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Note: This page contains sample records for the topic "fuel nuclear reactors" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
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We encourage you to perform a real-time search of NLEBeta
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1

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

2

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

3

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

4

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

5

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

6

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

7

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

8

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

9

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

10

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

11

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

12

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

13

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

14

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

15

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

16

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

17

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

18

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

19

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

20

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

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


21

Nuclear Fuels  

Science Journals Connector (OSTI)

The core of a nuclear reactor is composed of a controlled critical configuration of a fissile material, which in strict a sense is the fuel. This fissile material is contained in a matrix, normally a ceramic c...

Rudy J. M. Konings; Thierry Wiss…

2011-01-01T23:59:59.000Z

22

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

23

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

Science Journals Connector (OSTI)

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

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

24

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

25

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

26

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

27

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

28

Spark Plasma Sintering of Fuel Cermets for Nuclear Reactor Applications  

SciTech Connect (OSTI)

The feasibility of the fabrication of tungsten based nuclear fuel cermets via Spark Plasma Sintering (SPS) is investigated in this work. CeO2 is used to simulate fuel loadings of UO2 or Mixed-Oxide (MOX) fuels within tungsten-based cermets due to the similar properties of these materials. This study shows that after a short time sintering, greater than 90 % density can be achieved, which is suitable to possess good strength as well as the ability to contain fission products. The mechanical properties and the densities of the samples are also investigated as functions of the applied pressures during the sintering.

Yang Zhong; Robert C. O'Brien; Steven D. Howe; Nathan D. Jerred; Kristopher Schwinn; Laura Sudderth; Joshua Hundley

2011-11-01T23:59:59.000Z

29

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

30

Fresh and Spent Nuclear Fuel Repatriation from the IRT-2000 Research Reactor Facility, Sofia, Bulgaria  

SciTech Connect (OSTI)

The IRT 2000 research reactor, operated by the Bulgarian Institute for Nuclear Research and Nuclear Energy (INRNE), safely shipped all of their Russian-origin nuclear fuel from the Republic of Bulgaria to the Russian Federation beginning in 2003 and completing in 2008. These fresh and spent fuel shipments removed all highly enriched uranium (HEU) from Bulgaria. The fresh fuel was shipped by air in December 2003 using trucks and a commercial cargo aircraft. One combined spent fuel shipment of HEU and low enriched uranium (LEU) was completed in July 2008 using high capacity VPVR/M casks transported by truck, barge, and rail. The HEU shipments were assisted by the Russian Research Reactor Fuel Return Program (RRRFR) and the LEU spent fuel shipment was funded by Bulgaria. This report describes the work, approvals, organizations, equipment, and agreements required to complete these shipments and concludes with several major lessons learned.

K. J. