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


1

Nuclear Reactor Safety Design Criteria  

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

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

2

Computer simulations help design new nuclear reactors  

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

3

Neutron transport analysis for nuclear reactor design  

DOE Patents [OSTI]

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

Vujic, Jasmina L. (Lisle, IL)

1993-01-01T23:59:59.000Z

4

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

5

Achievements: Nuclear Reactors designed/built by Argonne National  

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

6

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

7

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

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

8

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

9

Nuclear Design of the HOMER-15 Mars Surface Fission Reactor  

SciTech Connect (OSTI)

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

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

2002-07-01T23:59:59.000Z

10

A Methodology for the Neutronics Design of Space Nuclear Reactors  

SciTech Connect (OSTI)

A methodology for the neutronics design of space power reactors is presented. This methodology involves balancing the competing requirements of having sufficient excess reactivity for the desired lifetime, keeping the reactor subcritical at launch and during submersion accidents, and providing sufficient control over the lifetime of the reactor. These requirements are addressed by three reactivity values for a given reactor design: the excess reactivity at beginning of mission, the negative reactivity at shutdown, and the negative reactivity margin in submersion accidents. These reactivity values define the control worth and the safety worth in submersion accidents, used for evaluating the merit of a proposed reactor type and design. The Heat Pipe-Segmented Thermoelectric Module Converters space reactor core design is evaluated and modified based on the proposed methodology. The final reactor core design has sufficient excess reactivity for 10 years of nominal operation at 1.82 MW of fission power and is subcritical at launch and in all water submersion accidents.

King, Jeffrey C.; El-Genk, Mohamed S. [Institute for Space and Nuclear Power Studies, University of New Mexico, Albuquerque, NM 87131 (United States); Chemical and Nuclear Engineering Department, University of New Mexico, Albuquerque, NM 87131 (United States)

2004-02-04T23:59:59.000Z

11

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...

12

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

Broader source: Energy.gov [DOE]

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

13

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

14

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)

15

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

SciTech Connect (OSTI)

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

16

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

Science Journals Connector (OSTI)

...methodology and software interfaces of...often used in radiation hydrodynamics...reactor design and safety analyses. In...enable appropriate software interfaces...Ragusa. 2007 Software design of SHARP...nuclear reactor safety: multi-scale...methods for the radiation-diffusion equations...

2014-01-01T23:59:59.000Z

17

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

18

Shielding design aspects of thermionic space nuclear reactors  

SciTech Connect (OSTI)

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

19

SCW Pressure-Channel Nuclear Reactors: Some Design Features and Concepts  

SciTech Connect (OSTI)

Concepts of nuclear reactors cooled with water at supercritical pressures were studied as early as the 1950's and 1960's in the USA and Russia. After a 30-year break, the idea of developing nuclear reactors cooled with supercritical water (SCW) became attractive again as the ultimate development path for water-cooling. The main objectives of using SCW in nuclear reactors are 1) to increase the thermal efficiency of modern nuclear power plants (NPPs) from 33 -- 35% to about 40 -- 45%, and 2) to decrease capital and operational costs and hence decrease electrical energy costs ({approx}$ 1000 US/kW). SCW NPPs will have much higher operating parameters compared to modern NPPs (pressure about 25 MPa and outlet temperature up to 625 deg. C), and a simplified flow circuit, in which steam generators, steam dryers, steam separators, etc., can be eliminated. Also, higher SCW temperatures allow direct thermo-chemical production of hydrogen at low cost, due to increased reaction rates. Pressure-channel SCW nuclear reactor concepts are being developed in Canada and Russia. Design features related to both channels and fuel bundles are discussed in this paper. Also, Russian experience with operating supercritical steam heaters at NPP is presented. The main conclusion is that development of SCW pressure-channel nuclear reactors is feasible and significant benefits can be expected over other thermal energy systems. (authors)

Duffey, R.B.; Pioro, I.L. [Atomic Energy of Canada, Ltd. (Canada); Gabaraev, B.A.; Kuznetsov, Yu. N. [Research and Development Institute of Power Engineering, ul.M. Krasnoselskaya, 2/8 Moscow, Moscow 107140 (Russian Federation)

2006-07-01T23:59:59.000Z

20

Spring design for use in the core of a nuclear reactor  

DOE Patents [OSTI]

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

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

1993-01-01T23:59:59.000Z

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


21

Design of a nuclear reactor system for lunar base applications  

E-Print Network [OSTI]

disadvantages. U02 and Pu02 fuels both have extremely poor ther mal conductivities, about 4 W/m K at 500 C, which would normally limit the maximum linear power in the reactor core to unacceptably low levels. For tunately, the ver y high melting temperatur es... conversion, however, high reactor exit temperatures are both necessary and desirable. The efficiency of the power conversion cycle is directly related to the difference between the high and low temperatur es in the system. Since the heat rejection...

Griffith, Richard Odell

2012-06-07T23:59:59.000Z

22

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

23

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

24

Reactor & Nuclear Systems Publications | ORNL  

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

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

25

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

26

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

27

Light Water Reactors Technology Development - Nuclear Reactors  

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

28

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

29

A Comparison of the Performance of Compact Neutrino Detector Designs for Nuclear Reactor Safeguards and Monitoring  

E-Print Network [OSTI]

There has been an increasing interest in the monitoring of nuclear fuel for power reactors by detecting the anti-neutrinos produced during operation. Small liquid scintillator detectors have already demonstrated sensitivity to operational power levels, but more sensitive monitoring requires improvements in the efficiency and uniformity of these detectors. In this work, we use a montecarlo simulation to investigate the detector performance of four different detector configurations. Based on the analysis of neutron detection efficiency and positron energy response, we find that the optimal detector design will depend on the goals and restrictions of the specific installation or application. We have attempted to present the relevant information so that future detector development can proceed in a profitable direction.

McKeown, R W

2006-01-01T23:59:59.000Z

30

Design and optimization of the heat rejection system for a liquid cooled thermionic space nuclear reactor power system  

SciTech Connect (OSTI)

The heat transport subsystem for a liquid metal cooled thermionic space nuclear power system was modelled using algorithms developed in support of previous nuclear power system study programs, which date back to the SNAP-10A flight system. The model was used to define the optimum dimensions of the various components in the heat transport subsystem subjected to the constraints of minimizing mass and achieving a launchable package that did not require radiator deployment. The resulting design provides for the safe and reliable cooling of the nuclear reactor in a proven lightweight design.

Moriarty, M.P. (Rocketdyne Division, Rockwell International Corporation, 6633 Canoga Avenue, P.O. Box 7922, Canoga Park, California 91309-7922 (United States))

1993-01-15T23:59:59.000Z

31

Advanced Nuclear Research Reactor  

SciTech Connect (OSTI)

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

32

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

SciTech Connect (OSTI)

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

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

2012-07-01T23:59:59.000Z

33

6 Nuclear Fuel Designs  

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

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

34

Conceptual design of the bimodal nuclear power system based on the ‘‘Romashka’’ type reactor with thermionic energy conversion system  

Science Journals Connector (OSTI)

The paper presents conceptual design of the bimodal space nuclear power system (NPS) based on the high?temperature reactor of ROMASHKA type with thermoninic energy conversion system. At the heart of the design is an employment of close?spaced thermionic diodes operating in a quasi?vacuum mode. The paper gives preliminary estimates of the NPS neutron?physical electric thermophysical and mass?dimensional parameters for the reactor electric power of 25 kW and propulsive thrust of about 80 N. Discussed are peculiarities of the combined mode wherein electric power is generated along with propulsive thrust. The paper contains results of the design studies performed by the Small Business ‘‘NP Energotech’’ under the Agreement with Rockwell International/Rocketdyne Division and according to the Rocketdyne Division provided Design Requirements. Involved in the work was the team of specialists of RRC ‘‘Kurchatov Institute’’ ‘‘Red Star’’ State Enterprise and Research Institute of SPA ‘‘Luch’’

Nikolai N. Ponmarev?Stepnoi; Veniamin A. Usov; Yuri V. Nikolaev; Stanislav A. Yeriemin; Yevgeny Ye. Zhabotinski; Anatoly Ya. Galkin; Yevgeny D. Avdoshyn

1995-01-01T23:59:59.000Z

35

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

DOE Patents [OSTI]

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

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

1999-03-23T23:59:59.000Z

36

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

DOE Patents [OSTI]

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

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

1999-01-01T23:59:59.000Z

37

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

38

13 - Generation IV reactor designs, operation and fuel cycle  

Science Journals Connector (OSTI)

Abstract: This chapter looks at Generation IV nuclear reactors, such as the very high-temperature reactor (VHTR), the supercritical water reactor (SCWR), the molten salt reactor (MSR), the sodium-cooled fast reactor (SFR), the lead-cooled fast reactor (LFR) and the gas-cooled fast reactor (GFR). Reactor designs and fuel cycles are also described.

N. Cerullo; G. Lomonaco

2012-01-01T23:59:59.000Z

39

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

40

Implementation of the SAM-CE Monte Carlo benchmark analysis capability for validating nuclear data and reactor design codes  

SciTech Connect (OSTI)

The National Nuclear Data Center is continuing its program to improve the nuclear data base used as input for commercial reactor analysis and design. In the most recent phase of this project the Monte Carlo program SAM-CE, developed by the Mathematical Applications Group, Inc. (MAGI), was made operational at BNL. This program was implemented on the BNL-CDC-7600 Computer, and also on the PDP-10 in-house computer. The NNDC made operational and developed techniques for processing ENDF/B-V cross sections for SAM-CE. A limited ENDF/B-V based library was produced. Use of the SAM-CE program in thermal reactor problems was validated using detailed comparisons of results with other Monte Carlo codes such as RECAP, RCP01 and VIM as well as with experimental data.

Beer, M.; Rose, P.

1981-04-01T23:59:59.000Z

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


41

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

42

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

Science Journals Connector (OSTI)

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

Samer D. Kahook; Edward T. Dugan

1991-01-01T23:59:59.000Z

43

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

SciTech Connect (OSTI)

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

Michael G. McKellar; Edwin A. Harvego

2010-05-01T23:59:59.000Z

44

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

45

The Argonaut Reactor - Reactors designed/built by Argonne National  

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

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

46

Nuclear divisional reactor  

SciTech Connect (OSTI)

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

47

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

48

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

49

Nuclear Reactor Severe Accident Experiments  

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

50

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

51

Heavy Liquid Metal Reactor Development - Nuclear Engineering Division  

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

52

Thermionic Reactor Design Studies  

SciTech Connect (OSTI)

Paper presented at the 29th IECEC in Monterey, CA in August 1994. The present paper describes some of the author's conceptual designs and their rationale, and the special analytical techniques developed to analyze their (thermionic reactor) performance. The basic designs, first published in 1963, are based on single-cell converters, either double-ended diodes extending over the full height of the reactor core or single-ended diodes extending over half the core height. In that respect they are similar to the thermionic fuel elements employed in the Topaz-2 reactor subsequently developed in the Soviet Union, copies of which were recently imported by the U.S. As in the Topaz-2 case, electrically heated steady-state performance tests of the converters are possible before fueling.

Schock, Alfred

1994-08-01T23:59:59.000Z

53

Nuclear Reactors and Technology; (USA)  

SciTech Connect (OSTI)

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

54

Design change management in regulation of nuclear fleets: World nuclear association's working groups on Cooperation in Reactor Design Evaluation and Licensing (CORDEL)  

SciTech Connect (OSTI)

The 60 year life of a reactor means that a plant will undergo change during its life. To ensure continuing safety, changes must be made with a full understanding of the design intent. With this aim, regulators require that each operating organisation should have a formally designated entity responsible for complete design knowledge in regard to plant safety. INSAG-19 calls such an entity 'Design Authority'. This requirement is difficult to achieve, especially as the number of countries and utilities operating plants increases. Some of these operating organisations will be new, and some will be small. For Gen III plants sold on a turnkey basis, it is even more challenging for the operating company to develop and retain the full knowledge needed for this role. CORDEL's Task Force entitled 'Design Change Management' is investigating options for effective design change management with the aim to support design standardization throughout a fleet's lifetime by means of enhanced international cooperation within industry and regulators. This paper starts with considering the causes of design change and identifies reasons for the increased beneficial involvement of the plant's original vendor in the design change process. A key central theme running through the paper is the definition of responsibilities for design change. Various existing mechanisms of vendor-operator interfaces over design change and how they are managed in different organisational and regulatory environments around the world are considered, with the functionality of Owners Groups and Design Authority being central. The roles played in the design change process by vendors, utilities, regulators, owners' groups and other organisations such as WANO are considered The aerospace industry approach to Design Authority has been assessed to consider what lessons might be learned. (authors)

Swinburn, R. [CORDEL DCM Task Force, Rolls-Royce Plc (United Kingdom); Borysova, I. [CORDEL, WNA, 22a St.James Sq., London SW1Y 4JH (United Kingdom); Waddington, J. [CORDEL Group (United Kingdom); Head, J. G. [CORDEL Group, GE-Hitachi Nuclear Energy (United Kingdom); Raidis, Z. [CORDEL Group, Candu Energy (United Kingdom)

2012-07-01T23:59:59.000Z

55

Reactor Physics and Fuel Cycle Analysis - Nuclear Engineering Division  

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

56

Design of a Receding Horizon Control System for Nuclear Reactor Power Distribution  

SciTech Connect (OSTI)

A receding horizon control method is applied to the axial power distribution control in a pressurized water reactor. The basic concept of receding horizon control is to solve on-line, at each sampling instant, an optimization problem for a finite future and to implement the first optimal control input as the current control input. Thus, it is a suitable control strategy for time-varying systems. The reactor model used for computer simulations is a two-point xenon oscillation model based on the nonlinear xenon and iodine balance equations and a one-group, one-dimensional, neutron diffusion equation with nonlinear power reactivity feedback that adequately describes axial oscillations and treats the nonlinearities explicitly. The reactor core is axially divided into two regions, and each region has one input and one output and is coupled with the other region. Through numerical simulations, it is shown that the proposed control algorithm exhibits very fast tracking responses due to the step and ramp changes of axial target shape and also works well in a time-varying parameter condition.

Na, Man Gyun [Chosun University (Korea, Republic of)

2001-07-15T23:59:59.000Z

57

Accuracy in temperature sensor response time estimation for new nuclear reactor designs  

Science Journals Connector (OSTI)

One method for measuring the response time of temperature sensors, the plunge test, verifies that the sensor has a suitable response time in the laboratory before installation. However, plunge test results cannot be extrapolated to the response time in an operating plant because response time is affected by multiple factors such as the ratio of internal heat-transfer resistance to the surface heat-transfer resistance, or Biot Modulus. The estimation method presented here can be used to extrapolate laboratory response-time measurements to determine sensor response time in another medium, in different test conditions, or in actual applications such as a nuclear power plant. However, experiments conducted at Oak Ridge National Laboratory have confirmed that the effect of temperature on sensor response time cannot be estimated confidently in an operating environment. Therefore, the loop current step response (LCSR) method was developed to measure the actual in-service response time of nuclear plant temperature sensors.

H.M. Hashemian; Jin Jiang

2011-01-01T23:59:59.000Z

58

Export possibilities for small nuclear reactors  

SciTech Connect (OSTI)

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

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

2007-07-01T23:59:59.000Z

59

Thermionic Reactor Design Studies  

SciTech Connect (OSTI)

During the 1960's and early 70's the author performed extensive design studies, analyses, and tests aimed at thermionic reactor concepts that differed significantly from those pursued by other investigators. Those studies, like most others under Atomic Energy Commission (AEC and DOE) and the National Aeronautics and Space Administration (NASA) sponsorship, were terminated in the early 1970's. Some of this work was previously published, but much of it was never made available in the open literature. U.S. interest in thermionic reactors resumed in the early 80's, and was greatly intensified by reports about Soviet ground and flight tests in the late 80's. This recent interest resulted in renewed U.S. thermionic reactor development programs, primarily under Department of Defense (DOD) and Department of Energy (DOE) sponsorship. Since most current investigators have not had an opportunity to study all of the author's previous work, a review of the highlights of that work may be of value to them. The present paper describes some of the author's conceptual designs and their rationale, and the special analytical techniques developed to analyze their performance. The basic designs, first published in 1963, are based on single-cell converters, either double-ended diodes extending over the full height of the reactor core or single-ended diodes extending over half the core height. In that respect they are similar to the thermionic fuel elements employed in the Topaz-2 reactor subsequently developed in the Soviet Union, copies of which were recently imported by the U.S. As in the Topaz-2 case, electrically heated steady-state performance tests of the converters are possible before fueling. Where the author's concepts differed from the later Topaz-2 design was in the relative location of the emitter and the collector. Placing the fueled emitter on the outside of the cylindrical diodes permits much higher axial conductances to reduce ohmic losses in the electrodes of full-core-height diodes. Moreover, placing the fuel on the outside of the diode makes possible reactors with much higher fuel volume fractions, which enable power-flattened fast reactors scalable to very low power levels without the need for life-limiting hydride moderators or the use of efficiency-limiting driver fuel. In addition, with the fuel on the outside its swelling does not increase the emitter diameter or reduce the interelectrode gap. This should permit long lifetimes even with closer spacings, which can significantly improve the system efficiences. This was confirmed by coupled neutronic, thermal, thermionic, and electrical system analyses - some of which are presented in this paper - and by subsequent experiments. A companion paper presented next describes the fabrication and testing of full-scale converter elements, both fueled and unfueled, and summarizes the test results obtained. There is a duplicate copy in the file.

Schock, Alfred

1994-06-01T23:59:59.000Z

60

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

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


61

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

62

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

63

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

64

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

65

Advanced Nuclear Reactors | Department of Energy  

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

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

66

Design and development of a special purpose SAFT system for nondestructive evaluation of nuclear reactor vessels and piping components  

SciTech Connect (OSTI)

This report describes the design details of a special purpose system for real-time nondestructive evaluation of reactor vessels and piping components. The system consists of several components and the report presents the results of the research aimed at the design of each component and recommendations based on the results. One major component of the NDE system, namely the real-time SAFT processor, was designed with sufficient details to enable the fabrications of a prototype by GARD Inc. under a subcontract from The University of Michigan and the report includes their results and conclusions.

Ganapathy, S.; Schmult, B.; Wu, W.S.; Dennehy, T.G.; Moayeri, N.; Kelly, P.

