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1

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

2

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

3

Advanced Nuclear Reactor Systems – An Indian Perspective  

Science Journals Connector (OSTI)

The Indian nuclear power programme envisages use of closed nuclear fuel cycle and thorium utilisation as its mainstay for its sustainable growth. The current levels of deployment of nuclear energy in India need to be multiplied nearly hundred fold to reach levels of electricity generation that would facilitate the country to achieve energy independence as well as a developed status. The Indian thorium based nuclear energy systems are being developed to achieve sustainability in respect of fuel resource along with enhanced safety and reduced waste generation. Advanced Heavy Water Reactor and its variants have been designed to meet these objectives. The Indian High Temperature Reactor programme also envisages use of thorium-based fuel with advanced levels of passive safety features.

Ratan Kumar Sinha

2011-01-01T23:59:59.000Z

4

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

E-Print Network (OSTI)

Programme A. Nuclear Power Subprogramme A.4 Technology Development for Advanced Reactor Lines the databases that will be produced in the course of the CRP and make them accessible through the IAEA's nuclear-Electrical Applications of Nuclear Power Project A.5.02: Nuclear hydrogen production CRP Title: Advances in nuclear power

De Cindio, Fiorella

5

Nuclear desalination in the Arab world â?? Part II: advanced inherent and passive safe nuclear reactors  

Science Journals Connector (OSTI)

Rapid increases in population levels have led to greater demands for fresh water and electricity in the Arab World. Different types of energies are needed to contribute to bridging the gap between increased demand and production. Increased levels of safeguards in nuclear power plants have became reliable due to their large operational experience, which now exceeds 11,000 years of operation. Thus, the nuclear power industry should be attracting greater attention. World electricity production from nuclear power has risen from 1.7% in 1970 to 17%-20% today. This ratio had increased in June 2002 to reach more than 30%, 33% and 42% in Europe, Japan, and South Korea respectively. In the Arab World, both the public acceptance and economic viability of nuclear power as a major source of energy are greatly dependent on the achievement of a high level of safety and environmental protection. An assessment of the recent generation of advanced reactor safety criteria requirements has been carried out. The promising reactor designs adapted for the Arab world and other similar developing countries are those that profit from the enhanced and passive safety features of the new generation of reactors, with a stronger focus on the effective use of intrinsic characteristics, simplified plant design, and easy construction, operation and maintenance. In addition, selected advanced reactors with a full spectrum from small to large capacities, and from evolutionary to radical types, which have inherent and passive safety features, are discussed. The relevant economic assessment of these reactors adapted for water/electricity cogeneration have been carried out and compared with non-nuclear desalination methods. This assessment indicates that, water/electricity cogeneration by the nuclear method with advanced inherent and passive safe nuclear power plants, is viable and competitive.

Aly Karameldin; Samer S. Mekhemar

2004-01-01T23:59:59.000Z

6

Advancing Small Modular Reactors: How We're Supporting Next-Gen Nuclear  

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

Advancing Small Modular Reactors: How We're Supporting Next-Gen Advancing Small Modular Reactors: How We're Supporting Next-Gen Nuclear Energy Technology Advancing Small Modular Reactors: How We're Supporting Next-Gen Nuclear Energy Technology December 12, 2013 - 4:00pm Addthis The basics of small modular reactor technology explained. | Infographic by Sarah Gerrity, Energy Department. The basics of small modular reactor technology explained. | Infographic by Sarah Gerrity, Energy Department. Assistant Secretary Lyons Assistant Secretary Lyons Assistant Secretary for Nuclear Energy Nuclear energy continues to be an important part of America's diverse energy portfolio, and the Energy Department is committed to supporting a domestic nuclear industry.

7

Advanced Test Reactor Tour  

SciTech Connect

The Advanced Test Reactor at Idaho National Laboratory is the foremost nuclear materials test reactor in the world. This virtual tour describes the reactor, how experiments are conducted, and how spent nuclear fuel is handled and stored. For more information about INL research, visit http://www.facebook.com/idahonationallaboratory.

Miley, Don

2011-01-01T23:59:59.000Z

8

Advanced Test Reactor Tour  

ScienceCinema (OSTI)

The Advanced Test Reactor at Idaho National Laboratory is the foremost nuclear materials test reactor in the world. This virtual tour describes the reactor, how experiments are conducted, and how spent nuclear fuel is handled and stored. For more information about INL research, visit http://www.facebook.com/idahonationallaboratory.

Miley, Don

2013-05-28T23:59:59.000Z

9

System modeling and reactor design studies of the Advanced Thermionic Initiative space nuclear reactor  

SciTech Connect

In-core thermionic space reactor design concepts that operate at a nominal power output range of 20 to 50 kW(electric) are described. Details of the neutronic, thermionic, thermal hydraulics, and shielding performance are presented. Because of the strong absorption of thermal neutrons by natural tungsten and the large amount of natural tungsten within the reactor core, two designs are considered. An overall system design code has been developed at Oregon State University to model advanced in-core thermionic energy conversion-based nuclear reactor systems for space applications. The results show that the driverless single-cell Advanced Thermionic Initiative (ATI) configuration, which does not have driver fuel rods, proved to be more efficient than the driven core, which has driver rods. The results also show that the inclusion of the true axial and radial power distribution decrease the overall conversion efficiency. The flattening of the radial power distribution by three different methods would lead to a higher efficiency. The results show that only one TFE works at the optimum emitter temperature; all other TFEs are off the optimum performance and result in a 40% decrease of the efficiency of the overall system. The true axial profile is significantly different as there is a considerable amount of neutron leakage out of the top and bottom of the reactor. The analysis reveals that the axial power profile actually has a chopped cosine shape. For this axial profile, the reactor core overall efficiency for the driverless ATI reactor version is found to be 5.84% with a total electrical power of 21.92 kW(electric). By considering the true axial power profile instead of the uniform power profile, each TFE loses {approximately}80 W(electric).

Lee, H.H.; Abdul-Hamid, S.; Klein, A.C. [Oregon State Univ., Corvallis, OR (United States). Dept. of Nuclear Engineering Radiation Center] [Oregon State Univ., Corvallis, OR (United States). Dept. of Nuclear Engineering Radiation Center

1996-07-01T23:59:59.000Z

10

Advanced nuclear reactors and tritium impacts. Modeling the aquatic pathway  

Science Journals Connector (OSTI)

The effective contribution of nuclear energy will depend on various factors related to economics, safety, public acceptance and sustainability. To assure, however, the nuclear energy development, reactor accident impacts, as Fukushima, must be evaluated in a predictive way. Environmental assessment models are used for evaluating the radiological impact of potential releases of radionuclides from nuclear reactors to the environment. It is important to evaluate, to the extent possible, the reliability of the predictions of such models, by comparing with measured values in the environment or by comparing with the predictions of other models. Tritium has a complex environmental behavior once released into the environment. It is essential to establish reference scenarios to allow the simulation of tritium aquatic pathway subsequent to accidental releases. For this purpose, two scenarios for seawater circulation were analyzed by hydrodynamic modeling. An inverse modeling procedure was successfully applied to estimate tide elevations on the borders, which are based on applying the harmonic constants and using the same overestimation percentage produced by model results to correct the border values. Simulations of validated model for postulated accidental releases of tritium inventory from heavy water reactors, whose doses could be relevant, were presented here. It was observed differences between the two scenarios for the transport modeling that were caused by the removal of large volume of polluted waters from the accident site and its dilution in the discharge area, which has minor tritium concentrations. Moreover, the processes involved in the dynamic transfer of tritium in the environment were analyzed in dependence on the environmental conditions of tropical coastal ecosystem.

Francisco Fernando Lamego Simões Filho; Abner Duarte Soares; André da Silva Aguiar; Celso Marcelo Franklin Lapa; Antonio Carlos Ferreira Guimarães

2013-01-01T23:59:59.000Z

11

NUCLEAR REACTORS.  

E-Print Network (OSTI)

??Nuclear reactors are devices containing fissionable material in sufficient quantity and so arranged as to be capable of maintaining a controlled, self-sustaining NUCLEAR FISSION chain… (more)

Belachew, Dessalegn

2010-01-01T23:59:59.000Z

12

Advanced Test Reactor National Scientific User Facility: Addressing advanced nuclear materials research  

SciTech Connect

The Advanced Test Reactor National Scientific User Facility (ATR NSUF), based at the Idaho National Laboratory in the United States, is supporting Department of Energy and industry research efforts to ensure the properties of materials in light water reactors are well understood. The ATR NSUF is providing this support through three main efforts: establishing unique infrastructure necessary to conduct research on highly radioactive materials, conducting research in conjunction with industry partners on life extension relevant topics, and providing training courses to encourage more U.S. researchers to understand and address LWR materials issues. In 2010 and 2011, several advanced instruments with capability focused on resolving nuclear material performance issues through analysis on the micro (10-6 m) to atomic (10-10 m) scales were installed primarily at the Center for Advanced Energy Studies (CAES) in Idaho Falls, Idaho. These instruments included a local electrode atom probe (LEAP), a field-emission gun scanning transmission electron microscope (FEG-STEM), a focused ion beam (FIB) system, a Raman spectrometer, and an nanoindentor/atomic force microscope. Ongoing capability enhancements intended to support industry efforts include completion of two shielded, irradiation assisted stress corrosion cracking (IASCC) test loops, the first of which will come online in early calendar year 2013, a pressurized and controlled chemistry water loop for the ATR center flux trap, and a dedicated facility intended to house post irradiation examination equipment. In addition to capability enhancements at the main site in Idaho, the ATR NSUF also welcomed two new partner facilities in 2011 and two new partner facilities in 2012; the Oak Ridge National Laboratory, High Flux Isotope Reactor (HFIR) and associated hot cells and the University California Berkeley capabilities in irradiated materials analysis were added in 2011. In 2012, Purdue University’s Interaction of Materials with Particles and Components Testing (IMPACT) facility and the Pacific Northwest Nuclear Laboratory (PNNL) Radiochemistry Processing Laboratory (RPL) and PIE facilities were added. The ATR NSUF annually hosts a weeklong event called User’s Week in which students and faculty from universities as well as other interested parties from regulatory agencies or industry convene in Idaho Falls, Idaho to see presentations from ATR NSUF staff as well as select researchers from the materials research field. User’s week provides an overview of current materials research topics of interest and an opportunity for young researchers to understand the process of performing work through ATR NSUF. Additionally, to increase the number of researchers engaged in LWR materials issues, a series of workshops are in progress to introduce research staff to stress corrosion cracking, zirconium alloy degradation, and uranium dioxide degradation during in-reactor use.

John Jackson; Todd Allen; Frances Marshall; Jim Cole

2013-03-01T23:59:59.000Z

13

Advanced Reactor Technologies | Department of Energy  

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

Advanced Reactor Advanced Reactor Technologies Advanced Reactor Technologies Advanced Reactor Technologies Advanced Reactor Technologies The Office of Advanced Reactor Technologies (ART) sponsors research, development and deployment (RD&D) activities through its Next Generation Nuclear Plant (NGNP), Advanced Reactor Concepts (ARC), and Advanced Small Modular Reactor (aSMR) programs to promote safety, technical, economical, and environmental advancements of innovative Generation IV nuclear energy technologies. The Office of Nuclear Energy (NE) will pursue these advancements through RD&D activities at the Department of Energy (DOE) national laboratories and U.S. universities, as well as through collaboration with industry and international partners. These activities will focus on advancing scientific

14

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

SciTech Connect

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

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

1991-09-01T23:59:59.000Z

15

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

16

Nuclear Reactor (atomic reactor)  

Science Journals Connector (OSTI)

A nuclear reactor splits Uranium or Plutonium nuclei, and the...235 is fissionable but more than 99% of the naturally occurring Uranium is U238 that makes enrichment mandatory. In some reactors U238 and Thorium23...

2008-01-01T23:59:59.000Z

17

System modeling for the advanced thermionic initiative single cell thermionic space nuclear reactor  

SciTech Connect

Incore thermionic space reactor design concepts which operate in a nominal power output range of 20 to 40 kWe are described. Details of the neutronics, thermionic, shielding, and heat rejection performance are presented. Two different designs, ATI-Driven and ATI-Driverless, are considered. Comparison of the core overall performance of these two configurations are described. The comparison of these two cores includes the overall conversion efficiency, reactor mass, shield mass, and heat rejection mass. An overall system design has been developed to model the advanced incore thermionic energy conversion based nuclear reactor systems for space applications in this power range.

Lee, H.H.; Lewis, B.R.; Klein, A.C. (Department of Nuclear Engineering, Oregon State University, Radiation Center, C116, Corvallis, Oregon 97331-5902 (United States)); Pawlowski, R.A. (Battelle Pacific Northwest Laboratories, Richland, Washington 99352 (United States))

1993-01-15T23:59:59.000Z

18

Energy Department Announces New Investments in Advanced Nuclear Power Reactors  

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

WASHINGTON – Today, as part of the President’s all-of-the-above energy approach and Climate Action Plan, the Energy Department announced awards for five companies to lead key nuclear energy...

19

Global Nuclear Energy Partnership Fact Sheet - Develop Advanced...  

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

Advanced Burner Reactors Global Nuclear Energy Partnership Fact Sheet - Develop Advanced Burner Reactors GNEP will develop and demonstrate Advanced Burner Reactors (ABRs) that...

20

Advanced Reactor Technology Documents | Department of Energy  

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

Nuclear Reactor Technologies » Advanced Reactor Nuclear Reactor Technologies » Advanced Reactor Technologies » Advanced Reactor Technology Documents Advanced Reactor Technology Documents January 30, 2013 Advanced Reactor Concepts Technical Review Panel Report This report documents the establishment of a technical review process and the findings of the Advanced Reactor Concepts (ARC) Technical Review Panel (TRP).1 The intent of the process is to identify R&D needs for viable advanced reactor concepts in order to inform DOE-NE R&D investment decisions. A goal of the process is to facilitate greater engagement between DOE and industry. The process involved establishing evaluation criteria, conducting a pilot review, soliciting concept inputs from industry entities, reviewing the concepts by TRP members and compiling the

Note: This page contains sample records for the topic "advanced nuclear reactor" 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

Energy Department Announces New Investments in Advanced Nuclear...  

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

Announces New Investments in Advanced Nuclear Power Reactors Energy Department Announces New Investments in Advanced Nuclear Power Reactors October 31, 2014 - 12:20pm Addthis NEWS...

22

Advances in Metallic Nuclear Fuel  

Science Journals Connector (OSTI)

Metallic nuclear fuels have generated renewed interest for advanced ... operations is excellent. Ongoing irradiation tests in Argonne-West’s Idaho-based Experimental Breeder Reactor ... fast reactor (IFR) concept...

B. R. Seidel; L. C. Walters; Y. I. Chang

1987-04-01T23:59:59.000Z

23

Parametric Evaluation of Large-Scale High-Temperature Electrolysis Hydrogen Production Using Different Advanced Nuclear Reactor Heat Sources  

SciTech Connect

High Temperature Electrolysis (HTE), when coupled to an advanced nuclear reactor capable of operating at reactor outlet temperatures of 800 °C to 950 °C, has the potential to efficiently produce the large quantities of hydrogen needed to meet future energy and transportation needs. To evaluate the potential benefits of nuclear-driven hydrogen production, the UniSim process analysis software was used to evaluate different reactor concepts coupled to a reference HTE process design concept. The reference HTE concept included an Intermediate Heat Exchanger and intermediate helium loop to separate the reactor primary system from the HTE process loops and additional heat exchangers to transfer reactor heat from the intermediate loop to the HTE process loops. The two process loops consisted of the water/steam loop feeding the cathode side of a HTE electrolysis stack, and the sweep gas loop used to remove oxygen from the anode side. The UniSim model of the process loops included pumps to circulate the working fluids and heat exchangers to recover heat from the oxygen and hydrogen product streams to improve the overall hydrogen production efficiencies. The reference HTE process loop model was coupled to separate UniSim models developed for three different advanced reactor concepts (a high-temperature helium cooled reactor concept and two different supercritical CO2 reactor concepts). Sensitivity studies were then performed to evaluate the affect of reactor outlet temperature on the power cycle efficiency and overall hydrogen production efficiency for each of the reactor power cycles. The results of these sensitivity studies showed that overall power cycle and hydrogen production efficiencies increased with reactor outlet temperature, but the power cycles producing the highest efficiencies varied depending on the temperature range considered.

Edwin A. Harvego; Michael G. McKellar; James E. O'Brien; J. Stephen Herring

2009-09-01T23:59:59.000Z

24

Advanced Gas Cooled Nuclear Reactor Materials Evaluation and Development Program. Progress report, January 1, 1980-March 31, 1980  

SciTech Connect

Results are presented of work performed on the Advanced Gas-Cooled Nuclear Reactor Materials Evaluation and Development Program. The objectives of this program are to evaluate candidate alloys for Very High Temperature Reactor (VHTR) Nuclear Process Heat (NPH) and Direct Cycle Helium Turbine (DCHT) applications, in terms of the effect of simulated reactor primary coolant (helium containing small amounts of various other gases), high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. A second objective is to select and recommend materials for future test facilities and more extensive qualification programs. Included are the activities associated with the status of the simulated reactor helium supply system, testing equipment and gas chemistry analysis instrumentation and equipment. The progress in the screening test program is described, including screening creep results and metallographic analysis for materials thermally exposed or tested at 750, 850, and 950/sup 0/C.

Not Available

1980-06-25T23:59:59.000Z

25

Advanced Gas-Cooled Nuclear Reactor Materials Evaluation and Development Program. Progress report, July 1, 1979-September 30, 1979  

SciTech Connect

The results of work performed from July 1, 1979 through September 30, 1979 on the Advanced Gas-Cooled Nuclear Reactor Materials Evaluation and Development Program are presented. The objectives of this program are to evaluate candidate alloys for Very High Temperature Reactor (VHTR) Nuclear Process Heat (NPH) and Direct Cycle Helium Turbine (DCHT) applications, in terms of the effect of simulated reactor primary coolant (helium containing small amounts of various other gases), high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. A second objective is to select and recommend materials for future test facilities and more extensive qualification programs. Work covered in this report includes the activities associated with the status of the simulated reactor helium supply system, testing equipment, and gas chemistry analysis instrumentation and equipment. The status of the data management system is presented. In addition, the progress in the screening test program is described.

Not Available

1980-03-07T23:59:59.000Z

26

Nuclear reactor  

DOE Patents (OSTI)

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

Thomson, Wallace B. (Severna Park, MD)

2004-03-16T23:59:59.000Z

27

nuclear reactor  

Science Journals Connector (OSTI)

...a complex atomic apparatus used to obtain energy from nuclear fission chain reaction. Used to produce nuclear energy, radioactive isotopes, and artificial elements.... atomic pile ...

2009-01-01T23:59:59.000Z

28

High Temperature Materials for Nuclear Fast Fission and Fusion Reactors and Advanced Fossil Power Plants  

Science Journals Connector (OSTI)

Development of materials plays a crucial role in the economic feasibility of fast nuclear fission and fusion power plant. In order to meet this objective, one of the methods is to extend the fuel burnup and decreasing doubling time. The burnup is largely limited by the void swelling and creep resistances of the fuel cladding and wrapping materials. India's 500 \\{MWe\\} Prototype Fast Breeder Reactor (PFBR) is in advanced stage of construction. The major structural materials chosen for PFBR with MOX fuel are alloy D9 as fuel clad and wrapper material, 316LN austenitic stainless steel for reactor components and piping and modified 9Cr-1Mo steel for steam generator. In order to improve the burnup further, titanium, phosphorous and silicon contents in alloy D9 have been optimized for better swelling and creep resistances to develop modified version of alloy D9 as IFAC-1. Creep resistance of inherently void swelling resistance 9Cr-ferritic steel has been improved with the dispersion of nano-size yttria to develop oxide dispersion strengthened (ODS) steel clad tube with long- term creep strength, similar to D9, for increasing the fuel burnup. Development of modified 9Cr-1Mo steel clad tube and 9Cr-1Mo steel wrapper for future metallic fuel reactors being developed for reducing the doubling time are in progress. Extensive studies on resistance of this new generation core materials to void swelling are also under progress along with material development. Improved versions of 316LN stainless steel with nitrogen content of about 0.14 wt.% having higher creep strength to increase the life of fast reactor and modified 9Cr-1Mo steel with reduced nitrogen content and controlled addition of boron to improve type IV cracking resistance for steam generator are other developments. India's participation in ITER programme necessitates the development of India-specific RAFM steel for Test Blanket Module (TBM). A comprehensive research programme is being carried out to develop India-specific 9Cr-W-Ta RAFM steel with the optimization of tungsten and tantalum contents for better combination of strength and toughness. Based of the extensive mechanical tests including impact, tensile, creep and fatigue on four heats of RAFM steels having tungsten in the range 1 – 2 wt. % and tantalum in the range 0.06 -.014 wt., the RAFM steel having 1.4 wt. % tungsten with 0.06 wt. % tantalum is found to possess better combination of strength and toughness. This steel is considered as India-specific RAFM steel and TBM is being manufactured by this RAFM steel. To limit the emission of green house gases, a research and development programme has been initiated to develop advanced ultra super critical fossil fuel fired thermal power plants working at temperature of around 973 K and pressure of 300 bar. High temperature creep strength super 304H austenitic steel and Inconel 617 superalloy tubes are indigenously developed for this purpose.

T. Jayakumar; M.D. Mathew; K. Laha

2013-01-01T23:59:59.000Z

29

Advanced neutron irradiation system using Texas A&M University Nuclear Science Center Reactor  

E-Print Network (OSTI)

was installed in the irradiation cell of the Texas A&M University Nuclear Science Center Reactor (NSCR). By increasing the thickness of the lead-bismuth alloy, the neutron spectra were shifted into lower energies by the scattering interactions of fast...

Jang, Si Young

2005-11-01T23:59:59.000Z

30

Reactor & Nuclear Systems Publications | ORNL  

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

Nuclear Science Home | Science & Discovery | Nuclear Science | Publications and Reports | Reactor and Nuclear Systems Publications SHARE Reactor and Nuclear Systems Publications...

31

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

SciTech Connect

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

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

2014-06-10T23:59:59.000Z

32

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

DOE Patents (OSTI)

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

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

2013-11-05T23:59:59.000Z

33

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

34

Nuclear Reactor Materials and Fuels  

Science Journals Connector (OSTI)

Nuclear reactor materials and fuels can be classified into six categories: Nuclear fuel materials Nuclear clad materials Nuclear coolant materials Nuclear poison materials Nuclear moderator materials

Dr. James S. Tulenko

2012-01-01T23:59:59.000Z

35

Advance Reactor Concepts Technical Review Panel Public Report  

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

The Office of Nuclear Energy supports research and development for advanced reactor technologies. This report documents the results of the 2014 Technical Review Panel (TRP) review of seven advanced reactor concepts. The intent of the process was to identify R&D needs for advanced reactor concepts in order to inform Department of Energy (DOE) Office of Nuclear Energy R&D investment decisions.

36

Radiation Damage in Nuclear Fuel for Advanced Burner Reactors: Modeling and Experimental Validation  

SciTech Connect

The consortium has completed its existence and we are here highlighting work and accomplishments. As outlined in the proposal, the objective of the work was to advance the theoretical understanding of advanced nuclear fuel materials (oxides) toward a comprehensive modeling strategy that incorporates the different relevant scales involved in radiation damage in oxide fuels. Approaching this we set out to investigate and develop a set of directions: 1) Fission fragment and ion trajectory studies through advanced molecular dynamics methods that allow for statistical multi-scale simulations. This work also includes an investigation of appropriate interatomic force fields useful for the energetic multi-scale phenomena of high energy collisions; 2) Studies of defect and gas bubble formation through electronic structure and Monte Carlo simulations; and 3) an experimental component for the characterization of materials such that comparisons can be obtained between theory and experiment.

Niels Gronbech Jensen; Mark Asta; Nigel Browning'Vidvuds Ozolins; Axel van de Walle; Christopher Wolverton

2011-12-29T23:59:59.000Z

37

Nuclear Research Reactors  

Science Journals Connector (OSTI)

... their countries for the advent of nuclear power. A few countries had built large research reactors for the production of isotopes and to study the behaviour of nuclear fuel, but ... production of isotopes and to study the behaviour of nuclear fuel, but the small training reactor had not been developed. Since then, research ...

T. E. ALLIBONE

1963-07-20T23:59:59.000Z

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

Functional issues and environmental qualification of digital protection systems of advanced light-water nuclear reactors  

SciTech Connect

Issues of obsolescence and lack of infrastructural support in (analog) spare parts, coupled with the potential benefits of digital systems, are driving the nuclear industry to retrofit analog instrumentation and control (I&C) systems with digital and microprocessor-based systems. While these technologies have several advantages, their application to safety-related systems in nuclear power plants raises key issues relating to the systems` environmental qualification and functional reliability. To bound the problem of new I&C system functionality and qualification, the authors focused this study on protection systems proposed for use in ALWRs. Specifically, both functional and environmental qualification issues for ALWR protection system I&C were addressed by developing an environmental, functional, and aging data template for a protection division of each proposed ALWR design. By using information provided by manufacturers, environmental conditions and stressors to which I&C equipment in reactor protection divisions may be subjected were identified. The resulting data were then compared to a similar template for an instrument string typically found in an analog protection division of a present-day nuclear power plant. The authors also identified fiber-optic transmission systems as technologies that are relatively new to the nuclear power plant environment and examined the failure modes and age-related degradation mechanisms of fiber-optic components and systems. One reason for the exercise of caution in the introduction of software into safety-critical systems is the potential for common-cause failure due to the software. This study, however, approaches the functionality problem from a systems point of view. System malfunction scenarios are postulated to illustrate the fact that, when dealing with the performance of the overall integrated system, the real issues are functionality and fault tolerance, not hardware vs. software.

Korsah, K.; Clark, R.L.; Wood, R.T. [Oak Ridge National Lab., TN (United States)

1994-04-01T23:59:59.000Z

40

Instrumentation of Nuclear Reactors  

Science Journals Connector (OSTI)

... s Lecture Theatre on January 8, a symposium of papers on the instrumentation of nuclear reactors was organized, at which about five hundred members and visitors attended, including guests from ... the Institution, took the chair and introduced Sir John Cockcroft, whose lecture on "Nuclear Reactors and their Applications" provided a general background for the three specialized papers which followed. ...

1953-03-07T23:59:59.000Z

Note: This page contains sample records for the topic "advanced nuclear reactor" 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

Global Nuclear Energy Partnership Fact Sheet - Develop Advanced Burner  

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

Develop Advanced Develop Advanced Burner Reactors Global Nuclear Energy Partnership Fact Sheet - Develop Advanced Burner Reactors GNEP will develop and demonstrate Advanced Burner Reactors (ABRs) that consume transuranic elements (plutonium and other long-lived radioactive material) while extracting their energy. The development of ABRs will allow us to build an improved nuclear fuel cycle that recycles used fuel. Accordingly, the U.S. will work with participating international partners on the design, development, and demonstration of ABRs as part of the GNEP. Global Nuclear Energy Partnership Fact Sheet - Develop Advanced Burner Reactors More Documents & Publications GNEP Element:Develop Advanced Burner Reactors Global Nuclear Energy Partnership Fact Sheet - Minimize Nuclear Waste

42

Improved best estimate plus uncertainty methodology including advanced validation concepts to license evolving nuclear reactors  

SciTech Connect

Many evolving nuclear energy programs plan to use advanced predictive multi-scale multi-physics simulation and modeling capabilities to reduce cost and time from design through licensing. Historically, the role of experiments was primary tool for design and understanding of nuclear system behavior while modeling and simulation played the subordinate role of supporting experiments. In the new era of multi-scale multi-physics computational based technology development, the experiments will still be needed but they will be performed at different scales to calibrate and validate models leading predictive simulations. Cost saving goals of programs will require us to minimize the required number of validation experiments. Utilization of more multi-scale multi-physics models introduces complexities in the validation of predictive tools. Traditional methodologies will have to be modified to address these arising issues. This paper lays out the basic aspects of a methodology that can be potentially used to address these new challenges in design and licensing of evolving nuclear technology programs. The main components of the proposed methodology are verification, validation, calibration, and uncertainty quantification. An enhanced calibration concept is introduced and is accomplished through data assimilation. The goal is to enable best-estimate prediction of system behaviors in both normal and safety related environments. To achieve this goal requires the additional steps of estimating the domain of validation and quantification of uncertainties that allow for extension of results to areas of the validation domain that are not directly tested with experiments, which might include extension of the modeling and simulation (M&S) capabilities for application to full-scale systems. The new methodology suggests a formalism to quantify an adequate level of validation (predictive maturity) with respect to required selective data so that required testing can be minimized for cost saving purposes by showing further testing wold not enhance the quality of the validation of predictive tools. The proposed methodology is at a conceptual level. When matured and if considered favorably by the stakeholders, it could serve as a new framework for the next generation of the best estimate plus uncertainty licensing methodology that USNRC developed previously. In order to come to that level of maturity it is necessary to communicate the methodology to scientific, design and regulatory stakeholders for discussion and debates. This paper is the first step to establish this communication.

Unal, Cetin [Los Alamos National Laboratory; Williams, Brian [Los Alamos National Laboratory; Mc Clure, Patrick [Los Alamos National Laboratory; Nelson, Ralph A [IDAHO NATIONAL LAB

2010-01-01T23:59:59.000Z

43

The Centralized Reliability Data Organization (CREDO); an Advanced Nuclear Reactor Reliability, Availability, and Maintainability Data Bank and Data Analysis Center  

Science Journals Connector (OSTI)

The Centralized Reliability Data Organization (CREDO) is a data bank and data analysis center, which since 1985 has been jointly ... of Technology Support Programs and Japan’s Power Reactor and Nuclear Fuel Devel...

