448

DESIGN OF A TOKAMAK FUSION REACTOR FIRST WALL ARMOR AGAINST NEUTRAL BEAM IMPINGEMENT  

E-Print Network (OSTI)

Hoffman, et. a1. , "Fusion Reactor First Wall Cooling foricance in Fusion Reactors," Fifth Conference Proceedings onfor a Thp.rmonuclear Reactor," Nu'clear Fusion, 26. H.A.B.

Myers, Richard Allen

2011-01-01T23:59:59.000Z

449

CATALYTIC RECOMBINER FOR A NUCLEAR REACTOR  

DOE Patents (OSTI)

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

King, L.D.P.

1960-07-01T23:59:59.000Z

450

System Design and Analysis of a 900-MW(thermal) Lead-Cooled Fast Reactor  

Science Conference Proceedings (OSTI)

Technical Paper / Special Issue on the 2008 International Congress on Advances in Nuclear Power Plants / Fission Reactors

Sang Ji Kim; Yonghee Kim; Sergi Hong; Chung Ho Cho; Jae-Hyuk Eoh; Jong Bum Kim; Myung Hwan Wi; Kwi Seok Ha; Eui Kwang Kim

451

Digital control of power transients in a nuclear reactor  

Science Conference Proceedings (OSTI)

An integrated, closed-loop, control system for on-line operations in nuclear power plants has been developed and demonstrated with an LSI-11/23 micro-processor on the 5 MWt fission reactor (MITR-II) that is operated by the Massachusetts Institute of Technology. This control system has inherent capabilities to perform on-line fault diagnosis, information display, sensor calibration, and measurement estimation. Recently, its scope has been extended to include the direct digital control of power changes ranging from 20-80% of the reactor's licensed limit. This controller differs from most of those discussed in theoretical and simulation studies by recognizing the non-linearity of reactor dynamics, calculating reactivity on-line, and controlling the rate of change of power by restricting both period and reactivity. The controller functions accurately using rods of non-linear worth in the presence of nonlinear feedback effects.

Bernard, J.A.; Lanning, D.D.; Ray, A.

1984-02-01T23:59:59.000Z

452

Software reliability and safety in nuclear reactor protection systems  

SciTech Connect

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

Lawrence, J.D. [Lawrence Livermore National Lab., CA (United States)

1993-11-01T23:59:59.000Z

453

COMSOL-based Nuclear Reactor Kinetics Studies at the HFIR  

Science Conference Proceedings (OSTI)

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

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

2011-01-01T23:59:59.000Z

454

Collecting and recirculating condensate in a nuclear reactor containment  

DOE Patents (OSTI)

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

Schultz, Terry L. (Murrysville Boro, PA)

1993-01-01T23:59:59.000Z

455

Passive heat-transfer means for nuclear reactors. [LMFBR  

DOE Patents (OSTI)

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

Burelbach, J.P.

1982-06-10T23:59:59.000Z

456

Collecting and recirculating condensate in a nuclear reactor containment  

DOE Patents (OSTI)

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

Schultz, T.L.

1993-10-19T23:59:59.000Z

457

CALMOS: Innovative device for the measurement of nuclear heating in material testing reactors  

Science Conference Proceedings (OSTI)

An R and D program has been carried out since 2002 in order to improve gamma heating measurements in the 70 MWth OSIRIS Material Testing Reactor operated by CEA's Nuclear Energy Div. at the Saclay research center. Throughout this program an innovative calorimetric probe associated to a specific handling system has been designed in order to make measurements both along the fissile height and on the upper part of the core, where nuclear heating rates still remain high. Two mock-ups of the probe were manufactured and tested in 2005 and 2009 in ex-core area of OSIRIS reactor for the process validation, whil