1985-08-01T23:59:59.000Z

67

Software: Reactor Physics and Fuel Cycle Analysis - Nuclear Engineering  

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

68

Nuclear Systems Modeling and Design Analysis - Nuclear Engineering Division  

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

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

69

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

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

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

70

Nuclear reactors and the nuclear fuel cycle  

SciTech Connect (OSTI)

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

71

Generic small modular reactor plant design.  

SciTech Connect (OSTI)

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

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

2012-12-01T23:59:59.000Z

72

Reactor and Nuclear Systems Division (RNSD)  

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

73

Solid0Core Heat-Pipe Nuclear Batterly Type Reactor  

SciTech Connect (OSTI)

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

Ehud Greenspan

2008-09-30T23:59:59.000Z

74

Simulation of a marine nuclear reactor  

SciTech Connect (OSTI)

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

75

Secretary Chu Statement on AP1000 Reactor Design Certification | Department  

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

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

76

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

77

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

78

Electric Power Produced from Nuclear Reactor | National Nuclear Security  

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

79

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

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

80

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

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


81

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

82

Direct conversion nuclear reactor space power systems  

SciTech Connect (OSTI)

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

83

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

84

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

85

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

86

Validation of nuclear design method by measured data obtained in the physics test at a small fast reactor  

SciTech Connect (OSTI)

The present paper discusses applicability of the measured data of Joyo cores from a view point of integral validation for the 4S nuclear design methodology. Through the evaluation of isothermal reactivity coefficients and reactivity losses due to burnup, the results confirm that those MK-I and MK-II database are effective in order to increase the dataset for uncertainty estimation for the prediction. Discussions on the 4S design method validation are also done through the analyses of criticality, power distributions and reactivity loss due to burn-up. The C/E values for criticality and reaction rate distributions are confirmed to be consistent with those obtained from the physics benchmark experiments. Through an analysis of burnup coefficient of the MK-I core by the detailed Monte Carlo calculations, the C/E value is 1.1, which is close to 1.06 obtained by the deterministic transport analysis. (authors)

Nagata, A.; Tsuboi, Y. [Advanced Energy Design and Engineering Dep., Toshiba Corporation, 8 Shinsugita-cho, Isogo-ku, Yokohama, 235-8523 (Japan); Moriki, Y. [Power and Industrial Systems Research and Development Center, Toshiba Corporation, 8 Shinsugita-cho, Isogo-ku, Yokohama, 235-8523 (Japan); Kawashima, M. [Nuclear Technology Application Dept., Toshiba Nuclear Engineering Services Corporation, 8 Shinsugita-cho, Isogo-ku, Yokohama, 235-8523 (Japan)

2012-07-01T23:59:59.000Z

87

Nuclear Reactor Technologies | Department of Energy  

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

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

88

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

89

Small Modular Nuclear Reactors | Department of Energy  

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

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

90

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

91

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

92

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

93

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

94

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

95

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

96

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...

97

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

98

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

99

Nuclear Facility Design  

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

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

100

Granular Dynamics in Pebble Bed Reactor Cores  

E-Print Network [OSTI]

pebble bed reactor,” Nuclear Engineering and Design, vol.the AVR reactor,” Nuclear Engineering and Design, vol. 121,Operating Experience,” Nuclear Engineering and Design, vol.

Laufer, Michael Robert

2013-01-01T23:59:59.000Z

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


101

Early Exploration - Reactors designed/built by Argonne National Laboratory  

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

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

102

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

103

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

104

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

105

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

106

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

107

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

108

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

109

Early Argonne reactor lit the way for worldwide nuclear industry -  

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

110

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

111

Sliding mode observer design for a PWR to estimate the xenon concentration & delayed neutrons precursor density based on the two point nuclear reactor model  

Science Journals Connector (OSTI)

Abstract One of the important operations in nuclear power plants is load-following in which imbalance of axial power distribution induces xenon oscillations. These oscillations must be maintained within acceptable limits otherwise the nuclear power plant could become unstable. Therefore, bounded xenon oscillation considered to be a constraint for the load-following operation. In other hands, precursors produce delayed neutrons which are most important in control of nuclear reactor, but xenon concentration & precursor density cannot be measured directly. In this paper, non-linear sliding mode observer which has the robust characteristics facing the parameters uncertainties and disturbances is proposed based on the two point nuclear reactor model to estimate the xenon concentration & delayed neutron precursor density of the Pressurized-Water Nuclear Reactor (PWR) using reactor power measurement. The stability analysis is given by means Lyapunov approach, thus the system is guaranteed to be stable within a large range. The employed method is easy to implement in practical applications. This estimation is done taking into account the effects of reactivity feedback due to temperature and xenon concentration. Simulation results clearly show that the sliding mode observer follows the actual system variables accurately and is satisfactory in the presence of the parameters uncertainties & disturbances.

G.R. Ansarifar; M.H. Esteki; M. Arghand

2015-01-01T23:59:59.000Z

112

Cost-Shared Development of Innovative Small Modular Reactor Designs |  

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

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

113

Cost-Shared Development of Innovative Small Modular Reactor Designs |  

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

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

114

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...

115

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

116

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

117

Generation III reactors safety requirements and the design solutions  

SciTech Connect (OSTI)

Nuclear energy's public acceptance, and hence its development, depends on its safety. As a reactor designer, we will first briefly remind the basic safety principles of nuclear reactors' design. We will then show how the industry, and in particular Areva with its EPR, made design evolution in the wake of the Three Miles Island accident in 1979. In particular, for this new generation of reactors, severe accidents are taken into account beyond the standard design basis accidents. Today, Areva's EPR meets all so-called 'generation III' safety requirements and was licensed by several nuclear safety authorities in the world. Many innovative solutions are integrated in the EPR, some of which will be introduced here.

Felten, P. [Areva NP (France)

2009-03-31T23:59:59.000Z

118

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

119

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

120

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

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


121

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

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

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

122

Advanced thermionic reactor systems design code  

SciTech Connect (OSTI)

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

123

Nuclear power reactor education and training at the Ford nuclear reactor  

SciTech Connect (OSTI)

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

124

Development of Improved Models and Designs for Coated-Particle Gas Reactor Fuels -- Final Report under the International Nuclear Energy Research Initiative (I-NERI)  

SciTech Connect (OSTI)

The objective of this INERI project was to develop improved fuel behavior models for gas reactor coated-particle fuels and to explore improved coated-particle fuel designs that could be used reliably at very high burnups and potentially in gas-cooled fast reactors. Project participants included the Idaho National Engineering Laboratory (INEEL), Centre Étude Atomique (CEA), and the Massachusetts Institute of Technology (MIT). To accomplish the project objectives, work was organized into five tasks.

David Petti; Philippe Martin; Mayeul Phélip; Ronald Ballinger; Petti does not have NT account

2004-12-01T23:59:59.000Z

125

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...

126

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...

127

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

128

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

129

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

130

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

131

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

132

Nuclear Data Measurements for 21st Century Reactor Physics Applications  

SciTech Connect (OSTI)

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

133

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

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

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

134

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

135

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

136

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...

137

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

138

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...

139

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

140

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


141

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

142

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

143

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

144

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

145

Industry Participation Sought for Design of Next Generation Nuclear Plant |  

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

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

146

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

147

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

148

Distributed computing and nuclear reactor analysis  

SciTech Connect (OSTI)

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

149

Ordered bed modular reactor design proposal  

SciTech Connect (OSTI)

The Ordered Bed Modular Reactor (OBMR) is a design as an advanced modular HTGR in which the annular reactor core is filled with an ordered bed of fuel spheres. This arrangement allows fuel elements to be poured into the core cavity which is shaped so that an ordered bed is formed and to be discharged from the core through the opening holes in the reactor top. These operations can be performed in a shutdown shorter time. The OBMR has the most of advantages from both the pebble bed reactor and block type reactor. Its core has great structural flexibility and stability, which allow increasing reactor output power and outlet gas temperature as well as decreasing core pressure drop. This paper introduces ordered packing bed characteristics, unloading and loading technique of the fuel spheres and predicted design features of the OBMR. (authors)

Tian, J. [Inst. of Nuclear Energy Technology, Tsinghua Univ., Beijing 100084 (China)

2006-07-01T23:59:59.000Z

150

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

151

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

152

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

153

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

154

Frequently Asked Questions Regarding DOE-STD-1195-2011, Design of Safety Significant Safety Instrumented Systems Used at DOE Non-Reactor Nuclear Facilities  

Broader source: Energy.gov [DOE]

Frequently Asked Questions Regarding DOE-STD-1195-2011 which provides requirements and guidance for the design, procurement, installation, testing, maintenance, operation, and quality assurance of safety instrumented systems (SIS) that may be used at Department of Energy (DOE) nonreactor nuclear facilities for safety significant (SS) functions.

155

Performance of Liquid Metals in Natural Circulation Cooled Nuclear Reactors  

SciTech Connect (OSTI)

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

156

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

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

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

157

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

158

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

159

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

160

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

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


161

Mirror Advanced Reactor Study interim design report  

SciTech Connect (OSTI)

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

Not Available

1983-04-01T23:59:59.000Z

162

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

163

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

164

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

165

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

SciTech Connect (OSTI)

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

166

Effect of Nuclear Data Libraries on Tritium Breeding in a (D–T) Fusion Driven Reactor  

Science Journals Connector (OSTI)

In design a Deuterium–Tritium (D–T) fusion driven hybrid reactor, neutronics and nuclear data libraries have an...2O, LiH, Li2TiO3, Li2ZrO3 and Li4SiO4 in a (D–T) driven fusion–fission (hybrid) reactor is investi...

Adem Ac?r

2008-12-01T23:59:59.000Z

167

Fast Reactor Subassembly Design Modifications for Increasing Electricity Generation Efficiency  

SciTech Connect (OSTI)

Suggested for Track 7: Advances in Reactor Core Design and In-Core Management _____________________________________________________________________________________ Fast Reactor Subassembly Design Modifications for Increasing Electricity Generation Efficiency R. Wigeland and K. Hamman Idaho National Laboratory Given the ability of fast reactors to effectively transmute the transuranic elements as are present in spent nuclear fuel, fast reactors are being considered as one element of future nuclear power systems to enable continued use and growth of nuclear power by limiting high-level waste generation. However, a key issue for fast reactors is higher electricity cost relative to other forms of nuclear energy generation. The economics of the fast reactor are affected by the amount of electric power that can be produced from a reactor, i.e., the thermal efficiency for electricity generation. The present study is examining the potential for fast reactor subassembly design changes to improve the thermal efficiency by increasing the average coolant outlet temperature without increasing peak temperatures within the subassembly, i.e., to make better use of current technology. Sodium-cooled fast reactors operate at temperatures far below the coolant boiling point, so that the maximum coolant outlet temperature is limited by the acceptable peak temperatures for the reactor fuel and cladding. Fast reactor fuel subassemblies have historically been constructed using a large number of small diameter fuel pins contained within a tube of hexagonal cross-section, or hexcan. Due to this design, there is a larger coolant flow area next to the hexcan wall as compared to flow area in the interior of the subassembly. This results in a higher flow rate near the hexcan wall, overcooling the fuel pins next to the wall, and a non-uniform coolant temperature distribution. It has been recognized for many years that this difference in sodium coolant temperature was detrimental to achieving greater thermal efficiency, since it causes the fuel pins in the center of the subassembly to operate at higher temperatures than those near the hexcan walls, and it is the temperature limit(s) for those fuel pins that limits the average coolant outlet temperature. Fuel subassembly design changes are being investigated using computational fluid dynamics (CFD) to quantify the effect that the design changes have on reducing the intra-subassembly coolant flow and temperature distribution. Simulations have been performed for a 19-pin test subassembly geometry using typical fuel pin diameters and wire wrap spacers. The results have shown that it may be possible to increase the average coolant outlet temperature by 20 C or more without changing the peak temperatures within the subassembly. These design changes should also be effective for reactor designs using subassemblies with larger numbers of fuel pins. R. Wigeland, Idaho National Laboratory, P.O. Box 1625, Mail Stop 3860, Idaho Falls, ID, U.S.A., 83415-3860 email – roald.wigeland@inl.gov fax (U.S.) – 208-526-2930

R. Wigeland; K. Hamman

2009-09-01T23:59:59.000Z

168

Design tool needs for space nuclear propulsion systems  

SciTech Connect (OSTI)

The interest in a return trip for humans to the moon and a pioneering voyage to Mars has rekindled interest in the use of nuclear reactors to provide propulsion for both piloted and robotic space vehicles. Two types of nuclear reactor-based propulsion systems are currently envisioned: nuclear thermal propulsion (NTP) and nuclear electric propulsion (NEP). The former relies on the direct heating and exhaust of a propellant within the core of the reactor, while the latter utilizes ion thruster engines for propulsion, and the nuclear reactor supplies the large amount of electrical power required to drive the engines. Another direct contrast between the NTP and NEP concepts is the length of reactor operation. The NTP nuclear rocket core is required to produce large amounts of thermal power for relatively short bursts (on the order of minutes to hours), and the NEP reactor core operates for a much longer period of time (on the order of days to months) with a steady-state electrical power output. The design of these types of nuclear reactor systems requires the use of specific analysis tools, some of which already exist and others that need considerable development. The general areas in which design tools are needed in the development of systems for space nuclear propulsion include the following: (1) neutronics design - both steady-state and transient applications including thermal feedback effects; (2) thermal-hydraulics design - again, both steady-state and transient applications with coupling to and from the neutronics design codes; (3) materials analysis tools - due to the high temperatures and high stresses required for efficient propulsion operation, increased importance will be placed on understanding the material responses; and (4) systems analysis - these codes allow optimizaiton of the entire propulsion system.

Klein, A.C. (Oregon State Univ., Corvallis (United States)); Lewis, B.R. (Atom Analysis, Inc., Portland, OR (United States))

1992-01-01T23:59:59.000Z

169

Analysis of nuclear reactor instability phenomena  

SciTech Connect (OSTI)

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

170

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

171

Design options for a bunsen reactor.  

SciTech Connect (OSTI)

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

Moore, Robert Charles

2013-10-01T23:59:59.000Z

172

Light weight space power reactors for nuclear electric propulsion  

SciTech Connect (OSTI)

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

173

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

174

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

175

Nuclear Archeology for CANDU Power Reactors  

SciTech Connect (OSTI)

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

176

CFD Analysis of Core Bypass Flow and Crossflow in the Prismatic Very High Temperature Gas-cooled Nuclear Reactor  

E-Print Network [OSTI]

Very High Temperature Rector (VHTR) had been designated as one of those promising reactors for the Next Generation (IV) Nuclear Plant (NGNP). For a prismatic core VHTR, one of the most crucial design considerations is the bypass flow and crossflow...

Wang, Huhu 1985-

2012-12-13T23:59:59.000Z

177

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

178

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

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

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

179

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

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

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

180

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

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


181

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

182

Coupled IVPs to Investigate a Nuclear Reactor Poison Burn Up  

SciTech Connect (OSTI)

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

183

University Research Reactor Task Force to the Nuclear Energy Research  

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

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

184

Modeling and Simulation for Nuclear Reactors Hub | Department of Energy  

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

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

185

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

SciTech Connect (OSTI)

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

Buden, D.

1993-08-11T23:59:59.000Z

186

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

SciTech Connect (OSTI)

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

Nick A. Altic

2011-11-11T23:59:59.000Z

187

Statement on Defense Nuclear Nonproliferation and Naval Reactors Activities  

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

188

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

189

Nuclear reactors built, being built, or planned 1992  

SciTech Connect (OSTI)

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

Not Available

1993-07-01T23:59:59.000Z

190

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

191

Overview of Sandia National Laboratories pulse nuclear reactors  

SciTech Connect (OSTI)

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

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

1994-10-01T23:59:59.000Z

192

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

193

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

SciTech Connect (OSTI)

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

194

Energy Secretary to Visit Georgia Nuclear Reactor Site and Tennessee  

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

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

195

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

196

MOOSE simulating nuclear reactor CRUD buildup  

SciTech Connect (OSTI)

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

197

REACTOR DOSIMETRY STUDY OF THE RHODE ISLAND NUCLEAR SCIENCE CENTER.  

SciTech Connect (OSTI)

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

198

CONSTRUCTION OF WEB-ACCESSIBLE MATERIALS HANDBOOK FORGENERATION IV NUCLEAR REACTORS  

SciTech Connect (OSTI)

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

Ren, Weiju [ORNL

2005-01-01T23:59:59.000Z

199

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

200

Preliminary design studies on a nuclear seawater desalination system  

SciTech Connect (OSTI)

Seawater desalination is one of the most promising technologies to provide fresh water especially in the arid region. The most used technology in seawater desalination are thermal desalination (MSF and MED) and membrane desalination (RO). Some developments have been done in the area of coupling the desalination plant with a nuclear reactor to reduce the cost of energy required in thermal desalination. The coupling a nuclear reactor to a desalination plant can be done either by using the co-generation or by using dedicated heat from a nuclear system. The comparison of the co-generation nuclear reactor with desalination plant, dedicated nuclear heat system, and fossil fueled system will be discussed in this paper using economical assessment with IAEA DEEP software. A newly designed nuclear system dedicated for the seawater desalination will also be suggested by KAIST (Korea Advanced Inst. of Science and Technology) research team and described in detail within this paper. The suggested reactor system is using gas cooled type reactor and in this preliminary study the scope of design will be limited to comparison of two cases in different operating temperature ranges. (authors)

Wibisono, A. F.; Jung, Y. H.; Choi, J.; Kim, H. S.; Lee, J. I.; Jeong, Y. H.; No, H. C. [KAIST, Dept. of Nuclear and Quantum Engineering, Korea Advanced Inst. of Science and Technology, 291 Daehak-ro, Yuseong-gu, Daejeon, 305-701 (Korea, Republic of)

2012-07-01T23:59:59.000Z

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


201

System aspects of a Space Nuclear Reactor Power System  

SciTech Connect (OSTI)

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

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

1988-01-01T23:59:59.000Z

202

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

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

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

203

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

SciTech Connect (OSTI)

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

None

2013-09-25T23:59:59.000Z

204

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

205

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

SciTech Connect (OSTI)

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

206

Idaho Site Obtains Patent for Nuclear Reactor Sodium Cleanup Treatment |  

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

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.

207

Idaho Site Obtains Patent for Nuclear Reactor Sodium Cleanup Treatment |  

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

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.

208

NUCLEAR STUDIES  

Science Journals Connector (OSTI)

Japanese nuclear power plant crisis sparks examination of U.S. REACTORS ... Calls are particularly zeroing in on reactors similar in location and design to those in Japan. ...