H. E. Knee

1991-01-01T23:59:59.000Z

44

Advancing Small Modular Reactors: How We're Supporting Next-Gen...  

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

Advancing Small Modular Reactors: How We're Supporting Next-Gen Nuclear Energy Technology Advancing Small Modular Reactors: How We're Supporting Next-Gen Nuclear Energy Technology...

45

Prospects for the development of advanced reactors  

SciTech Connect

Energy supply is an important prerequisite for further socio-economic development, especially in developing countries where the per capita energy use is only a very small fraction of that in industrialized countries. Nuclear energy is an essentially unlimited energy resource with the potential to provide this energy in the form of electricity, district heat and process heat under environmentally acceptable conditions. However, this potential will be realized only if nuclear power plants can meet the challenges of increasingly demanding safety requirements, economic competitiveness and public acceptance. Worldwide a tremendous amount of experience has been accumulated during development, licensing, construction and operation of nuclear power reactors. The experience forms a sound basis for further improvements. Nuclear programmes in many countries are addressing the development of advanced reactors which are intended to have better economics, higher reliability and improved safety in order to overcome the current concerns of nuclear power. Advanced reactors now being developed could help to meet the demand for new plants in developed and developing countries, not only for electricity generation, but also for district heating, desalination and for process heat. The IAEA, as the only global international governmental organization dealing with nuclear power, promotes international information exchange and international co-operation between all countries with their own advanced nuclear power programmes and offers assistance to countries with an interest in exploratory or research programmes.

Semenov, B.A.; Kupitz, J.; Cleveland, J. [International Atomic Energy Agency Vienna (Austria). Dept. of Nuclear Energy and Safety

1992-12-31T23:59:59.000Z

46

E-Print Network 3.0 - advanced reactor research Sample Search...  

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

to build the world's most advanced nuclear fusion reactor... ) International Thermonuclear Experimental Reactor (ITER), which will be built at Cadarache, near the...

47

E-Print Network 3.0 - advanced research reactor Sample Search...  

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

to build the world's most advanced nuclear fusion reactor... ) International Thermonuclear Experimental Reactor (ITER), which will be built at Cadarache, near the...

48

E-Print Network 3.0 - advanced marine reactor Sample Search Results  

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

to build the world's most advanced nuclear fusion reactor... ) International Thermonuclear Experimental Reactor (ITER), which will be built at Cadarache, near the...

49

E-Print Network 3.0 - advanced hanaro reactor Sample Search Results  

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

to build the world's most advanced nuclear fusion reactor... ) International Thermonuclear Experimental Reactor (ITER), which will be built at Cadarache, near the...

50

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

51

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

52

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

53

Advances in Nuclear Engineering  

Science Journals Connector (OSTI)

... door, closed for fifteen years, to scientific and technical information about fission and nuclear reactors. In spite of the 1,000 papers published then, there was an enormous amount ... Engineering and Science Conference held in Philadelphia, March 1957, and comprise 130 papers on reactors, fuel and a few other matters, almost all the papers being of American origin ...

T. E. ALLIBONE

1958-07-26T23:59:59.000Z

54

Advanced Safeguards Approaches for New Fast Reactors  

SciTech Connect

This third report in the series reviews possible safeguards approaches for new fast reactors in general, and the ABR in particular. Fast-neutron spectrum reactors have been used since the early 1960s on an experimental and developmental level, generally with fertile blanket fuels to “breed” nuclear fuel such as plutonium. Whether the reactor is designed to breed plutonium, or transmute and “burn” actinides depends mainly on the design of the reactor neutron reflector and the whether the blanket fuel is “fertile” or suitable for transmutation. However, the safeguards issues are very similar, since they pertain mainly to the receipt, shipment and storage of fresh and spent plutonium and actinide-bearing “TRU”-fuel. For these reasons, the design of existing fast reactors and details concerning how they have been safeguarded were studied in developing advanced safeguards approaches for the new fast reactors. In this regard, the design of the Experimental Breeder Reactor-II “EBR-II” at the Idaho National Laboratory (INL) was of interest, because it was designed as a collocated fast reactor with a pyrometallurgical reprocessing and fuel fabrication line – a design option being considered for the ABR. Similarly, the design of the Fast Flux Facility (FFTF) on the Hanford Site was studied, because it was a successful prototype fast reactor that ran for two decades to evaluate fuels and the design for commercial-scale fast reactors.

Durst, Philip C.; Therios, Ike; Bean, Robert; Dougan, A.; Boyer, Brian; Wallace, Rick L.; Ehinger, Michael H.; Kovacic, Don N.; Tolk, K.

2007-12-15T23:59:59.000Z

55

Idaho National Laboratory Advanced Test Reactor Probabilistic Risk  

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

Idaho National Laboratory Advanced Test Reactor Probabilistic Risk Idaho National Laboratory Advanced Test Reactor Probabilistic Risk Assessment Idaho National Laboratory Advanced Test Reactor Probabilistic Risk Assessment September 19, 2012 Presenter: Bentley Harwood, Advanced Test Reactor Nuclear Safety Engineer Battelle Energy Alliance Idaho National Laboratory Topics covered: PRA studies began in the late 1980s 1989, ATR PRA published as a summary report 1991, ATR PRA full report 1994 and 2004 various model changes 2011, Consolidation, update and improvement of previous PRA work 2012/2013, PRA risk monitor implementation Idaho National Laboratory Advanced Test Reactor Probabilistic Risk Assessment More Documents & Publications DOE's Approach to Nuclear Facility Safety Analysis and Management Nuclear Regulatory Commission Handling of Beyond Design Basis Events for

56

Advanced Nuclear Energy Projects Solicitation | Department of...  

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

Advanced Nuclear Energy Projects Solicitation Advanced Nuclear Energy Projects Solicitation INFORMATIONAL MATERIALS ADVANCED NUCLEAR ENERGY PROJECTS SOLICITATION Solicitation...

57

Draft Advanced Nuclear Energy Projects Solicitation | Department...  

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

Draft Advanced Nuclear Energy Projects Solicitation Draft Advanced Nuclear Energy Projects Solicitation INFORMATIONAL MATERIALS DRAFT ADVANCED NUCLEAR ENERGY PROJECTS SOLICITATION...

58

CESAR: Center for Exascale Simulation of Advanced Reactors | Argonne  

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

CESAR: Center for Exascale Simulation of Advanced Reactors CESAR: Center for Exascale Simulation of Advanced Reactors CESAR: Center for Exascale Simulation of Advanced Reactors CESAR is an interdisciplinary center for developing an innovative, next-generation nuclear reactor analysis tool that both utilizes and guides the development of exascale computing platforms. Existing reactor analysis codes are highly tuned and calibrated for commercial light-water reactors, but they lack the physics fidelity to seamlessly carry over to new classes of reactors with significantly different design characteristics-as, for example, innovative concepts such as TerraPower's Traveling Wave reactor and Small Modular Reactor concepts. Without vastly improved modeling capabilities, the economic and safety characteristics of these and other novel systems will require tremendous

59

Nuclear reactors in the United States  

Science Journals Connector (OSTI)

Nuclear reactors in the United States ... A chart listing the operating and planned nuclear reactors in the United States. ... Nuclear / Radiochemistry ...

Hubert N. Alyea

1956-01-01T23:59:59.000Z

60

Advanced Reactor Research and Development Funding Opportunity...  

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

Reactor Research and Development Funding Opportunity Announcement Advanced Reactor Research and Development Funding Opportunity Announcement The U.S. Department of Energy (DOE)...

Note: This page contains sample records for the topic "advanced nuclear reactor" 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

Nuclear divisional reactor  

SciTech Connect

A nuclear divisional reactor including a reactor core having side and top walls, a heat exchanger substantially surrounding the core, the heat exchanger including a plurality of separate fluid holding and circulating chambers each in contact with a portion of the core, control rod means associated with the core and external of the heat exchanger including control rods and means for moving said control rods, each of the chambers having separate means for delivering and removing fluid therefrom, separate means associated with each of the delivering and removing means for producing useable energy external of the chambers, each of the means for producing useable energy having separate variable capacity energy outputs thereby making available a plurality of individual sources of useable energy of varying degrees.

Administratrix, A.P.; Rugh, J.L.

1982-11-02T23:59:59.000Z

62

PIA - Advanced Test Reactor National Scientific User Facility...  

Energy Savers (EERE)

Advanced Test Reactor National Scientific User Facility Users Week 2009 PIA - Advanced Test Reactor National Scientific User Facility Users Week 2009 PIA - Advanced Test Reactor...

63

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

64

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.

65

Advanced Gas Cooled Nuclear Reactor Materials Evaluation and Development Program. Progress report, April 1, 1980-June 30, 1980  

SciTech Connect

Objectives of this program are to evaluate candidate alloys for Very High Temperature Reactor (VHTR) Nuclear Process Heat (NPH) and Direct Cycle Helium Turbine (DCHT) applications, in terms of the effect of simulated reactor primary coolant (helium containing small amounts of various other gases), high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. A second objective is to select and recommend materials for future test facilities and more extensive qualification programs. Work covered in this report includes the activities associated with the status of the simulated reactor helium supply system, testing equipment and gas chemistry analysis instrumentation and equipment. The progress in the screening test program is described; this includes: screening creep results and metallographic analysis for materials thermally exposed or tested at 750, 850 and 950/sup 0/C. The initiation of air creep-rupture testing in the intensive screening test program is discussed. In addition, the status of the data management system is described.

Not Available

1980-11-14T23:59:59.000Z

66

Argonne's pyroprocessing and advanced reactor research featured on WGN  

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

Argonne's pyroprocessing and advanced reactor research featured on WGN Argonne's pyroprocessing and advanced reactor research featured on WGN radio 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 Argonne's pyroprocessing and advanced reactor research featured on WGN radio Uranium dendrites These tiny branches, or "dendrites," of pure uranium form when engineers

67

Draft Advanced Nuclear Energy Projects Solicitation | Department...  

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

Projects Solicitation Draft Advanced Nuclear Energy Projects Solicitation Federal loan guarantee solicitation announcement -- Advanced Nuclear Energy Projects. Draft Advanced...

68

Operational Analysis of Multiregional Nuclear Reactor Kinetics  

Science Journals Connector (OSTI)

......Operational Analysis of Multiregional Nuclear Reactor Kinetics NASSAR H. S. HAIDAR...analytically for a multiregional nuclear reactor whose subregions are of arbitrary...Operational Analysis of Multiregional Nuclear Reactor Kinetics NASSAU H. S. HAIDAR......

NASSAR H. S. HAIDAR

1983-05-01T23:59:59.000Z

69

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

70

Advanced Reactor Research and Development Funding Opportunity Announcement  

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

Advanced Reactor Research and Development Funding Opportunity Advanced Reactor Research and Development Funding Opportunity Announcement Advanced Reactor Research and Development Funding Opportunity Announcement The U.S. Department of Energy (DOE) Office of Nuclear Energy (NE) sponsors a program of research, development, and demonstration related to advanced non-light water reactor concepts. A goal of the program is to facilitate greater engagement between DOE and industry. During FY12, DOE established a Technical Review Panel (TRP) process to identify R&D needs for viable advanced reactor concepts in order to inform DOE-NE R&D investment decisions. That process involved the use of a Request for Information (RFI) to solicit concept information from industry and engage technical experts to evaluate those concepts. Having completed this process, DOE desires to

71

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

72

Proceedings of the US Nuclear Regulatory Commission fifteenth water reactor safety information meeting: Volume 1, Plenary sessions, reactor licensing topics, NUREG-1150, risk analysis/PRA applications, innovative concepts for increased safety of advanced power reactors, severe accident modeling and analysis  

SciTech Connect

This six-volume report contains 140 papers out of the 164 that were presented at the Fifteenth Water Reactor Safety Information Meeting held at the National Bureau of Standards, Gaithersburg, Maryland, during the week of October 26-29, 1987. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. This report, Volume 1, discusses the following: plenary sessions; reactor licensing; NUREG-1150; risk analysis; innovative concepts for increased safety of advanced power reactors; and severe accident modeling and analysis. Thirty-two reports have been cataloged separately.

Weiss, A.J. (comp.)

1988-02-01T23:59:59.000Z

73

Advanced Reactor Research and Development Funding Opportunity Announcement  

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

Reactor Research and Development Funding Opportunity Reactor Research and Development Funding Opportunity Announcement Advanced Reactor Research and Development Funding Opportunity Announcement The U.S. Department of Energy (DOE) Office of Nuclear Energy (NE) sponsors a program of research, development, and demonstration related to advanced non-light water reactor concepts. A goal of the program is to facilitate greater engagement between DOE and industry. During FY12, DOE established a Technical Review Panel (TRP) process to identify R&D needs for viable advanced reactor concepts in order to inform DOE-NE R&D investment decisions. That process involved the use of a Request for Information (RFI) to solicit concept information from industry and engage technical experts to evaluate those concepts. Having completed this process, DOE desires to

74

Nuclear Reactor Safety Design Criteria  

Directives, Delegations, and Requirements

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

1993-01-19T23:59:59.000Z

75

Energy Department Announces New Investments in Advanced Nuclear Power  

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

Investments in Advanced Nuclear Investments in Advanced Nuclear Power Reactors Energy Department Announces New Investments in Advanced Nuclear Power Reactors June 27, 2013 - 2:20pm Addthis News Media Contact (202) 586-4940 WASHINGTON - As part of the Obama Administration's all-of-the-above strategy to accelerate clean energy leadership and to enable a low-carbon economy, the Energy Department today announced $3.5 million for four advanced nuclear reactor projects that go beyond traditional light water designs. These projects -- led by General Atomics, GE Hitachi, Gen4 Energy and Westinghouse -- will address key technical challenges to designing, building and operating the next generation of nuclear reactors. These steps support the President's plan to cut carbon pollution and spark innovation

76

Energy Department Announces New Investments in Advanced Nuclear Power  

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

Energy Department Announces New Investments in Advanced Nuclear Energy Department Announces New Investments in Advanced Nuclear Power Reactors Energy Department Announces New Investments in Advanced Nuclear Power Reactors June 27, 2013 - 2:20pm Addthis News Media Contact (202) 586-4940 WASHINGTON - As part of the Obama Administration's all-of-the-above strategy to accelerate clean energy leadership and to enable a low-carbon economy, the Energy Department today announced $3.5 million for four advanced nuclear reactor projects that go beyond traditional light water designs. These projects -- led by General Atomics, GE Hitachi, Gen4 Energy and Westinghouse -- will address key technical challenges to designing, building and operating the next generation of nuclear reactors. These steps support the President's plan to cut carbon pollution and spark innovation

77

Advanced Reactors Thermal Energy Transport for Process Industries  

SciTech Connect

The operation temperature of advanced nuclear reactors is generally higher than commercial light water reactors and thermal energy from advanced nuclear reactor can be used for various purposes such as liquid fuel production, district heating, desalination, hydrogen production, and other process heat applications, etc. Some of the major technology challenges that must be overcome before the advanced reactors could be licensed on the reactor side are qualification of next generation of nuclear fuel, materials that can withstand higher temperature, improvement in power cycle thermal efficiency by going to combined cycles, SCO2 cycles, successful demonstration of advanced compact heat exchangers in the prototypical conditions, and from the process side application the challenge is to transport the thermal energy from the reactor to the process plant with maximum efficiency (i.e., with minimum temperature drop). The main focus of this study is on doing a parametric study of efficient heat transport system, with different coolants (mainly, water, He, and molten salts) to determine maximum possible distance that can be achieved.

P. Sabharwall; S.J. Yoon; M.G. McKellar; C. Stoots; George Griffith

2014-07-01T23:59:59.000Z

78

Partnerships Help Advance Small Modular Reactor Technology | Department of  

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

Partnerships Help Advance Small Modular Reactor Technology Partnerships Help Advance Small Modular Reactor Technology Partnerships Help Advance Small Modular Reactor Technology March 5, 2012 - 12:00pm Addthis WASHINGTON, D.C. - DOE recently announced three public-private partnerships to develop deployment plans for small modular nuclear reactor (SMR) technologies at Savannah River Site (SRS) facilities near Aiken, S.C. Read the full story on the Memorandums of Agreement to help leverage SRS land assets, energy facilities and nuclear expertise to support potential private sector development, testing and licensing of prototype SMR technologies. Addthis Related Articles Energy Department Announces Small Modular Reactor Technology Partnerships at Savannah River Site The development of clean, affordable nuclear power options is a key element of the Energy Department's 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. | Photo by the Energy Department.

79

Nuclear reactor downcomer flow deflector  

DOE Patents (OSTI)

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

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

2011-02-15T23:59:59.000Z

80

Physics of nuclear reactor safety  

Science Journals Connector (OSTI)

Provides a concise review of the physical aspects of safety of nuclear fission reactors. It covers the developments of roughly the last decade. The introductory chapter contains an analysis of the changes in safety philosophy that are characteristic of the last decade and that have given rise to an increased importance of physical aspects because of the emphasis on passive or natural safety. The second chapter focuses on the basics of reactor safety, identifying the main risk sources and the main principles for a safe design. The third chapter concerns a systematic treatment of the physical processes that are fundamental for the properties of fission chain reacting processes and the control of those processes. Because of the rather specialized nature of the field of reactor physics, each paragraph contains a very concise description of the theory of the phenomenon under consideration, before presenting a review of the developments. Chapter 4 contains a short review of the thermal aspects of reactor safety, restricted to those aspects that are characteristic of the nuclear reactor field, because thermal hydraulics of fission reactors is not principally different from that of other physical systems. In chapter 5 the consequences of the physics treated in the preceding chapters for the dynamics and safety of actual reactors are reviewed. The systematics of the treatment is mainly based on a division of reactors into three categories according to the type of coolant, which to a large extent determines the safety properties of the reactors. The last chapter contains a physical analysis of the Chernobyl accident that occurred in 1986. The reason for an attempt to give a review of this accident, as complete as possible within the space limits set by the editors, is twofold: the Chernobyl accident is the most severe accident in history and physical properties of the reactor played a decisive role, thereby serving as an illustration of the material of the preceding chapters.

H van Dam

1992-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "advanced nuclear reactor" 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

ASME Material Challenges for Advanced Reactor Concepts  

SciTech Connect

This study presents the material Challenges associated with Advanced Reactor Concept (ARC) such as the Advanced High Temperature Reactor (AHTR). ACR are the next generation concepts focusing on power production and providing thermal energy for industrial applications. The efficient transfer of energy for industrial applications depends on the ability to incorporate cost-effective heat exchangers between the nuclear heat transport system and industrial process heat transport system. The heat exchanger required for AHTR is subjected to a unique set of conditions that bring with them several design challenges not encountered in standard heat exchangers. The corrosive molten salts, especially at higher temperatures, require materials throughout the system to avoid corrosion, and adverse high-temperature effects such as creep. Given the very high steam generator pressure of the supercritical steam cycle, it is anticipated that water tube and molten salt shell steam generators heat exchanger will be used. In this paper, the ASME Section III and the American Society of Mechanical Engineers (ASME) Section VIII requirements (acceptance criteria) are discussed. Also, the ASME material acceptance criteria (ASME Section II, Part D) for high temperature environment are presented. Finally, lack of ASME acceptance criteria for thermal design and analysis are discussed.

Piyush Sabharwall; Ali Siahpush

2013-07-01T23:59:59.000Z

82

Advanced thermionic reactor systems design code  

SciTech Connect

An overall systems design code is under development to model an advanced in-core thermionic nuclear reactor system for space applications at power levels of 10 to 50 kWe. The design code is written in an object-oriented programming environment that allows the use of a series of design modules, each of which is responsible for the determination of specific system parameters. The code modules include a neutronics and core criticality module, a core thermal hydraulics module, a thermionic fuel element performance module, a radiation shielding module, a module for waste heat transfer and rejection, and modules for power conditioning and control. The neutronics and core criticality module determines critical core size, core lifetime, and shutdown margins using the criticality calculation capability of the Monte Carlo Neutron and Photon Transport Code System (MCNP). The remaining modules utilize results of the MCNP analysis along with FORTRAN programming to predict the overall system performance.

Lewis, B.R.; Pawlowski, R.A.; Greek, K.J.; Klein, A.C. (Department of Nuclear Engineering, Radiation Center, C116, Oregon State University, Corvallis, Oregon 97331-5902 (US))

1991-01-01T23:59:59.000Z

83

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

84

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

85

Oklo natural nuclear reactor  

Science Journals Connector (OSTI)

Although the possibility of natural was considered in the 1950s, this prediction was not confirmed until 1972, when a routine analysis of a uranium ore standard at the nuclear-fuel processing plant in Pierre...

Mark J. Rigali; Bartholomew Nagy

1998-01-01T23:59:59.000Z

86

Simulation of a marine nuclear reactor  

SciTech Connect

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

Kusunoki, Tsuyoshi; Kyouya, Masahiko; Kobayashi, Hideo; Ochiai, Masaaki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Office of Nuclear Ship Research and Development

1995-02-01T23:59:59.000Z

87

Computer aided nuclear reactor modeling  

E-Print Network (OSTI)

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

Warraich, Khalid Sarwar

2012-06-07T23:59:59.000Z

88

Studies on steam condensation with non-condensable gases in a horizontal condenser tube for advanced nuclear reactors using RELAP5  

Science Journals Connector (OSTI)

Horizontal heat exchangers are used in advanced light water nuclear reactors in their passive cooling systems, such as Residual Heat Removal System (RHRS) and Passive Containment Cooling System (PCCS). Horizontal condensation studies of steam with non-condensable gases mixtures in these heat exchangers are very important. This work presents a comparison between simulation results and experimental data in steady state conditions for some inlet pressure, steam and non-condensable gases (air) inlet mass fractions. The test section was modelled and the simulations were performed with the RELAP5 code. Experimental tests were carried out for 200â??400 kPa inlet pressure and 5%, 10%, 15% and 20% of inlet air mass fractions. Comparisons between experimental data and simulation results are presented for 200 kPa and 400 kPa pressure conditions and showed good agreement. New correlations for heat transfer coefficients in these steam-air conditions must be theoretically and experimentally studied and implemented in the RELAP5 code.

L.A. Macedo; W.M. Torres

2011-01-01T23:59:59.000Z

89

Advanced gas cooled nuclear reactor materials evaluation and development program. Progress report, October 1, 1979-December 31, 1979  

SciTech Connect

This report presents the results of work performed from October 1, 1979 through December 31, 1979. Work covered in this report includes the activities associated with the status of the simulated reactor helium supply system, testing equipment and gas chemistry analysis instrumentation and equipment. The progress in the screening test program is described. This includes: screening creep results, weight gain and post-exposure mechanical properties for materials thermally exposed at 750/sup 0/ and 850/sup 0/C (1382/sup 0/ and 1562/sup 0/F). In addition, the status of the data management system is described.

Not Available

1980-04-18T23:59:59.000Z

90

Draft Advanced Nuclear Energy Solicitation Public Meeting Presentation...  

Office of Environmental Management (EM)

Draft Advanced Nuclear Energy Solicitation Public Meeting Presentation Draft Advanced Nuclear Energy Solicitation Public Meeting Presentation Draft Advanced Nuclear Solicitation...

91

Analysis of nuclear reactor instability phenomena  

SciTech Connect

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

Lahey, R.T. Jr.

1993-01-01T23:59:59.000Z

92

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

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

Electric Power Produced from Nuclear Reactor | National Nuclear Security Administration Facebook Twitter Youtube Flickr RSS People Mission Managing the Stockpile Preventing...

93

Nuclear reactors and the nuclear fuel cycle  

SciTech Connect

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

Pearlman, H

1989-11-01T23:59:59.000Z

94

Risk Informed Safety Margin Characterization (RISMC) Advanced Test Reactor  

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

Risk Informed Safety Margin Characterization (RISMC) Advanced Test Risk Informed Safety Margin Characterization (RISMC) Advanced Test Reactor Demonstration Case Study Risk Informed Safety Margin Characterization (RISMC) Advanced Test Reactor Demonstration Case Study Safety is central to the design, licensing, operation, and economics of Nuclear Power Plants (NPPs). Consequently, the ability to better characterize and quantify safety margin holds the key to improved decision making about light water reactor design, operation, and plant life extension. A systematic approach to characterization of safety margins and the subsequent margins management options represents a vital input to the licensee and regulatory analysis and decision making that will be involved. The purpose of the Risk Informed Safety Margin Characterization (RISMC)

95

Risk Informed Safety Margin Characterization (RISMC) Advanced Test Reactor  

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

(RISMC) Advanced Test (RISMC) Advanced Test Reactor Demonstration Case Study Risk Informed Safety Margin Characterization (RISMC) Advanced Test Reactor Demonstration Case Study Safety is central to the design, licensing, operation, and economics of Nuclear Power Plants (NPPs). Consequently, the ability to better characterize and quantify safety margin holds the key to improved decision making about light water reactor design, operation, and plant life extension. A systematic approach to characterization of safety margins and the subsequent margins management options represents a vital input to the licensee and regulatory analysis and decision making that will be involved. The purpose of the Risk Informed Safety Margin Characterization (RISMC) Pathway research and development (R&D) is to support plant decisions for

96

Advanced nuclear fuel  

SciTech Connect

Kurt Terrani uses his expertise in materials science to develop safer fuel for nuclear power plants.

Terrani, Kurt

2014-07-14T23:59:59.000Z

97

Advanced nuclear fuel  

ScienceCinema (OSTI)

Kurt Terrani uses his expertise in materials science to develop safer fuel for nuclear power plants.

Terrani, Kurt

2014-07-15T23:59:59.000Z

98

March 2006, Report of the ADVANCED NUCLEAR TRANSFORMATION TECHNOLOGY  

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

March 2006, Report of the ADVANCED NUCLEAR TRANSFORMATION March 2006, Report of the ADVANCED NUCLEAR TRANSFORMATION TECHNOLOGY SUBCOMMITTEE of the NUCLEAR ENERGY RESEARCH ADVISORY COMMITTEE March 2006, Report of the ADVANCED NUCLEAR TRANSFORMATION TECHNOLOGY SUBCOMMITTEE of the NUCLEAR ENERGY RESEARCH ADVISORY COMMITTEE The Global Nuclear Energy Partnership (GNEP) marks a major change in the direction of the DOE's nuclear energy R&D program. It is a coherent plan to test technologies that promise to markedly reduce the problem of nuclear waste treatment and to reduce the proliferation risk in a world with a greatly expanded nuclear power program. It brings the U.S. program into much closer alignment with that of the other major nuclear energy states. GNEP proposes to take spent fuel from existing light water reactors (LWRs),

99

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

100

Inherent safety concepts in nuclear power reactors  

Science Journals Connector (OSTI)

Different inherent safety concepts being considered in fast and thermal reactors are presented after outlining the basic goals of nuclear reactor safety, the ‘defence in depth’ philosophy to achieve these goal...

O M Pal Singh; R Shankar Singh

1989-06-01T23:59:59.000Z

Note: This page contains sample records for the topic "advanced nuclear reactor" 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

A future for nuclear energy: pebble bed reactors  

Science Journals Connector (OSTI)

Pebble Bed Reactors could allow nuclear plants to support the goal of reducing global climate change in an energy hungry world. They are small, modular, inherently safe, use a demonstrated nuclear technology and can be competitive with fossil fuels. Pebble bed reactors are helium cooled reactors that use small tennis ball size fuel balls consisting of only 9 grams of uranium per pebble to provide a low power density reactor. The low power density and large graphite core provide inherent safety features such that the peak temperature reached even under the complete loss of coolant accident without any active emergency core cooling system is significantly below the temperature that the fuel melts. This feature should enhance public confidence in this nuclear technology. With advanced modularity principles, it is expected that this type of design and assembly could lower the cost of new nuclear plants removing a major impediment to deployment.

Andrew C. Kadak

2005-01-01T23:59:59.000Z

102

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

103

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

104

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

105

Nuclear reactor multiphysics via bond graph formalism  

E-Print Network (OSTI)

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

Sosnovsky, Eugeny

2014-01-01T23:59:59.000Z

106

Consortium for Advanced Simulation of Light Water Reactors (CASL...  

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

Presentations 2015 back to top Smith, K., Advances in Reactor Physics and Computational Science, Physor 2014 International Conference, "The Role of Reactor Physics toward a...

107

Advanced Neutron Source Reactor thermal analysis of fuel plate defects  

SciTech Connect

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

Giles, G.E.

1995-08-01T23:59:59.000Z

108

Generating Unstructured Nuclear Reactor Core Meshes in Parallel  

Science Journals Connector (OSTI)

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

Rajeev Jain; Timothy J. Tautges

2014-01-01T23:59:59.000Z

109

Advanced Gas Reactor Fuel Program's TRISO Particle Fuel Sets A New World  

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

Advanced Gas Reactor Fuel Program's TRISO Particle Fuel Sets A New Advanced Gas Reactor Fuel Program's TRISO Particle Fuel Sets A New World Record For Irradiation Performance Advanced Gas Reactor Fuel Program's TRISO Particle Fuel Sets A New World Record For Irradiation Performance November 16, 2009 - 1:12pm Addthis As part of the Office of Nuclear Energy's Next Generation Nuclear Plant (NGNP) Program, the Advanced Gas Reactor (AGR) Fuel Development Program has achieved a new international record for irradiation testing of next-generation particle fuel for use in high temperature gas reactors (HTGRs). The AGR Fuel Development Program was initiated by the Department of Energy in 2002 to develop the advanced fabrication and characterization technologies, and provide irradiation and safety performance data required to license TRISO particle fuel for the NGNP and future HTGRs. The AGR

110

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, V.T.

1993-04-06T23:59:59.000Z

111

Nuclear Data Measurements for 21st Century Reactor Physics Applications  

SciTech Connect

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

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

2003-03-01T23:59:59.000Z

112

Consortium for Advanced Simulation of Light Water Reactors (CASL...  

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

behavior in AP1000 reactor core Test run signals emergence of the next generation in nuclear power reactor analysis tools OAK RIDGE, Tenn., Feb. 18, 2014 - Scientists and...

113

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

114

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

115

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

116

Development of advanced strain diagnostic techniques for reactor environments.  

SciTech Connect

The following research is operated as a Laboratory Directed Research and Development (LDRD) initiative at Sandia National Laboratories. The long-term goals of the program include sophisticated diagnostics of advanced fuels testing for nuclear reactors for the Department of Energy (DOE) Gen IV program, with the future capability to provide real-time measurement of strain in fuel rod cladding during operation in situ at any research or power reactor in the United States. By quantifying the stress and strain in fuel rods, it is possible to significantly improve fuel rod design, and consequently, to improve the performance and lifetime of the cladding. During the past year of this program, two sets of experiments were performed: small-scale tests to ensure reliability of the gages, and reactor pulse experiments involving the most viable samples in the Annulated Core Research Reactor (ACRR), located onsite at Sandia. Strain measurement techniques that can provide useful data in the extreme environment of a nuclear reactor core are needed to characterize nuclear fuel rods. This report documents the progression of solutions to this issue that were explored for feasibility in FY12 at Sandia National Laboratories, Albuquerque, NM.

Fleming, Darryn D.; Holschuh, Thomas Vernon,; Miller, Timothy J.; Hall, Aaron Christopher; Urrea, David Anthony,; Parma, Edward J.,

2013-02-01T23:59:59.000Z

117

ADVANCED NUCLEAR TRANSFORMATION TECHNOLOGY SUBCOMMITTEE  

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

Report of Report of ADVANCED NUCLEAR TRANSFORMATION TECHNOLOGY SUBCOMMITTEE of the NUCLEAR ENERGY RESEARCH ADVISORY COMMITTEE 24 October 2003 BURTON RICHTER, CHAIR DARLEANE C. HOFFMAN SEKAZI K. MTINGWA RONALD P. OMBERG SILVIE PILLON JOY L. REMPE I. INTRODUCTION AND SUMMARY The committee met in Washington on September 16 and 17 to review progress in the program with respect to a changed set of mission priorities. Our last meeting took place in December 2002 after the reorganization that had placed the Advanced Fuel Cycle Initiative (AFCI) and the GEN IV program together in the Advanced Nuclear Research Office (AN-20). Since mission priorities have been evolving, the committee felt that it should wait until they have settled down before we met again. We have kept in touch

118

Experimental investigations on decay heat removal in advanced nuclear reactors using single heater rod test facility: Air alone in the annular gap  

SciTech Connect

During a loss of coolant accident in nuclear reactors, radiation heat transfer accounts for a significant amount of the total heat transfer in the fuel bundle. In case of heavy water moderator nuclear reactors, the decay heat of a fuel bundle enclosed in the pressure tube and outer concentric calandria tube can be transferred to the moderator. Radiation heat transfer plays a significant role in removal of decay heat from the fuel rods to the moderator, which is available outside the calandria tube. A single heater rod test facility is designed and fabricated as a part of preliminary investigations. The objective is to anticipate the capability of moderator to remove decay heat, from the reactor core, generated after shut down. The present paper focuses mainly on the role of moderator in removal of decay heat, for situation with air alone in the annular gap of pressure tube and calandria tube. It is seen that the naturally aspirated air is capable of removing the heat generated in the system compared to the standstill air or stagnant water situations. It is also seen that the flowing moderator is capable of removing a greater fraction of heat generated by the heater rod compared to a stagnant pool of boiling moderator. (author)

Bopche, Santosh B.; Sridharan, Arunkumar [Department of Mechanical Engineering, Indian Institute of Technology Bombay, Powai, Mumbai 400 076 (India)

2010-11-15T23:59:59.000Z

119

October 2006, Report of the ADVANCED NUCLEAR TRANSFORMATION TECHNOLOGY  

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

6, Report of the ADVANCED NUCLEAR TRANSFORMATION 6, Report of the ADVANCED NUCLEAR TRANSFORMATION TECHNOLOGY SUBCOMMITTEE of the NUCLEAR ENERGY RESEARCH ADVISORY COMMITTEE October 2006, Report of the ADVANCED NUCLEAR TRANSFORMATION TECHNOLOGY SUBCOMMITTEE of the NUCLEAR ENERGY RESEARCH ADVISORY COMMITTEE The Global Nuclear Energy Partnership (GNEP) program is still evolving. Since our report of March 22, 2006 the DOE has sought to gauge industry interest in participation in the program from its very beginning. At the time the ANTT committee met, August 30- 31, 2006, responses had not yet been received from industry to the DOE's request for Expressions of Interest. This report is based on the assumption that the program outlined recently, which does not include an Advanced Burner Test Reactor, is what

120

Advanced Test Reactor National Scientific User Facility  

SciTech Connect

The Advanced Test Reactor (ATR), at the Idaho National Laboratory (INL), is a large test reactor for providing the capability for studying the effects of intense neutron and gamma radiation on reactor materials and fuels. The ATR is a pressurized, light-water, high flux test reactor with a maximum operating power of 250 MWth. The INL also has several hot cells and other laboratories in which irradiated material can be examined to study material irradiation effects. In 2007 the US Department of Energy (DOE) designated the ATR as a National Scientific User Facility (NSUF) to facilitate greater access to the ATR and the associated INL laboratories for material testing research by a broader user community. This paper highlights the ATR NSUF research program and the associated educational initiatives.

Frances M. Marshall; Jeff Benson; Mary Catherine Thelen

2011-08-01T23:59:59.000Z

Note: This page contains sample records for the topic "advanced nuclear reactor" 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

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

122

New Fuel Cycle and Fuel Management Options in Heavy Liquid Metal-Cooled Reactors  

Science Journals Connector (OSTI)

Technical Paper / Advances in Nuclear Fuel Management - Fuel Management of Reactors Other Than Light Water Reactors

Ehud Greenspan; Pavel Hejzlar; Hiroshi Sekimoto; Georgy Toshinsky; David Wade

123

Feasibility of Burning First- and Second-Generation Plutonium in Pebble Bed High-Temperature Reactors  

Science Journals Connector (OSTI)

Technical Paper / Advances in Nuclear Fuel Management - Fuel Management of Reactors Other Than Light Water Reactors

J. B. M. De Haas; J. C. Kuijper

124

Argonne step closer to safer nuclear reactor  

Science Journals Connector (OSTI)

Argonne step closer to safer nuclear reactor ... "A key technological link" toward development of meltdown-immune nuclear reactors is now in the demonstration phase at Argonne National Laboratory near Chicago. ... The technique is part of Argonne's continuing interest in the sodium-cooled integral fast reactor (IFR), whose immunity to meltdown derives from molten sodium's function as a heat sink and the use of metallic fuel that conducts heat better than conventional oxide fuels. ...

WARD WORTHY

1988-05-30T23:59:59.000Z

125

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

126

Sterile Neutrino Search Using China Advanced Research Reactor  

E-Print Network (OSTI)

We study the feasibility of a sterile neutrino search at the China Advanced Research Reactor by measuring $\\bar {\

Gang Guo; Fang Han; Xiangdong Ji; Jianglai Liu; Zhaoxu Xi; Huanqiao Zhang

2013-06-18T23:59:59.000Z

127

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

128

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

129

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

130

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

131

Nuclear reactor internals alignment configuration  

DOE Patents (OSTI)

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

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

2009-11-10T23:59:59.000Z

132

SciTech Connect: Nuclear power reactor instrumentation systems...  

Office of Scientific and Technical Information (OSTI)

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

133

Features of a subcritical nuclear reactor  

Science Journals Connector (OSTI)

Abstract A subcritical nuclear reactor is a device where the nuclear-fission chain reaction is initiated and maintained using an external neutron source. It is a valuable educational and research tool where in a safe way many reactor parameters can be measured. Here, we have used the six-factor formula to calculate the effective multiplication factor of a subcritical nuclear reactor Nuclear Chicago model 9000. Using the MCNP5 code, a three-dimensional model of the subcritical reactor was developed to estimate the effective multiplication factor, the neutron spectra, and the total and thermal neutron fluences along the radial and axial axis. The MCNP5 results of the effective multiplication factor were compared with those obtained from the six-factor formula. The effective dose and the Ambient dose equivalent, at three sites outside the reactor, were estimated; the Ambient dose equivalent was also measured and compared with the calculated values.

Hector Rene Vega-Carrillo; Isvi Ruben Esparza-Garcia; Alvaro Sanchez

2015-01-01T23:59:59.000Z

134

Mirror Advanced Reactor Study interim design report  

SciTech Connect

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

135

Advanced nuclear precleaner  

SciTech Connect

This Phase II Small Business Innovation Research (SBIR) program`s goal is to develop a dynamic, self-cleaning air precleaner for high-efficiency particulate air (HEPA) filtration systems that would extend significantly the life of HEPA filter banks by reducing the particulate matter that causes filter fouling and increased pack pressure. HEPA filters are widely used in DOE, Department of Defense, and a variety of commercial facilities. InnovaTech, Inc. (Formerly Micro Composite materials Corporation) has developed a proprietary dynamic separation device using a concept called Boundary Layer Momentum Transfer (BLMT) to extract particulate matter from fluid process streams. When used as a prefilter in the HVAC systems or downstream of waste vitrifiers in nuclear power plants, fuel processing facilities, and weapons decommissioning factories, the BLMT filter will dramatically extend the service life and increase the operation efficiency of existing HEPA filtration systems. The BLMT filter is self cleaning, so there will be no degraded flow or increased pressure drop. Because the BLMT filtration process is independent of temperature, it can be designed to work in ambient, medium, or high-temperature applications. During Phase II, the authors are continuing development of the computerized flow simulation model to include turbulence and incorporate expansion into a three-dimensional model that includes airflow behavior inside the filter housing before entering the active BLMT device. A full-scale (1000 ACFM) prototype filter is being designed to meet existing HEPA filter standards and will be fabricated for subsequent testing. Extensive in-house testing will be performed to determine a full range of performance characteristics. Final testing and evaluation of the prototype filter will be conducted at a DOE Quality Assurance Filter Test Station.

Wright, S.R. [InnovaTech, Inc., Durham, NC (United States)

1997-10-01T23:59:59.000Z

136

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

137

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

138

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

139

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

140

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

Note: This page contains sample records for the topic "advanced nuclear reactor" 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

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

142

Dynamic detection of nuclear reactor core incident  

Science Journals Connector (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

143

SAFETY AND RELIABILITY ANALYSIS OF NUCLEAR REACTORS  

Science Journals Connector (OSTI)

Abstract A survey of the various aspects of safety and reliability analysis of nuclear reactors is presented with particular emphasis on the interrelation between structural reliability and systems reliability. In reactor design this interrelation is of overriding importance since it is the task of the control, protective and containment systems to protect the mechanical system and the structure from accidental overloading.

T.A. JAEGER

1972-01-01T23:59:59.000Z

144

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

145

Advanced Test Reactor National Scientific User Facility Partnerships  

SciTech Connect

In 2007, the United States Department of Energy designated the Advanced Test Reactor (ATR), located at Idaho National Laboratory, as a National Scientific User Facility (NSUF). This designation made test space within the ATR and post-irradiation examination (PIE) equipment at INL available for use by researchers via a proposal and peer review process. The goal of the ATR NSUF is to provide researchers with the best ideas access to the most advanced test capability, regardless of the proposer's physical location. Since 2007, the ATR NSUF has expanded its available reactor test space, and obtained access to additional PIE equipment. Recognizing that INL may not have all the desired PIE equipment, or that some equipment may become oversubscribed, the ATR NSUF established a Partnership Program. This program enables and facilitates user access to several university and national laboratories. So far, seven universities and one national laboratory have been added to the ATR NSUF with capability that includes reactor-testing space, PIE equipment, and ion beam irradiation facilities. With the addition of these universities, irradiation can occur in multiple reactors and post-irradiation exams can be performed at multiple universities. In each case, the choice of facilities is based on the user's technical needs. Universities and laboratories included in the ATR NSUF partnership program are as follows: (1) Nuclear Services Laboratories at North Carolina State University; (2) PULSTAR Reactor Facility at North Carolina State University; (3) Michigan Ion Beam Laboratory (1.7 MV Tandetron accelerator) at the University of Michigan; (4) Irradiated Materials at the University of Michigan; (5) Harry Reid Center Radiochemistry Laboratories at University of Nevada, Las Vegas; (6) Characterization Laboratory for Irradiated Materials at the University of Wisconsin-Madison; (7) Tandem Accelerator Ion Beam. (1.7 MV terminal voltage tandem ion accelerator) at the University of Wisconsin-Madison; (8) Illinois Institute of Technology (IIT) Materials Research Collaborative Access Team (MRCAT) beamline at Argonne National Laboratory's Advanced Photon Source; and (9) Nanoindenter in the University of California at Berkeley (UCB) Nuclear Engineering laboratory Materials have been analyzed for ATR NSUF users at the Advanced Photon Source at the MRCAT beam, the NIST Center for Neutron Research in Gaithersburg, MD, the Los Alamos Neutron Science Center, and the SHaRE user facility at Oak Ridge National Laboratory (ORNL). Additionally, ORNL has been accepted as a partner facility to enable ATR NSUF users to access the facilities at the High Flux Isotope Reactor and related facilities.

Frances M. Marshall; Todd R. Allen; Jeff B. Benson; James I. Cole; Mary Catherine Thelen

2012-03-01T23:59:59.000Z

146

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

147

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

148

AdvAnced  

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

AdvAnced test reActor At the InL advanced Unlike large, commercial power reactors, ATR is a low- temperature, low-pressure reactor. A nuclear reactor is basically an elaborate...

149

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

150

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

Energy Savers (EERE)

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

151

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

152

Effects of an Advanced Reactor’s Design, Use of Automation, and Mission on Human Operators  

SciTech Connect

The roles, functions, and tasks of the human operator in existing light water nuclear power plants (NPPs) are based on sound nuclear and human factors engineering (HFE) principles, are well defined by the plant’s conduct of operations, and have been validated by years of operating experience. However, advanced NPPs whose engineering designs differ from existing light-water reactors (LWRs) will impose changes on the roles, functions, and tasks of the human operators. The plans to increase the use of automation, reduce staffing levels, and add to the mission of these advanced NPPs will also affect the operator’s roles, functions, and tasks. We assert that these factors, which do not appear to have received a lot of attention by the design engineers of advanced NPPs relative to the attention given to conceptual design of these reactors, can have significant risk implications for the operators and overall plant safety if not mitigated appropriately. This paper presents a high-level analysis of a specific advanced NPP and how its engineered design, its plan to use greater levels of automation, and its expanded mission have risk significant implications on operator performance and overall plant safety.

Jeffrey C. Joe; Johanna H. Oxstrand

2014-06-01T23:59:59.000Z

153

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

Science Journals Connector (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

154

The Basics of Nuclear Fusion Reactor Using Solid Pycnodeuterium as Nuclear Fuel  

Science Journals Connector (OSTI)

......Articles The Basics of Nuclear Fusion Reactor Using Solid Pycnodeuterium as...2004 241 The Basics of Nuclear Fusion Reactor Using Solid Pycnodeuterium as...2.1) The Basics of Nuclear Fusion Reactor 243 Fig. 3. Relation between......

Yoshiaki Arata; Yue Chang Zhang

2004-02-01T23:59:59.000Z

155

Advances in thermal hydraulic and neutronic simulation for reactor analysis and safety  

SciTech Connect

This paper describes several large-scale computational models developed at Argonne National Laboratory for the simulation and analysis of thermal-hydraulic and neutronic events in nuclear reactors and nuclear power plants. The impact of advanced parallel computing technologies on these computational models is emphasized.

Tentner, A.M.; Blomquist, R.N.; Canfield, T.R.; Ewing, T.F.; Garner, P.L.; Gelbard, E.M.; Gross, K.C.; Minkoff, M.; Valentin, R.A.

1993-03-01T23:59:59.000Z

156

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

157

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

158

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

159

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

SciTech Connect

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

D. E. Shropshire

2009-01-01T23:59:59.000Z

160

Nuclear power reactor education and training at the Ford nuclear reactor  

SciTech Connect

Since 1977, staff members of the University of Michigan's Ford nuclear reactor have provided courses and reactor laboratory training programs for reactor operators, engineers, and technicians from seven electric utilities, including Cleveland Electric Illuminating, Consumers Power, Detroit Edison, Indiana and Michigan Electric, Nebraska Public Power, Texas Utilities Generating Company, and Toledo Edison. Reactor laboratories, instrument technician training, and reactor physics courses have been conducted at the university. Courses conducted at plant sites include reactor physics, thermal sciences, materials sciences, and health physics and radiation protection.

Burn, R.R.

1989-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "advanced nuclear reactor" 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

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

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

TABLE 1. Nuclear Reactor, State, Type, Net Capacity, Generation, and Capacity Factor " "PlantReactor Name","Generator ID","State","Type","2009 Summer Capacity"," 2010 Annual...

162

Table 3. Nuclear Reactor Characteristics and Operational History  

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

3. Nuclear Reactor Characteristics and Operational History" "Plant Name","Generator ID","Type","Reactor Supplier and Model","Construction Start","Grid Connection","Commercial...

163

Advanced Reactor Concepts Technical Review Panel Report | Department of  

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

Advanced Reactor Concepts Technical Review Panel Report Advanced Reactor Concepts Technical Review Panel Report Advanced Reactor Concepts Technical Review Panel Report This report documents the establishment of a technical review process and the findings of the Advanced Reactor Concepts (ARC) Technical Review Panel (TRP).1 The intent of the process is to identify R&D needs for viable advanced reactor concepts in order to inform DOE-NE R&D investment decisions. A goal of the process is to facilitate greater engagement between DOE and industry. The process involved establishing evaluation criteria, conducting a pilot review, soliciting concept inputs from industry entities, reviewing the concepts by TRP members and compiling the results. The eight concepts received from industry spanned a range of reactor types

164

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-division standing and written consent of instructor. Textbook(s): Knief, Nuclear Engineering, 2 nd Edition. Other 361E ­ Nuclear Reactor Engineering Page 2 ABET EC2000 syllabus Contribution of Course to Meeting

Ben-Yakar, Adela

165

Consortium for Advanced Simulation of Light Water Reactors (CASL...  

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

Media Center News Obama highlights next generation nuclear reactors in the SOTU Posted: January 27, 2011 President Obama, in his State of the Union address Tuesday, cited work...

166

Consortium for Advanced Simulation of Light Water Reactors (CASL...  

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

has designed and operated 52 test reactors, including EBR-1, the world's first nuclear power plant Key Contributions System safety analysis Multiscale fuel performance...

167

Advanced nuclear plant control complex  

DOE Patents (OSTI)

An advanced control room complex for a nuclear power plant, including a discrete indicator and alarm system (72) which is nuclear qualified for rapid response to changes in plant parameters and a component control system (64) which together provide a discrete monitoring and control capability at a panel (14-22, 26, 28) in the control room (10). A separate data processing system (70), which need not be nuclear qualified, provides integrated and overview information to the control room and to each panel, through CRTs (84) and a large, overhead integrated process status overview board (24). The discrete indicator and alarm system (72) and the data processing system (70) receive inputs from common plant sensors and validate the sensor outputs to arrive at a representative value of the parameter for use by the operator during both normal and accident conditions, thereby avoiding the need for him to assimilate data from each sensor individually. The integrated process status board (24) is at the apex of an information hierarchy that extends through four levels and provides access at each panel to the full display hierarchy. The control room panels are preferably of a modular construction, permitting the definition of inputs and outputs, the man machine interface, and the plant specific algorithms, to proceed in parallel with the fabrication of the panels, the installation of the equipment and the generic testing thereof.

Scarola, Kenneth (Windsor, CT); Jamison, David S. (Windsor, CT); Manazir, Richard M. (North Canton, CT); Rescorl, Robert L. (Vernon, CT); Harmon, Daryl L. (Enfield, CT)

1993-01-01T23:59:59.000Z

168

Export possibilities for small nuclear reactors  

SciTech Connect

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

169

Oklo reactors and implications for nuclear science  

E-Print Network (OSTI)

We summarize the nuclear physics interests in the Oklo natural nuclear reactors, focusing particularly on developments over the past two decades. Modeling of the reactors has become increasingly sophisticated, employing Monte Carlo simulations with realistic geometries and materials that can generate both the thermal and epithermal fractions. The water content and the temperatures of the reactors have been uncertain parameters. We discuss recent work pointing to lower temperatures than earlier assumed. Nuclear cross sections are input to all Oklo modeling and we discuss a parameter, the $^{175}$Lu ground state cross section for thermal neutron capture leading to the isomer $^{176\\mathrm{m}}$ Lu, that warrants further investigation. Studies of the time dependence of dimensionless fundamental constants have been a driver for much of the recent work on Oklo. We critically review neutron resonance energy shifts and their dependence on the fine structure constant $\\alpha$ and the ratio $X_q=m_q/\\Lambda$ (where $m_...

Davis, E D; Sharapov, E I

2014-01-01T23:59:59.000Z

170

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

171

Compiled reports on the applicability of selected codes and standards to advanced reactors  

SciTech Connect

The following papers were prepared for the Office of Nuclear Regulatory Research of the U.S. Nuclear Regulatory Commission under contract DE-AC06-76RLO-1830 NRC FIN L2207. This project, Applicability of Codes and Standards to Advance Reactors, reviewed selected mechanical and electrical codes and standards to determine their applicability to the construction, qualification, and testing of advanced reactors and to develop recommendations as to where it might be useful and practical to revise them to suit the (design certification) needs of the NRC.

Benjamin, E.L.; Hoopingarner, K.R.; Markowski, F.J.; Mitts, T.M.; Nickolaus, J.R.; Vo, T.V.

1994-08-01T23:59:59.000Z

172

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

173

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

174

STATEMENT OF CONSIDERATIONS Advance Test Reactor Class Waiver  

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

Advance Test Reactor Class Waiver Advance Test Reactor Class Waiver W(C)-2008-004 The Advanced Test Reactor (A TR) is a pressurized water test reactor at the Idaho National Laboratory (INL) that operates at low pressure and temperature. The ATR was originally designed to study the effects of intense radiation on reactor material and fuels . It has a "Four Leaf Clover" design that allows a diverse array of testing locations. The unique design allows for different flux in various locations and specialized systems also allow for certain experiments to be run at their own temperature and pressure. The U.S. Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF) in April 2007. This designation will allow the ATR to

175

A probabilistic safety analysis of incidents in nuclear research reactors  

Science Journals Connector (OSTI)

......System for Research Reactor (IRSRR). Available...System for Research Reactor (IRSRR). Available...76. 7 Manual on reliability data collection for research reactor PSAs. (1992) IAEA...probabilistic safety analysis of incidents in nuclear......

Valdir Maciel Lopes; Gian Maria Agostinho Angelo Sordi; Mauricio Moralles; Tufic Madi Filho

2012-06-01T23:59:59.000Z

176

TABLE 2. U.S. Nuclear Reactor Ownership Data  

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

2. U.S. Nuclear Reactor Ownership Data" "PlantReactor Name","Generator ID","Utility Name - Operator","Owner Name","% Owned" "Arkansas Nuclear One",1,"Entergy Arkansas...

177

Gamma-ray Energy Spectra Observed around a Nuclear Reactor  

Science Journals Connector (OSTI)

......Energy Spectra Observed around a Nuclear Reactor Yoshiyuki Nakashima * Susumu Minato...Katsurayama ** * Department of Nuclear Engineering, Faculty of Engineering...Nagoya, Japan ** Reseach Reactor Institute, Kyoto Univ., Kumatori-cho......

Yoshiyuki Nakashima; Susumu Minato; Minoru Kawano; Tadashi Tsujimoto; Kousuke Katsurayama

1971-09-01T23:59:59.000Z

178

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

Energy Savers (EERE)

Nuclear Reactor Facility Operations - December 4, 2014 (EA CRAD 31-08, Rev. 0) CRAD, Nuclear Reactor Facility Operations - December 4, 2014 (EA CRAD 31-08, Rev. 0) December 4, 2014...

179

NUCLEAR POWER AND RESEARCH REACTORS 1939 1942 1943 1944  

E-Print Network (OSTI)

#12;#12;11 #12;2 NUCLEAR POWER AND RESEARCH REACTORS 1939 1942 1943 1944 Nuclear fission discovered Oak Ridge selected as site for World War II Manhattan Project First sustained and controlled nuclear 430 nuclear power reactors are operating in the world, and 103 nuclear power plants produce 20

Pennycook, Steve

180

Nuclear reactor alignment plate configuration  

DOE Patents (OSTI)

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

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

2014-01-28T23:59:59.000Z

Note: This page contains sample records for the topic "advanced nuclear reactor" 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

The siting of UK nuclear reactors  

Science Journals Connector (OSTI)

Choosing a suitable site for a nuclear power station requires the consideration and balancing of several factors. Some 'physical' site characteristics, such as the local climate and the potential for seismic activity, will be generic to all reactors designs, while others, such as the availability of cooling water, the area of land required and geological conditions capable of sustaining the weight of the reactor and other buildings will to an extent be dependent on the particular design of reactor chosen (or alternatively the reactor design chosen may to an extent be dependent on the characteristics of an available site). However, one particularly interesting tension is a human and demographic one. On the one hand it is beneficial to place nuclear stations close to centres of population, to reduce transmission losses and other costs (including to the local environment) of transporting electricity over large distances from generator to consumer. On the other it is advantageous to place nuclear stations some distance away from such population centres in order to minimise the potential human consequences of a major release of radioactive materials in the (extremely unlikely) event of a major nuclear accident, not only in terms of direct exposure but also concerning the management of emergency planning, notably evacuation.This paper considers the emergence of policies aimed at managing this tension in the UK. In the first phase of nuclear development (roughly speaking 1945–1965) there was a highly cautious attitude, with installations being placed in remote rural locations with very low population density. The second phase (1965–1985) saw a more relaxed approach, allowing the development of AGR nuclear power stations (which with concrete pressure vessels were regarded as significantly safer) closer to population centres (in 'semi-urban' locations, notably at Hartlepool and Heysham). In the third phase (1985–2005) there was very little new nuclear development, Sizewell B (the first and so far only PWR power reactor in the UK) being colocated with an early Magnox station on the rural Suffolk coast. Renewed interest in nuclear new build from 2005 onward led to a number of sites being identified for new reactors before 2025, all having previously hosted nuclear stations and including the semi-urban locations of the 1960s and 1970s. Finally, some speculative comments are made as to what a 'fifth phase' starting in 2025 might look like.

Malcolm Grimston; William J Nuttall; Geoff Vaughan

2014-01-01T23:59:59.000Z

182

Massachusetts Institute of Technology Department of Nuclear Engineering  

E-Print Network (OSTI)

Massachusetts Institute of Technology Department of Nuclear Engineering Advanced Reactor Technology of Technology Department of Nuclear Engineering Advanced Reactor Technology Pebble Bed Project MPBR-2 Student Department of Nuclear Engineering Advanced Reactor Technology Pebble Bed Project MPBR-3 Project Objective

183

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

184

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

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

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

185

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

186

Proceedings of the 1992 topical meeting on advances in reactor physics. Volume 1  

SciTech Connect

This document, Volume 1, presents proceedings of the 1992 Topical Meeting on Advances in Reactor Physics on March 8--11, 1992 at Charleston, SC. Session topics were as follows: Code Benchmarks and Validation; Fuel Management; Nodal Methods for Diffusion Theory; Criticality Safety and Applications and Waste; Core Computational Systems; Nuclear Data; Safety Aspects of Heavy Water Reactors; and Space-Time Core Kinetics. The individual papers have been cataloged separately. (FI)

Not Available

1992-04-01T23:59:59.000Z

187

Proceedings of the 1992 topical meeting on advances in reactor physics  

SciTech Connect

This document, Volume 1, presents proceedings of the 1992 Topical Meeting on Advances in Reactor Physics on March 8--11, 1992 at Charleston, SC. Session topics were as follows: Code Benchmarks and Validation; Fuel Management; Nodal Methods for Diffusion Theory; Criticality Safety and Applications and Waste; Core Computational Systems; Nuclear Data; Safety Aspects of Heavy Water Reactors; and Space-Time Core Kinetics. The individual papers have been cataloged separately. (FI)

Not Available

1992-01-01T23:59:59.000Z

188

ARTIGO INTERNET Professores visitam o maior reactor de Fuso Nuclear  

E-Print Network (OSTI)

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

Instituto de Sistemas e Robotica

189

APPLICATION OF DATA ANALYSIS TECHNIQUES TO NUCLEAR REACTOR  

E-Print Network (OSTI)

1 APPLICATION OF DATA ANALYSIS TECHNIQUES TO NUCLEAR REACTOR SYSTEMS CODE ACCURACY ASSESSMENT the inherent safety (and concomitant code reliability and licensing) concerns associated with nuclear reactors) has been developed by the authors to provide quantitative comparisons between nuclear reactor systems

Kunz, Robert Francis

190

New Reactor Concepts and New Nuclear Data Needed to Develop Them  

Science Journals Connector (OSTI)

Developments of new reactor designs for the utilization of thorium such as the Advanced Heavy Water Reactor especially demand creation of new nuclear data for all the isotopes of the thorium fuel cycle. Improved nuclear data are essential to support new initiatives such as the international project on innovative nuclear reactors and fuel cycles (INPRO) which aims to support the safe sustainable economic and proliferation?resistant use of nuclear technology to meet the global energy needs of the 21st century. The detailed pursuit of development of Generation IV nuclear energy systems that offer advantages in the areas of economics safety reliability and sustainability require significantly improved nuclear data. The development of Accelerator Driven Sub?critical Systems proposed by Carlo Rubbia and others require a significant amount of new nuclear data in extended energy regions and improvement of the presently available nuclear data. The quality assurance in design and safety studies in nuclear energy in the next few decades and centuries require new and improved nuclear data with high accuracy and energy resolution. The paper presents from the perspective of the author an overview of some of the improvements in nuclear data required for a sound scientific basis of advanced nuclear systems. Also from the perspective of benchmarking and integral validation of nuclear data presented briefly is the status of thorium irradiations performed in a Pressurized Heavy Water Reactor (PHWR) in India and new results of post?irradiation analyses available thus far.