JEFF JOHNSON

2011-04-04T23:59:59.000Z

209

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

210

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

211

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

212

Machine for monitoring nuclear reactor equipment at its storage station  

SciTech Connect (OSTI)

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

213

Reactor Design and Decommissioning - An Overview of International Activities in Post Fukushima Era1 - 12396  

SciTech Connect (OSTI)

Accidents at the Fukushima Dai-ichi reactors as a result of the devastating earthquake and tsunami of March 11, 2011 have not only dampened the nuclear renaissance but have also initiated a re-examination of the design and safety features for the existing and planned nuclear reactors. Even though failures of some of the key site features at Fukushima can be attributed to events that in the past would have been considered as beyond the design basis, the industry as well as the regulatory authorities are analyzing what features, especially passive features, should be designed into the new reactor designs to minimize the potential for catastrophic failures. It is also recognized that since the design of the Fukushima BWR reactors which were commissioned in 1971, many advanced safety features are now a part of the newer reactor designs. As the recovery efforts at the Fukushima site are still underway, decisions with respect to the dismantlement and decommissioning of the damaged reactors and structures have not yet been finalized. As it was with Three Mile Island, it could take several decades for dismantlement, decommissioning and clean up, and the project poses especially tough challenges. Near-term assessments have been issued by several organizations, including the IAEA, the USNRC and others. Results of such investigations will lead to additional improvements in system and site design measures including strengthening of the anti-tsunami defenses, more defense-in-depth features in reactor design, and better response planning and preparation involving reactor sites. The question also arises what would the effect be on the decommissioning scene worldwide, and what would the effect be on the new reactors when they are eventually retired and dismantled. This paper provides an overview of the US and international activities related to recovery and decommissioning including the decommissioning features in the reactor design process and examines these from a new perspective in the post Fukushima -accident era. Accidents at the Fukushima Daiichi reactors in the aftermath of the devastating earthquake and tsunami of March 11, 2011 have slowed down the nuclear renaissance world-wide and may have accelerated decommissioning either because some countries have decided to halt or reduce nuclear, or because the new safety requirements may reduce life-time extensions. Even in countries such as the UK and France that favor nuclear energy production existing nuclear sites are more likely to be chosen as sites for future NPPs. Even as the site recovery efforts continue at Fukushima and any decommissioning decisions are farther into the future, the accidents have focused attention on the reactor designs in general and specifically on the Fukushima type BWRs. The regulatory authorities in many countries have initiated a re-examination of the design of the systems, structures and components and considerations of the capability of the station to cope with beyond-design basis events. Enhancements to SSCs and site features for the existing reactors and the reactors that will be built will also impact the decommissioning phase activities. The newer reactor designs of today not only have enhanced safety features but also take into consideration the features that will facilitate future decommissioning. Lessons learned from past management and operation of reactors as well as the lessons from decommissioning are incorporated into the new designs. However, in the post-Fukushima era, the emphasis on beyond-design-basis capability may lead to significant changes in SSCs, which eventually will also have impact on the decommissioning phase. Additionally, where some countries decide to phase out the nuclear power, many reactors may enter the decommissioning phase in the coming decade. While the formal updating and expanding of existing guidance documents for accident cleanup and decommissioning would benefit by waiting until the Fukushima project has progressed sufficiently for that experience to be reliably interpreted, the development of structured on-li

Devgun, Jas S. [Nuclear Power Technologies, Sargent and Lundy LLC, Chicago, IL (United States); Laraia, Michele [private consultant, formerly from IAEA, Kolonitzgasse 10/2, 1030, Vienna (Austria); Pescatore, Claudio [OECD, Nuclear Energy Agency, Issy-les-Moulineaux, Paris (France); Dinner, Paul [International Atomic Energy Agency, Wagramerstrasse 5, A-1400 Vienna (Austria)

2012-07-01T23:59:59.000Z

214

Using reactor operating experience to improve the design of a new Broad Application Test Reactor  

SciTech Connect (OSTI)

Increasing regulatory demands and effects of plant aging are limiting the operation of existing test reactors. Additionally, these reactors have limited capacities and capabilities for supporting future testing missions. A multidisciplinary team of experts developed sets of preliminary safety requirements, facility user needs, and reactor design concepts for a new Broad Application Test Reactor (BATR). Anticipated missions for the new reactor include fuels and materials irradiation testing, isotope production, space testing, medical research, fusion testing, intense positron research, and transmutation doping. The early BATR design decisions have benefited from operating experiences with existing reactors. This paper discusses these experiences and highlights their significance for the design of a new BATR.

Fletcher, C.D.; Ryskamp, J.M.; Drexler, R.L.; Leyse, C.F.

1993-07-01T23:59:59.000Z

215

Nuclear Energy Enabling Technologies (NEET) Reactor Materials  

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

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

216

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

217

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

218

Spent nuclear fuel discharges from US reactors 1992  

SciTech Connect (OSTI)

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

219

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

SciTech Connect (OSTI)

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

220

Concept of nuclear reactor pumped laser for ICF  

SciTech Connect (OSTI)

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

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


221

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

SciTech Connect (OSTI)

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

222

Design of Complex Systems to Achieve Passive Safety: Natural Circulation Cooling of Liquid Salt Pebble Bed Reactors  

E-Print Network [OSTI]

a tool for reactor design optimization, and for design ofdesign tool for reactor design optimization, and for designdesign tool for reactor design optimization, and for design

Scarlat, Raluca Olga

2012-01-01T23:59:59.000Z

223

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

SciTech Connect (OSTI)

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

M. J. Russell

2006-06-01T23:59:59.000Z

224

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

E-Print Network [OSTI]

remote access to data from the reactor site is an important part of this project. The two goals of this monitoring system are to control the use of nuclear materials and to monitor the performance of the facility from a remote location. I have designed a...

Jiltchenkov, Dmitri Victorovich

2012-06-07T23:59:59.000Z

225

Choosing a standard reactor: International competition and domestic politics in Chinese nuclear policy  

Science Journals Connector (OSTI)

China has ambitious plans to expand its nuclear power capacity. One of the policy goals that high-level policymakers have desired is to base the nuclear program on a standardized reactor design. However, this has not materialized so far. By examining its nuclear reactor choices for individual projects, we argue that China’s policymaking process has been greatly influenced by international competition and domestic politics. Multiple international nuclear vendors are intent upon maintaining their respective niches in the expanding Chinese reactor market, and they have used various forms of economic and political pressure to achieve their objectives. On the other hand, China’s policymaking process is fragmented and the shifting power balances among powerful domestic actors do not allow a fixed path to be followed. Further, because of the high costs and potential profits involved, nuclear reactor choices in China have been driven not just by technical considerations but also by foreign and trade policy objectives. All of these make it unlikely that China will standardize the reactor type it constructs in the near future.

M.V. Ramana; Eri Saikawa

2011-01-01T23:59:59.000Z

226

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

Office of Scientific and Technical Information (OSTI)

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

227

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

228

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

SciTech Connect (OSTI)

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

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

2012-01-17T23:59:59.000Z

229

TOKAMAK REACTOR DESIGNS AS A FUNCTION OF ASPECT RATIO  

E-Print Network [OSTI]

GA­A23168 TOKAMAK REACTOR DESIGNS AS A FUNCTION OF ASPECT RATIO by C.P.C. WONG and R.D. STAMBAUGH or reflect those of the United States Government or any agency thereof. #12;GA­A23168 TOKAMAK REACTOR DESIGNS JULY 1999 #12;C.P.C. WONG AND R.D. STAMBAUGH TOKAMAK REACTOR DESIGNS AS A FUNCTION OF ASPECT RATIO

California at Los Angeles, University of

230

NRC review of the CANDU-3 reactor design  

SciTech Connect (OSTI)

This paper presents an overview of the US Nuclear Regulatory Commission's (NRC's) effort to complete an early review of the Canada deuterium uranium (CANDU)-3 reactor design prior to formal submittal of an application for standard design certification. The NRC is conducting a review of the CANDU-3 design in support of a request by AECL Technologics, the US sponsor of the design. The purpose of this review is to encourage early interactions by applicants, vendors, and government agencies with the NRC. The CANDU-3 design is being developed by Atomic Energy of Canada, Limited (AECL), whose CANDU operations are based in Mississauga, Ontario. AECL Technologies, a US subsidiary of AECL, Incorporated, informed the NRC of its intent to seek design certification of the CANDU-3 design under the provisions of 10CFR52 in a letter to the NRC dated May 25, 1989. This paper describes the commission's basis for this type of early review, its purposes and objectives, key elements of the review, the intended product, and the schedule.

Kennedy, J.L. (Nuclear Regulatory Commission, Washington, DC (United States))

1993-01-01T23:59:59.000Z

231

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

232

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

233

A Compact Nuclear Fusion Reactor for Space Flights  

SciTech Connect (OSTI)

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

234

Safety of Department of Energy-Owned Nuclear Reactors  

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

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

1986-09-23T23:59:59.000Z

235

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

236

Strengthening the nuclear-reactor fuel cycle against proliferation  

SciTech Connect (OSTI)

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

237

About Naval Reactors | National Nuclear Security Administration  

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

238

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

SciTech Connect (OSTI)

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

239

Novel Approach to Plasma Facing Materials in Nuclear Fusion Reactors  

SciTech Connect (OSTI)

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

Livramento, V.; Correia, J. B.; Shohoji, N.; Osawa, E. [INETI, Departamento de Materiais e Tecnologias de Producao, Estrada do Pacco do Lumiar, 1649-038 Lisboa (Portugal); Nunes, D. [Associacao Euratom/IST, Departamento de Engenharia de Materiais, Instituto Superior Tecnico, Av. Rovisco Pais, 1049-001 Lisboa (Portugal); Carvalho, P. A.; Fernandes, H.; Silva, C. [Associacao Euratom/IST, Centro de Fusao Nuclear, Instituto Superior Tecnico, Av. Rovisco Pais, 1049-001 Lisboa (Portugal); Hanada, K. [National Institute of Advanced Industrial Science and Technology (AIST), 1-2-1 Namiki, Tsukuba, Ibaraki 305-8564 (Japan)

2008-04-07T23:59:59.000Z

240

Small Modular Fast Reactor Design Description Joint Effort  

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

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

Note: This page contains sample records for the topic "nuclear reactor design" from the National Library of EnergyBeta (NLEBeta).
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241

Development of a molybdenum-rhenium alloy for space nuclear reactors  

SciTech Connect (OSTI)

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

242

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

243

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

244

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

245

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

246

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

247

Piezoelectric material for use in a nuclear reactor core  

SciTech Connect (OSTI)

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

248

Secondary heat exchanger design and comparison for advanced high temperature reactor  

SciTech Connect (OSTI)

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

249

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

250

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

251

Design Basis Threat | National Nuclear Security Administration  

National Nuclear Security Administration (NNSA)

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

252

Dynamics of reverse osmosis in a standalone cogenerative nuclear reactor (Part I: reactivity changes)  

Science Journals Connector (OSTI)

The present study considers the dynamic behaviour of the pressurised water reactor safety features, represented by the integrity of the fuel cladding, under some transient cases. A cosine-shaped heating through the fuel is taken with the corresponding coolant lumps, to simulate realistic cases encountered in nuclear reactors. A mathematical model was developed for the Westinghouse 3411 MWth pressurised water reactor, as an example of a familiar design with predominantly published data design. The model consists of two parts. The first part is concerned with the dynamics of the primary side of the reactor, which is described in this paper. The second part is concerned with the secondary side of the plant, which is described elsewhere in this issue. To study the dynamics of the reactor, a model of 17 lumped parameters was used, consisting of first-order differential equations deduced from the first principles considering six groups of delayed neutrons. A computer program was developed using the Runge-Kutta method to solve these equations and to predict the behaviour of the state variables with time. Two case studies were considered as examples for normal transients. The first case study, which represents Part 1 of this study, considers the effect of primary side transient on the system as the reactivity changes. Reactor reactivity changes, including movements of the reactor control rods, which are taken as an example for the effect of the reactor primary side conditions. These reactivity changes vary from 0.0005 up to 0.003, both for positive and negative reactivity. The results of the developed model, which describe the dynamic response of the reactor primary circuit, have been analysed and verified with the relevant models. These results indicate that the reactor components and the integrity of the fuel cladding were attained during different step changes of reactivity.

Aly Karameldin; M.M. Shamloul; M.R. Shaalan; M.H. Esawy

2007-01-01T23:59:59.000Z

253

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

SciTech Connect (OSTI)

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

Trond Bjornard; John Hockert

2011-08-01T23:59:59.000Z

254

Deployment of a three-dimensional array of micro-pocket fission detector triads (MPFD[superscript]3) for real-time, in-core neutron flux measurements in the Kansas State University TRIGA Mark-II Nuclear Reactor.  

E-Print Network [OSTI]

??A Micro-Pocket Fission Detector (MPFD) is a miniaturized type of fission chamber developed for use inside a nuclear reactor. Their unique design allows them to… (more)

Ohmes, Martin Francis

2012-01-01T23:59:59.000Z

255

Conversion and upgrade of the Rhode Island Nuclear Science Center Reactor  

SciTech Connect (OSTI)

The Rhode Island Atomic Energy Commission, an agency of the state of Rhode Island, has operated a 2-MW swimming pool research reactor at the Rhode Island Nuclear Science Center (RINSC) since 1964. The reactor, which utilizes plate-type materials test reactor fuel elements, is used primarily by facility and research scientists from the University of Rhode Island for neutron scattering, using the beam tubes and activation analysis programs that use irradiation facilities both inside and adjacent to the core. Along with most other university research reactors, the RINSC reactor is now required, pursuant to 10CFR50.64, to convert from the use of high-enrichment uranium fuel elements to the use of low-enriched uranium (LEU) fuel elements. It is apparent that the US Nuclear Regulatory Commission mandate to convert the RINSC reactor to the use of LEU will result in a new core, designed to use the standard fuel plate and at the same time enhance the available neutron flux and spectrum for research using neutron scattering and activation analysis.

DiMeglio, A.F.

1987-01-01T23:59:59.000Z

256

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

SciTech Connect (OSTI)

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

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

2013-07-01T23:59:59.000Z

257

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

258

Design of a 25-kWe Surface Reactor System Based on SNAP Reactor Technologies  

SciTech Connect (OSTI)

A Hastelloy-X clad, sodium-potassium (NaK-78) cooled, moderated spectrum reactor using uranium zirconium hydride (UZrH) fuel based on the SNAP program reactors is a promising design for use in surface power systems. This paper presents a 98 kWth reactor for a power system the uses multiple Stirling engines to produce 25 kWe-net for 5 years. The design utilizes a pin type geometry containing UZrHx fuel clad with Hastelloy-X and NaK-78 flowing around the pins as coolant. A compelling feature of this design is its use of 49.9% enriched U, allowing it to be classified as a category III-D attractiveness and reducing facility costs relative to highly-enriched space reactor concepts. Presented below are both the design and an analysis of this reactor's criticality under various safety and operations scenarios.

Dixon, David D. [Nuclear Systems Design Group, Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Department of Nuclear Engineering, North Carolina State University, Raleigh, NC 27695 (United States); Hiatt, Matthew T. [Nuclear Systems Design Group, Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Department of Nuclear Engineering, Texas A and M University, College Station, TX 77843 (United States); Poston, David I.; Kapernick, Richard J. [Nuclear Systems Design Group, Los Alamos National Laboratory, Los Alamos, NM 87545 (United States)

2006-01-20T23:59:59.000Z

259

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

SciTech Connect (OSTI)

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

Simos, N.

2011-05-01T23:59:59.000Z

260

Small modular reactors and the future of nuclear power in the United States  

Science Journals Connector (OSTI)

Abstract Small modular reactors are the latest “new” technology that nuclear advocates tout as the game changer that will overcome previous economic failures of nuclear power. The debate over \\{SMRs\\} has been particularly intense because of the rapid failure of large “nuclear renaissance” reactors in market economies, the urgent need to address climate change, and the dramatic success of alternative, decentralized resources in lowering costs and increasing deployment. This paper assesses the prospects for SMR technology from three perspectives: the implications of the history of cost escalation in nuclear reactor construction for learning, economies of scale and other process that SMR advocates claim will lower cost; the challenges SMR technology faces in terms of high costs resulting from lost economies of scale, long lead time needed to develop a new design, the size of the task to create assembly lines for modular reactors and intense concern about safety; and the cost and other characteristics – e.g. scalability, speed to market, flexibility, etc. – of available alternatives compared SMR technology. The paper concludes that the decision of the major vendors (Westinghouse and B&W) to dramatically reduce SMR development efforts reflects the severe disadvantages that SMR technology faces in the next several decades.

Mark Cooper

2014-01-01T23:59:59.000Z

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


261

Nuclear plant-aging research on reactor protection systems  

SciTech Connect (OSTI)

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

262

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

263

Modal approximation of xenon oscillations in nuclear reactors — an FEM-based approach  

Science Journals Connector (OSTI)

A modal expansion method for modelling dynamic space-dependent effects (xenon oscillations) in large nuclear reactors is presented. Based on the finite element method (FEM), modal expansion functions for various cases with practical significance can be calculated. In this way it is possible to derive a low order state space model describing the considered effect. The use of this model in on-line controller design and simulation is intended. Modal methods in connection with modern numerical algorithms turned out to be effective to treat distributed parameter systems governed by elliptic differential equations. This will be demonstrated by application to a VVER-1000 pressurized water reactor.

J.Chr. Kalkkuhl; M.G. Döring

1993-01-01T23:59:59.000Z

264

design basis threat | National Nuclear Security Administration  

National Nuclear Security Administration (NNSA)

design basis threat | National Nuclear Security Administration Facebook Twitter Youtube Flickr RSS People Mission Managing the Stockpile Preventing Proliferation Powering the...

265

Design of chemical reactors of the heat exchanger type  

E-Print Network [OSTI]

Operating Profile - Example I 23 , 53 Heat Rate Comparison - Example I Operating Profile - Example 2 Operating Profile - Example 3 Operating Profile - Example 4 Equations (113) and (114) at 790 Reactor Profile - Exan piss 5 and 6 Heat of Reaction.... simple inathematical function of time. While his work was a step forward, it is not directly applicable to the problem of reactor design. Hougen and Watsor. (3), and recently Fair and Rase (4), illustra- ted an exact non-machine method of reactor...