S. Ganesan

2005-01-01T23:59:59.000Z

191

Advanced Nuclear Medicine Initiative Owen Lowe  

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

Isotopes for Life Isotopes for Life Isotopes for Life Advanced Nuclear Medicine Initiative Owen Lowe Office of Isotopes for Medicine and Science Office of Nuclear Energy, Science and Technology October 1, 2002 Isotopes for Life Isotopes for Life Isotopes for Life Office of Nuclear Energy, Science and Technology Lowe/Oct01_02 ANMI to NERAC.ppt (2) Advanced Nuclear Medicine Initiative Advanced Nuclear Medicine Initiative 6 Sponsor nuclear medical science research using a peer-review selection process * 9 three-year research grants awarded 6 Sponsor the training of individuals in nuclear medical science * 5 three-year education grants awarded 6 Continue research and education programs to completion; however, no additional funds for new grants is in the FY 2003 budget Isotopes for Life Isotopes for Life

192

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

DOE Patents (OSTI)

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

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

2013-09-03T23:59:59.000Z

193

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

DOE Patents (OSTI)

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

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

2011-03-01T23:59:59.000Z

194

Advanced ceramic materials for next-generation nuclear applications  

Science Journals Connector (OSTI)

The nuclear industry is at the eye of a 'perfect storm' with fuel oil and natural gas prices near record highs, worldwide energy demands increasing at an alarming rate, and increased concerns about greenhouse gas (GHG) emissions that have caused many to look negatively at long-term use of fossil fuels. This convergence of factors has led to a growing interest in revitalization of the nuclear power industry within the United States and across the globe. Many are surprised to learn that nuclear power provides approximately 20% of the electrical power in the US and approximately 16% of the world-wide electric power. With the above factors in mind, world-wide over 130 new reactor projects are being considered with approximately 25 new permit applications in the US. Materials have long played a very important role in the nuclear industry with applications throughout the entire fuel cycle; from fuel fabrication to waste stabilization. As the international community begins to look at advanced reactor systems and fuel cycles that minimize waste and increase proliferation resistance, materials will play an even larger role. Many of the advanced reactor concepts being evaluated operate at high-temperature requiring the use of durable, heat-resistant materials. Advanced metallic and ceramic fuels are being investigated for a variety of Generation IV reactor concepts. These include the traditional TRISO-coated particles, advanced alloy fuels for 'deep-burn' applications, as well as advanced inert-matrix fuels. In order to minimize wastes and legacy materials, a number of fuel reprocessing operations are being investigated. Advanced materials continue to provide a vital contribution in 'closing the fuel cycle' by stabilization of associated low-level and high-level wastes in highly durable cements, ceramics, and glasses. Beyond this fission energy application, fusion energy will demand advanced materials capable of withstanding the extreme environments of high-temperature plasma systems. Fusion reactors will likely depend on lithium-based ceramics to produce tritium that fuels the fusion plasma, while high-temperature alloys or ceramics will contain and control the hot plasma. All the while, alloys, ceramics, and ceramic-related processes continue to find applications in the management of wastes and byproducts produced by these processes.

John Marra

2011-01-01T23:59:59.000Z

195

DOE - Office of Legacy Management -- Westinghouse Advanced Reactors Div  

Office of Legacy Management (LM)

Advanced Reactors Div Advanced Reactors Div Plutonium and Advanced Fuel Labs - PA 10 FUSRAP Considered Sites Site: WESTINGHOUSE ADVANCED REACTORS DIV., PLUTONIUM FUEL LABORATORIES, AND THE ADVANCED FUEL LAB (PA.10 ) Eliminated from further consideration under FUSRAP Designated Name: Not Designated Alternate Name: None Location: Cheswick , Pennsylvania PA.10-1 Evaluation Year: Circa 1987 PA.10-1 PA.10-4 Site Operations: 1960s and 1970s - Produced light water and fast breeder reactor fuels on a development and pilot plant scale. Closed in 1979. PA.10-2 PA.10-3 Site Disposition: Eliminated - Decommissioned and decontaminated under another Federal program. Release condition confirmed by radiological surveys. PA.10-1 PA.10-2 PA.10-3 PA.10-4 PA.10-5 Radioactive Materials Handled: Yes

196

Analysis of nuclear reactor instability phenomena. Progress report  

SciTech Connect

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

Lahey, R.T. Jr.

1993-03-01T23:59:59.000Z

197

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

198

Nuclear reactor control room construction  

DOE Patents (OSTI)

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

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

1993-01-01T23:59:59.000Z

199

Distributed computing and nuclear reactor analysis  

SciTech Connect

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

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

1994-03-01T23:59:59.000Z

200

Advanced Small Modular Reactor (SMR) Probabilistic Risk Assessment (PRA) Demonstration  

SciTech Connect

A key area of the Advanced Small Modular Reactor (SMR) Probabilistic Risk Assessment (PRA) strategy is the development of methodologies and tools that will be used to predict the safety, security, safeguards, performance, and deployment viability of SMRs. The goal of the SMR PRA activity will be to develop quantitative methods and tools and the associated analysis framework for assessing a variety of risks. Development and implementation of SMR-focused safety assessment methods may require new analytic methods or adaptation of traditional methods to the advanced design and operational features of SMRs. We will need to move beyond the current limitations such as static, logic-based models in order to provide more integrated, scenario-based models based upon predictive modeling which are tied to causal factors. The development of SMR-specific safety models for margin determination will provide a safety case that describes potential accidents, design options (including postulated controls), and supports licensing activities by providing a technical basis for the safety envelope. This report documents the progress that was made to implement the PRA framework, specifically by way of demonstration of an advanced 3D approach to representing, quantifying and understanding flooding risks to a nuclear power plant.

Curtis Smith; Steven Prescott; Tony Koonce

2014-04-01T23:59:59.000Z

Note: This page contains sample records for the topic "advanced nuclear reactor" 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

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

202

Draft Advanced Nuclear Energy Solicitation Fact Sheet | Department...  

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

Fact Sheet Draft Advanced Nuclear Energy Solicitation Fact Sheet Draft Advanced Nuclear Energy Projects Solicitation Fact Sheet (September 2014) More Documents & Publications Draft...

203

Energy Department Invests $67 Million to Advanced Nuclear Technology...  

Office of Environmental Management (EM)

Energy Department Invests 67 Million to Advanced Nuclear Technology Energy Department Invests 67 Million to Advanced Nuclear Technology August 20, 2014 - 12:00pm Addthis News...

204

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

SciTech Connect

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

205

A comparison of nuclear reactor control room display panels  

E-Print Network (OSTI)

complex and time consuming task. It is expected that the control room of future commercial nuclear reactor power plants will change considerably as a result of these studies. Currently there are literally hundreds of displays and controls...: Dr. Rodger S. Koppa A study was conducted to investigate the use of computer generated displays to operate nuclear reactor power plants. The AGN-201 reactor at Texas A&M university was the reactor studied. After observing several licensed reactor...

Bowers, Frances Renae

2012-06-07T23:59:59.000Z

206

Development of a system model for advanced small modular reactors.  

SciTech Connect

This report describes a system model that can be used to analyze three advance small modular reactor (SMR) designs through their lifetime. Neutronics of these reactor designs were evaluated using Monte Carlo N-Particle eXtended (MCNPX/6). The system models were developed in Matlab and Simulink. A major thrust of this research was the initial scoping analysis of Sandia's concept of a long-life fast reactor (LLFR). The inherent characteristic of this conceptual design is to minimize the change in reactivity over the lifetime of the reactor. This allows the reactor to operate substantially longer at full power than traditional light water reactors (LWRs) or other SMR designs (e.g. high temperature gas reactor (HTGR)). The system model has subroutines for lifetime reactor feedback and operation calculations, thermal hydraulic effects, load demand changes and a simplified SCO2 Brayton cycle for power conversion.

Lewis, Tom Goslee,; Holschuh, Thomas Vernon,

2014-01-01T23:59:59.000Z

207

A probabilistic safety analysis of incidents in nuclear research reactors  

Science Journals Connector (OSTI)

......ICRP-64. INTRODUCTION Nuclear research reactors are considered important tools in nuclear science. For more than...as well as prevention policy, have stimulated the development...level 3 in the International Nuclear Events Scale (INES) of......

Valdir Maciel Lopes; Gian Maria Agostinho Angelo Sordi; Mauricio Moralles; Tufic Madi Filho

2012-06-01T23:59:59.000Z

208

Secondary heat exchanger design and comparison for advanced high temperature reactor  

SciTech Connect

Next generation nuclear reactors such as the advanced high temperature reactor (AHTR) are designed to increase energy efficiency in the production of electricity and provide high temperature heat for industrial processes. The efficient transfer of energy for industrial applications depends on the ability to incorporate effective heat exchangers between the nuclear heat transport system and the industrial process heat transport system. This study considers two different types of heat exchangers - helical coiled heat exchanger and printed circuit heat exchanger - as possible options for the AHTR secondary heat exchangers with distributed load analysis and comparison. Comparison is provided for all different cases along with challenges and recommendations. (authors)

Sabharwall, P. [Idaho National Laboratory, Idaho Falls, ID 83415-3860 (United States); Kim, E. S. [Seoul National Univ., P.O. Box 1625, Idaho Falls, ID 83415-3860 (United States); Siahpush, A.; McKellar, M.; Patterson, M. [Idaho National Laboratory, Idaho Falls, ID 83415-3860 (United States)

2012-07-01T23:59:59.000Z

209

Oklo reactors and implications for nuclear science  

E-Print Network (OSTI)

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

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

2014-04-19T23:59:59.000Z

210

Advanced LWR Nuclear Fuel Cladding System Development Trade-Off Study  

SciTech Connect

The Advanced Light Water Reactor (LWR) Nuclear Fuel Development Research and Development (R&D) Pathway encompasses strategic research focused on improving reactor core economics and safety margins through the development of an advanced fuel cladding system. To achieve significant operating improvements while remaining within safety boundaries, significant steps beyond incremental improvements in the current generation of nuclear fuel are required. Fundamental improvements are required in the areas of nuclear fuel composition, cladding integrity, and the fuel/cladding interaction to allow power uprates and increased fuel burn-up allowance while potentially improving safety margin through the adoption of an “accident tolerant” fuel system that would offer improved coping time under accident scenarios. With a development time of about 20 – 25 years, advanced fuel designs must be started today and proven in current reactors if future reactor designs are to be able to use them with confidence.

Kristine Barrett; Shannon Bragg-Sitton

2012-09-01T23:59:59.000Z

211

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

212

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

213

LIMITED POWER BURSTS IN DISTRIBUTED MODELS OF NUCLEAR REACTORS  

E-Print Network (OSTI)

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

Bazhenov, Maxim

214

16N ?-Ray Diagnostics of a Nuclear Reactor in a Nuclear Power Plant  

Science Journals Connector (OSTI)

The AIRM system, which uses the 16N activity in the VVÉR-1000 reactor at the Kalinin nuclear power plant to measure the thermal power of the reactor and the coolant flow rate, and similar systems used in nuclear

S. G. Tsypin; V. V. Lysenko; A. I. Musorin; L. N. Bogachek; V. F. Bai…

2003-09-01T23:59:59.000Z

215

CONSTRUCTION OF WEB-ACCESSIBLE MATERIALS HANDBOOK FORGENERATION IV NUCLEAR REACTORS  

SciTech Connect

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

216

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

1993-01-01T23:59:59.000Z

217

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

218

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

219

Light Water Reactors A DOE Energy Innovation Hub for Modeling...  

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

Consortium for Advanced Simulation of Light Water Reactors A DOE Energy Innovation Hub for Modeling and Simulation of Nuclear Reactors CASL is focused on three issues for nuclear...

220

ADVANCED REACTOR SAFETY PROGRAM – STAKEHOLDER INTERACTION AND FEEDBACK  

SciTech Connect

In the Spring of 2013, we began discussions with our industry stakeholders on how to upgrade our safety analysis capabilities. The focus of these improvements would primarily be on advanced safety analysis capabilities that could help the nuclear industry analyze, understand, and better predict complex safety problems. The current environment in the DOE complex is such that recent successes in high performance computer modeling could lead the nuclear industry to benefit from these advances, as long as an effort to translate these advances into realistic applications is made. Upgrading the nuclear industry modeling analysis capabilities is a significant effort that would require substantial participation and coordination from all industry segments: research, engineering, vendors, and operations. We focus here on interactions with industry stakeholders to develop sound advanced safety analysis applications propositions that could have a positive impact on industry long term operation, hence advancing the state of nuclear safety.

Benjamin W. Spencer; Hai Huang

2014-08-01T23:59:59.000Z

Note: This page contains sample records for the topic "advanced nuclear reactor" 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

Advanced nuclear reactor safety analysis: the simulation of a small break loss of coolant accident in the simplified boiling water reactor using RELAP5/MOD3.1.1  

E-Print Network (OSTI)

The thermal hydraulic simulation code RELAP5/MOD3.1.1 was utilized to model General Electric's Simplified Boiling Water Reactor plant. The model of the plant was subjected to a small break loss of coolant accident occurring from a guillotine shear...

Faust, Christophor Randall

1995-01-01T23:59:59.000Z

222

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.

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

1994-01-01T23:59:59.000Z

223

Polynomial regression with derivative information in nuclear reactor uncertainty quantification*  

E-Print Network (OSTI)

, Argonne National Laboratory, Argonne, IL, USA b Nuclear Engineering Division, Argonne National Laboratory1 Polynomial regression with derivative information in nuclear reactor uncertainty quantification, Argonne, IL, USA Abstract. We introduce a novel technique of uncertainty quantification using polynomial

Anitescu, Mihai

224

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

225

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

226

Direct conversion nuclear reactor space power systems  

SciTech Connect

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

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

1982-08-01T23:59:59.000Z

227

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] [ORNL

2011-01-01T23:59:59.000Z

228

Reference site selection report for the advanced liquid metal reactor at the Idaho National Engineering Laboratory  

SciTech Connect

This Reference Site Selection Report was prepared by EG G, Idaho Inc., for General Electric (GE) to provide information for use by the Department of Energy (DOE) in selecting a Safety Test Site for an Advanced Liquid Metal Reactor. Similar Evaluation studies are planned to be conducted at other potential DOE sites. The Power Reactor Innovative Small Module (PRISM) Concept was developed for ALMR by GE. A ALMR Safety Test is planned to be performed on a DOE site to demonstrate features and meet Nuclear Regulatory Commission Requirements. This study considered possible locations at the Idaho National Engineering Laboratory that met the ALMR Prototype Site Selection Methodology and Criteria. Four sites were identified, after further evaluation one site was eliminated. Each of the remaining three sites satisfied the criteria and was graded. The results were relatively close. Thus concluding that the Idaho National Engineering Laboratory is a suitable location for an Advanced Liquid Metal Reactor Safety Test. 23 refs., 13 figs., 9 tabs.

Sivill, R.L.

1990-03-01T23:59:59.000Z

229

Radioactive target needs for nuclear reactor physics and nuclear astrophysics , G. Barreau1  

E-Print Network (OSTI)

Radioactive target needs for nuclear reactor physics and nuclear astrophysics B.Jurado1* , G Gradignan, France 2 IPN, CNRS/IN2P3, Univ. Paris-Sud, 91405 Orsay, France Abstract: Nuclear reaction cross sections of short-lived nuclei are key inputs for new generation nuclear reactor simulations and for models

Boyer, Edmond

230

Office of Nuclear Energy | Department of Energy  

Office of Environmental Management (EM)

Office of Nuclear Energy Small Modular Reactors Small Modular Reactors The Small Modular Reactor program advances the licensing and commercialization of this next-generation...

231

Nuclear Safety Information | Department of Energy  

Office of Environmental Management (EM)

Safety Information Nuclear Safety Information Idaho National Laboratory's Advanced Test Reactor (ATR) | April 8, 2009 Idaho National Laboratory's Advanced Test Reactor (ATR) |...

232

Inherently Safe Reactors and a Second Nuclear Era  

Science Journals Connector (OSTI)

...Institute of Nuclear Power Operations. Improvements of this...reactors appear to be as safe as large dams, most...that, yes, inherently safe re-actors are surprisingly...more coal under utility boilers. But the environmental...parks. Would inherently safe reactors render existing...

Alvin M. Weinberg; Irving Spiewak

1984-06-29T23:59:59.000Z

233

International Effort to Design Nuclear Fusion Reactor Launched  

Science Journals Connector (OSTI)

International Effort to Design Nuclear Fusion Reactor Launched ... Their mission is to draw up a design concept for a thermonuclear fusion reactor by December 1990. ... The work at Garching is a direct outgrowth of the recently signed International Thermonuclear Experimental Reactor (ITER) pact involving the European Community, Japan, the Soviet Union, and the U.S. (C&EN, April 25, page 19). ...

DERMOT A. O'SULLLVAN

1988-05-23T23:59:59.000Z

234

E-Print Network 3.0 - advanced fast reactor Sample Search Results  

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

ANNULAR FAST REACTOR (3000 MWth) Fuel... and NRE Design Class., "Advances in the Subcritical, Gas-Cooled Fast Transmutation Reactor Concept", Nucl... Cooled, Fast, ... Source:...

235

E-Print Network 3.0 - advanced reactors coupled Sample Search...  

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

ANNULAR FAST REACTOR (3000 MWth) Fuel... and NRE Design Class., "Advances in the Subcritical, Gas-Cooled Fast Transmutation Reactor Concept", Nucl... Cooled, Fast, Subcritical...

236

E-Print Network 3.0 - advanced reactor analyses Sample Search...  

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

ANNULAR FAST REACTOR (3000 MWth) Fuel... and NRE Design Class., "Advances in the Subcritical, Gas-Cooled Fast Transmutation Reactor Concept", Nucl... Cooled, Fast, Subcritical...

237

High-resolution coupled physics solvers for analysing fine-scale nuclear reactor design problems  

Science Journals Connector (OSTI)

...solvers for analysing fine-scale nuclear reactor design problems Vijay S. Mahadevan...analysis of current and future nuclear reactor models is being investigated...in radiation hydrodynamics, nuclear reactor analysis, fluid-structure...

2014-01-01T23:59:59.000Z

238

Strategic Need for Multi-Purpose Thermal Hydraulic Loop for Support of Advanced Reactor Technologies  

SciTech Connect

This report presents a conceptual design for a new high-temperature multi fluid, multi loop test facility for the INL to support thermal hydraulic, materials, and thermal energy storage research for nuclear and nuclear-hybrid applications. In its initial configuration, the facility will include a high-temperature helium loop, a liquid salt loop, and a hot water/steam loop. The three loops will be thermally coupled through an intermediate heat exchanger (IHX) and a secondary heat exchanger (SHX). Research topics to be addressed with this facility include the characterization and performance evaluation of candidate compact heat exchangers such as printed circuit heat exchangers (PCHEs) at prototypical operating conditions, flow and heat transfer issues related to core thermal hydraulics in advanced helium-cooled and salt-cooled reactors, and evaluation of corrosion behavior of new cladding materials and accident-tolerant fuels for LWRs at prototypical conditions. Based on its relevance to advanced reactor systems, the new facility has been named the Advanced Reactor Technology Integral System Test (ARTIST) facility. Research performed in this facility will advance the state of the art and technology readiness level of high temperature intermediate heat exchangers (IHXs) for nuclear applications while establishing the INL as a center of excellence for the development and certification of this technology. The thermal energy storage capability will support research and demonstration activities related to process heat delivery for a variety of hybrid energy systems and grid stabilization strategies. Experimental results obtained from this research will assist in development of reliable predictive models for thermal hydraulic design and safety codes over the range of expected advanced reactor operating conditions. Proposed/existing IHX heat transfer and friction correlations and criteria will be assessed with information on materials compatibility and instrumentation needs. The experimental database will guide development of appropriate predictive methods and be available for code verification and validation (V&V) related to these systems.

James E. O'Brien; Piyush Sabharwall; Su-Jong Yoon; Gregory K. Housley

2014-09-01T23:59:59.000Z

239

Coupled IVPs to Investigate a Nuclear Reactor Poison Burn Up  

SciTech Connect

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

Faghihi, F. [Department of Nuclear Engineering, School of Engineering, Shiraz University 71348-51154, Shiraz (Iran, Islamic Republic of)

2009-09-09T23:59:59.000Z

240

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

Note: This page contains sample records for the topic "advanced nuclear reactor" 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

REACTOR DOSIMETRY STUDY OF THE RHODE ISLAND NUCLEAR SCIENCE CENTER.  

SciTech Connect

The Rhode Island Nuclear Science Center (RINSC), located on the Narragansett Bay Campus of the University of Rhode Island, is a state-owned and US NRC-licensed nuclear facility constructed for educational and industrial applications. The main building of RINSC houses a two-megawatt (2 MW) thermal power critical reactor immersed in demineralized water within a shielded tank. As its original design in 1958 by the Rhode Island Atomic Energy Commission focused on the teaching and research use of the facility, only a minimum of 3.85 kg fissile uranium-235 was maintained in the fuel elements to allow the reactor to reach a critical state. In 1986 when RINSC was temporarily shutdown to start US DOE-directed core conversion project for national security reasons, all the U-Al based Highly-Enriched Uranium (HEU, 93% uranium-235 in the total uranium) fuel elements were replaced by the newly developed U{sub 3}Si{sub 2}-Al based Low Enriched Uranium (LEU, {le}20% uranium-235 in the total uranium) elements. The reactor first went critical after the core conversion was achieved in 1993, and feasibility study on the core upgrade to accommodate Boron Neutron-Captured Therapy (BNCT) was completed in 2000 [3]. The 2-MW critical reactor at RINSC which includes six beam tubes, a thermal column, a gamma-ray experimental station and two pneumatic tubes has been extensive utilized as neutron-and-photon dual source for nuclear-specific research in areas of material science, fundamental physics, biochemistry, and radiation therapy. After the core conversion along with several major system upgrade (e.g. a new 3-MW cooling tower, a large secondary piping system, a set of digitized power-level instrument), the reactor has become more compact and thus more effective to generate high beam flux in both the in-core and ex-core regions for advance research. If not limited by the manpower and operating budget in recent years, the RINSC built ''in concrete'' structure and control systems should have been systematically upgraded to a 5 Mw power facility to further enhance its experimental capability while still maintaining its safe margin as designed.

HOLDEN, N.E.,; RECINIELLO, R.N.; HU, J.-P.

2005-05-08T23:59:59.000Z

242

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

243

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

244

Nuclear reactors built, being built, or planned 1992  

SciTech Connect

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

245

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

246

Insights from the WGRISK workshop on the PSA of advanced and new reactors  

SciTech Connect

Probabilistic Safety Assessment /Probabilistic Risk Assessment for new and advanced reactors is recognized as an essential complement of the deterministic approaches to achieve improved safety and performances of new nuclear power plants, comparing to the operating plants. However, the development of PSA to these reactors is encountered to concurrent challenges, mainly due to the limited available design information, as well as due to potentially new initiating events, accident sequences and phenomena. The use of PSA in the decision making process is also challenging since the resulting PSA may not sufficiently reflect the future as-built, as-operated plant information. In order to address these aspects, the OECD/NEA/WGRISK initiated two coordinated tasks on 'PSA for Advanced Reactors' and 'PSA in the frame of Design and Commissioning of New NPPs'. In this context, a joint workshop was organized by OECD, during which related subjects were presented and discussed, including PSA for generation IV reactors, PSA for evolutionary reactors, PSA for small modular reactors, severe accidents and Level 2 PSA, Level 3 PSA and consequences analysis, digital I and C modeling, passive systems reliability, safety-security interface, as well as the results of the surveys performed in the frame of theses WGRISK tasks. (authors)

Georgescu, G. [Inst. for Radioprotection and Nuclear Safety IRSN, BP17, 92262 Fontenay aux Roses (France); Ahn, K. I. [Korea Atomic Energy Research Inst. KAERI, 150-1 Deokjin-Dong, 1045 Daedeokdaero, Yuseong, Daejeon (Korea, Republic of); Amri, A. [OCED/NEA, Le Seine St.Germain, Bd des Iles, 92130 Issy-les-Moulineaux (France)

2012-07-01T23:59:59.000Z

247

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

248

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

249

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

250

MOOSE simulating nuclear reactor CRUD buildup  

SciTech Connect

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

None

2014-02-06T23:59:59.000Z

251

Directions for advanced use of nuclear power in century XXI  

SciTech Connect

Nuclear power can provide a significant contribution to electricity generation and meet other needs of the world and the US during the next century provided that certain directions are taken to achieve its public acceptance. These directions include formulation of projections of population, energy consumption, and energy resources over a responsible period of time. These projections will allow assessment of cumulative effects on the environment and on fixed resources. Use of fossil energy resources in a century of growing demand for energy must be considered in the context of long-term environmental damage and resource depletion. Although some question the validity of these consequences, they can be mitigated by use of advanced fast reactor technology. It must be demonstrated that nuclear power technology is safe, resistant to material diversion for weapon use, and economical. An unbiased examination of all the issues related to energy use, especially of electricity, is an essential direction to take.

Walter, C E

1999-05-01T23:59:59.000Z

252

E-Print Network 3.0 - advanced test reactor critical facility...  

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

Powered by Explorit Topic List Advanced Search Sample search results for: advanced test reactor critical facility Page: << < 1 2 3 4 5 > >> 1 Engineers at Western are...

253

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

254

Computational Design of Advanced Nuclear Fuels  

SciTech Connect

The objective of the project was to develop a method for theoretical understanding of nuclear fuel materials whose physical and thermophysical properties can be predicted from first principles using a novel dynamical mean field method for electronic structure calculations. We concentrated our study on uranium, plutonium, their oxides, nitrides, carbides, as well as some rare earth materials whose 4f eletrons provide a simplified framework for understanding complex behavior of the f electrons. We addressed the issues connected to the electronic structure, lattice instabilities, phonon and magnon dynamics as well as thermal conductivity. This allowed us to evaluate characteristics of advanced nuclear fuel systems using computer based simulations and avoid costly experiments.

Savrasov, Sergey; Kotliar, Gabriel; Haule, Kristjan

2014-06-03T23:59:59.000Z

255

Critical Facility for lattice physics experiments for the Advanced Heavy Water Reactor and the 500 MWe pressurized heavy water reactors  

Science Journals Connector (OSTI)

Bhabha Atomic Research Centre (BARC), Mumbai, is embarking on a broad based program for thorium utilization in power production to achieve all-round capability in the entire thorium cycle. As a step in this direction, a low power Critical Facility is under construction at BARC. The facility will greatly contribute to the understanding and validation of the calculational models and nuclear data used in the design of thorium based Advanced Heavy Water Reactor. The facility is also designed to cater to the experimental requirements of future lattice studies related to 500 MWe pressurized heavy water reactors. This paper covers the basic design features, safety aspects and the planned experimental program of the new facility.

V.K. Raina; R. Srivenkatesan; D.C. Khatri; D.K. Lahiri

2006-01-01T23:59:59.000Z

256

E-Print Network 3.0 - advanced hybrid nuclear Sample Search Results  

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

nuclear Search Powered by Explorit Topic List Advanced Search Sample search results for: advanced hybrid nuclear Page: << < 1 2 3 4 5 > >> 1 Nuclear Chemical EngineeringNuclear...

257

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

Science Journals Connector (OSTI)

......Radio-toxicity of spent fuel of the advanced heavy water reactor S. Anand * K. D. S...Mumbai 400085, India The Advanced Heavy Water Reactor (AHWR) is a new power...PHWR. INTRODUCTION The Advanced Heavy Water Reactor (AHWR)(1, 2), currently......

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

2010-01-01T23:59:59.000Z

258

Mirror Advanced Reactor Study (MARS) final report summary  

SciTech Connect

The Mirror Advanced Reactor Study (MARS) has resulted in an overview of a first-generation tandem mirror reactor. The central cell fusion plasma is self-sustained by alpha heating (ignition), while electron-cyclotron resonance heating and negative ion beams maintain the electrostatic confining potentials in the end plugs. Plug injection power is reduced by the use of high-field choke coils and thermal barriers, concepts to be tested in the Tandem Mirror Experiment-Upgrade (TMX-U) and Mirror Fusion Test Facility (MFTF-B) at Lawrence Livermore National Laboratory.

Henning, C.D.; Logan, B.G.; Carlson, G.A.; Perkins, L.J.; Werner, R.W.; Gordon, J.D.; Parmer, J.F.; Bilton, J.R.; Glancy, J.E.; Kulcinski, G.L.

1983-12-08T23:59:59.000Z

259

Integrated intelligent systems in advanced reactor control rooms  

SciTech Connect

An intelligent, reactor control room, information system is designed to be an integral part of an advanced control room and will assist the reactor operator's decision making process by continuously monitoring the current plant state and providing recommended operator actions to improve that state. This intelligent system is an integral part of, as well as an extension to, the plant protection and control systems. This paper describes the interaction of several functional components (intelligent information data display, technical specifications monitoring, and dynamic procedures) of the overall system and the artificial intelligence laboratory environment assembled for testing the prototype. 10 refs., 5 figs.

Beckmeyer, R.R.

1989-01-01T23:59:59.000Z

260

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

Note: This page contains sample records for the topic "advanced nuclear reactor" 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

Advanced Fuels Campaign Light Water Reactor Accident Tolerant Fuel Performance Metrics  

SciTech Connect

The safe, reliable and economic operation of the nation’s nuclear power reactor fleet has always been a top priority for the United States’ nuclear industry. As a result, continual improvement of technology, including advanced materials and nuclear fuels, remains central to industry’s success. Decades of research combined with continual operation have produced steady advancements in technology and yielded an extensive base of data, experience, and knowledge on light water reactor (LWR) fuel performance under both normal and accident conditions. In 2011, following the Great East Japan Earthquake, resulting tsunami, and subsequent damage to the Fukushima Daiichi nuclear power plant complex, enhancing the accident tolerance of LWRs became a topic of serious discussion. As a result of direction from the U.S. Congress, the U.S. Department of Energy Office of Nuclear Energy (DOE-NE) initiated an Accident Tolerant Fuel (ATF) Development program. The complex multiphysics behavior of LWR nuclear fuel makes defining specific material or design improvements difficult; as such, establishing qualitative attributes is critical to guide the design and development of fuels and cladding with enhanced accident tolerance. This report summarizes a common set of technical evaluation metrics to aid in the optimization and down selection of candidate designs. As used herein, “metrics” describe a set of technical bases by which multiple concepts can be fairly evaluated against a common baseline and against one another. Furthermore, this report describes a proposed technical evaluation methodology that can be applied to assess the ability of each concept to meet performance and safety goals relative to the current UO2 – zirconium alloy system and relative to one another. The resultant ranked evaluation can then inform concept down-selection, such that the most promising accident tolerant fuel design option(s) can continue to be developed for lead test rod or lead test assembly insertion into a commercial reactor within the desired timeframe (by 2022).

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

2014-02-01T23:59:59.000Z

262

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.

263

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.

264

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

265

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

266

Solid0Core Heat-Pipe Nuclear Batterly Type Reactor  

SciTech Connect

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

267

Managerial Aspects of BNCT at a Nuclear Research Reactor  

Science Journals Connector (OSTI)

All BNCT facilities worldwide, performing clinical trials, are presently located at a nuclear research reactor. They are nevertheless, to all ... 43/EURATOM which stipulates that radiotherapy quality assurance programmes

Wolfgang A. G. Sauerwein; Ray Moss

2012-01-01T23:59:59.000Z

268

Machine for monitoring nuclear reactor equipment at its storage station  

SciTech Connect

Machine for monitoring internal equipment of a nuclear reactor, when stored outside the reactor vessel in the swimming pool of the reactor, comprising a first longitudinal carriage movable on rails on the bottom of the swimming pool, a transverse beam movable on the first carriage, a second transverse carriage bearing monitoring instruments and movable on the beam, and an arrangement for moving the beam in the predetermined steps on the first carriage, and the second carriage on the beam.

Castrec, Y. M.; Launay, J. P.

1985-02-12T23:59:59.000Z

269

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

SciTech Connect

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

270

Distillation and Dehydro Reactors Advanced Process Conrol Freeport Texas PLant  

E-Print Network (OSTI)

Distillation and Dehydro Reactors Advanced Process Control Freeport Texas Plant ESL-IE-14-05-16 Proceedings of the Thrity-Sixth Industrial Energy Technology Conference New Orleans, LA. May 20-23, 2014 G-KTI, Polyamide and Intermediates Distillation... APC 6/2/2014 INTERNAL; CONFIDENTIAL 2 APC is a collection of two different control and automation technologies • Multivariable Predictive Control (MPC). In this approach, an empirical, dynamic, plant model is used in combination with both a steady...

Eisele, D.

2014-01-01T23:59:59.000Z

271

A Blueprint for GNEP Advanced Burner Reactor Startup Fuel Fabrication Facility  

SciTech Connect

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

S. Khericha

2010-12-01T23:59:59.000Z

272

NEAMS: The Nuclear Energy Advanced Modeling and Simulation Program  

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

NEAMS: The Nuclear Energy Advanced NEAMS: The Nuclear Energy Advanced Modeling and Simulation Program The Nuclear Energy Advanced Modeling and Simulation (NEAMS) Program is developing a simulation tool kit using leading-edge computational methods that will accelerate the development and deployment of nuclear power technologies that employ enhanced safety and security features, produce power more cost-effectively, and utilize natural resources more efficiently. The NEAMS ToolKit

273

Advanced pressurized water reactor for improved resource utilization, part II - composite advanced PWR concept  

SciTech Connect

This report evaluates the enhanced resource utilization in an advanced pressurized water reactor (PWR) concept using a composite of selected improvements identified in a companion study. The selected improvements were in the areas of reduced loss of neutrons to control poisons, reduced loss of neutrons in leakage from the core, and improved blanket/reflector concepts. These improvements were incorporated into a single composite advanced PWR. A preliminary assessment of resource requirements and costs and impact on safety are presented.

Turner, S.E.; Gurley, M.K.; Kirby, K.D.; Mitchell, W III

1981-09-15T23:59:59.000Z

274

NRC review of Electric Power Research Institute's Advanced Light Reactor Utility Requirements Document - Program summary, Project No. 669  

SciTech Connect

The staff of the US Nuclear Regulatory Commission has prepared Volume 1 of a safety evaluation report (SER), NRC Review of Electric Power Research Institute's Advanced Light Water Reactor Utility Requirements Document -- Program Summary,'' to document the results of its review of the Electric Power Research Institute's Advanced Light Water Reactor Utility Requirements Document.'' This SER provides a discussion of the overall purpose and scope of the Requirements Document, the background of the staff's review, the review approach used by the staff, and a summary of the policy and technical issues raised by the staff during its review.

Not Available

1992-08-01T23:59:59.000Z

275

January 2003, Report of the ADVANCED NUCLEAR TRANSFORMATION TECHNOLOGY  

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

January 2003, Report of the ADVANCED NUCLEAR TRANSFORMATION January 2003, Report of the ADVANCED NUCLEAR TRANSFORMATION TECHNOLOGY SUBCOMMITTEE of the NUCLEAR ENERGY RESEARCH ADVISORY COMMITTEE January 2003, Report of the ADVANCED NUCLEAR TRANSFORMATION TECHNOLOGY SUBCOMMITTEE of the NUCLEAR ENERGY RESEARCH ADVISORY COMMITTEE The ANTT Subcommittee met in Washington on Dec 4-5, 2002 to review progress in the transmutation program, and to learn about major organizational changes that affect the management of the program. The NE's new Advanced Nuclear Research Office (NE-20) now oversees both the transmutation program (ANTT) and the Generation-IV program (GEN-IV). antt14Jan_03.pdf More Documents & Publications October 2003, Report of the ADVANCED NUCLEAR TRANSFORMATION TECHNOLOGY SUBCOMMITTEE of the NUCLEAR ENERGY RESEARCH ADVISORY COMMITTEE

276

Advanced Instrumentation and Control Methods for Small and Medium Reactors with IRIS Demonstration  

SciTech Connect

Development and deployment of small-scale nuclear power reactors and their maintenance, monitoring, and control are part of the mission under the Small Modular Reactor (SMR) program. The objectives of this NERI-consortium research project are to investigate, develop, and validate advanced methods for sensing, controlling, monitoring, diagnosis, and prognosis of these reactors, and to demonstrate the methods with application to one of the proposed integral pressurized water reactors (IPWR). For this project, the IPWR design by Westinghouse, the International Reactor Secure and Innovative (IRIS), has been used to demonstrate the techniques developed under this project. The research focuses on three topical areas with the following objectives. Objective 1 - Develop and apply simulation capabilities and sensitivity/uncertainty analysis methods to address sensor deployment analysis and small grid stability issues. Objective 2 - Develop and test an autonomous and fault-tolerant control architecture and apply to the IRIS system and an experimental flow control loop, with extensions to multiple reactor modules, nuclear desalination, and optimal sensor placement strategy. Objective 3 - Develop and test an integrated monitoring, diagnosis, and prognosis system for SMRs using the IRIS as a test platform, and integrate process and equipment monitoring (PEM) and process and equipment prognostics (PEP) toolboxes. The research tasks are focused on meeting the unique needs of reactors that may be deployed to remote locations or to developing countries with limited support infrastructure. These applications will require smaller, robust reactor designs with advanced technologies for sensors, instrumentation, and control. An excellent overview of SMRs is described in an article by Ingersoll (2009). The article refers to these as deliberately small reactors. Most of these have modular characteristics, with multiple units deployed at the same plant site. Additionally, the topics focus on meeting two of the eight needs outlined in the recently published 'Technology Roadmap on Instrumentation, Control, and Human-Machine Interface (ICHMI) to Support DOE Advanced Nuclear Energy Programs' which was created 'to provide a systematic path forward for the integration of new ICHMI technologies in both near-term and future nuclear power plants and the reinvigoration of the U.S. nuclear ICHMI community and capabilities.' The research consortium is led by The University of Tennessee (UT) and is focused on three interrelated topics: Topic 1 (simulator development and measurement sensitivity analysis) is led by Dr. Mike Doster with Dr. Paul Turinsky of North Carolina State University (NCSU). Topic 2 (multivariate autonomous control of modular reactors) is led by Dr. Belle Upadhyaya of the University of Tennessee (UT) and Dr. Robert Edwards of Penn State University (PSU). Topic 3 (monitoring, diagnostics, and prognostics system development) is led by Dr. Wes Hines of UT. Additionally, South Carolina State University (SCSU, Dr. Ken Lewis) participated in this research through summer interns, visiting faculty, and on-campus research projects identified throughout the grant period. Lastly, Westinghouse Science and Technology Center (Dr. Mario Carelli) was a no-cost collaborator and provided design information related to the IRIS demonstration platform and defining needs that may be common to other SMR designs. The results of this research are reported in a six-volume Final Report (including the Executive Summary, Volume 1). Volumes 2 through 6 of the report describe in detail the research and development under the topical areas. This volume serves to introduce the overall NERI-C project and to summarize the key results. Section 2 provides a summary of the significant contributions of this project. A list of all the publications under this project is also given in Section 2. Section 3 provides a brief summary of each of the five volumes (2-6) of the report. The contributions of SCSU are described in Section 4, including a summary of undergraduate research exper

J. Wesley Hines; Belle R. Upadhyaya; J. Michael Doster; Robert M. Edwards; Kenneth D. Lewis; Paul Turinsky; Jamie Coble

2011-05-31T23:59:59.000Z

277

US approves first new nuclear reactors in 30 years  

Science Journals Connector (OSTI)

... giant based in Atlanta, Georgia, can now move forward with a pair of Westinghouse AP1000 nuclear reactors at its Vogtle nuclear station near Waynesboro. But whereas many within government ... to hasten a historically slow, expensive and, some say, unpredictable regulatory process).  The AP1000 (fuzzy artistic rendering, top right) is one of a new generation of pressurized- ...

Jeff Tollefson

2012-02-10T23:59:59.000Z

278

The Advanced Test Reactor National Scientific User Facility  

SciTech Connect

In 2007, the Advanced Test Reactor (ATR), located at Idaho National Laboratory (INL), was designated by the Department of Energy (DOE) as a National Scientific User Facility (NSUF). This designation made test space within the ATR and post-irradiation examination (PIE) equipment at INL available for use by approved researchers via a proposal and peer review process. The goal of the ATR NSUF is to provide those researchers with the best ideas access to the most advanced test capability, regardless of the proposer’s physical location. Since 2007, the ATR NSUF has expanded its available reactor test space, obtained access to additional PIE equipment, taken steps to enable the most advanced post-irradiation analysis possible, and initiated an educational program and digital learning library to help potential users better understand the critical issues in reactor technology and how a test reactor facility could be used to address this critical research. Recognizing that INL may not have all the desired PIE equipment, or that some equipment may become oversubscribed, the ATR NSUF established a Partnership Program. This program invited universities to nominate their capability to become part of a broader user facility. Any university is eligible to self-nominate. Any nomination is then peer reviewed to ensure that the addition of the university facilities adds useful capability to the NSUF. Once added to the NSUF team, the university capability is then integral to the NSUF operations and is available to all users via the proposal process. So far, six universities have been added to the ATR NSUF with capability that includes reactor-testing space, PIE equipment, and ion beam irradiation facilities. With the addition of these university capabilities, irradiation can occur in multiple reactors and post-irradiation exams can be performed at multiple universities. In each case, the choice of facilities is based on the user’s technical needs. The current NSUF partners are shown in Figure 1. This article describes the ATR as well as the expanded capabilities, partnerships, and services that allow researchers to take full advantage of this national resource.

Todd R. Allen; Collin J. Knight; Jeff B. Benson; Frances M. Marshall; Mitchell K. Meyer; Mary Catherine Thelen

2011-08-01T23:59:59.000Z

279

The Path to Sustainable Nuclear Energy. Basic and Applied Research Opportunities for Advanced Fuel Cycles  

SciTech Connect

The objective of this report is to identify new basic science that will be the foundation for advances in nuclear fuel-cycle technology in the near term, and for changing the nature of fuel cycles and of the nuclear energy industry in the long term. The goals are to enhance the development of nuclear energy, to maximize energy production in nuclear reactor parks, and to minimize radioactive wastes, other environmental impacts, and proliferation risks. The limitations of the once-through fuel cycle can be overcome by adopting a closed fuel cycle, in which the irradiated fuel is reprocessed and its components are separated into streams that are recycled into a reactor or disposed of in appropriate waste forms. The recycled fuel is irradiated in a reactor, where certain constituents are partially transmuted into heavier isotopes via neutron capture or into lighter isotopes via fission. Fast reactors are required to complete the transmutation of long-lived isotopes. Closed fuel cycles are encompassed by the Department of Energy?s Advanced Fuel Cycle Initiative (AFCI), to which basic scientific research can contribute. Two nuclear reactor system architectures can meet the AFCI objectives: a ?single-tier? system or a ?dual-tier? system. Both begin with light water reactors and incorporate fast reactors. The ?dual-tier? systems transmute some plutonium and neptunium in light water reactors and all remaining transuranic elements (TRUs) in a closed-cycle fast reactor. Basic science initiatives are needed in two broad areas: ? Near-term impacts that can enhance the development of either ?single-tier? or ?dual-tier? AFCI systems, primarily within the next 20 years, through basic research. Examples: Dissolution of spent fuel, separations of elements for TRU recycling and transmutation Design, synthesis, and testing of inert matrix nuclear fuels and non-oxide fuels Invention and development of accurate on-line monitoring systems for chemical and nuclear species in the nuclear fuel cycle Development of advanced tools for designing reactors with reduced margins and lower costs ? Long-term nuclear reactor development requires basic science breakthroughs: Understanding of materials behavior under extreme environmental conditions Creation of new, efficient, environmentally benign chemical separations methods Modeling and simulation to improve nuclear reaction cross-section data, design new materials and separation system, and propagate uncertainties within the fuel cycle Improvement of proliferation resistance by strengthening safeguards technologies and decreasing the attractiveness of nuclear materials A series of translational tools is proposed to advance the AFCI objectives and to bring the basic science concepts and processes promptly into the technological sphere. These tools have the potential to revolutionize the approach to nuclear engineering R&D by replacing lengthy experimental campaigns with a rigorous approach based on modeling, key fundamental experiments, and advanced simulations.

Finck, P.; Edelstein, N.; Allen, T.; Burns, C.; Chadwick, M.; Corradini, M.; Dixon, D.; Goff, M.; Laidler, J.; McCarthy, K.; Moyer, B.; Nash, K.; Navrotsky, A.; Oblozinsky, P.; Pasamehmetoglu, K.; Peterson, P.; Sackett, J.; Sickafus, K. E.; Tulenko, J.; Weber, W.; Morss, L.; Henry, G.

2005-09-01T23:59:59.000Z

280

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

Note: This page contains sample records for the topic "advanced nuclear reactor" 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
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281

Enhanced in-pile instrumentation at the advanced test reactor  

SciTech Connect

Many of the sensors deployed at materials and test reactors cannot withstand the high flux/high temperature test conditions often requested by users at U.S. test reactors, such as the Advanced Test Reactor (ATR) at the Idaho National Laboratory. To address this issue, an instrumentation development effort was initiated as part of the ATR National Scientific User Facility in 2007 to support the development and deployment of enhanced in-pile sensors. This paper reports results from this effort. Specifically, this paper identifies the types of sensors currently available to support in-pile irradiations and those sensors currently available to ATR users. Accomplishments from new sensor technology deployment efforts are highlighted by describing new temperature and thermal conductivity sensors now available to ATR users. Efforts to deploy enhanced in-pile sensors for detecting elongation and realtime flux detectors are also reported, and recently-initiated research to evaluate the viability of advanced technologies to provide enhanced accuracy for measuring key parameters during irradiation testing are noted. (authors)

Rempe, J. L.; Knudson, D. L.; Daw, J. E.; Unruh, T.; Chase, B. M.; Palmer, J.; Condie, K. G.; Davis, K. L. [Idaho National Laboratory, MS 3840, P.O. Box 1625, Idaho Falls, ID 83415 (United States)

2011-07-01T23:59:59.000Z

282

Concept of nuclear reactor pumped laser for ICF  

SciTech Connect

It is well known that attempts of civil utilization of fusion energy encounter many difficulties. At the same time we know that creation of thermonuclear weapon had been possible by using of the nuclear fission reaction as ignition of the nuclear fusion. The question arises{emdash}can help us similar idea in civil case and how that can be realized? In paper, it is shown that such idea is useful in this case and can be realized using nuclear reactor pumped laser. Contemporary state of research in nuclear reactor pumped laser for ICF field is considered. Progress by IPPE (Obninsk, Russia) in the development of the energy model of pulse reactor pumped laser system with waiting output energy about 50 kJ is reported. {copyright} {ital 1996 American Institute of Physics.}

Dyachenko, P.P. [Institute of Physics and Power Engineering, 1, Bondarenko Sq., Obninsk, 249020, Kaluga Reg. (Russia)

1996-05-01T23:59:59.000Z

283

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

284

Physics-based multiscale coupling for full core nuclear reactor simulation  

Science Journals Connector (OSTI)

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

Derek R. Gaston; Cody J. Permann; John W. Peterson; Andrew E. Slaughter; David Andrš; Yaqi Wang; Michael P. Short; Danielle M. Perez; Michael R. Tonks; Javier Ortensi; Ling Zou; Richard C. Martineau

2014-01-01T23:59:59.000Z

285

21 - Plant life management (PLiM) practices for pressurised heavy water nuclear reactors (PHWR)  

Science Journals Connector (OSTI)

Abstract: The chapter begins with the history of evolution of pressurised heavy water reactor (PHWR) technology in Canada and India and its importance to the three stage Indian Nuclear Power Programme. An insight into the technology and its variants in use in Canada and India has been provided. Regulatory practices followed in India for renewal of operating licences and also for re-licensing of older plants have been highlighted. Several technological advancements, both in the inspection technology and reactor design concepts have been briefly described to give a glimpse of development trends in future.

R.K. Sinha; S.K. Sinha; K.B. Dixit; A.K. Chakrabarty; D.K. Jain

2010-01-01T23:59:59.000Z

286

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

SciTech Connect

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

Miller, J. M. [AECL, Chalk River Laboratories, Chalk River, ON K0J 1J0 (Canada)

2008-07-15T23:59:59.000Z

287

Advanced High-Temperature, High-Pressure Transport Reactor Gasification  

SciTech Connect

The U.S. Department of Energy (DOE) National Energy Technology Laboratory Office of Coal and Environmental Systems has as its mission to develop advanced gasification-based technologies for affordable, efficient, zero-emission power generation. These advanced power systems, which are expected to produce near-zero pollutants, are an integral part of DOE's Vision 21 Program. DOE has also been developing advanced gasification systems that lower the capital and operating costs of producing syngas for chemical production. A transport reactor has shown potential to be a low-cost syngas producer compared to other gasification systems since its high-throughput-per-unit cross-sectional area reduces capital costs. This work directly supports the Power Systems Development Facility utilizing the KBR transport reactor located at the Southern Company Services Wilsonville, Alabama, site. Over 2800 hours of operation on 11 different coals ranging from bituminous to lignite along with a petroleum coke has been completed to date in the pilot-scale transport reactor development unit (TRDU) at the Energy & Environmental Research Center (EERC). The EERC has established an extensive database on the operation of these various fuels in both air-blown and oxygen-blown modes utilizing a pilot-scale transport reactor gasifier. This database has been useful in determining the effectiveness of design changes on an advanced transport reactor gasifier and for determining the performance of various feedstocks in a transport reactor. The effects of different fuel types on both gasifier performance and the operation of the hot-gas filter system have been determined. It has been demonstrated that corrected fuel gas heating values ranging from 90 to 130 Btu/scf have been achieved in air-blown mode, while heating values up to 230 Btu/scf on a dry basis have been achieved in oxygen-blown mode. Carbon conversions up to 95% have also been obtained and are highly dependent on the oxygen-coal ratio. Higher-reactivity (low-rank) coals appear to perform better in a transport reactor than the less reactive bituminous coals. Factors that affect TRDU product gas quality appear to be coal type, temperature, and air/coal ratios. Testing with a higher-ash, high-moisture, low-rank coal from the Red Hills Mine of the Mississippi Lignite Mining Company has recently been completed. Testing with the lignite coal generated a fuel gas with acceptable heating value and a high carbon conversion, although some drying of the high-moisture lignite was required before coal-feeding problems were resolved. No ash deposition or bed material agglomeration issues were encountered with this fuel. In order to better understand the coal devolatilization and cracking chemistry occurring in the riser of the transport reactor, gas and solid sampling directly from the riser and the filter outlet has been accomplished. This was done using a baseline Powder River Basin subbituminous coal from the Peabody Energy North Antelope Rochelle Mine near Gillette, Wyoming.

Michael Swanson; Daniel Laudal

2008-03-31T23:59:59.000Z

288

R. Shane Johnson, Associate Director Office of Advanced Nuclear Research  

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

Advanced Nuclear Research Advanced Nuclear Research September 30, 2002 Generation IV International Forum Generation IV International Forum Presentation to the Nuclear Energy Research Advisory Committee Office of Nuclear Energy, Science and Technology Generation IV International Forum Generation IV International Forum 6 Government-sanctioned organization working together to plan the future of nuclear energy * Chartered in July 2002 * Conduct joint R&D on next-generation nuclear energy systems * Voluntary member participation in specific projects 6 Observer Organizations * OECD-NEA * IAEA * Euratom South Korea U.S.A. Argentina Brazil Canada France Japan South Africa United Kingdom Switzerland Office of Nuclear Energy, Science and Technology

289

Advanced nuclear plant control room complex  

DOE Patents (OSTI)

An advanced control room complex for a nuclear power plant, including a discrete indicator and alarm system (72) which is nuclear qualified for rapid response to changes in plant parameters and a component control system (64) which together provide a discrete monitoring and control capability at a panel (14-22, 26, 28) in the control room (10). A separate data processing system (70), which need not be nuclear qualified, provides integrated and overview information to the control room and to each panel, through CRTs (84) and a large, overhead integrated process status overview board (24). The discrete indicator and alarm system (72) and the data processing system (70) receive inputs from common plant sensors and validate the sensor outputs to arrive at a representative value of the parameter for use by the operator during both normal and accident conditions, thereby avoiding the need for him to assimilate data from each sensor individually. The integrated process status board (24) is at the apex of an information hierarchy that extends through four levels and provides access at each panel to the full display hierarchy. The control room panels are preferably of a modular construction, permitting the definition of inputs and outputs, the man machine interface, and the plant specific algorithms, to proceed in parallel with the fabrication of the panels, the installation of the equipment and the generic testing thereof.

Scarola, Kenneth (Windsor, CT); Jamison, David S. (Windsor, CT); Manazir, Richard M. (North Canton, CT); Rescorl, Robert L. (Vernon, CT); Harmon, Daryl L. (Enfield, CT)

1993-01-01T23:59:59.000Z

290

KRNFYSIK, FRDJUPNINGSKURS FKF 021 Nuclear Physics, Advanced Course I  

E-Print Network (OSTI)

K�RNFYSIK, F�RDJUPNINGSKURS FKF 021 Nuclear Physics, Advanced Course I Antal poäng: 5.0. Valfri för. Partikelfysik. Laborationerna är obligatoriska. Litteratur Krane, K.S.: Introductory Nuclear Physics

291

KRNFYSIK, FRDJUPNINGSKURS FKF021 Nuclear Physics, Advanced Course I  

E-Print Network (OSTI)

K�RNFYSIK, F�RDJUPNINGSKURS FKF021 Nuclear Physics, Advanced Course I Poäng: 5.0 Betygskala: TH. Partikelfysik. Laborationerna är obligatoriska. Litteratur: Krane, K.S.: Introductory Nuclear Physics

292

Nuclear Energy Advanced Modeling and Simulation (NEAMS) Software...  

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

Software Verification and Validation (V&V) Plan Requirements Nuclear Energy Advanced Modeling and Simulation (NEAMS) Software Verification and Validation (V&V) Plan Requirements...

293

Nuclear Energy Advanced Modeling and Simulation (NEAMS) Program...  

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

Program Plan Nuclear Energy Advanced Modeling and Simulation (NEAMS) Program Plan The NEAMS program plan includes information on the program vision, objective, scope, schedule and...

294

University Program in Advanced Technology | National Nuclear Security  

National Nuclear Security Administration (NNSA)

University Program in Advanced Technology | National Nuclear Security University Program in Advanced Technology | 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 University Program in Advanced Technology Home > About Us > Our Programs > Defense Programs > Future Science & Technology Programs > Office of Advanced Simulation and Computing and

295

Metal fires and their implications for advanced reactors.  

SciTech Connect

This report details the primary results of the Laboratory Directed Research and Development project (LDRD 08-0857) Metal Fires and Their Implications for Advance Reactors. Advanced reactors may employ liquid metal coolants, typically sodium, because of their many desirable qualities. This project addressed some of the significant challenges associated with the use of liquid metal coolants, primary among these being the extremely rapid oxidation (combustion) that occurs at the high operating temperatures in reactors. The project has identified a number of areas for which gaps existed in knowledge pertinent to reactor safety analyses. Experimental and analysis capabilities were developed in these areas to varying degrees. In conjunction with team participation in a DOE gap analysis panel, focus was on the oxidation of spilled sodium on thermally massive surfaces. These are spills onto surfaces that substantially cool the sodium during the oxidation process, and they are relevant because standard risk mitigation procedures seek to move spill environments into this regime through rapid draining of spilled sodium. While the spilled sodium is not quenched, the burning mode is different in that there is a transition to a smoldering mode that has not been comprehensively described previously. Prior work has described spilled sodium as a pool fire, but there is a crucial, experimentally-observed transition to a smoldering mode of oxidation. A series of experimental measurements have comprehensively described the thermal evolution of this type of sodium fire for the first time. A new physics-based model has been developed that also predicts the thermal evolution of this type of sodium fire for the first time. The model introduces smoldering oxidation through porous oxide layers to go beyond traditional pool fire analyses that have been carried out previously in order to predict experimentally observed trends. Combined, these developments add significantly to the safety analysis capabilities of the advanced-reactor community for directly relevant scenarios. Beyond the focus on the thermally-interacting and smoldering sodium pool fires, experimental and analysis capabilities for sodium spray fires have also been developed in this project.

Nowlen, Steven Patrick; Figueroa, Victor G.; Olivier, Tara Jean; Hewson, John C.; Blanchat, Thomas K.

2010-10-01T23:59:59.000Z

296

Neutrino mass hierarchy from nuclear reactor experiments  

Science Journals Connector (OSTI)

Ten years from now reactor neutrino experiments will attempt to determine which neutrino mass eigenstate is the most massive. In this paper we present the results of more than seven million detailed simulations of such experiments, studying the dependence of the probability of successfully determining the mass hierarchy upon the analysis method, the neutrino mass matrix parameters, reactor flux models, geoneutrinos and, in particular, combinations of baselines. We show that a recently reported spurious dependence of the data analysis upon the high energy tail of the reactor spectrum can be removed by using a weighted Fourier transform. We determine the optimal baselines and corresponding detector locations. For most values of the CP-violating, leptonic Dirac phase ?, a degeneracy prevents NO?A and T2K from determining either ? or the hierarchy. We determine the confidence with which a reactor experiment can determine the hierarchy, breaking the degeneracy.

Emilio Ciuffoli; Jarah Evslin; Xinmin Zhang

2013-08-29T23:59:59.000Z

297

Fiscal year 1999 multi-year work plan, advanced reactors transition program  

SciTech Connect

The Advanced Reactors Transition (ART) has two missions. One, funded by DOE-EM is to transition assigned, surplus facilities to a safe and compliant, low-cost stable, deactivated condition (requiring minimal surveillance and maintenance) pending eventual reuse or D and D. Facilities to be transitioned include the 309 Building/Plutonium Recycle Test Reactor (PRTR) and Nuclear Energy (NE) Legacy Facilities. The second mission, funded by DOE-NE, is to maintain the Fast Flux Test Facility (FFTF) and affiliated 400 Area buildings in a safe and compliant standby condition. The condition of the plant hardware, software and personnel is to be preserved in a manner not to preclude a plant restart.

Gantt, D.A.