McBeth, Lloyd Theodore

2012-06-07T23:59:59.000Z

266

SUPPLEMENT ANALYSIS OF FOREIGN RESEARCH REACTOR srENT NUCLEAR FUEL  

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

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

267

Synergistic smart fuel for in-pile nuclear reactor measurements  

SciTech Connect (OSTI)

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

268

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

269

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

270

The effect of pulse reactivity for stochastic neutron point kinetic equation in nuclear reactor dynamics  

Science Journals Connector (OSTI)

In this present analysis, the numerical simulation methods are applied to calculate the solution for Stochastic Neutron Point Kinetic Equations (SNPKE) with pulse reactivity in dynamical system of a nuclear reactor. The resulting systems of differential equations are solved for each time step-size. Using experimental data, the methods are investigated with pulse reactivity. The computational results designate that these numerical approximation methods are straightforward, effective and easy for solving stochastic point kinetic equations.

A. Patra; S. Saha Ray

2014-01-01T23:59:59.000Z

271

Novel electromagnetic technique for repositioning of coolant tube spacers in CANDU nuclear reactors  

Science Journals Connector (OSTI)

A novel electromagnetic technique to reposition the coolant tube spacers in the fuel channels of CANDU nuclear reactors was successfully developed in the fall of 1983 at Ontario Hydro Research Division. The need to reposition dislocated spacers in noncommissioned reactors was discovered subsequent to the rupture of a pressure tube in one reactor at the Pickering Nuclear Generator Station in Ontario. A contributing factor to the failure of the tube was the fact that the annular spacers (garter springs) used to maintain the coaxial configuration between the pressure tube and its surrounding calandria tube had been displaced longitudinally for a number of years. Subsequent to this finding it was discovered that a number of garter springs in noncommissioned nuclear reactors were displaced due to vibration induced by various sources during the construction stage. Since the garter springs are not directly accessible by mechanical means extensive dismantling of the fuel channels would have been necessary to reposition the springs in their designated locations. This paper describes a novel method to reposition the garter springs without dismantling the fuel channels. The method consists of exerting a force on the springs in the direction of the required displacement by applying a large electromagnetic impulse (generated by a 200?kJ capacitor bank) to a drive coil inserted into the pressure tube opposite the spacer. The repositioning of displaced garter springs in five new reactors in Ontario has been carried out successfully in 1984. The saving in reactor repair cost interest charges and replacement energy cost was on the order of hundreds of millions of dollars. Equally large benefits and savings will be realized if the need to use this technique in commissioned reactors arises. Also the related development of strong compact coils and low?resistance pulse power cable have significant implications and advantages in various other applications related to the pulse power industry in general and to electromagnetic metal forming and fusion technologies specifically.

Joseph H. Dableh

1986-01-01T23:59:59.000Z

272

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

273

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

274

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

275

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

SciTech Connect (OSTI)

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

276

Sec. Moniz to Georgia, Energy Department Scheduled to Close on Loan Guarantees to Construct New Nuclear Power Plant Reactors  

Broader source: Energy.gov [DOE]

Project represents first new nuclear reactors to begin construction in the United States in three decades

277

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

278

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

279

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

280

Helium circulator design considerations for modular high temperature gas-cooled reactor plant  

SciTech Connect (OSTI)

Efforts are in progress to develop a standard modular high temperature gas-cooled reactor (MHTGR) plant that is amenable to design certification and serial production. The MHTGR reference design, based on a steam cycle power conversion system, utilizes a 350 MW(t) annular reactor core with prismatic fuel elements. Flexibility in power rating is afforded by utilizing a multiplicity of the standard module. The circulator, which is an electric motor-driven helium compressor, is a key component in the primary system of the nuclear plant, since it facilitates thermal energy transfer from the reactor core to the steam generator; and, hence, to the external turbo-generator set. This paper highlights the helium circulator design considerations for the reference MHTGR plant and includes a discussion on the major features of the turbomachine concept, operational characteristics, and the technology base that exists in the U.S.

McDonald, C.F.; Nichols, M.K.

1987-01-01T23:59:59.000Z

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


281

Helium circulator design considerations for modular high temperature gas-cooled reactor plant  

SciTech Connect (OSTI)

Efforts are in progress to develop a standard modular high temperature gas-cooled reactor (MHTGR) plant that is amenable to design certification and serial production. The MHTGR reference design, based on a steam cycle power conversion system, utilizes a 350 MW(t) annular reactor core with prismatic fuel elements. Flexibility in power rating is afforded by utilizing a multiplicity of the standard module. The circulator, which is an electric motor-driven helium compressor, is a key component in the primary system of the nuclear plant, since it facilitates thermal energy transfer from the reactor core to the steam generator; and, hence, to the external turbo-generator set. This paper highlights the helium circulator design considerations for the reference MHTGR plant and includes a discussion on the major features of the turbomachine concept, operational characteristics, and the technology base that exists in the US.

McDonald, C.F.; Nichols, M.K.

1986-12-01T23:59:59.000Z

282

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

283

Dynamics of reverse osmosis in a standalone cogenerative nuclear reactor (Part II: load changes)  

Science Journals Connector (OSTI)

The coupling of desalination units with nuclear power plants has been studied in the present work â?? in this respect, the dynamic behaviour of the pressurised water reactor safety features, represented by the integrity of the fuel cladding, under some transient cases caused by secondary circuit load changes. A cosine-shaped heating through the reactor fuel is taken with its corresponding coolant lumps, to simulate realistic cases encountered in nuclear reactors. As an example, the mathematical model for the Westinghouse 3411 MWth pressurised water reactor, a familiar design with widely published design data was developed. The model consists of two parts; the first one is concerned with the dynamics of the primary side of the reactor; and the second, with the secondary side of the plant. To study the dynamics of the reactor, a 17-lumped parameters model was used. This is a first-order differential equation deduced from the first principles considering six groups of delayed neutrons. A computer program was developed using the Runge-Kutta method to solve these equations and to predict the behaviour of the state variables with time. Two case studies were considered as examples for normal transients. The developed model, which describes the dynamic response of the reactor, primary circuit and secondary circuit, has been analysed and verified with the relevant models. The first case, represented in Part 1 of this study, is concerned with the effect of changes of primary side transient reactivity, including the movement of the reactor control rods. The second one, represented in Part 2 of this study, considers the effect of the secondary side transient reactivity, as the load changes, on the system behaviour. As an example for the secondary side transient, load perturbations, such as load variations in standalone RO desalination units, are selected to study the effect of changing the secondary side conditions on the plant behaviour. The results showed that the reactor components and the fuel matrix should not be affected, and the fuel-cladding integrity is maintained within the safe limits, in all scram cases of RO desalination units coupled with nuclear power plants in the case of either a planned shutdown or an accidental shutdown.

Aly Karameldin; M.M. Shamloul; M.R. Shaalan; M.H. Esawy

2007-01-01T23:59:59.000Z

284

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

285

System Engineering Design [Nuclear Waste Management using  

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

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

286

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

E-Print Network [OSTI]

et. a1. , "A Conceptual Tokamak Reactor Design, umtAK-II,"et. al.. "A Non-Circular Tokamak 'Power Reactor Design,"Contt'ol in Near Term Tokamak Reactors," Proceedings of the

Myers, Richard Allen

2011-01-01T23:59:59.000Z

287

High temperature gas cooled reactor steam-methane reformer design  

SciTech Connect (OSTI)

The concept of the long distance transportation of process heat energy from a High Temperature Gas Cooled Reactor (HTGR) heat source, based on the steam-methane reforming reaction, is being evaluated by the Department of Energy as an energy source/application for use early in the 21st century. This paper summaries the design of a helium heated steam reformer utilized in conjunction with an intermediate loop, 850/degree/C reactor outlet temperature, HTGR process heat plant concept. This paper also discusses various design considerations leading to the mechanical design features, the thermochemical performance, the materials selection and the structural design analysis. 12 refs.

Impellezzeri, J.R.; Drendel, D.B.; Odegaard, T.K.

1981-01-01T23:59:59.000Z

288

Conceptual Design of a Large, Passive Pressure-Tube Light Water Reactor  

E-Print Network [OSTI]

A design for a large, passive, light water reactor has been developed. The proposed concept is a pressure tube reactor of similar design to CANDU reactors, but differing in three key aspects. First, a solid SiC-coated ...

Hejzlar, P.

289

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

E-Print Network [OSTI]

reactor core design 7 Assembly Design and Optimization codePhysics Optimization of Breed and Burn Fast Reactor Systems.OPTIMIZATION CODE (ADOPT) Table 7.4: SWR B&B Reference Reactor

Qvist, Staffan Alexander

2013-01-01T23:59:59.000Z

290

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

291

Research in nondestructive evaluation techniques for nuclear reactor concrete structures  

SciTech Connect (OSTI)

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

292

Foreign Research Reactor Spent Nuclear Fuel Acceptance Program  

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

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

293

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

SciTech Connect (OSTI)

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

Determan, W.R. (Rocketdyne Division, Rockwell International, 6633 Canoga Avenue, P.O. Box 7922, anoga Park, California 91309-7922 (United States)); Van Hagan, T.H. (General Atomics, P.O. Box 85608, San Diego, California 92186-9784 (United States))

1993-01-20T23:59:59.000Z

294

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

295

Development of Improved Burnable Poisons for Commercial Nuclear Power Reactors  

SciTech Connect (OSTI)

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

296

Office of Nuclear Safety Basis and Facility Design  

Broader source: Energy.gov [DOE]

The Office of Nuclear Safety Basis & Facility Design establishes safety basis and facility design requirements and expectations related to analysis and design of nuclear facilities to ensure protection of workers and the public from the hazards associated with nuclear operations.

297

Evaluation of LLNL's Nuclear Accident Dosimeters at the CALIBAN Reactor September 2010  

SciTech Connect (OSTI)

The Lawrence Livermore National Laboratory uses neutron activation elements in a Panasonic TLD holder as a personnel nuclear accident dosimeter (PNAD). The LLNL PNAD has periodically been tested using a Cf-252 neutron source, however until 2009, it was more than 25 years since the PNAD has been tested against a source of neutrons that arise from a reactor generated neutron spectrum that simulates a criticality. In October 2009, LLNL participated in an intercomparison of nuclear accident dosimeters at the CEA Valduc Silene reactor (Hickman, et.al. 2010). In September 2010, LLNL participated in a second intercomparison of nuclear accident dosimeters at CEA Valduc. The reactor generated neutron irradiations for the 2010 exercise were performed at the Caliban reactor. The Caliban results are described in this report. The procedure for measuring the nuclear accident dosimeters in the event of an accident has a solid foundation based on many experimental results and comparisons. The entire process, from receiving the activated NADs to collecting and storing them after counting was executed successfully in a field based operation. Under normal conditions at LLNL, detectors are ready and available 24/7 to perform the necessary measurement of nuclear accident components. Likewise LLNL maintains processing laboratories that are separated from the areas where measurements occur, but contained within the same facility for easy movement from processing area to measurement area. In the event of a loss of LLNL permanent facilities, the Caliban and previous Silene exercises have demonstrated that LLNL can establish field operations that will very good nuclear accident dosimetry results. There are still several aspects of LLNL's nuclear accident dosimetry program that have not been tested or confirmed. For instance, LLNL's method for using of biological samples (blood and hair) has not been verified since the method was first developed in the 1980's. Because LLNL and the other DOE participants were limited in what they were allowed to do at the Caliban and Silene exercises and testing of various elements of the nuclear accident dosimetry programs cannot always be performed as guests at other sites, it has become evident that DOE needs its own capability to test nuclear accident dosimeters. Angular dependence determination and correction factors for NADs desperately need testing as well as more evaluation regarding the correct determination of gamma doses. It will be critical to properly design any testing facility so that the necessary experiments can be performed by DOE laboratories as well as guest laboratories. Alternate methods of dose assessment such as using various metals commonly found in pockets and clothing have yet to be evaluated. The DOE is planning to utilize the Godiva or Flattop reactor for testing nuclear accident dosimeters. LLNL has been assigned the primary operational authority for such testing. Proper testing of nuclear accident dosimeters will require highly specific characterization of the pulse fields. Just as important as the characterization of the pulsed fields will be the design of facilities used to process the NADs. Appropriate facilities will be needed to allow for early access to dosimeters to test and develop quick sorting techniques. These facilities will need appropriate laboratory preparation space and an area for measurements. Finally, such a facility will allow greater numbers of LLNL and DOE laboratory personnel to train on the processing and interpretation of nuclear accident dosimeters and results. Until this facility is fully operational for test purposes, DOE laboratories may need to continue periodic testing as guests of other reactor facilities such as Silene and Caliban.

Hickman, D P; Wysong, A R; Heinrichs, D P; Wong, C T; Merritt, M J; Topper, J D; Gressmann, F A; Madden, D J

2011-06-21T23:59:59.000Z

298

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

299

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

300

Bonus -- Cameras Designed to Strengthen Nuclear Security Can...  

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

Bonus -- Cameras Designed to Strengthen Nuclear Security Can Also Detect Cancer Bonus -- Cameras Designed to Strengthen Nuclear Security Can Also Detect Cancer November 17, 2010 -...

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


301

Passing good judgment, part 1: weapons designers with nuclear...  

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

2014 All Issues submit Passing good judgment, part 1: weapons designers with nuclear testing experience The nuclear weapons designers who developed their skills during...

302

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

303

Iterative methods for solving nonlinear problems of nuclear reactor criticality  

SciTech Connect (OSTI)

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

304

Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR)  

SciTech Connect (OSTI)

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

305

Analysis of nuclear reactor instability phenomena. Progress report  

SciTech Connect (OSTI)

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

306

Status of Fusion Reactor Blanket Design  

Science Journals Connector (OSTI)

Blanket Design and Evaluation / Proceedings of the Seveth Topical Meeting on the Technology of Fusion Energy (Reno, Nevada, June 15–19, 1986)

D. L. Smith; D.-K. Sze

307

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

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

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

308

Power Reactor Progress  

Science Journals Connector (OSTI)

Argonne kicks off EBWR; Allis-Chalmers plans power reactor using both nuclear and conventional fuels ... NUCLEAR POWER took two giant steps last week. ... Just as the first nuclear power system in the U. S. designed and built solely for the generation of electric power went into full operation at Argonne, Allis-Chalmers came up with a new twist in power reactors—a controlled recirculation boiling reactor (CRBR) using both nuclear and conventional fuels (C&EN, Feb. 18, page 7). ...

1957-02-25T23:59:59.000Z

309

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

310

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

SciTech Connect (OSTI)

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

311

RAMI Analysis Program Design and Research for CFETR (Chinese Fusion Engineering Testing Reactor) Tokamak Machine  

Science Journals Connector (OSTI)

Chinese Fusion Engineering Testing Reactor (CFETR) is a test reactor which shall be constructed by National Integration Design Group for Magnetic Confinement Fusion Reactor of China with an ambitious scientific ...

Shijun Qin; Yuntao Song; Damao Yao; Yuanxi Wan; Songtao Wu…

2014-10-01T23:59:59.000Z

312

Heat pipe based passive emergency core cooling system for safe shutdown of nuclear power reactor  

Science Journals Connector (OSTI)

Abstract On March 11th, 2011, a natural disaster created by earthquakes and Tsunami caused a serious potential of nuclear reactor meltdown in Fukushima due to the failure of Emergency Core Cooling System (ECCS) powered by diesel generators. In this paper, heat pipe based ECCS has been proposed for nuclear power plants. The designed loop type heat pipe ECCS is composed of cylindrical evaporator with 62 vertical tubes, each 150 mm diameter and 6 m length, mounted around the circumference of nuclear fuel assembly and 21 m × 10 m × 5 m naturally cooled finned condenser installed outside the primary containment. Heat pipe with overall thermal resistance of 1.44 × 10?5 °C/W will be able to reduce reactor temperature from initial working temperature of 282 °C to below 250 °C within 7 h. The overall ECCS also includes feed water flooding of the core using elevated water tank for initial 10 min which will accelerate cooling of the core, replenish core coolant during loss of coolant accident and avoids heat transfer crisis phenomena during heat pipe start-up process. The proposed heat pipe system will operate in fully passive mode with high runtime reliability and therefore provide safer environment to nuclear power plants.

Masataka Mochizuki; Randeep Singh; Thang Nguyen; Tien Nguyen

2014-01-01T23:59:59.000Z

313

Nuclear processes in magnetic fusion reactors with polarized fuel  

E-Print Network [OSTI]

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

Michail P. Rekalo; Egle Tomasi-Gustafsson

2000-10-16T23:59:59.000Z

314

Modularity in design of the MIT Pebble Bed Reactor  

E-Print Network [OSTI]

The future of new nuclear power plant construction will depend in large part on the ability of designers to reduce capital, operations, and maintenance costs. One of the methods proposed, is to enhance the modularity of ...

Berte, Marc Vincent, 1977-

2004-01-01T23:59:59.000Z

315

The Evolution of Nuclear Power Plant Design: Synopsis  

Science Journals Connector (OSTI)

1 April 1961 research-article The Evolution of Nuclear Power Plant Design: Synopsis Christopher Hinton

1961-01-01T23:59:59.000Z

316

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

317

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

318

Business Case for Early Orders of New Nuclear Reactors Executive Overview  

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

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.

319

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

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

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

320

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

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


321

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

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

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.

322

Economic Analysis of a Nuclear Reactor Powered High-Temperature Electrolysis Hydrogen Production Plant  

SciTech Connect (OSTI)

A reference design for a commercial-scale high-temperature electrolysis (HTE) plant for hydrogen production was developed to provide a basis for comparing the HTE concept with other hydrogen production concepts. The reference plant design is driven by a high-temperature helium-cooled nuclear reactor coupled to a direct Brayton power cycle. The reference design reactor power is 600 MWt, with a primary system pressure of 7.0 MPa, and reactor inlet and outlet fluid temperatures of 540°C and 900°C, respectively. The electrolysis unit used to produce hydrogen includes 4,009,177 cells with a per-cell active area of 225 cm2. The optimized design for the reference hydrogen production plant operates at a system pressure of 5.0 MPa, and utilizes an air-sweep system to remove the excess oxygen that is evolved on the anode (oxygen) side of the electrolyzer. The inlet air for the air-sweep system is compressed to the system operating pressure of 5.0 MPa in a four-stage compressor with intercooling. The alternating-current, AC, to direct-current, DC, conversion efficiency is 96%. The overall system thermal-to-hydrogen production efficiency (based on the lower heating value of the produced hydrogen) is 47.12% at a hydrogen production rate of 2.356 kg/s. An economic analysis of this plant was performed using the standardized H2A Analysis Methodology developed by the Department of Energy (DOE) Hydrogen Program, and using realistic financial and cost estimating assumptions. The results of the economic analysis demonstrated that the HTE hydrogen production plant driven by a high-temperature helium-cooled nuclear power plant can deliver hydrogen at a competitive cost. A cost of $3.23/kg of hydrogen was calculated assuming an internal rate of return of 10%.