1998-09-17T23:59:59.000Z

298

Major Safety Aspects of Advanced Candu Reactor and Associated Research and Development  

SciTech Connect

The Advanced Candu{sup R} Reactor design is built on the proven technology of existing Candu plants and on AECL's knowledge base acquired over decades of nuclear power plant design, engineering, construction and research. Two prime objectives of ACR-700TM1 are cost reduction and enhanced safety. To achieve them some new features were introduced and others were improved from the previous Candu 6 and Candu 9 designs. The ACR-700 reactor design is based on the modular concept of horizontal fuel channels surrounded by a heavy water moderator, the same as with all Candu reactors. The major novelty in the ACR-700 is the use of slightly enriched fuel and light water as coolant circulating in the fuel channels. This results in a more compact reactor design and a reduction of heavy water inventory, both contributing to a significant decrease in cost compared to Candu reactors, which employ natural uranium as fuel and heavy water as coolant. The reactor core design adopted for ACR-700 also has some features that have a bearing on inherent safety, such as negative power and coolant void reactivity coefficient. Several improvements in engineered safety have been made as well, such as enhanced separation of the safety support systems. Since the ACR-700 design is an evolutionary development of the currently operating Candu plants, limited research is required to extend the validation database for the design and the supporting safety analysis. A program of safety related research and development has been initiated to address the areas where the ACR-700 design is significantly different from the Candu designs. This paper describes the major safety aspects of the ACR-700 with a particular focus on novel features and improvements over the existing Candu reactors. It also outlines the key areas where research and development efforts are undertaken to demonstrate the effectiveness and robustness of the design. (authors)

Bonechi, M.; Wren, D.J.; Hopwood, J.M. [Atomic Energy of Canada Limited, 2251 Speakman Drive, Mississauga, Ontario, L5K 1B2 (Canada)

2002-07-01T23:59:59.000Z

299

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

300

October 2003, Report of the ADVANCED NUCLEAR TRANSFORMATION TECHNOLOGY  

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

3, Report of the ADVANCED NUCLEAR TRANSFORMATION 3, Report of the ADVANCED NUCLEAR TRANSFORMATION TECHNOLOGY SUBCOMMITTEE of the NUCLEAR ENERGY RESEARCH ADVISORY COMMITTEE October 2003, Report of the ADVANCED NUCLEAR TRANSFORMATION TECHNOLOGY SUBCOMMITTEE of the NUCLEAR ENERGY RESEARCH ADVISORY COMMITTEE The committee met in Washington in Sept 16-17 to review progress in the program with respect to a changed set of mission priorities. Our last meeting took place in Dec 2002 after the reorganization that had place the Advanced Fuel Cycle Initiative (AFCI) and GEN IV program together in the Advanced Nuclear Reserach Office (AN-20). Since mission priorities have been evolving, the committee felt that it should wait unti they have settled down before we met again. We have kept in touch during the process,

Note: This page contains sample records for the topic "advanced nuclear reactor" 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

Formation of a nuclear reactor’s molten core bath in a crucible-type corium catcher for a nuclear power station equipped with VVER reactors  

Science Journals Connector (OSTI)

Results from a calculation study on analyzing the formation of a melt bath in a crucible-type catcher for the conditions of a severe accident at a nuclear power station equipped with VVER-1000 reactors are presen...

S. V. Beshta; S. A. Vitol’; V. S. Granovskii; E. K. Kalyago…

2011-05-01T23:59:59.000Z

302

A Compact Nuclear Fusion Reactor for Space Flights  

SciTech Connect

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

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

2006-05-02T23:59:59.000Z

303

Spectral Structure of Electron Antineutrinos from Nuclear Reactors  

E-Print Network (OSTI)

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

D. A. Dwyer; T. J. Langford

2014-07-04T23:59:59.000Z

304

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

305

Fabrication development for the Advanced Neutron Source Reactor  

SciTech Connect

This report presents the fuel fabrication development for the Advanced Neutron Source (ANS) reactor. The fuel element is similar to that successfully fabricated and used in the High Flux Isotope Reactor (HFIR) for many years, but there are two significant differences that require some development. The fuel compound is U{sub 3}Si{sub 2} rather than U{sub 3}O{sub 8}, and the fuel is graded in the axial as well as the radial direction. Both of these changes can be accomplished with a straightforward extension of the HFIR technology. The ANS also requires some improvements in inspection technology and somewhat more stringent acceptance criteria. Early indications were that the fuel fabrication and inspection technology would produce a reactor core meeting the requirements of the ANS for the low volume fraction loadings needed for the highly enriched uranium design (up to 1.7 Mg U/m{sup 3}). Near the end of the development work, higher volume fractions were fabricated that would be required for a lower- enrichment uranium core. Again, results look encouraging for loadings up to {approx}3.5 Mg U/m{sup 3}; however, much less evaluation was done for the higher loadings.

Pace, B.W. [Babcock and Wilcox, Lynchburg, VA (United States); Copeland, G.L. [Oak Ridge National Lab., TN (United States)

1995-08-01T23:59:59.000Z

306

Reliability assessment of Passive Containment Cooling System of an Advanced Reactor using APSRA methodology  

Science Journals Connector (OSTI)

Abstract Passive Systems are increasingly playing a prominent role in the advanced nuclear reactor systems and are being utilised in normal operations as well as safety systems of the reactors following an accident. The Passive Containment Cooling System (PCCS) is one of the several passive safety features in an Advanced Reactor (AHWR). In this paper, the APSRA methodology has been employed for reliability evaluation of the PCCS of AHWR. Performance assessment of the PCCS is initially performed during a postulated design basis LOCA using the best-estimate code RELAP5/Mod 3.2. The parameters affecting the system performance are then identified and considered for further analysis. Based on some pre-determined failure criterion, the failure surface for the system is predicted using the best-estimate code taking into account the deviations of the identified parameters from their nominal states as well as the model uncertainties inherent to the best estimate code. Root diagnosis is then carried out to determine the various failure causes, which occurs mainly due to malfunctioning of mechanical components. The failure probabilities of the various components are assessed through a classical PSA treatment using generic data. The reliability of the PCCS is then evaluated from the probability of availability of these components.

Mukesh Kumar; Aranyak Chakravarty; A.K. Nayak; Hari Prasad; V. Gopika

2014-01-01T23:59:59.000Z

307

About Naval Reactors | National Nuclear Security Administration  

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

fleet to remain on station is further enhanced by our ability to forward-deploy repair and maintenance activities. Schools and Training Facilities The Naval Nuclear...

308

E-Print Network 3.0 - advanced water-cooled reactors Sample Search...  

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

in the Subcritical, Gas-Cooled Fast Transmutation Reactor Concept", Nucl... Cooled, Fast, Subcritical Advanced Burner ... Source: MIT Plasma Science and Fusion Center Collection:...

309

Solar Thermochemical Advanced Reactor System, Wins R&D 100 Award...  

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

National Laboratory, the Solar Thermochemical Advanced Reactor System, or STARS, converts natural gas and sunlight into a more energy-rich fuel called syngas, which power plants...

310

E-Print Network 3.0 - advanced reactor technology Sample Search...  

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

advanced countries like France, Canada, the USA. Expansion... to the greenhouse effect. New Generation reactors to achieve the reuse of spent fuel. Fusion...

311

E-Print Network 3.0 - advanced converter reactors Sample Search...  

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

computer network ring for advance science and education cooperation in Beijing... thermonuclear experimental reactor (ITER) project have an opportunity to offer technical...

312

Maximum Fuel Utilization in Advanced Fast Reactors without Actinides Separation  

E-Print Network (OSTI)

Physics Optimization of Breed and Burn Fast Reactor Systems.reactors: Fabrication and properties and their optimization.

Heidet, Florent

2010-01-01T23:59:59.000Z

313

Status Report on Efforts to Enhance Instrumentation to Support Advanced Test Reactor Irradiations  

SciTech Connect

The Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF) in April 2007 to support the growth of nuclear science and technology in the United States (US). By attracting new research users - universities, laboratories, and industry - the ATR NSUF facilitates basic and applied nuclear research and development, further advancing the nation's energy security needs. A key component of the ATR NSUF effort at the Idaho National Laboratory (INL) is to design, develop, and deploy new in-pile instrumentation techniques that are capable of providing real-time measurements of key parameters during irradiation. To address this need, an assessment of instrumentation available and under-development at other test reactors was completed. Based on this initial review, recommendations were made with respect to what instrumentation is needed at the ATR, and a strategy was developed for obtaining these sensors. In 2009, a report was issued documenting this program’s strategy and initial progress toward accomplishing program objectives. Since 2009, annual reports have been issued to provide updates on the program strategy and the progress made on implementing the strategy. This report provides an update reflecting progress as of January 2014.

J. Rempe; D. Knudson; J. Daw; T. Unruh; B. Chase; R. Schley; J. Palmer; K. Condie

2014-01-01T23:59:59.000Z

314

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

SciTech Connect

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

Eric C. Woolstenhulme; Dana M. Meyer

2007-04-01T23:59:59.000Z

315

E-Print Network 3.0 - advanced test idaho reactor Sample Search...  

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

Powered by Explorit Topic List Advanced Search Sample search results for: advanced test idaho reactor Page: << < 1 2 3 4 5 > >> 1 Point of Contact: Doug Kothe CASL Director...

316

A modeling and control approach to advanced nuclear power plants with gas turbines  

Science Journals Connector (OSTI)

Abstract Advanced nuclear power plants are currently being proposed with a number of various designs. However, there is a lack of modeling and control strategies to deal with load following operations. This research investigates a possible modeling approach and load following control strategy for gas turbine nuclear power plants in order to provide an assessment way to the concept designs. A load frequency control strategy and average temperature control mechanism are studied to get load following nuclear power plants. The suitability of the control strategies and concept designs are assessed through linear stability analysis methods. Numerical results are presented on an advanced molten salt reactor concept as an example nuclear power plant system to demonstrate the validity and effectiveness of the proposed modeling and load following control strategies.

Günyaz Ablay

2013-01-01T23:59:59.000Z

317

E-Print Network 3.0 - advanced light-water reactors Sample Search...  

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

of low-energy antineutrino detectors, together... Nuclear reactor safeguards and monitoring with ... Source: Gratta, Giorgio - Kavli Institute for Particle Astrophysics...

318

Development of a molybdenum-rhenium alloy for space nuclear reactors  

SciTech Connect

Information is presented on the fabrication, properties, and use of molybdenum-rhenium alloys for space nuclear reactors.

Lundberg, L.B.

1986-01-01T23:59:59.000Z

319

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

SciTech Connect

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

320

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

Note: This page contains sample records for the topic "advanced nuclear reactor" 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

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

322

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

323

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

324

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

325

Piezoelectric material for use in a nuclear reactor core  

SciTech Connect

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

Parks, D. A.; Reinhardt, Brian; Tittmann, B. R. [EES Department, Penn State University, University Park, PA 16802 (United States)

2012-05-17T23:59:59.000Z

326

Advancing Global Nuclear Security | Department of Energy  

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

OF THE NUCLEAR ENERGY AND NUCLEAR SECURITY WORKING GROUP OF THE BILATERAL U.S. - RUSSIA PRESIDENTIAL COMMISSION Secretary Moniz's Remarks at the 2014 IAEA General Conference...

327

A NOVEL MICROMEGAS DETECTOR FOR IN-CORE NUCLEAR REACTOR NEUTRON FLUX MEASUREMENTS  

E-Print Network (OSTI)

1 A NOVEL MICROMEGAS DETECTOR FOR IN-CORE NUCLEAR REACTOR NEUTRON FLUX MEASUREMENTS S. ANDRIAMONJE Talence Cedex, France Future fast nuclear reactors designed for energy production and transmutation to neutron detection inside nuclear reactor is given. The advantage of this detector over conventional

Paris-Sud XI, Université de

328

XPS Investigations of Ruthenium Deposited onto Representative Inner Surfaces of Nuclear Reactor Containment Buildings  

E-Print Network (OSTI)

XPS Investigations of Ruthenium Deposited onto Representative Inner Surfaces of Nuclear Reactor in a nuclear power plant, interactions of gaseous RuO4 with reactor containment building surfaces (stainless, during nuclear reactor operation, the fission-product ruthenium will accumulate in the fuel. The quantity

Paris-Sud XI, Université de

329

Safety Assurance for Irradiating Experiments in the Advanced Test Reactor  

SciTech Connect

The Advanced Test Reactor (ATR), located at the Idaho National Engineering and Environmental Laboratory (INEEL), was specifically designed to provide a high neutron flux test environment for conducting a variety of experiments. This paper addresses the safety assurance process for two general types of experiments conducted in the ATR facility and how the safety analyses for experiments are related to the ATR safety basis. One type of experiment is more routine and generally represents greater risks; therefore, this type of experiment is addressed in more detail in the ATR safety basis. This allows the individual safety analysis for this type of experiment to be more standardized. The second type of experiment is defined in more general terms in the ATR safety basis and is permitted under more general controls. Therefore, the individual safety analysis for the second type of experiment tends to be more unique and is tailored to each experiment.

T. A. Tomberlin; S. B. Grover

2004-11-01T23:59:59.000Z

330

Advanced neutron source reactor probabilistic flow blockage assessment  

SciTech Connect

The Phase I Level I Probabilistic Risk Assessment (PRA) of the conceptual design of the Advanced Neutron Source (ANS) Reactor identified core flow blockage as the most likely internal event leading to fuel damage. The flow blockage event frequency used in the original ANS PRA was based primarily on the flow blockage work done for the High Flux Isotope Reactor (HFIR) PRA. This report examines potential flow blockage scenarios and calculates an estimate of the likelihood of debris-induced fuel damage. The bulk of the report is based specifically on the conceptual design of ANS with a 93%-enriched, two-element core; insights to the impact of the proposed three-element core are examined in Sect. 5. In addition to providing a probability (uncertainty) distribution for the likelihood of core flow blockage, this ongoing effort will serve to indicate potential areas of concern to be focused on in the preliminary design for elimination or mitigation. It will also serve as a loose-parts management tool.

Ramsey, C.T.

1995-08-01T23:59:59.000Z

331

Fuel qualification plan for the Advanced Neutron Source Reactor  

SciTech Connect

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

Copeland, G.L.

1995-07-01T23:59:59.000Z

332

Advanced Reactor Innovation Evaluation Study (ARIES) Properties Archive  

DOE Data Explorer (OSTI)

ARIES stands for Advanced Reactor Innovation Evaluation Study. It is a program and a team that explores the commercial potential of fusion as an energy resource. Though it is a multi-institutional program, ARIES is led by the University of California at San Diego. ARIES studies both magnetic fusion energy (MFE) and inertial fusion energy (IFE), using an approach that integrates theory, experiments, and technology. The ARIES team proposes fusion reactor designs and works to understand how technology, materials and plasma physics processes interact and influence each other. A 2005 report to the Fusion Energy Sciences Advisory Committee ("Scientific Challenges, Opportunities, and Priorities for the U.S. Fusion Energy Sciences Program") noted on page 98 an example of the importance of this materials properties aspect: "For instance, effects on plasma edge by various plasma facing materials and effects on various plasma stabilization and control techniques by highly conducting liquid metal blankets are being considered by physicists." This web page is an archive of material properties collected here for the use of the ARIES Fusion Power Plant Studies Team.

333

Nuclear Fuel in a Reactor Accident  

Science Journals Connector (OSTI)

...Three Mile Island: A report to the commissioners and to the public” (Nuclear Regulatory Commission, Washington, DC, 1980). 5...Podcast The contents of this podcast interview represent the opinion of the author and may go beyond the content of the published...

Peter C. Burns; Rodney C. Ewing; Alexandra Navrotsky

2012-03-09T23:59:59.000Z

334

Light Water Reactor Sustainability Program Risk Informed Safety Margin Characterization (RISMC) Advanced Test Reactor Demonstration Case Study  

SciTech Connect

Safety is central to the design, licensing, operation, and economics of Nuclear Power Plants (NPPs). Consequently, the ability to better characterize and quantify safety margin holds the key to improved decision making about LWR design, operation, and plant life extension. A systematic approach to characterization of safety margins and the subsequent margins management options represents a vital input to the licensee and regulatory analysis and decision making that will be involved. The purpose of the RISMC Pathway R&D is to support plant decisions for risk-informed margins management with the aim to improve economics, reliability, and sustain safety of current NPPs. Goals of the RISMC Pathway are twofold: (1) Develop and demonstrate a risk-assessment method coupled to safety margin quantification that can be used by NPP decision makers as part of their margin recovery strategies. (2) Create an advanced “RISMC toolkit” that enables more accurate representation of NPP safety margin. This report describes the RISMC methodology demonstration where the Advanced Test Reactor (ATR) was used as a test-bed for purposes of determining safety margins. As part of the demonstration, we describe how both the thermal-hydraulics and probabilistic safety calculations are integrated and used to quantify margin management strategies.

Curtis Smith; David Schwieder; Cherie Phelan; Anh Bui; Paul Bayless

2012-08-01T23:59:59.000Z

335

Nuclear plant-aging research on reactor protection systems  

SciTech Connect

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

Meyer, L.C.

1988-01-01T23:59:59.000Z

336

E-Print Network 3.0 - advanced nuclear materials Sample Search...  

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

Powered by Explorit Topic List Advanced Search Sample search results for: advanced nuclear materials Page: << < 1 2 3 4 5 > >> 1 Enabling a Sustainable Nuclear Energy Future...

337

E-Print Network 3.0 - advanced nuclear power Sample Search Results  

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

Powered by Explorit Topic List Advanced Search Sample search results for: advanced nuclear power Page: << < 1 2 3 4 5 > >> 1 Nuclear Engineering Graduate Program Summary: Power...

338

E-Print Network 3.0 - advanced nuclear engineering Sample Search...  

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

Powered by Explorit Topic List Advanced Search Sample search results for: advanced nuclear engineering Page: << < 1 2 3 4 5 > >> 1 Nuclear Engineering Graduate Program Summary:...

339

Department of Energy Issues Final $12.5 Billion Advanced Nuclear...  

Energy Savers (EERE)

Final 12.5 Billion Advanced Nuclear Energy Loan Guarantee Solicitation Department of Energy Issues Final 12.5 Billion Advanced Nuclear Energy Loan Guarantee Solicitation December...

340

Proceedings of the 1992 topical meeting on advances in reactor physics. Volume 2  

SciTech Connect

This document, Volume 2, presents proceedings of the 1992 Topical Meeting on Advances in Reactor Physics on March 8--11, 1992 at Charleston, SC. Session topics were as follows: Transport Theory; Fast Reactors; Plant Analyzers; Integral Experiments/Measurements & Analysis; Core Computational Systems; Reactor Physics; Monte Carlo; Safety Aspects of Heavy Water Reactors; and Space-Time Core Kinetics. The individual reports have been cataloged separately. (FI)

Not Available

1992-04-01T23:59:59.000Z

Note: This page contains sample records for the topic "advanced nuclear reactor" 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

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

342

Advanced LWR Nuclear Fuel Cladding System Development Trade-off Study |  

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

LWR Nuclear Fuel Cladding System Development Trade-off LWR Nuclear Fuel Cladding System Development Trade-off Study Advanced LWR Nuclear Fuel Cladding System Development Trade-off Study The LWR Sustainability (LWRS) Program activities must support the timeline dictated by utility life extension decisions to demonstrate a lead test rod in a commercial reactor within 10 years. In order to maintain the demanding development schedule that must accompany this aggressive timeline, the LWRS Program focuses on advanced fuel cladding systems that retain standard UO2 fuel pellets for deployment in currently operating LWR power plants. The LWRS work scope focuses on fuel system components outside of the fuel pellet, allowing for alteration of the existing zirconium-based clad system through coatings, addition of ceramic sleeves, or complete replacement

343

Space–time kinetics modeling of Advanced Heavy Water Reactor for control studies  

Science Journals Connector (OSTI)

The paper presents the mathematical modeling of the space–time kinetics phenomena in Advanced Heavy Water Reactor (AHWR), a 920 MW (thermal), vertical pressure tube type thorium based nuclear reactor. The physical dimensions and the internal feedback effects of the AHWR are such that it is susceptible to xenon induced spatial oscillations. For the study of spatial effects and design of suitable control strategy, the need for a suitable mathematical model which is not of a very large order arises. In this paper, a mathematical model of the reactor within the framework of nodal modeling is derived with the two group neutron diffusion equation as the basis. A linear model in standard state space form is formulated from the set of equations so obtained. It has been shown that comparison of linear system properties could be helpful in deciding upon an appropriate nodalization scheme and thus obtaining a reasonably accurate model. For validation, the transient response of the simplified model has been compared with those from a rigorous finite-difference model.

S.R. Shimjith; A.P. Tiwari; M. Naskar; B. Bandyopadhyay

2010-01-01T23:59:59.000Z

344

Control of xenon oscillations in Advanced Heavy Water Reactor via two-stage decomposition  

Science Journals Connector (OSTI)

Abstract Xenon induced spatial oscillations developed in large nuclear reactors, like Advanced Heavy Water Reactor (AHWR) need to be controlled for safe operation. Otherwise, a serious situation may arise in which different regions of the core may undergo variations in neutron flux in opposite phase. If these oscillations are left uncontrolled, the power density and rate of change of power at some locations in the reactor core may exceed their respective thermal limits, resulting in fuel failure. In this paper, a state feedback based control strategy is investigated for spatial control of AHWR. The nonlinear model of AHWR including xenon and iodine dynamics is characterized by 90 states, 5 inputs and 18 outputs. The linear model of AHWR, obtained by linearizing the nonlinear equations is found to be highly ill-conditioned. This higher order model of AHWR is first decomposed into two comparatively lower order subsystems, namely, 73rd order ‘slow’ subsystem and 17th order ‘fast’ subsystem using two-stage decomposition. Composite control law is then derived from individual subsystem feedback controls and applied to the vectorized nonlinear model of AHWR. Through the dynamic simulations it is observed that the controller is able to suppress xenon induced spatial oscillations developed in AHWR and the overall performance is found to be satisfactory.

R.K. Munje; J.G. Parkhe; B.M. Patre

2015-01-01T23:59:59.000Z

345

Synergistic smart fuel for in-pile nuclear reactor measurements  

SciTech Connect

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

Smith, J.A.; Kotter, D.K. [Idaho National Laboratories, Idaho Falls (United States); Ali, R.A.; Garrett, S.L. [Penn State University, University Park, State College, PA 16801 (United States)

2013-07-01T23:59:59.000Z

346

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

347

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

348

Study of plutonium disposition using existing GE advanced Boiling Water Reactors  

SciTech Connect

The end of the cold war and the resulting dismantlement of nuclear weapons has resulted in the need for the US to dispose of 50 to 100 metric tons of excess of plutonium in a safe and proliferation resistant manner. A number of studies, including the recently released National Academy of Sciences (NAS) study, have recommended conversion of plutonium into spent nuclear fuel with its high radiation barrier as the best means of providing permanent conversion and long-term diversion resistance to this material. The NAS study ``Management and Disposition of Excess Weapons Plutonium identified Light Water Reactor spent fuel as the most readily achievable and proven form for the disposition of excess weapons plutonium. The study also stressed the need for a US disposition program which would enhance the prospects for a timely reciprocal program agreement with Russia. This summary provides the key findings of a GE study where plutonium is converted into Mixed Oxide (MOX) fuel and a typical 1155 MWe GE Boiling Water Reactor (BWR) is utilized to convert the plutonium to spent fuel. A companion study of the Advanced BWR has recently been submitted. The MOX core design work that was conducted for the ABWR enabled GE to apply comparable fuel design concepts and consequently achieve full MOX core loading which optimize plutonium throughput for existing BWRs.

Not Available

1994-06-01T23:59:59.000Z

349

Relap5-3d model validation and benchmark exercises for advanced gas cooled reactor application  

E-Print Network (OSTI)

HTTR High Temperature engineering Test Reactor INET Institute of Nuclear Energy Technology LWR Light Water Reactor OKBM Test Design Bureau for Machine Building ORNL Oak Ridge National Laboratory RCCS Reactor Cavity Cooling System... to be at right angles to each other, ignoring an angular distribution of radiant heat.7 MORECA, used by ORNL, simulates accident scenarios for certain gas-cooled reactor types.7 INET conducts their analysis using Thermix, which performs two...

Moore, Eugene James Thomas

2006-08-16T23:59:59.000Z

350

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

Science Journals Connector (OSTI)

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

Charles S. Olsen; Henry Welland; James Abraschoff; Kenneth Thoms

1993-01-01T23:59:59.000Z

351

E-Print Network 3.0 - austrian research reactor Sample Search...  

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

to build the world's most advanced nuclear fusion reactor... ) International Thermonuclear Experimental Reactor (ITER), which will be built at Cadarache, near the...

352

E-Print Network 3.0 - anuclear research reactor Sample Search...  

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

to build the world's most advanced nuclear fusion reactor... ) International Thermonuclear Experimental Reactor (ITER), which will be built at Cadarache, near the...

353

E-Print Network 3.0 - atomic reactors nouvelles Sample Search...  

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

to build the world's most advanced nuclear fusion reactor... ) International Thermonuclear Experimental Reactor (ITER), which will be built at Cadarache, ... Source:...

354

Removable check valve for use in a nuclear reactor  

DOE Patents (OSTI)

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

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

1988-01-01T23:59:59.000Z

355

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

356

Systems and methods for dismantling a nuclear reactor  

DOE Patents (OSTI)

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

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

2014-10-28T23:59:59.000Z

357

Performance of Liquid Metals in Natural Circulation Cooled Nuclear Reactors  

SciTech Connect

The inherent safety capability of natural circulation makes reactor design more reliable. Additionally, the construction and operation of a nuclear power plant with natural circulation in the primary cooling circuit is an interesting alternative for nuclear plant designers, due to their lower operational and investment costs obtained by simplifying systems and controls. This paper deals with the feasibility of application of natural circulation in the primary cooling circuit of a liquid metal fast reactor. The methodology employed is a non-dimensional analysis, which describes the relationship between the physical properties and system variables. The performance criterion is bounded by a safety argument, referring to the maximum cladding temperature allowed during operation. The study considers several coolants, which can play a part in reactor cooling systems, such as lead, lead-bismuth and sodium. Bismuth and gallium are included in this analysis, in order to extend the range of properties for reference purposes. The results present a characterization of natural circulation flow in a reactor and compare the cooling capabilities from different liquid metals coolants. (authors)

Ceballos, Carlos; Lathouwers, Danny; Verkooijen, Adrian [Interfacultair Reactor Instituut, Technische Universiteit Delft, Mekelweg 15, Delft (Netherlands)

2004-07-01T23:59:59.000Z

358
359

Detachable connection for a nuclear reactor fuel assembly  

DOE Patents (OSTI)

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

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

1986-01-01T23:59:59.000Z

360

Search for Neutrino Oscillations at the Palo Verde Nuclear Reactors  

E-Print Network (OSTI)

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

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

1999-12-22T23:59:59.000Z

Note: This page contains sample records for the topic "advanced nuclear reactor" 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

Expert system for online surveillance of nuclear reactor coolant pumps  

DOE Patents (OSTI)

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

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

1993-01-01T23:59:59.000Z

362

The DOE Advanced Gas Reactor (AGR) Fuel Development and Qualification Program  

SciTech Connect

The Department of Energy has established the Advanced Gas Reactor Fuel Development and Qualification Program to address the following overall goals: Provide a baseline fuel qualification data set in support of the licensing and operation of the Next Generation Nuclear Plant (NGNP). Gas-reactor fuel performance demonstration and qualification comprise the longest duration research and development (R&D) task for the NGNP feasibility. The baseline fuel form is to be demonstrated and qualified for a peak fuel centerline temperature of 1250°C. Support near-term deployment of an NGNP by reducing market entry risks posed by technical uncertainties associated with fuel production and qualification. Utilize international collaboration mechanisms to extend the value of DOE resources. The Advanced Gas Reactor Fuel Development and Qualification Program consists of five elements: fuel manufacture, fuel and materials irradiations, postirradiation examination (PIE) and safety testing, fuel performance modeling, and fission product transport and source term evaluation. An underlying theme for the fuel development work is the need to develop a more complete fundamental understanding of the relationship between the fuel fabrication process, key fuel properties, the irradiation performance of the fuel, and the release and transport of fission products in the NGNP primary coolant system. Fuel performance modeling and analysis of the fission product behavior in the primary circuit are important aspects of this work. The performance models are considered essential for several reasons, including guidance for the plant designer in establishing the core design and operating limits, and demonstration to the licensing authority that the applicant has a thorough understanding of the in-service behavior of the fuel system. The fission product behavior task will also provide primary source term data needed for licensing. An overview of the program and recent progress will be presented.

David Petti; Hans Gougar; Gary Bell

2005-05-01T23:59:59.000Z

363

Monte Carlo Domain Decomposition for Robust Nuclear Reactor Analyses  

Science Journals Connector (OSTI)

Abstract Monte Carlo (MC) neutral particle transport codes are considered the gold-standard for nuclear simulations, but they cannot be robustly applied to high-fidelity nuclear reactor analysis without accommodating several terabytes of materials and tally data. While this is not a large amount of aggregate data for a typical high performance computer, MC methods are only embarrassingly parallel when the key data structures are replicated for each processing element, an approach which is likely infeasible on future machines. The present work explores the use of spatial domain decomposition to make full-scale nuclear reactor simulations tractable with Monte Carlo methods, presenting a simple implementation in a production-scale code. Good performance is achieved for mesh-tallies of up to 2.39TB distributed across 512 compute nodes while running a full-core reactor benchmark on the Mira Blue Gene/Q supercomputer at the Argonne National Laboratory. In addition, the effects of load imbalances are explored with an updated performance model that is empirically validated against observed timing results. Several load balancing techniques are also implemented to demonstrate that imbalances can be largely mitigated, including a new and efficient way to distribute extra compute resources across coarse domain meshes.