E. A. Harvego; M. G. McKellar; M. S. Sohal; J. E. O'Brien; J. S. Herring

2008-08-01T23:59:59.000Z

323

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

E-Print Network [OSTI]

Actinides MOX Mixed OXide MSR Molten-Salt Reactors NERI Nuclear Energy Research Initiative vii PWR Pressurized Water Reactor RGPu Reactor-Grade Plutonium SCNES Self-Consistent Nuclear Energy System STAR Secure Transportable Autonomous Reactor... of LWR?s, the drastic increase of Am and Cm inventories are observed after uranium fuel irradiation and the second recycling of MOX fuel.1 Therefore, partitioning and transmutation of the recovered MA?s could significantly reduce the long...

Tsvetkova, Galina Valeryevna

2004-09-30T23:59:59.000Z

324

Thermal analysis and design of a passive reflux condenser for the simplified boiling water reactor  

SciTech Connect (OSTI)

At present, the advanced light water reactors (ALWRS) in the United States are being designed to remove reactor decay heat for a period of 72 h following a postulated loss-of-coolant accident (LOCA). The water in the pools external to the containment is evaporated or boiled off to remove the decay heat. It is presumed that the water in the pools can be replenished within 72 h through operator actions or outside assistance. Some countries in Europe require that the plant be designed to remove the reactor decay heat for a much longer duration than 72 h without external assistance. This paper presents an analysis and design of a passive heat exchanger called a reflux condenser (RC), which was considered for an ALWR-the 600-MW(electric) simplified boiling water reactor. The RC is required to condense the steam formed when the water in the pool in which the passive containment cooling system (PCCS) is immersed boils following a LOCA. The RCs are nuclear non-safety related. This paper presents steady-state performance of an RC at various outdoor air dry-bulb temperatures under still air conditions.

Bijlani, C.; Patti, F. (Burns Roe Inc., Oradell, NJ (United States)); Prasad, V. (SUNY, Stony Brook, NY (United States))

1993-01-01T23:59:59.000Z

325

Analysis of Granular Flow in a Pebble-Bed Nuclear Reactor  

E-Print Network [OSTI]

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

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

2006-02-16T23:59:59.000Z

326

Analysis of granular flow in a pebble-bed nuclear reactor  

Science Journals Connector (OSTI)

Pebble-bed nuclear reactor technology, which is currently being revived around the world, raises fundamental questions about dense granular flow in silos. A typical reactor core is composed of graphite fuel pebbles, which drain very slowly in a continuous refueling process. Pebble flow is poorly understood and not easily accessible to experiments, and yet it has a major impact on reactor physics. To address this problem, we perform full-scale, discrete-element simulations in realistic geometries, with up to 440000 frictional, viscoelastic 6-cm-diam spheres draining in a cylindrical vessel of diameter 3.5m and height 10m with bottom funnels angled at 30° or 60°. We also simulate a bidisperse core with a dynamic central column of smaller graphite moderator pebbles and show that little mixing occurs down to a 1:2 diameter ratio. We analyze the mean velocity, diffusion and mixing, local ordering and porosity (from Voronoi volumes), the residence-time distribution, and the effects of wall friction and discuss implications for reactor design and the basic physics of granular flow.

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

2006-08-24T23:59:59.000Z

327

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

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

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

328

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

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

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

329

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

330

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

331

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

332

A case study on effectiveness of structural reliability analysis in nuclear reactor safety assessment  

Science Journals Connector (OSTI)

Problems on reliability of structural integrity occupy an important position in various aspects of nuclear reactor safety. In the present paper, an effective method for quantitative evaluation of structural reliability based on ‘stress strength model’ is developed with the objectives of taking a larger number of factors into the evaluation than before and giving useful results within moderate computing time. The method is applied to the reliability analysis of PWR pressure vessels. The results show the relative importance of inspection as well as the parameter uncertainty for assuring the reliability of the structure, although analysis is limited within the scope of linear elastic fracture mechanics (LEFM). This case study also shows that the analysis of structural reliability is effective for safety assessment of nuclear power plants in general and possibly for the improvements of the consistency in the design code.

A. Yamaguchi; S. Kondo; Y. Togo

1983-01-01T23:59:59.000Z

333

Design Options for Ultra-compact Nuclear Driven Power Sources for Field Applications  

E-Print Network [OSTI]

be compact, able to discharge its waste heat into the environment without a local water source, and have low logistical overhead. An air-cooled fast-spectrum nuclear reactor coupled to a direct Brayton cycle would be a viable and suitable design concept...

Mathis, Dean

2014-04-21T23:59:59.000Z

334

Fusion Engineering and Design 42 (1998) 307318 Experimental measurement of nuclear heating in a  

E-Print Network [OSTI]

Fusion Engineering and Design 42 (1998) 307­318 Experimental measurement of nuclear heating , Yoshimi Kasugai b a School of Engineering and Applied Science, Uni6ersity of California (UCLA), Los Angeles, CA 90095-159710, USA b Department of Reactor Engineering, Japan Atomic Energy Research Institute

Abdou, Mohamed

335

Assigning Seismic Design Category to Large Reactors: A Case Study of the ATR  

Broader source: Energy.gov [DOE]

Assigning Seismic Design Category to Large Reactors: A Case Study of the ATR Stuart Jensen October 21, 2014

336

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

SciTech Connect (OSTI)

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

337

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

SciTech Connect (OSTI)

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

338

IoT-Based Maintenance Process Design for Fusion Reactor Remote Handling System  

Science Journals Connector (OSTI)

Nuclear fusion is one of the most important ways ... to risk of ionizing radiation, the nuclear fusion reactor will relay on remote handling maintenance to ... the maintenance efficiency. The basic structure of fusion

Ruonan Zhang; Xinbao Liu

2014-12-01T23:59:59.000Z

339

Safeguards-by-Design: Guidance for High Temperature Gas Reactors (HTGRs) With Pebble Fuel  

SciTech Connect (OSTI)

The following is a guidance document from a series prepared for the U.S. Department of Energy (DOE) National Nuclear Security Administration (NNSA), under the Next Generation Safeguards Initiative (NGSI), to assist facility designers and operators in implementing international Safeguards-by-Design (SBD). SBD has two main objectives: (1) to avoid costly and time consuming redesign work or retrofits of new nuclear fuel cycle facilities and (2) to make the implementation of international safeguards more effective and efficient at such facilities. In the long term, the attainment of these goals would save industry and the International Atomic Energy Agency (IAEA) time, money, and resources and be mutually beneficial. This particular safeguards guidance document focuses on pebble fuel high temperature gas reactors (HTGR). The purpose of the IAEA safeguards system is to provide credible assurance to the international community that nuclear material and other specified items are not diverted from peaceful nuclear uses. The safeguards system consists of the IAEA’s statutory authority to establish safeguards; safeguards rights and obligations in safeguards agreements and additional protocols; and technical measures implemented pursuant to those agreements. Of foremost importance is the international safeguards agreement between the country and the IAEA, concluded pursuant to the Treaty on the Non-Proliferation of Nuclear Weapons (NPT). According to a 1992 IAEA Board of Governors decision, countries must: notify the IAEA of a decision to construct a new nuclear facility as soon as such decision is taken; provide design information on such facilities as the designs develop; and provide detailed design information based on construction plans at least 180 days prior to the start of construction, and on "as-built" designs at least 180 days before the first receipt of nuclear material. Ultimately, the design information will be captured in an IAEA Design Information Questionnaire (DIQ), prepared by the facility operator, typically with the support of the facility designer. The IAEA will verify design information over the life of the project. This design information is an important IAEA safeguards tool. Since the main interlocutor with the IAEA in each country is the State Regulatory Authority/SSAC (or Regional Regulatory Authority, e.g. EURATOM), the responsibility for conveying this design information to the IAEA falls to the State Regulatory Authority/SSAC.

Philip Casey Durst; Mark Schanfein

2012-08-01T23:59:59.000Z

340

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

SciTech Connect (OSTI)

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

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


341

Design, development, and applications of a low-cost, dynamic neutron radiography system utilizing the TAMU NSC TRIGA reactor  

E-Print Network [OSTI]

partial fulfilment of the requirements for the degree of MASTER OF SC'IENCE May 1990 Major Subject: Nuclear Engineering DESIGN, DEVELOPMENT. AND APPLICATIONS OF A LOW ? COST, DYNAMIC NEUTRON RADIOGRAPHY SYSTEM UTILIZING THE TAMU NSC TRIGA REACTOR A...DESIGN, DEVELOPMENT. AND APPLICATIONS OF A LOW ? COST, DYNAMIC NEUTRON RADIOGRAPHY SYSTEM UTILIZING THE TAMU NSC TRIGA REACTOR A Thesis SC'OTT PATRIC'If ItIIDGETT Submitted to the Ofhce of Graduate Studies of Texas AklVI I!niversity rn...

Midgett, Scott Patrick

2012-06-07T23:59:59.000Z

342

Novel Reactor Design for Solid Fuel Chemical Looping Combustion  

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

Novel Reactor Design for Solid Fuel Novel Reactor Design for Solid Fuel Chemical Looping Combustion Opportunity Research is active on the patent pending technology, titled "Apparatus and Method for Solid Fuel Chemical Looping Combustion." This technology is available for licensing and/or further collaborative research from the U.S. Department of Energy's National Energy Technology Laboratory. Overview The removal of CO2 from power plants is challenging because existing methods to separate CO2 from the gas mixture requires a significant fraction of the power plant output. Chemical-looping combustion (CLC) is a novel technology that utilizes a metal oxide oxygen carrier to transport oxygen to the fuel thereby avoiding direct contact between fuel and air. The use of CLC has the advantages of reducing the energy penalty while

343

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

344

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

SciTech Connect (OSTI)

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

345

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

346

Nuclear options  

Science Journals Connector (OSTI)

... sad if transient commercial expediency led the country to take risks now by adopting nuclear reactor designs from abroad which are in some respects technically less sound than those produced at ... much lower priority. It can be anticipated, however; that although the types of nuclear reactor selected as the best that Britain could build next may be suitable for unit designs ...

G. R. Bainbridge

1974-06-21T23:59:59.000Z

347

Management of Hanford Site non-defense production reactor spent nuclear fuel, Hanford Site, Richland, Washington  

SciTech Connect (OSTI)

The US Department of Energy (DOE) needs to provide radiologically, and industrially safe and cost-effective management of the non-defense production reactor spent nuclear fuel (SNF) at the Hanford Site. The proposed action would place the Hanford Site`s non-defense production reactor SNF in a radiologically- and industrially-safe, and passive storage condition pending final disposition. The proposed action would also reduce operational costs associated with storage of the non-defense production reactor SNF through consolidation of the SNF and through use of passive rather than active storage systems. Environmental, safety and health vulnerabilities associated with existing non-defense production reactor SNF storage facilities have been identified. DOE has determined that additional activities are required to consolidate non-defense production reactor SNF management activities at the Hanford Site, including cost-effective and safe interim storage, prior to final disposition, to enable deactivation of facilities where the SNF is now stored. Cost-effectiveness would be realized: through reduced operational costs associated with passive rather than active storage systems; removal of SNF from areas undergoing deactivation as part of the Hanford Site remediation effort; and eliminating the need to duplicate future transloading facilities at the 200 and 400 Areas. Radiologically- and industrially-safe storage would be enhanced through: (1) removal from aging facilities requiring substantial upgrades to continue safe storage; (2) utilization of passive rather than active storage systems for SNF; and (3) removal of SNF from some storage containers which have a limited remaining design life. No substantial increase in Hanford Site environmental impacts would be expected from the proposed action. Environmental impacts from postulated accident scenarios also were evaluated, and indicated that the risks associated with the proposed action would be small.

NONE

1997-03-01T23:59:59.000Z

348

Relative Movements for Design of Commodities in Nuclear Power Plants  

Broader source: Energy.gov [DOE]

Relative Movements for Design of Commodities in Nuclear Power Plants Javad Moslemian, Vice President, Nuclear Power Technologies, Sargent & Lundy LLC Nezar Abraham, Senior Associate II, Nuclear Power Technologies, Sargent & Lundy LLC

349

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

350

Safeguards-by-Design:Guidance for High Temperature Gas Reactors (HTGRs) With Prismatic Fuel  

SciTech Connect (OSTI)

Introduction and Purpose The following is a guidance document from a series prepared for the U.S. Department of Energy (DOE) National Nuclear Security Administration (NNSA), under the Next Generation Safeguards Initiative (NGSI), to assist facility designers and operators in implementing international Safeguards-by-Design (SBD). SBD has two main objectives: (1) to avoid costly and time consuming redesign work or retrofits of new nuclear fuel cycle facilities and (2) to make the implementation of international safeguards more effective and efficient at such facilities. In the long term, the attainment of these goals would save industry and the International Atomic Energy Agency (IAEA) time, money, and resources and be mutually beneficial. This particular safeguards guidance document focuses on prismatic fuel high temperature gas reactors (HTGR). The purpose of the IAEA safeguards system is to provide credible assurance to the international community that nuclear material and other specified items are not diverted from peaceful nuclear uses. The safeguards system consists of the IAEA’s statutory authority to establish safeguards; safeguards rights and obligations in safeguards agreements and additional protocols; and technical measures implemented pursuant to those agreements. Of foremost importance is the international safeguards agreement between the country and the IAEA, concluded pursuant to the Treaty on the Non-Proliferation of Nuclear Weapons (NPT). According to a 1992 IAEA Board of Governors decision, countries must: notify the IAEA of a decision to construct a new nuclear facility as soon as such decision is taken; provide design information on such facilities as the designs develop; and provide detailed design information based on construction plans at least 180 days prior to the start of construction, and on "as-built" designs at least 180 days before the first receipt of nuclear material. Ultimately, the design information will be captured in an IAEA Design Information Questionnaire (DIQ), prepared by the facility operator, typically with the support of the facility designer. The IAEA will verify design information over the life of the project. This design information is an important IAEA safeguards tool. Since the main interlocutor with the IAEA in each country is the State Regulatory Authority/SSAC (or Regional Regulatory Authority, e.g. EURATOM), the responsibility for conveying this design information to the IAEA falls to the State Regulatory Authority/SSAC. For the nuclear industry to reap the benefits of SBD (i.e. avoid cost overruns and construction schedule slippages), nuclear facility designers and operators should work closely with the State Regulatory Authority and IAEA as soon as a decision is taken to build a new nuclear facility. Ideally, this interaction should begin during the conceptual design phase and continue throughout construction and start-up of a nuclear facility. Such early coordination and planning could influence decisions on the design of the nuclear material processing flow-sheet, material storage and handling arrangements, and facility layout (including safeguards equipment), etc.

Mark Schanfein; Casey Durst

2012-11-01T23:59:59.000Z

351

Numerical solution for stochastic point-kinetics equations with sinusoidal reactivity in dynamical system of nuclear reactor  

Science Journals Connector (OSTI)

In this present analysis, the numerical simulation methods are applied to calculate the solution for stochastic point-kinetic equations with sinusoidal reactivity in dynamical system of nuclear reactor. The resulting system of differential equations is solved for each time step-size. Using experimental data, the methods are investigated over initial conditions and with sinusoidal reactivity. The computational results designate that these numerical approximation methods are straightforward, effective and easy for solving stochastic point-kinetic equations.

S. Saha Ray; A. Patra

2013-01-01T23:59:59.000Z

352

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

353

Nuclear Deployment Scorecards | Department of Energy  

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

Initiatives » Nuclear Reactor Technologies » Nuclear Deployment Initiatives » Nuclear Reactor Technologies » Nuclear Deployment Scorecards Nuclear Deployment Scorecards October 31, 2013 Quarterly Nuclear Deployment Scorecard - October 2013 The scorecard includes news updates, regulatory status, reactor design certification, early site permits, new plant construction progress, and expected operation dates. August 8, 2013 Quarterly Nuclear Deployment Scorecard - July 2013 The scorecard includes news updates, regulatory status, reactor design certification, early site permits, new plant construction progress, and expected operation dates. May 1, 2013 Quarterly Nuclear Power Deployment Scorecard - April 2013 The scorecard includes news updates, regulatory status, reactor design certification, early site permits, and new plant construction progress.

354

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

355

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

356

High Temperature Reactor (HTR) Deep Burn Core and Fuel Analysis: Design Selection for the Prismatic Block Reactor  

SciTech Connect (OSTI)

The Deep Burn (DB) Project is a U.S. Department of Energy sponsored feasibility study of Transuranic Management using high burnup fuel in the high temperature helium cooled reactor (HTR). The DB Project consists of seven tasks: project management, core and fuel analysis, spent fuel management, fuel cycle integration, TRU fuel modeling, TRU fuel qualification, and HTR fuel recycle. In the Phase II of the Project, we conducted nuclear analysis of TRU destruction/utilization in the HTR prismatic block design (Task 2.1), deep burn fuel/TRISO microanalysis (Task 2.3), and synergy with fast reactors (Task 4.2). The Task 2.1 covers the core physics design, thermo-hydraulic CFD analysis, and the thermofluid and safety analysis (low pressure conduction cooling, LPCC) of the HTR prismatic block design. The Task 2.3 covers the analysis of the structural behavior of TRISO fuel containing TRU at very high burnup level, i.e. exceeding 50% of FIMA. The Task 4.2 includes the self-cleaning HTR based on recycle of HTR-generated TRU in the same HTR. Chapter IV contains the design and analysis results of the 600MWth DB-HTR core physics with the cycle length, the average discharged burnup, heavy metal and plutonium consumptions, radial and axial power distributions, temperature reactivity coefficients. Also, it contains the analysis results of the 450MWth DB-HTR core physics and the analysis of the decay heat of a TRU loaded DB-HTR core. The evaluation of the hot spot fuel temperature of the fuel block in the DB-HTR (Deep-Burn High Temperature Reactor) core under full operating power conditions are described in Chapter V. The investigated designs are the 600MWth and 460MWth DB-HTRs. In Chapter VI, the thermo-fluid and safety of the 600MWth DB-HTRs has been analyzed to investigate a thermal-fluid design performance at the steady state and a passive safety performance during an LPCC event. Chapter VII describes the analysis results of the TRISO fuel microanalysis of the 600MWth and 450MWth DB-HTRs. The TRISO fuel microanalysis covers the gas pressure buildup in a coated fuel particle including helium production, the thermo-mechanical behavior of a CFP, the failure probabilities of CFPs, the temperature distribution in a CPF, and the fission product (FP) transport in a CFP and a graphite. In Chapter VIII, it contains the core design and analysis of sodium cooled fast reactor (SFR) with deep burn HTR reactor. It considers a synergistic combination of the DB-MHR and an SFR burner for a safe and efficient transmutation of the TRUs from LWRs. Chapter IX describes the design and analysis results of the self-cleaning (or self-recycling) HTR core. The analysis is considered zero and 5-year cooling time of the spent LWR fuels.

Francesco Venneri; Chang-Keun Jo; Jae-Man Noh; Yonghee Kim; Claudio Filippone; Jonghwa Chang; Chris Hamilton; Young-Min Kim; Ji-Su Jun; Moon-Sung Cho; Hong-Sik Lim; MIchael A. Pope; Abderrafi M. Ougouag; Vincent Descotes; Brian Boer

2010-09-01T23:59:59.000Z

357

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

SciTech Connect (OSTI)

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

Policke, Timothy A; Nygaard, Eric T

2014-05-06T23:59:59.000Z

358

Conceptual design analysis of an MHD power conversion system for droplet-vapor core reactors. Final report  

SciTech Connect (OSTI)

A nuclear driven magnetohydrodynamic (MHD) generator system is proposed for the space nuclear applications of few hundreds of megawatts. The MHD generator is coupled to a vapor-droplet core reactor that delivers partially ionized fissioning plasma at temperatures in range of 3,000 to 4,000 K. A detailed MHD model is developed to analyze the basic electrodynamics phenomena and to perform the design analysis of the nuclear driven MHD generator. An incompressible quasi one dimensional model is also developed to perform parametric analyses.

Anghaie, S.; Saraph, G.

1995-12-31T23:59:59.000Z

359

Uncertainties in the Anti-neutrino Production at Nuclear Reactors  

E-Print Network [OSTI]

Output and Performance of Nuclear Power Plants by Im- provedPower, Tomari Nuclear Power Plant (Units 1, 2) License

Djurcic, Zelimir

2009-01-01T23:59:59.000Z

360

6 - Other nuclear energy applications: Hydrogen for transport desalination ships space research reactors for radioisotopes  

Science Journals Connector (OSTI)

Publisher Summary This chapter describes several nuclear energy applications. Hydrogen itself is likely to be an important future fuel; like electricity, it is an energy carrier. Nuclear energy can be used to make hydrogen electrolytically; and in the future, high-temperature reactors are likely to be used for thermochemical production. Desalination is energy-intensive. Nuclear energy is already being used for desalination, and nuclear energy has the potential for much greater use. Nuclear power has also revolutionized the navy; it is particularly suitable for vessels that need to be at sea for long periods without refueling, or for powerful submarine propulsion. After a gap of several years, there is a revival of interest in the use of nuclear fission power for space missions as well. Many of the world's nuclear reactors are used for research and training, materials testing, or the production of radioisotopes for medicine and industry. Research reactors are much smaller than power reactors or those propelling ships, and many are on university campuses. Research reactors are simpler than power reactors and operate at lower temperatures.

Ian Hore-Lacy

2007-01-01T23:59:59.000Z

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


361

Role of fast reactor and its cycle to reduce nuclear waste burden  

SciTech Connect (OSTI)

The role of the metal fuel fast reactor with recycling of actinides and the five long-lived fission products based on the concept of the Self-Consistent Nuclear Energy System has been examined by evaluating the reduction of nuclear wastes during the transition period to this reactor system. The evaluation was done in comparison to an LWR once-through case and a conventional actinide recycling oxide fast reactor. As a result, it is quantitatively clarified that a metal fuel fast reactor with actinide and the five long-lived fission products (I{sup 129}, Tc{sup 99}, Zr{sup 93}, Cs{sup 135} and Sn{sup 126}) recycling could play a significant role in reducing the nuclear waste burden including the current LWR wastes. This can be achieved by using a fast neutron spectrum reactor enhanced with metal fuel that brings high capability as a 'waste burner'. (authors)

Arie, Kazuo; Oomori, Takashi; Okita, Takeshi [Toshiba Corporation, 8, Shinsugita-cho, Isogo-ku, Yokohama 235-8523 (Japan); Kawashima, Masatoshi [Toshiba Nuclear Engineering Services Corporation, 8, Shinsugita-cho, Isogo-ku, Yokohama, 235-8523 (Japan); Kotake, Shoji [The Japan Atomic Power Company, 1-1, Kanda-Mitoshiro-cho, Chiyoda-ku, Tokyo 101-0053 (Japan); Fuji-ie, Yoichi [Nuclear Salon Fuji-ie, 1-11-10, Yushima, Bunkyo-ku, Tokyo 113-0034 (Japan)

2013-07-01T23:59:59.000Z

362

Summary of Prometheus Radiation Shielding Nuclear Design Analysis  

SciTech Connect (OSTI)

This report transmits a summary of radiation shielding nuclear design studies performed to support the Prometheus project. Together, the enclosures and references associated with this document describe NRPCT (KAPL & Bettis) shielding nuclear design analyses done for the project.

J. Stephens

2006-01-13T23:59:59.000Z

363

Design of Complex Systems to Achieve Passive Safety: Natural Circulation Cooling of Liquid Salt Pebble Bed Reactors  

E-Print Network [OSTI]

Safety. The Accident at TEPCO’s Fukushima Nuclear Power2: Accident and Thermal Fluids Analysis PIRTs. (Nuclearmolten nuclear reactor core debris following accidents such

Scarlat, Raluca Olga

2012-01-01T23:59:59.000Z

364

A visit to the bombed nuclear reactor at Tuwaitha, Iraq  

Science Journals Connector (OSTI)

... stallations and of people with one exception; they did not want cameras inside the Soviet reactor building.

1983-03-31T23:59:59.000Z

365

Modular Inspection System for a Complete IN-Service Examination of Nuclear Reactor Pressure Vessel, Including Beltline Region  

SciTech Connect (OSTI)

Final Report for a DOE Phase II Contract Describing the design and fabrication of a reactor inspection modular rover prototype for reactor vessel inspection.

David H. Bothell

2000-04-30T23:59:59.000Z

366

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

E-Print Network [OSTI]

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

367

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

E-Print Network [OSTI]

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

Aldrich, David C.

1979-01-01T23:59:59.000Z

368

Heterogeneous nuclear reactor models for optimal xenon control  

SciTech Connect (OSTI)

In this study a thermal nuclear reactor is modeled as a two-dimensional lattice of fuel and control rods placed in an infinite-moderator in plane geometry. The two-group diffusion theory approximation is used for neutron transport. Space-time neutron balance equations are written for two groups and reduced to one space-time algebraic equation by using the two-dimensional Fourier transform. This equation is written at all fuel and control rod locations. Iodine-xenon and promethium-samarium dynamic equations are also written at fuel rod locations only. These equations are then linearized about an equilibrium point which is determined from the steady-state form of the original nonlinear system equations. After studying poisonless criticality, with and without control, and the stability of the open-loop system and after checking its controllability, a performance criterion is defined for the xenon-induced spatial flux oscillation problem in the form of a functional to be minimized. Linear-quadratic optimal control theory is then applied to solve the problem.

Gondal, I.A.

1984-01-01T23:59:59.000Z

369

Record of Decision for the Final EIS on Proposed Nuclear Weapons Nonproliferation Policy Concerning Foreign Research Reactor Spent Nuclear Fuel  

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

5091 5091 Friday May 17, 1996 Part IV Department of Energy Record of Decision for the Final Environmental Impact Statement on a Proposed Nuclear Weapons Nonproliferation Policy Concerning Foreign Research Reactor Spent Nuclear Fuel; Notice 25092 Federal Register / Vol. 61, No. 97 / Friday, May 17, 1996 / Notices DEPARTMENT OF ENERGY Record of Decision for the Final Environmental Impact Statement on a Proposed Nuclear Weapons Nonproliferation Policy Concerning Foreign Research Reactor Spent Nuclear Fuel AGENCY: Department of Energy. ACTION: Record of decision. SUMMARY: DOE, in consultation with the Department of State, has decided to implement a new foreign research reactor spent fuel acceptance policy as specified in the Preferred Alternative contained in the Final Environmental Impact Statement on a Proposed

370

Coolant flows in prismatic fuel and particle bed nuclear reactors for rocket applications  

SciTech Connect (OSTI)

Semiempirical expressions for pressure losses in prismatic and particle bed reactors for nuclear propulsion are combined with the geometric characteristics of core configurations and coolant flow patterns. The results are used to illustrate a limitation on the coolant velocity and to develop a unified approach to a quantitative comparison of merits and demerits of different reactor core concepts intended for space applications.

Bohachevsky, I.O. (Rocketdyne Division FA44, Rockwell International Corporation, 6633 Canoga Avenue, Canoga Park, California 91309-7922 (United States))

1993-01-20T23:59:59.000Z

371

OPTIMAL DESIGN OF A HIGH PRESSURE ORGANOMETALLIC CHEMICAL VAPOR DEPOSITION REACTOR  

E-Print Network [OSTI]

OPTIMAL DESIGN OF A HIGH PRESSURE ORGANOMETALLIC CHEMICAL VAPOR DEPOSITION REACTOR K.J. BACHMANN of computer simulations as an optimal design tool which lessens the costs in time and effort in experimental vapor deposition (HPOMCVD) reactor for use in thin film crystal growth. The advantages of such a reactor

372

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

SciTech Connect (OSTI)

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

Snoj, L. [Josef Stefan Inst., Jamova cesta 39, SI-1000 Ljubljana (Slovenia); Sklenka, L.; Rataj, J. [Dept. of Nuclear Reactor, Czech Technical Univ. in Prague, V Holesovickach 2, 180 00 Prague 8 (Czech Republic); Boeck, H. [Vienna Univ. of Technology/Atominstitut, Stadionallee 2, 1020 Vienna (Austria)

2012-07-01T23:59:59.000Z

373

Design, optimization and evaluation of a free-fall biomass fast pyrolysis reactor and its products.  

E-Print Network [OSTI]

??The focus of this work is a radiatively heated, free-fall, fast pyrolysis reactor. The reactor was designed and constructed for the production of bio-oil from… (more)

Ellens, Cody James

2009-01-01T23:59:59.000Z

374

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

375

Sliding Mode Control for Pressurized-Water Nuclear Reactors in load following operations with bounded xenon oscillations  

Science Journals Connector (OSTI)

Abstract One of the important operations in nuclear power plants is load-following in which imbalance of axial power distribution induces xenon oscillations. These oscillations must be maintained within acceptable limits otherwise the nuclear power plant could become unstable. Therefore, bounded xenon oscillation considered to be a constraint for the load-following operation. In this paper, sliding mode control (SMC) which is a robust nonlinear controller is designed to control the Pressurized-Water Nuclear Reactor (PWR) power for the load-following operation problem that ensures xenon oscillations are kept bounded within acceptable limits. The proposed controller uses constant axial offset (AO) strategy to maintain xenon oscillations to be bounded. The constant AO is a robust state constraint for load-following problem. The reactor core is simulated based on the two-point nuclear reactor model and one delayed neutron group. The stability analysis is given by means Lyapunov approach, thus the control system is guaranteed to be stable within a large range. The employed method is easy to implement in practical applications and moreover, the sliding mode control exhibits the desired dynamic properties during the entire output-tracking process independent of perturbations. Simulation results are presented to demonstrate the effectiveness of the proposed controller in terms of performance, robustness and stability. Results show that the proposed controller for the load-following operation is sufficiently effective so that the xenon oscillations are kept bounded in the considered region.

G.R. Ansarifar; S. Saadatzi

2015-01-01T23:59:59.000Z

376

Westinghouse Small Modular Reactor balance of plant and supporting systems design  

SciTech Connect (OSTI)

The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (>225 MWe) integral pressurized water reactor (iPWR), in which all of the components typically associated with the nuclear steam supply system (NSSS) of a nuclear power plant are incorporated within a single reactor pressure vessel. This paper is the second in a series of four papers which describe the design and functionality of the Westinghouse SMR. It focuses, in particular, upon the supporting systems and the balance of plant (BOP) designs of the Westinghouse SMR. Several Westinghouse SMR systems are classified as safety, and are critical to the safe operation of the Westinghouse SMR. These include the protection and monitoring system (PMS), the passive core cooling system (PXS), and the spent fuel cooling system (SFS) including pools, valves, and piping. The Westinghouse SMR safety related systems include the instrumentation and controls (I and C) as well as redundant and physically separated safety trains with batteries, electrical systems, and switch gears. Several other incorporated systems are non-safety related, but provide functions for plant operations including defense-in-depth functions. These include the chemical volume control system (CVS), heating, ventilation and cooling (HVAC) systems, component cooling water system (CCS), normal residual heat removal system (RNS) and service water system (SWS). The integrated performance of the safety-related and non-safety related systems ensures the safe and efficient operation of the Westinghouse SMR through various conditions and transients. The turbine island consists of the turbine, electric generator, feedwater and steam systems, moisture separation systems, and the condensers. The BOP is designed to minimize assembly time, shipping challenges, and on-site testing requirements for all structures, systems, and components. (authors)

Memmott, M. J.; Stansbury, C.; Taylor, C. [Westinghouse Electric Company LLC, 600 Cranberry Woods Drive, Cranberry Twp. PA 16066 (United States)

2012-07-01T23:59:59.000Z

377

Advanced Core Design And Fuel Management For Pebble-Bed Reactors  

SciTech Connect (OSTI)

A method for designing and optimizing recirculating pebble-bed reactor cores is presented. At the heart of the method is a new reactor physics computer code, PEBBED, which accurately and efficiently computes the neutronic and material properties of the asymptotic (equilibrium) fuel cycle. This core state is shown to be unique for a given core geometry, power level, discharge burnup, and fuel circulation policy. Fuel circulation in the pebble-bed can be described in terms of a few well?defined parameters and expressed as a recirculation matrix. The implementation of a few heat?transfer relations suitable for high-temperature gas-cooled reactors allows for the rapid estimation of thermal properties critical for safe operation. Thus, modeling and design optimization of a given pebble-bed core can be performed quickly and efficiently via the manipulation of a limited number key parameters. Automation of the optimization process is achieved by manipulation of these parameters using a genetic algorithm. The end result is an economical, passively safe, proliferation-resistant nuclear power plant.

Hans D. Gougar; Abderrafi M. Ougouag; William K. Terry

2004-10-01T23:59:59.000Z

378

RELIABLE CONTROL AND ONLINE MONITORING DESIGN FOR NUCLEAR POWER PLANT WITH IRIS DEMONSTRATION.  

E-Print Network [OSTI]

??Development and deployment of small-scale nuclear power reactors and their autonomous control and monitoring is part of the mission under the Global Nuclear Energy Partnership… (more)

Guo, Yin

2010-01-01T23:59:59.000Z

379

Fast reactor power plant design having heat pipe heat exchanger  

DOE Patents [OSTI]

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

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

1984-08-30T23:59:59.000Z

380

Fast reactor power plant design having heat pipe heat exchanger  

DOE Patents [OSTI]

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

Huebotter, Paul R. (Western Springs, IL); McLennan, George A. (Downers Grove, IL)

1985-01-01T23:59:59.000Z

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


381

Computer analyses for the design, operation and safety of new isotope production reactors: A technology status review  

SciTech Connect (OSTI)

A review is presented on the currently available technologies for nuclear reactor analyses by computer. The important distinction is made between traditional computer calculation and advanced computer simulation. Simulation needs are defined to support the design, operation, maintenance and safety of isotope production reactors. Existing methods of computer analyses are categorized in accordance with the type of computer involved in their execution: micro, mini, mainframe and supercomputers. Both general and special-purpose computers are discussed. Major computer codes are described, with regard for their use in analyzing isotope production reactors. It has been determined in this review that conventional systems codes (TRAC, RELAP5, RETRAN, etc.) cannot meet four essential conditions for viable reactor simulation: simulation fidelity, on-line interactive operation with convenient graphics, high simulation speed, and at low cost. These conditions can be met by special-purpose computers (such as the AD100 of ADI), which are specifically designed for high-speed simulation of complex systems. The greatest shortcoming of existing systems codes (TRAC, RELAP5) is their mismatch between very high computational efforts and low simulation fidelity. The drift flux formulation (HIPA) is the viable alternative to the complicated two-fluid model. No existing computer code has the capability of accommodating all important processes in the core geometry of isotope production reactors. Experiments are needed (heat transfer measurements) to provide necessary correlations. It is important for the nuclear community, both in government, industry and universities, to begin to take advantage of modern simulation technologies and equipment. 41 refs.

Wulff, W.

1990-01-01T23:59:59.000Z

382

Routine and post-accident sampling in nuclear reactors  

SciTech Connect (OSTI)

Review of the Three Mile Island accident by NRC has resulted in new post-accident-sampling-capability requirements for utilities that operate pressurized water reactors and/or boiling water reactors. Several vendors are offering equipment that they hope will suffice to met both the new NRC regulations and an operational deadline of January 1, 1981. The advantages and disadvantages of these systems and projected future-new-system needs for TVA reactors are being evaluated in light of TMI experience.

Armento, W.J.; Kitts, F.G.; German, G.E.

1980-01-01T23:59:59.000Z

383

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

SciTech Connect (OSTI)

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

Not Available

1994-11-01T23:59:59.000Z

384

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

SciTech Connect (OSTI)

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

NONE

1995-12-01T23:59:59.000Z

385

Methods of estimating the reliability of nuclear reactor installations  

Science Journals Connector (OSTI)

Domestic and foreign experience indicate that quantitative analysis of the reliability of reactor systems during their development is a rather effective means of increasing their reliability.