Nicholas Horelik; Andrew Siegel; Benoit Forget; Kord Smith

2014-01-01T23:59:59.000Z

364

Secondary Heat Exchanger Design and Comparison for Advanced High Temperature Reactor  

SciTech Connect

The goals of next generation nuclear reactors, such as the high temperature gas-cooled reactor and advance high temperature reactor (AHTR), are to increase energy efficiency in the production of electricity and provide high temperature heat for industrial processes. The efficient transfer of energy for industrial applications depends on the ability to incorporate effective heat exchangers between the nuclear heat transport system and the industrial process heat transport system. The need for efficiency, compactness, and safety challenge the boundaries of existing heat exchanger technology, giving rise to the following study. Various studies have been performed in attempts to update the secondary heat exchanger that is downstream of the primary heat exchanger, mostly because its performance is strongly tied to the ability to employ more efficient conversion cycles, such as the Rankine super critical and subcritical cycles. This study considers two different types of heat exchangers—helical coiled heat exchanger and printed circuit heat exchanger—as possible options for the AHTR secondary heat exchangers with the following three different options: (1) A single heat exchanger transfers all the heat (3,400 MW(t)) from the intermediate heat transfer loop to the power conversion system or process plants; (2) Two heat exchangers share heat to transfer total heat of 3,400 MW(t) from the intermediate heat transfer loop to the power conversion system or process plants, each exchanger transfers 1,700 MW(t) with a parallel configuration; and (3) Three heat exchangers share heat to transfer total heat of 3,400 MW(t) from the intermediate heat transfer loop to the power conversion system or process plants. Each heat exchanger transfers 1,130 MW(t) with a parallel configuration. A preliminary cost comparison will be provided for all different cases along with challenges and recommendations.

Piyush Sabharwall; Ali Siahpush; Michael McKellar; Michael Patterson; Eung Soo Kim

2012-06-01T23:59:59.000Z

365

Advanced Fusion Reactors for Space Propulsion and Power Systems  

SciTech Connect

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

Chapman, John J.

2011-06-15T23:59:59.000Z

366

INVESTIGATIONS ON NUCLEAR SPECTROSCOPY AT THE REACTOR AND THEIR APPLICATIONS1  

E-Print Network (OSTI)

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

Titov, Anatoly

367

Qualification issues associated with the use of advanced instrumentation and control systems hardware in nuclear power plants  

SciTech Connect

The instrumentation and control (I&C) systems in advanced reactors will make extensive use of digital controls, microprocessors, multiplexing, and Tiber-optic transmission. Elements of these advances in I&C have been implemented on some current operating plants. However, the widespread use of the above technologies, as well as the use of artificial intelligence with minimum reliance on human operator control of reactors, highlights the need to develop standards for qualifying I&C used in the next generation of nuclear power plants. As a first step in this direction, the protection system I&C for present-day plants was compared to that proposed for advanced light water reactors (ALWRs). An evaluation template was developed by assembling a configuration of a safety channel instrument string for a generic ALWR, then comparing the impact of environmental stressors on that string to their effect on an equivalent instrument string from an existing light water reactor. The template was then used to address reliability issues for microprocessor-based protection systems. Standards (or lack thereof) for the qualification of microprocessor-based safety I&C systems were also identified. This approach addresses in part issues raised in Nuclear Regulatory Commission policy document SECY-91-292. which recognizes that advanced I&C systems for the nuclear industry are ``being developed without consensus standards, as the technology available for design is ahead of the technology that is well understood through experience and supported by application standards.``

Korsah, K. [Oak Ridge National Lab., TN (United States); Antonescu, C. [Nuclear Regulatory Commission, Rockville, MD (United States). Office of Nuclear Regulatory Research

1993-10-01T23:59:59.000Z

368

Nuclear Energy Advanced Modeling and Simulation (NEAMS) Software  

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

Advanced Modeling and Simulation (NEAMS) Software Advanced Modeling and Simulation (NEAMS) Software Verification and Validation (V&V) Plan Requirements Nuclear Energy Advanced Modeling and Simulation (NEAMS) Software Verification and Validation (V&V) Plan Requirements The purpose of the NEAMS Software V&V Plan is to define what the NEAMS program expects in terms of V&V for the computational models that are developed under NEAMS. NEAMS Software Verification and Validation Plan Requirements Version 0.pdf More Documents & Publications NEAMS Quarterly Report for January-March 2013 Nuclear Energy Advanced Modeling and Simulation (NEAMS) Program Plan CRAD, Assessment Criteria and Guidelines for Determining the Adequacy of Software Used in the Safety Analysis and Design of Defense Nuclear Facilities

369

Advances in instrumentation for nuclear astrophysics  

SciTech Connect

The study of the nuclear physics properties which govern energy generation and nucleosynthesis in the astrophysical phenomena we observe in the universe is crucial to understanding how these objects behave and how the chemical history of the universe evolved to its present state. The low cross sections and short nuclear lifetimes involved in many of these reactions make their experimental determination challenging, requiring developments in beams and instrumentation. A selection of developments in nuclear astrophysics instrumentation is discussed, using as examples projects involving the nuclear astrophysics group at Oak Ridge National Laboratory. These developments will be key to the instrumentation necessary to fully exploit nuclear astrophysics opportunities at the Facility for Rare Isotope Beams which is currently under construction.

Pain, S. D. [Physics Division, Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States)] [Physics Division, Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States)

2014-04-15T23:59:59.000Z

370

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

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.

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

1991-01-01T23:59:59.000Z

371

Technology Readiness Levels for Advanced Nuclear Fuels and Materials Development  

SciTech Connect

The Technology Readiness Level (TRL) process is used to quantitatively assess the maturity of a given technology. The TRL process has been developed and successfully used by the Department of Defense (DOD) for development and deployment of new technology and systems for defense applications. In addition, NASA has also successfully used the TRL process to develop and deploy new systems for space applications. Advanced nuclear fuels and materials development is a critical technology needed for closing the nuclear fuel cycle. Because the deployment of a new nuclear fuel forms requires a lengthy and expensive research, development, and demonstration program, applying the TRL concept to the advanced fuel development program is very useful as a management and tracking tool. This report provides definition of the technology readiness level assessment process as defined for use in assessing nuclear fuel technology development for the Advanced Fuel Campaign (AFC).

Jon Carmack

2014-01-01T23:59:59.000Z

372

Proceedings of the US Nuclear Regulatory Commission fourteenth water reactor safety information meeting: Volume 1, Plenary session, Severe accident sequence analysis, Risk analysis/PRA applications, Reference plant risk analysis - NUREG-1150, Innovative concepts for increased safety of advanced power reactors  

SciTech Connect

This six-volume report contains 156 papers out of the 175 that were presented at the Fourteenth Water Reactor Safety Information Meeting held at the National Bureau of Standards, Gaithersburg, Maryland, during the week of October 27-31, 1986. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included thirty-four different papers presented by researchers from Canada, Czechoslovakia, Finland, Germany, Italy, Japan, Mexico, Spain, Sweden, Switzerland and the United Kingdom. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting.

Weiss, A.J. (comp.)

1987-02-01T23:59:59.000Z

373

Research in nondestructive evaluation techniques for nuclear reactor concrete structures  

SciTech Connect

The purpose of the Materials Aging and Degradation (MAaD) Pathway of the Department of Energy's Light Water Reactor Sustainability (LWRS) Program is to develop the scientific basis for understanding and predicting longterm environmental degradation behavior of material in nuclear power plants and to provide data and methods to assess the performance of systems, structures, and components (SSCs) essential to safe and sustained nuclear power plant operations. The understanding of aging-related phenomena and their impacts on SSCs is expected to be a significant issue for any nuclear power plant planning for long-term operations (i.e. service beyond the initial license renewal period). Management of those phenomena and their impacts during long-term operations can be better enable by improved methods and techniques for detection, monitoring, and prediction of SSC degradation. The MAaD Pathway R and D Roadmap for Concrete, 'Light Water Reactor Sustainability Nondestructive Evaluation for Concrete Research and Development Roadmap', focused initial research efforts on understanding the recent concrete issues at nuclear power plants and identifying the availability of concrete samples for NDE techniques evaluation and testing. [1] An overview of the research performed by ORNL in these two areas is presented here.

Clayton, Dwight; Smith, Cyrus [Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831 (United States)

2014-02-18T23:59:59.000Z

374

Dynamic Complexity Study of Nuclear Reactor and Process Heat Application Integration  

SciTech Connect

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

J'Tia Patrice Taylor; David E. Shropshire

2009-09-01T23:59:59.000Z

375

Foreign Research Reactor Spent Nuclear Fuel Acceptance Program  

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

Global Threat Reduction Initiative: Global Threat Reduction Initiative: U.S. Nuclear Remove Program Foreign Research Reactor Spent Nuclear Fuel (FRR SNF) Acceptance 2007 DOE TEC Meeting Chuck Messick DOE/NNSA/SRS 2 Contents * Program Objective and Policy * Program implementation status * Shipment Information * Operational Logistics * Lessons Learned * Conclusion 3 U.S. Nuclear Remove Program Objective * To play a key role in the Global Threat Reduction Remove Program supporting permanent threat reduction by accepting program eligible material. * Works in conjunction with the Global Threat Reduction Convert Program to accept program eligible material as an incentive to core conversion providing a disposition path for HEU and LEU during the life of the Acceptance Program. 4 Reasons for the Policy

376

Solar Thermochemical Advanced Reactor System, Wins R&D 100 Award  

Office of Energy Efficiency and Renewable Energy (EERE)

Solar Thermochemical Advanced Reactor System, or STARS, converts natural gas and sunlight into a more energy-rich fuel called syngas, which power plants can burn to make electricity.

377

U.N. report concludes that Syrian site destroyed in 2007 was a nuclear reactor  

E-Print Network (OSTI)

U.N. report concludes that Syrian site destroyed in 2007 was a nuclear reactor Joby Warrick, 24 May.N. claims was a nuclear plant before and after a Sept. 6 Israeli airstrike. The left image is from 5 Aug that Syria "very likely" was building a secret nuclear reactor in 2007 when the partially completed project

378

Advanced Heat Exchanger Development for Molten Salts in Nuclear and Non Nuclear Systems  

SciTech Connect

This study addresses present work concerned with advanced heat exchanger development for molten salt in nuclear and non nuclear thermal systems. The molten salt systems discussed herein use alloys, such as Hastelloy N and 242, which show corrosion resistance to molten salt at nominal operating temperatures up to 700°C. These alloys were diffusion welded, and the corresponding information is presented. Test specimens were prepared for exposing diffusion welds to molten salt environments. Hastelloy N and 242 were found to be weldable by diffusion welding, with ultimate tensile strengths about 90% of base metal values. Both diffusion welds and sheet material in Hastelloy N were corrosion tested in?58 mol% KF and 42 mol% ZrF4 at 650, 700, and 850°C for 200, 500, and 1,000 hours. Corrosion rates found were similar between welded and nonwelded materials, typically <10 mils per year. For materials of construction, nickel and alloys with dense nickel coatings are effectively inert to corrosion in fluorides, but not so in chlorides. Hence, additional testing of selected alloys for resistance to intergranular corrosion is needed, as is a determination of corrosion rate as a function of contaminant type and alloy composition with respect to chromium and carbon to better define the optimal chromium and carbon composition, independent of galvanic or differential solubility effects. Also presented is the division of the nuclear reactor and high temperature components per ASME standards, along with design requirements for a subcritical Rankine power cycle heat exchanger that has to overcome pressure difference of about 17 MPa.

Piyush Sabharwall; Denis Clark; Kumar Sridharan; Guiqiu Zheng; Mark Anderson

2014-10-01T23:59:59.000Z

379

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

SciTech Connect

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

Cooke, Conrad; Spann, Holger [Siempelkamp Nuclear Services: 5229 Sunset Blvd., (Suite M), West Columbia, SC, 29169 (United States)] [Siempelkamp Nuclear Services: 5229 Sunset Blvd., (Suite M), West Columbia, SC, 29169 (United States)

2013-07-01T23:59:59.000Z

380

Light weight space power reactors for nuclear electric propulsion  

SciTech Connect

A Nuclear Electric Propulsion (NEP) unit capable of propelling a manned vehicle to MARS will be required to have a value of {alpha} (kg/kWe) which is less than five. In order to meet this goal the reactor mass, and thus its contribution to the value of {alpha} will have to be minimized. In this paper a candidate for such a reactor is described. It consists of a gas cooled Particle Bed Reactor (PBR), with specially chosen materials which allow it to operate at an exit temperature of approximately 2000 K. One of the unique features of a PBR is the direct cooling of particulate fuel by the working fluid. This feature allows for high power densities, highest possible gas exit temperatures, for a given fuel temperature and because of the thin particle bed a low pressure drop. The PBR's described in this paper will have a ceramic moderator (Be{sub 2}C), ZrC coated fuel particles and a carbon/carbon hot frit. All the reactors will be designed with sufficient fissile loading to operate at full power for seven years. The burn up possible with particulate fuel is approximately 30%--50%. These rector designs achieve a value of {alpha} less than unity in the power range of interest (5 MWe). 5 refs., 3 figs.

Ludewig, H.; Mughabghab, S.; Lazareth, O.; Perkins, K.; Schmidt, E.; Powell, J.R.

1991-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "advanced nuclear reactor" 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

Development of Improved Burnable Poisons for Commercial Nuclear Power Reactors  

SciTech Connect

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

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

2003-09-30T23:59:59.000Z

382

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

SciTech Connect

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

J. Blair Briggs; Anatoly Tsibulya; Yevgeniy Rozhikhin

2012-03-01T23:59:59.000Z

383

Consortium for Advanced Simulation of Light Water Reactors (CASL...  

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

the better understanding of the system uncertainties and sensitivities afforded by the virtual reactor will identify improvements in both the operation and design of the fuel...

384

E-Print Network 3.0 - advanced light reactor Sample Search Results  

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

2 3 4 5 > >> Page: << < 1 2 3 4 5 > >> 21 Nuclear reactor safeguards and monitoring with antineutrino detectors A. Bernsteina) Summary: at the end of 2000.2 Most are light water...

385

Roadmap for development of an advanced head-end reactor  

SciTech Connect

A novel dry treatment process for used nuclear fuel (UNF) using nitrogen dioxide is being developed to remove volatile and semi-volatile fission products and convert the monolithic fuel material to a fine powder suitable as a feed to many different separations processes. The process may be considered an advanced form of voloxidation, which was envisioned to remove tritium from the fuel prior to introduction of the fuel into the aqueous separations systems, where subsequent separation of tritium from the water would be difficult and expensive. The product from NO{sub 2} reaction can be selectively chosen to be U{sub 3}O{sub 8}, UO{sub 3}, or a nitrate by adjusting the processing conditions; all products are generated at temperatures lower than those used in standard voloxidation. All the fundamental tenants of the process have been successfully demonstrated as a proof of principle, and many aspects have been corroborated multiple times at laboratory scale. The goal of this roadmap is to define the activities required to develop the process to a technology-readiness level sufficient to an engineering-scale implementation. (authors)

Del Cul, G.D.; Johnson, J.A.; Spencer, B.B.; Collins, E.D. [Oak Ridge National Laboratory, Oak Ridge, Tennessee 37831-6243 (United States)

2013-07-01T23:59:59.000Z

386

Proceedings of the Twenty-First Water Reactor Safety Information Meeting: Volume 1, Plenary session; Advanced reactor research; advanced control system technology; advanced instrumentation and control hardware; human factors research; probabilistic risk assessment topics; thermal hydraulics; thermal hydraulic research for advanced passive LWRs  

SciTech Connect

This three-volume report contains 90 papers out of the 102 that were presented at the Twenty-First Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, during the week of October 25--27, 1993. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from France, Germany, Japan, Russia, Switzerland, Taiwan, and United Kingdom. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting. Individual papers have been cataloged separately. This document, Volume 1 covers the following topics: Advanced Reactor Research; Advanced Instrumentation and Control Hardware; Advanced Control System Technology; Human Factors Research; Probabilistic Risk Assessment Topics; Thermal Hydraulics; and Thermal Hydraulic Research for Advanced Passive Light Water Reactors.

Monteleone, S. [Brookhaven National Lab., Upton, NY (United States)] [comp.

1994-04-01T23:59:59.000Z

387

Assessment of Silicon Carbide Composites for Advanced Salt-Cooled Reactors  

SciTech Connect

The Advanced High-Temperature Reactor (AHTR) is a new reactor concept that uses a liquid fluoride salt coolant and a solid high-temperature fuel. Several alternative fuel types are being considered for this reactor. One set of fuel options is the use of pin-type fuel assemblies with silicon carbide (SiC) cladding. This report provides (1) an initial viability assessment of using SiC as fuel cladding and other in-core components of the AHTR, (2) the current status of SiC technology, and (3) recommendations on the path forward. Based on the analysis of requirements, continuous SiC fiber-reinforced, chemically vapor-infiltrated SiC matrix (CVI SiC/SiC) composites are recommended as the primary option for further study on AHTR fuel cladding among various industrially available forms of SiC. Critical feasibility issues for the SiC-based AHTR fuel cladding are identified to be (1) corrosion of SiC in the candidate liquid salts, (2) high dose neutron radiation effects, (3) static fatigue failure of SiC/SiC, (4) long-term radiation effects including irradiation creep and radiation-enhanced static fatigue, and (5) fabrication technology of hermetic wall and sealing end caps. Considering the results of the issues analysis and the prospects of ongoing SiC research and development in other nuclear programs, recommendations on the path forward is provided in the order or priority as: (1) thermodynamic analysis and experimental examination of SiC corrosion in the candidate liquid salts, (2) assessment of long-term mechanical integrity issues using prototypical component sections, and (3) assessment of high dose radiation effects relevant to the anticipated operating condition.

Katoh, Yutai [ORNL; Wilson, Dane F [ORNL; Forsberg, Charles W [ORNL

2007-09-01T23:59:59.000Z

388

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

Office of Science (SC) Website

instrumentation: Improved Design of Nuclear Reactor Control System Developed at: Oak Ridge National Laboratory, Holifield Radioactive Ion Beam Facility (HRIBF) Developed...

389

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.… (more)

Jiltchenkov, Dmitri Victorovich

2012-01-01T23:59:59.000Z

390

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

E-Print Network (OSTI)

??Heat transfer, neutronics, and accident analyses are presented to support an increase of the power level of the Texas A&M University Nuclear Science Center Reactor.… (more)

Rearden, Bradley Thomas

2012-01-01T23:59:59.000Z

391

Iterative methods for solving nonlinear problems of nuclear reactor criticality  

SciTech Connect

The paper presents iterative methods for calculating the neutron flux distribution in nonlinear problems of nuclear reactor criticality. Algorithms for solving equations for variations in the neutron flux are considered. Convergence of the iterative processes is studied for two nonlinear problems in which macroscopic interaction cross sections are functionals of the spatial neutron distribution. In the first problem, the neutron flux distribution depends on the water coolant density, and in the second one, it depends on the fuel temperature. Simple relationships connecting the vapor content and the temperature with the neutron flux are used.

Kuz'min, A. M., E-mail: mephi.kam@mail.ru [National Research Nuclear University MEPhI (Russian Federation)

2012-12-15T23:59:59.000Z

392

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

393

Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR)  

SciTech Connect

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

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

1992-10-01T23:59:59.000Z

394

DOE-STD-0100T; DOE Standard Licensed Reactor Nuclear Safety Criteria Applicable to DOE Reactors  

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

00T 00T November 1993 Superseding DOE/NE-0100T April 1991 DOE STANDARD LICENSED REACTOR NUCLEAR SAFETY CRITERIA APPLICABLE TO DOE REACTORS U.S. Department of Energy Washington, D.C. 20585 AREA SAFT DISTRIBUTION STATEMENT A. Approved for public release; distribution is unlimited. DISTRIBUTION STATEMENT A. Approved for public release; distribution is unlimited. This document has been reproduced directly frorn the best available copy. Available to DOE and DOE contractors from the Office of Scientific and Technical Information, P.O. Box 62, Oak Ridge, TN 37831; (615) 576-8401. Available to the public from the U.S. Department of Commerce, Technology Administration, National Technical Information Service, Springfield, VA 22161; (703) 487-4650. Order No. DE94005221 CONTENTS

395

Presented by CASL: The Consortium for Advanced  

E-Print Network (OSTI)

against 60% of existing U.S. reactor fleet (PWRs), using data from TVA reactors · Base M&S LWR capabilityPresented by Nuclear Energy CASL: The Consortium for Advanced Simulation of Light Water Reactors A DOE Energy Innovation Hub for Modeling and Simulation of Nuclear Reactors Doug Kothe Director, CASL

396

Uncertainty analysis of LBLOCA for Advanced Heavy Water Reactor  

Science Journals Connector (OSTI)

The main objective of safety analysis is to demonstrate in a robust way that all safety requirements are met, i.e. sufficient margins exist between real values of important parameters and their threshold values at which damage of the barriers against release of radioactivity would occur. As stated in the IAEA Safety Requirements for Design of \\{NPPs\\} “a safety analysis of the plant design shall be conducted in which methods of both deterministic and probabilistic analysis shall be applied”. It is required that “the computer programs, analytical methods and plant models used in the safety analysis shall be verified and validated, and adequate consideration shall be given to uncertainties”. Uncertainties are present in calculations due to the computer codes, initial and boundary conditions, plant state, fuel parameters, scaling and numerical solution algorithm. All conservative approaches, still widely used, were introduced to cover uncertainties due to limited capability for modelling and understanding of physical phenomena at the early stages of safety analysis. The results obtained by this approach are quite unrealistic and the level of conservatism is not fully known. Another approach is the use of Best Estimate (BE) codes with realistic initial and boundary conditions. If this approach is selected, it should be based on statistically combined uncertainties for plant initial and boundary conditions, assumptions and code models. The current trends are going into direction of the best estimate code with some conservative assumptions of the system with realistic input data with uncertainty analysis. The BE analysis with evaluation of uncertainties offers, in addition, a way to quantify the existing plant safety margins. Its broader use in the future is therefore envisaged, even though it is not always feasible because of the difficulty of quantifying code uncertainties with sufficiently narrow range for every phenomenon and for each accident sequence. In this paper, uncertainty analysis for the Large Break LOCA (200% Inlet Header Break) of Advanced Heavy Water Reactor (AHWR) has been carried out. The uncertainty analysis was carried out for the peak cladding temperature (PCT), based on the two different methods i.e., Wilk’s method and the response surface technique. Their findings have also been compared.

A. Srivastava; H.G. Lele; A.K. Ghosh; H.S. Kushwaha

2008-01-01T23:59:59.000Z

397

Status of the NGNP Graphite Creep Experiments AGC-1 and AGC-2 Irradiated in the Advanced Test Reactor  

SciTech Connect

The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Program will be irradiating six nuclear graphite creep experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The graphite experiments will be irradiated over the next six to eight years to support development of a graphite irradiation performance data base on the new nuclear grade graphites now available for use in high temperature gas reactors. The goals of the irradiation experiments are to obtain irradiation performance data, including irradiation creep, at different temperatures and loading conditions to support design of the Next Generation Nuclear Plant (NGNP) Very High Temperature Gas Reactor, as well as other future gas reactors. The experiments will each consist of a single capsule that will contain six peripheral stacks of graphite specimens, with half of the graphite specimens in each stack under a compressive load, while the other half of the specimens will not be subjected to a compressive load during irradiation. The six peripheral stacks will have different compressive loads applied to the top half of each pair of specimen stacks, while a seventh stack will not have a compressive load. The specimens will be irradiated in an inert sweep gas atmosphere with on-line temperature and compressive load monitoring and control. There will also be sampling the sweep gas effluent to determine if any oxidation or off-gassing of the specimens occurs during irradiation of the experiment. The first experiment, AGC-1, started its irradiation in September 2009, and the irradiation was completed in January 2011. The second experiment, AGC-2, started its irradiation in April 2011 and completed its irradiation in May 2012. This paper will briefly discuss the design of the experiment and control systems, and then present the irradiation results for each experiment to date.

Blaine Grover

2012-10-01T23:59:59.000Z

398

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)

399

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

SciTech Connect

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

J. K. Wright; R. N. Wright

2008-04-01T23:59:59.000Z

400

Maximum Fuel Utilization in Advanced Fast Reactors without Actinides Separation  

E-Print Network (OSTI)

According to the DOE handbook [57], a shutdown margin largerp. 812. Not, A. , DOE fundamentals handbook: Nuclear physics

Heidet, Florent

2010-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "advanced nuclear reactor" 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

Nuclear bremsstrahlung and its radiation effects in fusion reactors  

Science Journals Connector (OSTI)

Nucleons, i.e. protons or neutrons, are emitted in many fusion processes of light nuclei. The fusion-generated nucleons can in turn fuse with or be captured by an un-reacted nuclear fuel, for example deuterium. The fusion reaction occurs at an average center of mass (COM) energy of 10?keV (thousand electron volts) or more in a typical fusion reactor. At such a relative low COM energy, the proton and deuteron are in a state where the relative angular momentum approaches zero, or an s-wave state. The single gamma radiation process is thus suppressed due to the conservation of parity. Instead, the gamma ray (typically of the order of a few MeV or more) released is likely to be accompanied by soft x-ray photons from a nuclear bremsstrahlung process, which generates continuous x-ray radiation peaked around a few hundred eV to a few keV. The average photon energy and spectrum properties of such a process is calculated with a semiclassical approach. This phenomenon may cause additional parasitic issues to a fusion reactor, while also opening up the possibility of new plasma diagnostics.

Nie Luo; Magdi Ragheb; George H Miley

2009-01-01T23:59:59.000Z

402

Novel Fabrication of SiC Based Ceramics for Nuclear Applications.  

E-Print Network (OSTI)

??Advances in nuclear reactor technology and the use of gas-cooled fast reactors require the development of new materials that can operate at the higher temperatures… (more)

Singh, Abhishek Kumar

2009-01-01T23:59:59.000Z

403

Nuclear Design of the HOMER-15 Mars Surface Fission Reactor  

SciTech Connect

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

404

MIT - Center for Advanced Nuclear Energy Systems | Open Energy Information  

Open Energy Info (EERE)

MIT - Center for Advanced Nuclear Energy Systems MIT - Center for Advanced Nuclear Energy Systems Jump to: navigation, search Logo: MIT - Center for Advanced Nuclear Energy Systems Name MIT - Center for Advanced Nuclear Energy Systems Address 77 Massachusetts Avenue, 24-215 Place Cambridge, Massachusetts Zip 02139-4307 Phone number (617) 452-2660 Coordinates 42.3613041°, -71.0967653° Loading map... {"minzoom":false,"mappingservice":"googlemaps3","type":"ROADMAP","zoom":14,"types":["ROADMAP","SATELLITE","HYBRID","TERRAIN"],"geoservice":"google","maxzoom":false,"width":"600px","height":"350px","centre":false,"title":"","label":"","icon":"","visitedicon":"","lines":[],"polygons":[],"circles":[],"rectangles":[],"copycoords":false,"static":false,"wmsoverlay":"","layers":[],"controls":["pan","zoom","type","scale","streetview"],"zoomstyle":"DEFAULT","typestyle":"DEFAULT","autoinfowindows":false,"kml":[],"gkml":[],"fusiontables":[],"resizable":false,"tilt":0,"kmlrezoom":false,"poi":true,"imageoverlays":[],"markercluster":false,"searchmarkers":"","locations":[{"text":"","title":"","link":null,"lat":42.3613041,"lon":-71.0967653,"alt":0,"address":"","icon":"","group":"","inlineLabel":"","visitedicon":""}]}

405

Education in nuclear science at IPEN - CNEN, São Paulo, Brazil: Advanced School of Nuclear Energy—EAEN  

Science Journals Connector (OSTI)

EAEN (Advanced School of Nuclear Energy, 2010) is an annual school that ... a week of activities in the area of Nuclear Physics, Radiochemistry and uses of Nuclear Energy for a public made of high school students...

R. Semmler; M. G. M. Catharino…

2012-01-01T23:59:59.000Z

406

E-Print Network 3.0 - advanced nuclear research Sample Search...  

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

Powered by Explorit Topic List Advanced Search Sample search results for: advanced nuclear research Page: << < 1 2 3 4 5 > >> 1 Los Alamos National Laboratory DOE NNSA...

407

E-Print Network 3.0 - advancing nuclear security Sample Search...  

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

Powered by Explorit Topic List Advanced Search Sample search results for: advancing nuclear security Page: << < 1 2 3 4 5 > >> 1 Los Alamos National Laboratory DOE NNSA...

408

Review of the proposed materials of construction for the SBWR and AP600 advanced reactors  

SciTech Connect

Two advanced light water reactor (LWR) concepts, namely the General Electric Simplified Boiling Water Reactor (SBWR) and the Westinghouse Advanced Passive 600 MWe Reactor (AP600), were reviewed in detail by Argonne National Laboratory. The objectives of these reviews were to (a) evaluate proposed advanced-reactor designs and the materials of construction for the safety systems, (b) identify all aging and environmentally related degradation mechanisms for the materials of construction, and (c) evaluate from the safety viewpoint the suitability of the proposed materials for the design application. Safety-related systems selected for review for these two LWRs included (a) reactor pressure vessel, (b) control rod drive system and reactor internals, (c) coolant pressure boundary, (d) engineered safety systems, (e) steam generators (AP600 only), (f) turbines, and (g) fuel storage and handling system. In addition, the use of cobalt-based alloys in these plants was reviewed. The selected materials for both reactors were generally sound, and no major selection errors were found. It was apparent that considerable thought had been given to the materials selection process, making use of lessons learned from previous LWR experience. The review resulted in the suggestion of alternate an possibly better materials choices in a number of cases, and several potential problem areas have been cited.