I. Ya. Emel'yanov; A. I. Klemin; E. F. Polyakov

1974-11-01T23:59:59.000Z

386

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

Broader source: Energy.gov [DOE]

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

387

Method for loading, operating, and unloading a ball-bed nuclear reactor  

SciTech Connect (OSTI)

This patent describes a method of operating a ball-bed nuclear reactor with fuel element balls. Some have a fissionable material content different from that of others of the balls. It consists of: initially partly filling a reactor core with fuel balls of sufficient fissionable material content for establishing criticality and a desired level of power production at the completion of the partial filling and then, without any further filling of the reactor cavern, starting reactor operation; thereafter without any removal of fuel balls from the reactor cavern, filling fuel balls continually or in groups at relatively short intervals into the reactor cavern during increasing burning up of the fuel balls already, for compensation of the diminishing fissionable material content of the reactor core constituted by the fuel balls until a final total quantity of filling is reached; after the final filling quantity is reached and burning up has occurred, shutting down the reactor, cooling it off, releasing the pressure in the cavern, and thereafter unloading all the fuel balls from the reactor cavern, unloading being begun when the reactor is shut down and being completed before the reactor is restarted.

Teuchert, E.; Haas, K.A.; Gerwin, H.

1987-09-22T23:59:59.000Z

388

Nuclear reactor having a polyhedral primary shield and removable vessel insulation  

DOE Patents [OSTI]

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

Ekeroth, D.E.; Orr, R.

1993-12-07T23:59:59.000Z

389

Nuclear reactor having a polyhedral primary shield and removable vessel insulation  

DOE Patents [OSTI]

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

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

1993-01-01T23:59:59.000Z

390

Nuclear waste transmutation performance assessment of an accelerator driven subcritical reactor for waste transmutation (ADS-NWT)  

Science Journals Connector (OSTI)

Abstract As a concept of lead–bismuth cooled accelerator driven sub-critical reactor, ADS-NWT was proposed for nuclear waste transmutation by FDS Team. In this paper, preliminary transmutation performance assessment on ADS-NWT was performed by using metallic dispersion fuel and metal alloy fuel. The studies were committed to improving the Discharge Transmutation Support Ratio (DTSR) under three key design constraints. Subject to the design constraints on the effective multiplication factor, the peak linear power and the peak fast neutron fluence, the design studies had been performed for ADS-NWT blanket. From these studies, the results showed that the maximum DTSR was primarily related to the MA/Pu ratio, the peak fast neutron fluence and fuel design under imposed design constraints. The maximum DTSR could be ?47 in one refueling cycle time when the thermal power was fixed to 1000 MW and the effective multiplication factor was 0.98 based on the specific design.

Zhong Chen; Yican Wu; Baoxin Yuan; Dongmei Pan

2015-01-01T23:59:59.000Z

391

Materials Research Needs for Near-Term Nuclear Reactors  

Science Journals Connector (OSTI)

Technical Paper / NSF Workshop on the Research Needs of the Next Generation Nuclear Power Technology / Material

John R. Weeks

392

Safeguards Issues at Nuclear Reactors and Enrichment Plants  

SciTech Connect (OSTI)

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

Boyer, Brian D [Los Alamos National Laboratory

2012-08-15T23:59:59.000Z

393

Inherently Safe Reactors and a Second Nuclear Era  

Science Journals Connector (OSTI)

...system pumps and the steam generators are also...of active safety systems or reactor operators...must be proven. Tools and proce-dures...must be devised. Steam generators must be...deploy a new reactor system such as PIUS or even...proba-bilistic risk assessment, is perhaps 10 to...

Alvin M. Weinberg; Irving Spiewak

1984-06-29T23:59:59.000Z

394

Light Water Reactors [Corrosion and Mechanics of Materials] - Nuclear  

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

Light Water Reactors Light Water Reactors Capabilities Materials Testing Environmentally Assisted Cracking (EAC) of Reactor Materials Corrosion Performance/Metal Dusting Overview Light Water Reactors Fatigue Testing of Carbon Steels and Low-Alloy Steels Environmentally Assisted Cracking of Ni-Base Alloys Irradiation-Induced Stress Corrosion Cracking of Austenitic Stainless Steels Steam Generator Tube Integrity Program Air Oxidation Kinetics for Zr-based Alloys Fossil Energy Fusion Energy Metal Dusting Publications List Irradiated Materials Steam Generator Tube Integrity Other Facilities Work with Argonne Contact us For Employees Site Map Help Join us on Facebook Follow us on Twitter NE on Flickr Corrosion and Mechanics of Materials Light Water Reactors Bookmark and Share To continue safe operation of current LWRs, the aging degradation of the

395

High Flux Isotope Reactor (HFIR) | Nuclear Science | ORNL  

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

High Flux Isotope Reactor High Flux Isotope Reactor May 30, 2013 The High Flux Isotope Reactor (HFIR) first achieved criticality on August 25, 1965, and achieved full power in August 1966. It is a versatile 85-MW isotope production, research, and test reactor with the capability and facilities for performing a wide variety of irradiation experiments and a world-class neutron scattering science program. HFIR is a beryllium-reflected, light water-cooled and moderated flux-trap type swimming pool reactor that uses highly enriched uranium-235 as fuel. HFIR typically operates seven 23-to-27 day cycles per year. Irradiation facility capabilities include Flux trap positions: Peak thermal flux of 2.5X1015 n/cm2/s with similar epithermal and fast fluxes (Highest thermal flux available in the

396

Coupling the nickel-iodine-sulphur cycle with a nuclear reactor  

Science Journals Connector (OSTI)

Nuclear hydrogen production is a technically feasible and economically viable option for addressing future energy needs. Several projects have been started on the co-generation of hydrogen and electricity from nuclear energy. In this report, the nickel sulphur iodine (NIS) cycle, a thermochemical water splitting cycle originally developed in ENEA for solar hydrogen production, was studied to be coupled with a new generation nuclear reactor for massive hydrogen production.

Pier Paolo Prosini

2013-01-01T23:59:59.000Z

397

Design Windows for a He Cooled Fusion Reactor* Dai-Kai Sze and Ahmed Hassanein  

E-Print Network [OSTI]

Design Windows for a He Cooled Fusion Reactor* Dai-Kai Sze and Ahmed Hassanein Argonne National Laboratory 9700 South Cass Avenue, Argonne, IL 60439 EQUATIONDERIVATION ABSTRACT A design window concept is developed for a He-cooled fusion reactor blanket and divertor design. This concept allows study

Harilal, S. S.

398

Innovative fusion reactor design analysis: Annual performance report, May 15, 1988--January 31, 1989  

SciTech Connect (OSTI)

This report discusses the following topics on fusion reactor component design: FLiBe intermediate heat exchanger design analysis; FLiBe properties; design methodology; FLiBe system steam generator freezeup; FLiBe reactor systems studies; tritium breeding ratio control; analysis of original objectives; and budget analysis. 15 refs., 13 figs., 3 tabs. (LSP)

Klein, A.C.

1989-01-31T23:59:59.000Z

399

Neural net controlled tag gas sampling system for nuclear reactors  

DOE Patents [OSTI]

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

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

1997-01-01T23:59:59.000Z

400

Applicability of base-isolation R D in non-reactor facilities to a nuclear reactor plant  

SciTech Connect (OSTI)

Seismic isolation is gaining increased attention worldwide for use in a wide spectrum of critical facilities, ranging from hospitals and computing centers to nuclear power plants. While the fundamental principles and technology are applicable to all of these facilities, the degree of assurance that the actual behavior of the isolation systems is as specified varies with the nature of the facility involved. Obviously, the level of effort to provide such assurance for a nuclear power plant will be much greater than that required for, say, a critical computer facility. The question, therefore, is to what extent can research and development (R D) for non-nuclear use be used to provide technological data needed for seismic isolation of a nuclear power plant. This question, of course is not unique to seismic isolation. Virtually every structural component, system, or piece of equipment used in nuclear power plants is also used in non- nuclear facilities. Experience shows that considerable effort is needed to adapt conventional technology into a nuclear power plant. Usually, more thorough analysis is required, material and fabrication quality-control requirements are more stringent as are controls on field installation. In addition, increased emphasis on maintainability and inservice inspection throughout the life of the plant is generally required to gain acceptance in nuclear power plant application. This paper reviews the R D programs ongoing for seismic isolation in non-nuclear facilities and related experience and makes a preliminary assessment of the extent to which such R D and experience can be used for nuclear power plant application. Ways are suggested to improve the usefulness of such non-nuclear R D in providing the high level of confidence required for the use of seismic isolation in a nuclear reactor plant. 2 refs.

Seidensticker, R.W.; Chang, Y.W.

1990-01-01T23:59:59.000Z

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


401

Control rod worth and related nuclear characteristics of an axially heterogeneous liquidmetal fast breeder reactor core  

SciTech Connect (OSTI)

An axially heterogeneous core (AHC) concept is applied to a 1000-MW(electric)-class tank-type liquidmetal fast breeder reactor (LMFBR). This AHC is characterized by a disk-shaped internal blanket with a radial thickness adjustment at the core midplane. The nuclear characteristics connected with control rod worth of the AHC are analyzed and compared with those of a homogeneous core (HOC) of the same power rating. The neutronics analysis shows that the reactivity insertion due to the vertical displacement of control rods relative to the core, which is an important safety characteristic of a tank-type LMFBR core, is significantly decreased in the AHC because of the reduced control worth requirement and smaller peakto-average differential worth in the primary control system. This allows the AHC to have a vertical displacement 50% greater than that of the HOC and may offer less rigid design conditions for roof slab stiffness and core support systems in a tank-type LMFBR. A reduction in the number of primary control rods is also possible because of a smaller control worth requirement and a better power peaking factor. The work was sponsored by a federation of Japanese electric power companies and performed under the guidance of the fast breeder reactor project office.

Kawashima, K.; Inayaki, T.; Inoue, K.; Kaneto, K.

1985-08-01T23:59:59.000Z

402

Radiochemical characteristics of tritium to be considered in fusion reactor facility design  

Science Journals Connector (OSTI)

The results of research and development related to radiochemical characteristics of tritium to be considered in a fusion reactor facility design are summarized. Reactions induced by...

S. O’hira; T. Hayashi; W. Shu; T. Yamanishi

2007-06-01T23:59:59.000Z

403

Nuclear reactor with low-level core coolant intake  

DOE Patents [OSTI]

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

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

1993-01-01T23:59:59.000Z

404

Mossbauer Spectroscopic Study of Gamma Irradiation on the Structural Properties of Hematite, Magnetite and Limonite Concrete for Nuclear Reactor Shielding  

Science Journals Connector (OSTI)

This work investigate the effect of gamma irradiation on a heavy type of concrete, constructed for nuclear reactor shield. The effect of gamma irradiation was...

N.A. Eissa; M.S.I. Kany; A.S. Mohamed; A.A. Sallam; M.H. El Fouly

405

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

DOE Patents [OSTI]

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

Mariani, Robert Dominick

2014-09-09T23:59:59.000Z

406

Analysis of Nuclear Reactor Siting Policy Using Multiobjective Programming  

Science Journals Connector (OSTI)

Decisions on power plant sites have emphasized the minimization of location-dependent costs. Recent events, particularly the incident at the Three Mile Island nuclear power plant, have led to calls for a nuclear ...

Jared L. Cohon; Thomas W. Eagles…

1983-01-01T23:59:59.000Z

407

Effect of Fuel Type on the Attainable Power of the Encapsulated Nuclear Heat Source Reactor  

SciTech Connect (OSTI)

The Encapsulated Nuclear Heat Source (ENHS) is a small liquid metal cooled fast reactor that features uniform composition core, at least 20 effective full power years of operation without refueling, nearly zero burnup reactivity swing and heat removal by natural circulation. A number of cores have been designed over the last few years to provide the first three of the above features. The objective of this work is to find to what extent use of nitride fuel, with either natural or enriched nitrogen, affects the attainable power as compared to the reference metallic fueled core. All the compared cores use the same fuel rod diameter, D, and length but differ in the lattice pitch, P, and Pu weight percent. Whereas when using Pb-Bi eutectic for both primary and intermediate coolants the P/D of the metallic fueled core is 1.36, P/D for the nitride cores are, respectively, 1.21 for natural nitrogen and 1.45 for enriched nitrogen. A simple one-dimensional thermal hydraulic model has been developed for the ENHS reactor. Applying this model to the different designs it was found that when the IHX length is at its reference value of 10.4 m, the power that can be removed by natural circulation using nitride fuel with natural nitrogen is 65% of the reference power of 125 MWth that is attainable using metallic fuel. However, using enriched nitrogen the attainable power is 110% of the reference. To get 125 MWth the effective IHX length need be 8.7 m and 30.5 m for, respectively, enriched and natural nitrogen nitride fuel designs. (authors)

Okawa, Tsuyoshi; Greenspan, Ehud [Department of Nuclear Engineering, University of California, Berkeley, CA 94720 (United States)

2006-07-01T23:59:59.000Z

408

High Flux Isotope Reactor named Nuclear Historic Landmark | ornl...  

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

late 1950s as a production reactor to meet anticipated demand for transuranic isotopes ("heavy" elements such as plutonium and curium). HFIR today is a DOE Office of Science User...

409

The Neutrino Mass Hierarchy from Nuclear Reactor Experiments  

E-Print Network [OSTI]

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

Emilio Ciuffoli; Jarah Evslin; Xinmin Zhang

2013-02-04T23:59:59.000Z

410

A Participatory Design Framework: Incorporating Public Views into the Design of Nuclear Power Plants.  

E-Print Network [OSTI]

??This thesis presents a participatory systems design framework for the design of a nuclear power plant. The work begins with a review of the so-called… (more)

Goodfellow, Martin J

2013-01-01T23:59:59.000Z

411

High Flux Isotope Reactor cold neutron source reference design concept  

SciTech Connect (OSTI)

In February 1995, Oak Ridge National Laboratory`s (ORNL`s) deputy director formed a group to examine the need for upgrades to the High Flux Isotope Reactor (HFIR) system in light of the cancellation of the Advanced neutron Source Project. One of the major findings of this study was that there was an immediate need for the installation of a cold neutron source facility in the HFIR complex. In May 1995, a team was formed to examine the feasibility of retrofitting a liquid hydrogen (LH{sub 2}) cold source facility into an existing HFIR beam tube. The results of this feasibility study indicated that the most practical location for such a cold source was the HB-4 beam tube. This location provides a potential flux environment higher than the Institut Laue-Langevin (ILL) vertical cold source and maximizes the space available for a future cold neutron guide hall expansion. It was determined that this cold neutron beam would be comparable, in cold neutron brightness, to the best facilities in the world, and a decision was made to complete a preconceptual design study with the intention of proceeding with an activity to install a working LH{sub 2} cold source in the HFIR HB-4 beam tube. During the development of the reference design the liquid hydrogen concept was changed to a supercritical hydrogen system for a number of reasons. This report documents the reference supercritical hydrogen design and its performance. The cold source project has been divided into four phases: (1) preconceptual, (2) conceptual design and testing, (3) detailed design and procurement, and (4) installation and operation. This report marks the conclusion of the conceptual design phase and establishes the baseline reference concept.

Selby, D.L.; Lucas, A.T.; Hyman, C.R. [and others

1998-05-01T23:59:59.000Z

412

Optimized core design of a supercritical carbon dioxide-cooled fast reactor  

E-Print Network [OSTI]

Spurred by the renewed interest in nuclear power, Gas-cooled Fast Reactors (GFRs) have received increasing attention in the past decade. Motivated by the goals of the Generation-IV International Forum (GIF), a GFR cooled ...

Handwerk, Christopher S. (Christopher Stanley), 1974-

2007-01-01T23:59:59.000Z

413

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

SciTech Connect (OSTI)

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

414

Dual annular rotating [open quotes]windowed[close quotes] 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. 4 figures.

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

1994-03-29T23:59:59.000Z

415

System pressure effect on the nuclear reactor limiting criterion. Revision 1  

SciTech Connect (OSTI)

The acceptable operating limits of a nuclear reactor are set to prevent fuel cladding damage. Critical Heat Flux (CHF) is the limiting criterion for the high pressure systems such as the BWRs (6.9 MPa) and the PWRs (13.8 MPa). However, the Onset of Flow Instability (OFI) is the limiting criterion of the low pressure system such as the existing Savannah River Site (SRS) production reactors (0.2 MPa). The physical basis of this difference is presented. 3 refs.

Chen, Kuo-Fu

1990-12-31T23:59:59.000Z

416

Canister design for deep borehole disposal of nuclear waste .  

E-Print Network [OSTI]

??The objective of this thesis was to design a canister for the disposal of spent nuclear fuel and other high-level waste in deep borehole repositories… (more)

Hoag, Christopher Ian.

2006-01-01T23:59:59.000Z

417

Detecting a Nuclear Fission Reactor at the Center of the Earth  

E-Print Network [OSTI]

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

R. S. Raghavan

2002-08-23T23:59:59.000Z

418

Hanging core support system for a nuclear reactor  

DOE Patents [OSTI]

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

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

1987-01-01T23:59:59.000Z

419

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

420

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

SciTech Connect (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.

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

2007-09-01T23:59:59.000Z

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


421

Power generation from nuclear reactors in aerospace applications  

SciTech Connect (OSTI)

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

English, R.E.

1982-01-01T23:59:59.000Z

422

Monthly Nuclear Utility Generation by State and Reactor, 2004  

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

194,1095428,1211897,1374523,1347985,1262403,1340639,1347577,1387346,15503218 "Arkansas Nuclear One 1",840,,639800,598183,639443,403029,346066,489260,629320,629209,610996,637714,619...

423

Monthly Nuclear Utility Generation by State and Reactor, 2008  

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

2889,1359120,1330655,1334646,1369191,1335231,1264610,725939,1043730,14168091 "Arkansas Nuclear One 1",842,,640210,332640,639812,611739,611473,610763,626182,625966,612739,514485,-32...

424

Monthly Nuclear Utility Generation by State and Reactor, 2005  

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

049398,1385659,1329546,1365357,1364550,1315097,826104,724688,842506,13689571 "Arkansas Nuclear One 1",836,,642446,580408,640573,616442,635753,607660,621404,622382,608955,74815,-336...

425

Monthly Nuclear Utility Generation by State and Reactor, 2003  

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

0183,1370443,1333879,1333266,1232981,970683,1026829,1008188,1033852,14689416 "Arkansas Nuclear One 1",840,,638732,576736,637898,614123,622312,611199,598045,569278,582588,637484,278...

426

Monthly Nuclear Utility Generation by State and Reactor, 2007  

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

2613,979493,1330868,1372259,1363174,1328057,1383098,1348837,1393879,15486102 "Arkansas Nuclear One 1",843,,642145,579996,631611,425625,233279,610804,629284,623738,610379,634652,620...

427

Fuel characterization for hydrogen-producing nuclear reactors.  