Diercks, D.R.; Shack, W.J.; Chung, H.M.; Kassner, T.F. [Argonne National Lab., IL (United States)

1994-06-01T23:59:59.000Z

409

Business Case for Early Orders of New Nuclear Reactors Executive Overview  

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

Early Orders of New Nuclear Reactors Early Orders of New Nuclear Reactors Executive Overview Page EX - 1 Business Case for New Business Case for New Nuclear Power Plants Nuclear Power Plants Bringing Public and Private Resources Together for Nuclear Energy Mitigating Critical Risks on Early Orders for New Reactors Briefing for NERAC October 1, 2002 Disclaimer: This draft report was prepared to help the Department of Energy determine the barriers related to the deployment of new nuclear power plants but does not necessarily represent the views or policy of the Department. Business Case for Early Orders of New Nuclear Reactors Executive Overview Page EX - 2 Integrated Project Team Process * Integrated project team (IPT) approach facilitated consideration of complex issues involved in the project and to ensure contractor access to important data from NE.

410

EA-0912: Urgent-Relief Acceptance of Foreign Research Reactor Spent Nuclear  

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

2: Urgent-Relief Acceptance of Foreign Research Reactor Spent 2: Urgent-Relief Acceptance of Foreign Research Reactor Spent Nuclear Fuel EA-0912: Urgent-Relief Acceptance of Foreign Research Reactor Spent Nuclear Fuel SUMMARY This EA evaluates the environmental impacts of a proposal to accept 409 spent fuel elements from eight foreign research reactors in seven European countries. The spent fuel would be shipped across the ocean in spent fuel transportation casks from the country of origin to one or more United States eastern seaboard ports. PUBLIC COMMENT OPPORTUNITIES None available at this time. DOCUMENTS AVAILABLE FOR DOWNLOAD April 22, 1994 EA-0912: Finding of No Significant Impact Urgent-Relief Acceptance of Foreign Research Reactor Spent Nuclear Fuel April 22, 1994 EA-0912: Final Environmental Assessment Urgent-Relief Acceptance of Foreign Research Reactor Spent Nuclear Fuel

411

Maximum Fuel Utilization in Advanced Fast Reactors without Actinides Separation  

E-Print Network (OSTI)

Oxford ; New York ; Oxford University Press. Fuel- Trac,Spent Fuel / Reprocessing, in Nuclear Industry Statusto Burn Non-Fissile Fuels. 2008. GA. Energy Multiplier

Heidet, Florent

2010-01-01T23:59:59.000Z

412

Consortium for Advanced Simulation of Light Water Reactors (CASL...  

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

on specific industry issues, CASL engages an Industry Council chaired by the Electric Power Research Institute. CASL is committed to engaging the nuclear power industry to...

413

Consortium for Advanced Simulation of Light Water Reactors (CASL...  

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

in nuclear energy R&D, as well as a record of accomplishment in leading large-scale scientific collaborations. The participation of Idaho National Laboratory (INL), Los...

414

Shielding design aspects of thermionic space nuclear reactors  

SciTech Connect

It has been well documented that nuclear power sources will be required for the future exploration of space. Higher power levels (>10 kW (electric)) will be enabling, if not absolutely necessary, for the continued expansion of a human presence in the solar system and beyond. Space missions that will directly benefit continued life on Earth, including the monitoring for climate change and global warming, high-capacity communication satellites, and large, space-based radar systems to monitor the flow of airline traffic, will require progressively larger amounts of electrical power. Military applications, even with the ending of the Cold War, will continue to be needed for treaty verification activities. A thermionic energy conversion-based nuclear reactor system is one of the many different technologies proposed for the utilization of nuclear energy in space. How the energy conversion is accomplished and the equipment requiring shielding have a profound effect on the overall shielding requirements for the system. There exist two configurations of this technology that can be exploited and will have a significant effect on shielding needs. The paper discusses in-core thermionic conversion and out-of-core conversion concepts.

Klein, A.C.

1991-01-01T23:59:59.000Z

415

Incorporation of a risk analysis approach for the nuclear fuel cycle advanced transparency framework.  

SciTech Connect

Proliferation resistance features that reduce the likelihood of diversion of nuclear materials from the civilian nuclear power fuel cycle are critical for a global nuclear future. A framework that monitors process information continuously can demonstrate the ability to resist proliferation by measuring and reducing diversion risk, thus ensuring the legitimate use of the nuclear fuel cycle. The automation of new nuclear facilities requiring minimal manual operation makes this possible by generating instantaneous system state data that can be used to track and measure the status of the process and material at any given time. Sandia National Laboratories (SNL) and the Japan Atomic Energy Agency (JAEA) are working in cooperation to develop an advanced transparency framework capable of assessing diversion risk in support of overall plant transparency. The ''diversion risk'' quantifies the probability and consequence of a host nation diverting nuclear materials from a civilian fuel cycle facility. This document introduces the details of the diversion risk quantification approach to be demonstrated in the fuel handling training model of the MONJU Fast Reactor.

Mendez, Carmen Margarita (Sociotecnia Solutions, LLC); York, David L.; Inoue, Naoko (Japan Atomic Energy Agency); Kitabata, Takuya (Japan Atomic Energy Agency); Vugrin, Eric D.; Vugrin, Kay White; Rochau, Gary Eugene; Cleary, Virginia D.

2007-05-01T23:59:59.000Z

416

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

E-Print Network (OSTI)

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

V. V. Sinev

2009-02-22T23:59:59.000Z

417

Advanced Resin Cleaning System (ARCS) at Grand Gulf Nuclear Station  

SciTech Connect

Steam generation system in-core components can undergo serious material degradation by a variety of corrosion-related phenomena. These phenomena are largely controlled by boiler water (i.e. reactor water) chemistry which is strongly impacted by the performance of the condensate system mixed bed ion exchange units. In Boiling Water Reactors (BWR), the mixed bed ion exchange units not only provide protection from ionic contaminants, but also remove insoluble corrosion products by filtration/adsorption. These insoluble corrosion products removed by the ion exchange units must then be periodically cleaned from the resin bed by some process external to the BWR primary water loop. A unique resin cleaning process called the {open_quotes}Advanced Resin Cleaning System{close_quotes} (ARCS) was developed in the late 1980`s by members of CENTEC-XXI, located in Santa Clara, CA. This system, which has been successfully operated for several years at a Pressurized Water Reactor is highly efficient for removal of both insoluble corrosion products and anion/cation resin fines, and generates significantly less waste water than other cleaning methods. The ARCS was considered the most attractive method for meeting the demanding and costly resin cleaning needs of a BWR. A {open_quotes}Tailored Collaboration{close_quotes} project was initiated between EPRI, Entergy Operations (Grand Gulf Station), and CENTEC-XXI to demonstrate the {open_quotes}Advanced Resin Cleaning System{close_quotes} in a BWR.

Asay, R.H.; Earls, J.E.; Naughton, M.D. [Centec 21, Inc., Santa Clara, CA (United States)

1996-10-01T23:59:59.000Z

418

E-Print Network 3.0 - advanced light-water nuclear Sample Search...  

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

nuclear Search Powered by Explorit Topic List Advanced Search Sample search results for: advanced light-water nuclear Page: << < 1 2 3 4 5 > >> 1 1 Managed by UT-Battelle for the...

419

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

E-Print Network (OSTI)

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

Dittmar, Michael

2009-01-01T23:59:59.000Z

420

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

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

Chu Visits Site of America's First New Nuclear Reactor in Three 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 WASHINGTON, D.C. - Just two days after the Department of Energy requested more than $770 million for nuclear energy in 2013, U.S. Secretary of Energy Steven Chu visited the Vogtle nuclear power plant in Waynesboro, Georgia and Oak Ridge National Laboratory to highlight the steps the Obama Administration is taking to restart America's nuclear industry as part of an all-of-the-above American energy strategy. During remarks to more than 500 workers at the Vogtle nuclear power plant, Secretary Chu highlighted the wide variety of steps the Obama Administration has taken to help restart America's nuclear energy

Note: This page contains sample records for the topic "advanced nuclear reactor" 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

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

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

Chu Visits Site of America's First New Nuclear Reactor in Three 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 Just two days after the Department of Energy requested more than $770 million for nuclear energy in 2013, U.S. Secretary of Energy Steven Chu visited the Vogtle nuclear power plant in Waynesboro, Georgia and Oak Ridge National Laboratory to highlight the steps the Obama Administration is taking to restart America's nuclear industry as part of an all-of-the-above American energy strategy. During remarks to more than 500 workers at the Vogtle nuclear power plant, Secretary Chu highlighted the wide variety of steps the Obama Administration has taken to help restart America's nuclear energy

422

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

DOE Patents (OSTI)

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

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

1980-06-06T23:59:59.000Z

423

Self-actuating and locking control for nuclear reactor  

DOE Patents (OSTI)

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

Chung, Dong K. (Chatsworth, CA)

1982-01-01T23:59:59.000Z

424

Hybrid reliability model for nuclear reactor safety system  

Science Journals Connector (OSTI)

The dependability of critical safety systems needs to be quantitatively determined in order to verify their effectiveness, e.g. with regard to regulatory requirements. Since modular redundant safety systems are not required for normal operation, their reliability is strongly dependent on periodic inspection. Several modeling methods for the quantitative assessment of dependability are described in the literature, with a broad variation in complexity and modeling power. Static modeling techniques such as fault tree analysis (FTA) or reliability block diagrams (RBD) are not capable of capturing redundancy and repair or test activities. Dynamic state space based models such as continuous time Markov chains (CTMC) are more powerful but often result in very large, intractable models. Moreover, exponentially distributed state residence times are not a correct representation of actual residence times associated with repair activities or periodic inspection. In this study, a hybrid model combines a system level RBD with a CTMC to describe the dynamics. The effects of periodic testing are modeled by redistributing state probabilities at deterministic test times. Applying the method to the primary safety shutdown system of the BR2(Belgian Reactor 2)—nuclear research reactor, resulted in a quantitative as well as a qualitative assessment of its reliability.

Steven Verlinden; Geert Deconinck; Bernard Coupé

2012-01-01T23:59:59.000Z

425

Piping benchmark problems for the General Electric Advanced Boiling Water Reactor  

SciTech Connect

To satisfy the need for verification of the computer programs and modeling techniques that will be used to perform the final piping analyses for an advanced boiling water reactor standard design, three benchmark problems were developed. The problems are representative piping systems subjected to representative dynamic loads with solutions developed using the methods being proposed for analysis for the advanced reactor standard design. It will be required that the combined license holders demonstrate that their solutions to these problems are in agreement with the benchmark problem set.

Bezler, P.; DeGrassi, G.; Braverman, J.; Wang, Y.K. [Brookhaven National Lab., Upton, NY (US)

1993-08-01T23:59:59.000Z

426

E-Print Network 3.0 - advanced reactors advanced Sample Search...  

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

emit to the environment significant quantities of heat, contributing... to the greenhouse effect. New Generation reactors to achieve the reuse of spent fuel. Fusion...

427

Advanced Computational Thermal Fluid Physics (CTFP) and Its Assessment for Light Water Reactors and Supercritical Reactors  

SciTech Connect

Background: The ultimate goal of the study is the improvement of predictive methods for safety analyses and design of Generation IV reactor systems such as supercritical water reactors (SCWR) for higher efficiency, improved performance and operation, design simplification, enhanced safety and reduced waste and cost. The objective of this Korean / US / laboratory / university collaboration of coupled fundamental computational and experimental studies is to develop the supporting knowledge needed for improved predictive techniques for use in the technology development of Generation IV reactor concepts and their passive safety systems. The present study emphasizes SCWR concepts in the Generation IV program.

D.M. McEligot; K. G. Condie; G. E. McCreery; H. M. McIlroy; R. J. Pink; L.E. Hochreiter; J.D. Jackson; R.H. Pletcher; B.L. Smith; P. Vukoslavcevic; J.M. Wallace; J.Y. Yoo; J.S. Lee; S.T. Ro; S.O. Park

2005-10-01T23:59:59.000Z

428

Status of advanced fuel candidates for Sodium Fast Reactor within the Generation IV International Forum  

SciTech Connect

The main challenge for fuels for future Sodium Fast Reactor systems is the development and qualification of a nuclear fuel sub-assembly which meets the Generation IV International Forum goals. The Advanced Fuel project investigates high burn-up minor actinide bearing fuels as well as claddings and wrappers to withstand high neutron doses and temperatures. The R&D outcome of national and collaborative programs has been collected and shared between the AF project members in order to review the capability of sub-assembly material and fuel candidates, to identify the issues and select the viable options. Based on historical experience and knowledge, both oxide and metal fuels emerge as primary options to meet the performance and the reliability goals of Generation IV SFR systems. There is a significant positive experience on carbide fuels but major issues remain to be overcome: strong in-pile swelling, atmosphere required for fabrication as well as Pu and Am losses. The irradiation performance database for nitride fuels is limited with longer term R&D activities still required. The promising core material candidates are Ferritic/Martensitic (F/M) and Oxide Dispersed Strengthened (ODS) steels.

F. Delage; J. Carmack; C. B. Lee; T. Mizuno; M. Pelletier; J. Somers

2013-10-01T23:59:59.000Z

429

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

SciTech Connect

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

Shannon Bragg-Sitton

2014-02-01T23:59:59.000Z

430

Safety Design Strategy for the Advanced Test Reactor Primary Coolant Pump and Motor Replacement Project  

SciTech Connect

In accordance with the requirements of U.S. Department of Energy (DOE) Order 413.3B, “Program and Project Management for the Acquisition of Capital Assets,” safety must be integrated into the design process for new or major modifications to DOE Hazard Category 1, 2, and 3 nuclear facilities. The intended purpose of this requirement involves the handling of hazardous materials, both radiological and chemical, in a way that provides adequate protection to the public, workers, and the environment. Requirements provided in DOE Order 413.3B and DOE Order 420.1B, “Facility Safety,” and the expectations of DOE-STD-1189-2008, “Integration of Safety into the Design Process,” provide for identification of hazards early in the project and use of an integrated team approach to design safety into the facility. This safety design strategy provides the basic safety-in-design principles and concepts that will be used for the Advanced Test Reactor Reliability Sustainment Project. While this project does not introduce new hazards to the ATR, it has the potential for significant impacts to safety-related systems, structures, and components that are credited in the ATR safety basis and are being replaced. Thus the project has been determined to meet the definition of a major modification and is being managed accordingly.

Noel Duckwitz

2011-06-01T23:59:59.000Z

431

Safety Design Strategy for the Advanced Test Reactor Emergency Firewater Injection System Replacement Project  

SciTech Connect

In accordance with the requirements of U.S. Department of Energy (DOE) Order 413.3B, “Program and Project Management for the Acquisition of Capital Assets,” safety must be integrated into the design process for new or major modifications to DOE Hazard Category 1, 2, and 3 nuclear facilities. The intended purpose of this requirement involves the handling of hazardous materials, both radiological and chemical, in a way that provides adequate protection to the public, workers, and the environment. Requirements provided in DOE Order 413.3B and DOE Order 420.1B, “Facility Safety,” and the expectations of DOE-STD-1189-2008, “Integration of Safety into the Design Process,” provide for identification of hazards early in the project and use of an integrated team approach to design safety into the facility. This safety design strategy provides the basic safety-in-design principles and concepts that will be used for the Advanced Test Reactor Reliability Sustainment Project. While this project does not introduce new hazards to the ATR, it has the potential for significant impacts to safety-related systems, structures, and components that are credited in the ATR safety basis and are being replaced. Thus the project has been determined to meet the definition of a major modification and is being managed accordingly.

Noel Duckwitz

2011-06-01T23:59:59.000Z

432

Safety Design Strategy for the Advanced Test Reactor Diesel Bus (E-3) and Switchgear Replacement Project  

SciTech Connect

In accordance with the requirements of U.S. Department of Energy (DOE) Order 413.3B, “Program and Project Management for the Acquisition of Capital Assets,” safety must be integrated into the design process for new or major modifications to DOE Hazard Category 1, 2, and 3 nuclear facilities. The intended purpose of this requirement involves the handling of hazardous materials, both radiological and chemical, in a way that provides adequate protection to the public, workers, and the environment. Requirements provided in DOE Order 413.3B and DOE Order 420.1B, “Facility Safety,” and the expectations of DOE-STD-1189-2008, “Integration of Safety into the Design Process,” provide for identification of hazards early in the project and use of an integrated team approach to design safety into the facility. This safety design strategy provides the basic safety-in-design principles and concepts that will be used for the Advanced Test Reactor Reliability Sustainment Project. While this project does not introduce new hazards to the ATR, it has the potential for significant impacts to safety-related systems, structures, and components that are credited in the ATR safety basis and are being replaced. Thus the project has been determined to meet the definition of a major modification and is being managed accordingly.

Noel Duckwitz

2011-06-01T23:59:59.000Z

433

Single channel flow blockage accident phenomena identification and ranking table (PIRT) for the advanced Candu reactor  

SciTech Connect

The Advanced Candu Reactor (ACRTM) is an evolutionary advancement of the current Candu 6{sup R} reactor, aimed at producing electrical power for a capital cost and at a unit-energy cost significantly less than that of the current reactor designs. The ACR retains the modular concept of horizontal fuel channels surrounded by a heavy water moderator, as with all Candu reactors. However, ACR uses slightly enriched uranium (SEU) fuel, compared to the natural uranium used in Candu 6. This achieves the twin goals of improved economics (e.g., via reductions in the heavy water requirements and the use of a light water coolant), as well as improved safety. This paper documents the results of Phenomena Identification and Ranking Table (PIRT) results for a very limited frequency, beyond design basis event of the ACR design. This PIRT is developed in a highly structured process of expert elicitation that is well supported by experimental data and analytical results. The single-channel flow blockage event in an ACR reactor assumes a severe flow blockage of one of the reactor fuel channels, which leads to a reduction of the flow in the affected channel, leading to fuel cladding and fuel temperature increase. The paper outlines the design characteristics of the ACR reactor that impact the PIRT process and computer code applicability. It also describes the flow blockage phenomena, lists all components and systems that have an important role during the event, discusses the PIRT process and results, and presents the finalized PIRT tables. (authors)

Popov, N.K.; Abdul-Razzak, A.; Snell, V.G.; Langman, V. [Atomic Energy of Canada Ltd., 2251 Speakman Drive, Mississauga, Ontario, L5K 1B2 (Canada); Sills, H. [Consultant, Deep River, Ontario (Canada)

2004-07-01T23:59:59.000Z

434

Radiation Resistance of XLPE Nano-dielectrics for Advanced Reactor Applications  

SciTech Connect

Recently there has been renewed interest in nuclear reactor safety, particularly as commercial reactors are approaching 40 years service and lifetime extensions are considered, as well as for new reactor building projects around the world. The materials that are currently used in cabling for instrumentation, reactor control, and communications include cross-linked polyethylene (XLPE), ethylene propylene rubber (EPR), polyvinyl chloride (PVC), neoprene, and chlorosulfonated polyethylene. While these materials show suitable radiation tolerance in laboratory tests, failures before their useful lifetime occur due to the combined environmental effects of radiation, temperature and moisture, or operation under abnormal conditions. In addition, the extended use of commercial reactors beyond their original service life places a greater demand on insulating materials to perform beyond their current ratings in these nuclear environments. Nanocomposite materials that are based on XLPE and other epoxy resins incorporating TiO2, MgO, SiO2, and Al2O3 nanoparticles are being fabricated using a novel in-situ method established at ORNL to demonstrate materials with increased resistance to radiation. As novel nanocomposite dielectric materials are developed, characterization of the non-irradiated and irradiated nanodielectrics will lead to a knowledge base that allow for dielectric materials to be engineered with specific nanoparticle additions for maximum benefit to wide-variety of radiation environments found in nuclear reactors. This paper presents the initial findings on the development of XLPE-based SiO2 nano-composite dielectrics in the context of electrical performance and radiation degradation.

Duckworth, Robert C [ORNL; Polyzos, Georgios [ORNL; Paranthaman, Mariappan Parans [ORNL; Aytug, Tolga [ORNL; Leonard, Keith J [ORNL; Sauers, Isidor [ORNL

2014-01-01T23:59:59.000Z

435

Fusion–fission hybrids for nuclear waste transmutation: A synergistic step between Gen-IV fission and fusion reactors  

Science Journals Connector (OSTI)

Energy demand and GDP per capita are strongly correlated, while public concern over the role of energy in climate change is growing. Nuclear power plants produce 16% of world electricity demands without greenhouse gases. Generation-IV advanced nuclear energy systems are being designed to be safe and economical. Minimizing the handling and storage of nuclear waste is important. NIF and ITER are bringing sustainable fusion energy closer, but a significant gap in fusion technology development remains. Fusion–fission hybrids could be a synergistic step to a pure fusion economy and act as a technology bridge. We discuss how a pulsed power-driven Z-pinch hybrid system producing only 20 MW of fusion yield can drive a sub-critical transuranic blanket that transmutes 1280 kg of actinide wastes per year and produces 3000 MW. These results are applicable to other inertial and magnetic fusion energy systems. A hybrid system could be introduced somewhat sooner because of the modest fusion yield requirements and can provide both a safe alternative to fast reactors for nuclear waste transmutation and a maturation path for fusion technology. The development and demonstration of advanced materials that withstand high-temperature, high-irradiation environments is a fundamental technology issue that is common to both fusion–fission hybrids and Generation-IV reactors.

T.A. Mehlhorn; B.B. Cipiti; C.L. Olson; G.E. Rochau

2008-01-01T23:59:59.000Z

436

Fusion-fission hybrids for nuclear waste transmutation : a synergistic step between Gen-IV fission and fusion reactors.  

SciTech Connect

Energy demand and GDP per capita are strongly correlated, while public concern over the role of energy in climate change is growing. Nuclear power plants produce 16% of world electricity demands without greenhouse gases. Generation-IV advanced nuclear energy systems are being designed to be safe and economical. Minimizing the handling and storage of nuclear waste is important. NIF and ITER are bringing sustainable fusion energy closer, but a significant gap in fusion technology development remains. Fusion-fission hybrids could be a synergistic step to a pure fusion economy and act as a technology bridge. We discuss how a pulsed power-driven Z-pinch hybrid system producing only 20 MW of fusion yield can drive a sub-critical transuranic blanket that transmutes 1280 kg of actinide wastes per year and produces 3000 MW. These results are applicable to other inertial and magnetic fusion energy systems. A hybrid system could be introduced somewhat sooner because of the modest fusion yield requirements and can provide both a safe alternative to fast reactors for nuclear waste transmutation and a maturation path for fusion technology. The development and demonstration of advanced materials that withstand high-temperature, high-irradiation environments is a fundamental technology issue that is common to both fusion-fission hybrids and Generation-IV reactors.

Olson, Craig Lee; Mehlhorn, Thomas Alan; Cipiti, Benjamin B.; Rochau, Gary Eugene

2007-09-01T23:59:59.000Z

437

Development of Modeling Techniques for A Generation IV Gas Fast Reactor  

E-Print Network (OSTI)

Worldwide, multiple countries are investing a great deal of time and energy towards developing a new class of technologically advanced nuclear reactors. These new reactors have come to be known as the Generation IV (Gen IV) class of nuclear...

Dercher, Andrew Steven

2012-10-19T23:59:59.000Z

438

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

SciTech Connect

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

Curtis, D.B. [Los Alamos National Lab., NM (US)

1992-12-31T23:59:59.000Z

439

System aspects of a Space Nuclear Reactor Power System  

SciTech Connect

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

440

Advances in the Hopkinson bar testing of irradiated/non-irradiated nuclear materials and large specimens  

Science Journals Connector (OSTI)

...the twentieth century, research activity in the nuclear field of the Joint Research Centre (JRC) was concentrated on problems of nuclear reactor safety, especially...dynamic material testing programme for the high ductility...

2014-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "advanced nuclear reactor" 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

Graduate School of Advanced Science and Engineering Cooperative Major in Nuclear Energy  

E-Print Network (OSTI)

Graduate School of Advanced Science and Engineering Cooperative Major in Nuclear Energy Master in Nuclear Energy Summary of Research Instruction Research Instruction Application Code Name Major in Nuclear Energy Master's Program Doctoral Program Summary of Research Instruction

Kaji, Hajime

442

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

SciTech Connect

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

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

2008-01-01T23:59:59.000Z

443

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

National Nuclear Security Administration (NNSA)

Domestic U.S. Reactor Conversions: Fact Sheet | National Nuclear Security Domestic U.S. Reactor Conversions: Fact Sheet | 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 > Fact Sheets > Domestic U.S. Reactor Conversions: Fact Sheet Fact Sheet Domestic U.S. Reactor Conversions: Fact Sheet Mar 23, 2012 The National Nuclear Security Administration (NNSA) helps convert research

444

Statistical Methods Handbook for Advanced Gas Reactor Fuel Materials  

SciTech Connect

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

J. J. Einerson

2005-05-01T23:59:59.000Z

445

A Framework for Human Performance Criteria for Advanced Reactor Operational Concepts  

SciTech Connect

This report supports the determination of new Operational Concept models needed in support of the operational design of new reactors. The objective of this research is to establish the technical bases for human performance and human performance criteria frameworks, models, and guidance for operational concepts for advanced reactor designs. The report includes a discussion of operating principles for advanced reactors, the human performance issues and requirements for human performance based upon work domain analysis and current regulatory requirements, and a description of general human performance criteria. The major findings and key observations to date are that there is some operating experience that informs operational concepts for baseline designs for SFR and HGTRs, with the Experimental Breeder Reactor-II (EBR-II) as a best-case predecessor design. This report summarizes the theoretical and operational foundations for the development of a framework and model for human performance criteria that will influence the development of future Operational Concepts. The report also highlights issues associated with advanced reactor design and clarifies and codifies the identified aspects of technology and operating scenarios.

Jacques V Hugo; David I Gertman; Jeffrey C Joe

2014-08-01T23:59:59.000Z

446

THE RADIATION SAFETY INFORMATION COMPUTATIONAL CENTER: A RESOURCE FOR REACTOR DOSIMETRY SOFTWARE AND NUCLEAR DATA  

SciTech Connect

The Radiation Safety Information Computational Center (RSICC) was established in 1963 to collect and disseminate computational nuclear technology in the form of radiation transport, shielding and safety software and corresponding nuclear cross sections. Approximately 1700 nuclear software and data packages are in the RSICC collection, and the majority are applicable to reactor dosimetry.

Kirk, Bernadette Lugue [ORNL] [ORNL

2009-01-01T23:59:59.000Z

447

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

E-Print Network (OSTI)

. Popovic. K. W. Neale. S. Rizkalla. and G. Tadros Synopsis: The Gentilly· l nuclear power plant, in Quebec support. Keywords: concrete repair; fiber optic sensors; nuclear power plant; structural monitoring. 1. HISTORY 1.1 Decommissioning of the Reactor The Gentilly-I nuclear power plant, located

448

2013 Radiological Monitoring Results Associated with the Advanced Test Reactor Complex Cold Waste Pond  

SciTech Connect

This report summarizes radiological monitoring performed of the Idaho National Laboratory Site’s Advanced Test Reactor Complex Cold Waste wastewater prior to discharge into the Cold Waste Pond and of specific groundwater monitoring wells associated with the Industrial Wastewater Reuse Permit (#LA-000161-01, Modification B). All radiological monitoring is performed to fulfill Department of Energy requirements under the Atomic Energy Act.

Mike Lewis

2014-02-01T23:59:59.000Z

449

Incorporation of Hydride Nuclear Fuels in Commercial Light Water Reactors  

E-Print Network (OSTI)

of hydride fueled BWRs. Nuclear Engineering and Design, 239:Fueled PWR Cores. Nuclear Engineering and Design, 239:1489–Hydride Fueled LWRs. Nuclear Engineering and Design, 239:

Terrani, Kurt Amir

2010-01-01T23:59:59.000Z

450

Above-ground antineutrino detection for nuclear reactor monitoring  

Science Journals Connector (OSTI)

Abstract Antineutrino monitoring of nuclear reactors has been demonstrated many times (Klimov et al., 1994 [1]; Bowden et al., 2009 [2]; Oguri et al., 2014 [3]), however the technique has not as of yet been developed into a useful capability for treaty verification purposes. The most notable drawback is the current requirement that detectors be deployed underground, with at least several meters-water-equivalent of shielding from cosmic radiation. In addition, the deployment of liquid-based detection media presents a challenge in reactor facilities. We are currently developing a detector system that has the potential to operate above ground and circumvent deployment problems associated with a liquid detection media: the system is composed of segments of plastic scintillator surrounded by 6LiF/ZnS:Ag. ZnS:Ag is a radio-luminescent phosphor used to detect the neutron capture products of 6Li. Because of its long decay time compared to standard plastic scintillators, pulse-shape discrimination can be used to distinguish positron and neutron interactions resulting from the inverse beta decay (IBD) of antineutrinos within the detector volume, reducing both accidental and correlated backgrounds. Segmentation further reduces backgrounds by identifying the positron?s annihilation gammas, a signature that is absent for most correlated and uncorrelated backgrounds. This work explores different configurations in order to maximize the size of the detector segments without reducing the intrinsic neutron detection efficiency. We believe that this technology will ultimately be applicable to potential safeguards scenarios such as those recently described by Huber et al. (2014) [4,5].

M. Sweany; J. Brennan; B. Cabrera-Palmer; S. Kiff; D. Reyna; D. Throckmorton

2015-01-01T23:59:59.000Z

451

A Generalized Adjoint Framework for Sensitivity and Global Error Estimation in Time-Dependent Nuclear Reactor Simulations 1  

E-Print Network (OSTI)

-Dependent Nuclear Reactor Simulations 1 H. F. Striplinga, , M. Anitescub , M. L. Adamsa aNuclear Engineering Bateman and transport equations, which govern the time-dependent neutronic behavior of a nuclear reactor framework for computing the adjoint variable to nuclear engineering problems gov- erned by a set

Anitescu, Mihai

452

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

DOE Patents (OSTI)

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

Hampel, Viktor E. (Pleasanton, CA)

1989-01-01T23:59:59.000Z

453

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

DOE Patents (OSTI)

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

Hampel, V.E.

1988-05-17T23:59:59.000Z