E-Print Network [OSTI]

??Reatores nucleares de 4 geração do tipo HTGR (reatores de alta temperatura refrigerados a gás) apresentam vantagens em relação a um reator a água pressurizada,… (more)

Kelly Cristina Martins Faêda

2011-01-01T23:59:59.000Z

428

Conceptual design of the bimodal nuclear power and propulsion system based on the ‘‘TOPAZ?2’’ type thermionic reactor?converter with the modernized single?cell thermionic fuel elements  

Science Journals Connector (OSTI)

The paper presents characteristics and conceptual design of the bimodal space thermionic system with the single?cell thermionic fuel elements intended for operation in two modes: steady?state power mode wherein the electric power is generated required for supply of the spaceborne systems at a level of 20 kWe and forced power and propulsion mode. The paper contains results of the design studies performed by the Small Business ‘‘NP Energotech’’ under the Agreement with Rockwell International/Rocketdyne Division and according to the Rocketdyne Division provided Design Requirements. Involved in the work was the team of specialists of RRC ‘‘Kurchatov Institute’’ CDBMB and Research Institute of SPA ‘‘Luch’’.

Nikolai N. Ponomarev?Stepnoi; Veniamin A. Usov; Yuri V. Nikolaev; Aleksander S. Gontar; Boris G. Oglobin; Aleksei N. Luppov; Aleksander V. Klimov; Yevgeny D. Avdoshyn

1995-01-01T23:59:59.000Z

429

Nuclear nonproliferation: Concerns with US delays in accepting foregin research reactors` spent fuel  

SciTech Connect (OSTI)

One key US nonproliferation goal is to discourage use of highly enriched uranium fuel (HEU), which can be used to make nuclear bombs, in civilian nuclear programs worldwide. DOE`s Off-Site Fuels Policy for taking back spent HEU from foreign research reactors was allowed to expire due to environmental reasons. This report provides information on the effects of delays in renewing the Off-Site Fuels Policy on US nonproliferation goals and programs (specifically the reduced enrichment program), DOE`s efforts to renew the fuels policy, and the price to be charged to the operators of foreign reactors for DOE`s activities in taking back spent fuel.

NONE

1994-03-25T23:59:59.000Z

430

High flux isotope reactor cold source preconceptual design study report  

SciTech Connect (OSTI)

In February 1995, the deputy director of Oak Ridge National Laboratory (ORNL) formed a group to examine the need for upgrades to the High Flux Isotope Reactor (HFIR) system in light of the cancellation of the Advanced Neutron Source Project. One of the major findings of this study was that there was an immediate need for the installation of a cold neutron source facility in the HFIR complex. The anticipated cold source will consist of a cryogenic LH{sub 2} moderator plug, a cryogenic pump system, a refrigerator that uses helium gas as a refrigerant, a heat exchanger to interface the refrigerant with the hydrogen loop, liquid hydrogen transfer lines, a gas handling system that includes vacuum lines, and an instrumentation and control system to provide constant system status monitoring and to maintain system stability. The scope of this project includes the development, design, safety analysis, procurement/fabrication, testing, and installation of all of the components necessary to produce a working cold source within an existing HFIR beam tube. This project will also include those activities necessary to transport the cold neutron beam to the front face of the present HFIR beam room. The cold source project has been divided into four phases: (1) preconceptual, (2) conceptual design and research and development (R and D), (3) detailed design and procurement, and (4) installation and operation. This report marks the conclusion of the preconceptual phase and establishes the concept feasibility. The information presented includes the project scope, the preliminary design requirements, the preliminary cost and schedule, the preliminary performance data, and an outline of the various plans for completing the project.

Selby, D.L.; Bucholz, J.A.; Burnette, S.E. [and others

1995-12-01T23:59:59.000Z

431

Generation IV Nuclear Energy Systems Construction Cost Reductions through the Use of Virtual Environments - Task 5 Report: Generation IV Reactor Virtual Mockup Proof-of-Principle Study  

SciTech Connect (OSTI)

Task 5 report is part of a 3 year DOE NERI-sponsored effort evaluating immersive virtual reality (CAVE) technology for design review, construction planning, and maintenance planning and training for next generation nuclear power plants. Program covers development of full-scale virtual mockups generated from 3D CAD data presented in a CAVE visualization facility. Created a virtual mockup of PBMR reactor cavity and discussed applications of virtual mockup technology to improve Gen IV design review, construction planning, and maintenance planning.

Timothy Shaw; Anthony Baratta; Vaughn Whisker

2005-02-28T23:59:59.000Z

432

Conceptual design of an annular-fueled superheat boiling water reactor  

E-Print Network [OSTI]

The conceptual design of an annular-fueled superheat boiling water reactor (ASBWR) is outlined. The proposed design, ASBWR, combines the boiler and superheater regions into one fuel assembly. This ensures good neutron ...

Ko, Yu-Chih, Ph. D. Massachusetts Institute of Technology

2011-01-01T23:59:59.000Z

433

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

SciTech Connect (OSTI)

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

434

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

SciTech Connect (OSTI)

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

435

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

SciTech Connect (OSTI)

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

436

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

SciTech Connect (OSTI)

The Department of Energy's Environmental Management (DOE-EM) Program is responsible for the receipt and storage of aluminum research reactor spent nuclear fuel or used fuel until ultimate disposition. Aluminum research reactor used fuel is currently being stored or is anticipated to be returned to the U.S. and stored at DOE-EM storage facilities at the Savannah River Site and the Idaho Nuclear Technology and Engineering Center. This paper assesses the technologies and the options for safe transportation/receipt and interim storage of aluminum research reactor spent fuel and reviews the comprehensive strategy for its management. The U.S. Department of Energy uses the Appendix A, Spent Nuclear Fuel Acceptance Criteria, to identify the physical, chemical, and isotopic characteristics of spent nuclear fuel to be returned to the United States under the Foreign Research Reactor Spent Nuclear Fuel Acceptance Program. The fuel is further evaluated for acceptance through assessments of the fuel at the foreign sites that include corrosion damage and handleability. Transport involves use of commercial shipping casks with defined leakage rates that can provide containment of the fuel, some of which are breached. Options for safe storage include wet storage and dry storage. Both options must fully address potential degradation of the aluminum during the storage period. This paper focuses on the various options for safe transport and storage with respect to technology maturity and application.

Vinson, D.

2010-07-11T23:59:59.000Z

437

Constraining potential nuclear-weapons proliferation from civilian reactors  

SciTech Connect (OSTI)

Cessation of the Cold War and renewed international attention to the proliferation of weapons of mass destruction are leading to national policies aimed at restraining nuclear-weapons proliferation that could occur through the nuclear-fuel cycle. Argonne, which has unique experience, technology, and capabilities, is one of the US national laboratories contributing to this nonproliferation effort.

Travelli, A.; Gaines, L.L.; Minkov, V.; Olson, A.P.; Snelgrove, J.

1993-11-01T23:59:59.000Z

438

GCRA review and appraisal of HTGR reactor-core-design program. [HTGR-SC, -R, -NHSDR  

SciTech Connect (OSTI)

The reactor-core-design program has as its principal objective and responsibility the design and resolution of major technical issues for the reactor core and core components on a schedule consistent with the plant licensing and construction program. The task covered in this review includes three major design areas: core physics, core thermal and hydraulic performance fuel element design, and in-core fuel performance evaluation.

Not Available

1980-09-01T23:59:59.000Z

439

Nuclear breeder reactor fuel element with silicon carbide getter  

DOE Patents [OSTI]

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

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

1987-01-01T23:59:59.000Z

440

The Programmable Logic Controller and its application in nuclear reactor systems  

SciTech Connect (OSTI)

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

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

1993-09-01T23:59:59.000Z

Note: This page contains sample records for the topic "nuclear reactor design" from the National Library of EnergyBeta (NLEBeta).
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441

GEN-IV Reactors  

Science Journals Connector (OSTI)

Generation-IV reactors are a set of nuclear reactors currently being developed under international collaborations targeting ... economics, proliferation resistance, and physical protection of nuclear energy. Nuclear

Taek K. Kim

2013-01-01T23:59:59.000Z

442

Experimental and Computational Study of a Scaled Reactor Cavity Cooling System  

E-Print Network [OSTI]

The Very High Temperature Gas-Cooled Reactor (VHTR) is one of the next generation nuclear reactors designed to achieve high temperatures to support industrial applications and power generation. The Reactor Cavity Cooling System (RCCS) is a passive...

Vaghetto, Rodolfo

2013-11-25T23:59:59.000Z

443

Dry Storage of Research Reactor Spent Nuclear Fuel - 13321  

SciTech Connect (OSTI)

Spent fuel from domestic and foreign research reactors is received and stored at the Savannah River Site's L Area Material Storage (L Basin) Facility. This DOE-owned fuel consists primarily of highly enriched uranium in metal, oxide or silicide form with aluminum cladding. Upon receipt, the fuel is unloaded and transferred to basin storage awaiting final disposition. Disposition alternatives include processing via the site's H Canyon facility for uranium recovery, or packaging and shipment of the spent fuel to a waste repository. A program has been developed to provide a phased approach for dry storage of the L Basin fuel. The initial phase of the dry storage program will demonstrate loading, drying, and storage of fuel in twelve instrumented canisters to assess fuel performance. After closure, the loaded canisters are transferred to pad-mounted concrete overpacks, similar to those used for dry storage of commercial fuel. Unlike commercial spent fuel, however, the DOE fuel has high enrichment, very low to high burnup, and low decay heat. The aluminum cladding presents unique challenges due to the presence of an oxide layer that forms on the cladding surface, and corrosion degradation resulting from prolonged wet storage. The removal of free and bound water is essential to the prevention of fuel corrosion and radiolytic generation of hydrogen. The demonstration will validate models predicting pressure, temperature, gas generation, and corrosion performance, provide an engineering scale demonstration of fuel handling, drying, leak testing, and canister backfill operations, and establish 'road-ready' storage of fuel that is suitable for offsite repository shipment or retrievable for onsite processing. Implementation of the Phase I demonstration can be completed within three years. Phases II and III, leading to the de-inventory of L Basin, would require an additional 750 canisters and 6-12 years to complete. Transfer of the fuel from basin storage to dry storage requires integration with current facility operations, and selection of equipment that will allow safe operation within the constraints of existing facility conditions. Examples of such constraints that are evaluated and addressed by the dry storage program include limited basin depth, varying fuel lengths up to 4 m, (13 ft), fissile loading limits, canister closure design, post-load drying and closure of the canisters, instrument selection and installation, and movement of the canisters to storage casks. The initial pilot phase restricts the fuels to shorter length fuels that can be loaded to the canister directly underwater; subsequent phases will require use of a shielded transfer system. Removal of the canister from the basin, followed by drying, inerting, closure of the canister, and transfer of the canister to the storage cask are completed with remotely operated equipment and appropriate shielding to reduce personnel radiation exposure. (authors)

Adams, T.M.; Dunsmuir, M.D.; Leduc, D.R.; Severynse, T.F.; Sindelar, R.L. [Savannah River National Laboratory (United States)] [Savannah River National Laboratory (United States); Moore, E.N. [Moore Nuclear Energy, LLC (United States)] [Moore Nuclear Energy, LLC (United States)

2013-07-01T23:59:59.000Z

444

Energy Department Announces New Investments in Advanced Nuclear Power Reactors  

Broader source: Energy.gov [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...

445

Advanced Light Water Reactor Plants System 80+{trademark} Design Certification Program. Annual progress report, October 1, 1992--September 30, 1993  

SciTech Connect (OSTI)

The purpose of this report is to provide a status of the progress that was made towards Design Certification of System 80+{trademark} during the US government`s 1993 fiscal year. The System 80+ Advanced Light Water Reactor (ALWR) is a 3931 MW{sub t} (1350 MWe) Pressurized Water Reactor (PWR). The design consists of an essentially complete plant. It is based on evolutionary improvements to the Standardized System 80 nuclear steam supply system in operation at Palo Verde Units 1, 2, and 3, and the Duke Power Company P-81 balance-of-plant (BOP) that was designed and partially constructed at the Cherokee plant site. The System 80/P-81 original design has been substantially enhanced to increase conformance with the EPRI ALWR Utility Requirements Document (URD). Some design enhancements incorporated in the System 80+ design are included in the four units currently under construction in the Republic of Korea. These units form the basis of the Korean standardization program. The full System 80+ standard design has been offered to the Republic of China, in response to their recent bid specification. The ABB-CE Standard Safety Analysis Report (CESSAR-DC) was submitted to the NRC and a Draft Safety Evaluation Report was issued by the NRC in October 1992. CESSAR-DC contains the technical basis for compliance with the EPRI URD for simplified emergency planning. The Nuclear Steam Supply System (NSSS) is the standard ABB-Combustion Engineering two-loop arrangement with two steam generators, two hot legs and four cold legs each with a reactor coolant pump. The System 80+ standard plant includes a sperical steel containment vessel which is enclosed in a concrete shield building, thus providing the safety advantages of a dual containment.

Not Available

1993-12-31T23:59:59.000Z

446

Long Term Storage with Surveillance of Canadian Prototype Nuclear Power Reactors  

SciTech Connect (OSTI)

Atomic Energy of Canada (AECL) was originally formed by the government of Canada in 1952 to perform research in radiation and nuclear areas. In the mid 1950's Canada decided to limit itself to peaceful uses of nuclear energy and AECL embarked on several research and development programs, one of them being the development of nuclear power plants. This led to the development of the CANDU{sup TM} design of heavy water power reactors, of which there are now 29 operating around the world. This presentation discusses the present state of the first two CANDU{sup TM} prototype reactors and a prototype boiling light water reactor and lessons learnt after being in a long-term storage with surveillance state for more than 20 years. AECL facilities undergo decommissioning by either a prompt or a deferred removal approach. Both approaches are initiated after an operating facility has been declared redundant and gone through final operational shutdown. For the deferred approach, initial decommissioning activities are performed to put the facility into a sustainable, safe, shutdown state to minimize the hazards and costs of the ensuing extended storage with surveillance (SWS) or Safestor phase. At the appropriate time, the facility is dismantled and removed, or put into a suitable condition for re-use. AECL has a number of facilities that were built during its history, and some of these are now redundant or will become redundant in the near future. The deferred removal approach is part of AECL's decommissioning strategy for several reasons: 1. Reduction in radiation doses to workers during the final dismantling, 2. No facilities are available yet in Canada for the management of quantity of wastes arising from decommissioning, 3. Financial constraints presented by the number of facilities that will undergo decommissioning, compared to the availability of funds to carry out the work. This has led to the development of a comprehensive decommissioning plan that includes all of AECL's redundant and presently operating facilities. Several significant issues have arisen over the decades these reactors have been in the SWS phase. With the long time frames encompassed by this approach to decommissioning, the storage and maintenance of facility information for future decommissioning, and the knowledge and training of successive generations of staff to perform facility inspections and maintenance are major issues. Complacency of both staff and management is a potential issue. The problem arises primarily because these facilities have been put into a comparatively low hazard state, are remotely located, and not much changes over time and changes are slow. During the period that these facilities have been in this state, regulatory scrutiny and expectations have increased. This along with continuing changes in regulatory staff requires a continuous education and communication activity with the regulator. Because of the long time frames involved, the building structures continue to deteriorate slowly, and repairs and maintenance are required upon occasion. The costs can be significant, for example, to replace a roof. When these occasions arise, the balance of cost and benefit is always questioned, i.e., 'Wouldn't it be better to spend the money on dismantling rather than fixing?'. One positive note is that ancillary buildings that either have very low radiation hazards or were decontaminated during the initial decommissioning activities have been used for several alternate uses. This has helped to defray costs, and also keeps an interest in maintaining the building structures and systems in good condition during the SWS phase. Over the past few years, the lessons learnt from these and other facilities in SWS, have been addressed in a holistic manner by moving from a short-term, project-oriented approach to a comprehensive, long term stewardship approach. Key to this approach was the development of an integrated plan to decommission all of AECL's facilities. One of the prime results of the implementation of this plan was the development of an organiza

Janzen, Rick [Atomic Energy of Canada Ltd., Chalk River Laboratories, Chalk River, Ontario, K0J 1J0 (Canada)

2008-01-15T23:59:59.000Z

447

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

E-Print Network [OSTI]

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

L. A. Mikaelyan

2002-10-03T23:59:59.000Z

448

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

449

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

SciTech Connect (OSTI)

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

Not Available

1993-11-01T23:59:59.000Z

450

Reactor Physics Parametric and Depletion Studies in Support of TRISO Particle Fuel Specification for the Next Generation Nuclear Plant  

SciTech Connect (OSTI)

Reactor physics calculations were initiated to answer several major questions related to the proposed TRISO-coated particle fuel that is to be used in the prismatic Very High Temperature Reactor (VHTR) or the Next Generation Nuclear Plant (NGNP). These preliminary design evaluation calculations help ensure that the upcoming fuel irradiation tests will test appropriate size and type of fuel particles for a future NGNP reactor design. Conclusions from these calculations are expected to confirm and suggest possible modifications to the current particle fuel parameters specified in the evolving Fuel Specification. Calculated results dispel the need for a binary fuel particle system, which is proposed in the General Atomics GT-MHR concept. The GT-MHR binary system is composed of both a fissile and fertile particle with 350- and 500- micron kernel diameters, respectively. For the NGNP reactor, a single fissile particle system (single UCO kernel size) can meet the reactivity and power cycle length requirements demanded of the NGNP. At the same time, it will provide substantial programmatic cost savings by eliminating the need for dual particle fabrication process lines and dual fuel particle irradiation tests required of a binary system. Use of a larger 425-micron kernel diameter single fissile particle (proposed here), as opposed to the 350-micron GT-MHR fissile particle size, helps alleviate current compact particle packing fractions fabrication limitations (<35%), improves fuel block loading for higher n-batch reload options, and tracks the historical correlation between particle size and enrichment (10 and 14 wt% U-235 particle enrichments are proposed for the NGNP). Overall, the use of the slightly larger kernel significantly broadens the NGNP reactor core design envelope and provides increased design margin to accommodate the (as yet) unknown final NGNP reactor design. Maximum power-peaking factors are calculated for both the initial and equilibrium NGNP cores. Radial power-peaking can be fully controlled with particle packing fraction zoning (no enrichment zoning required) and discrete burnable poison rods. Optimally loaded NGNP cores can expect radial powerpeaking factors as low as 1.14 at beginning of cycle (BOC), increasing slowly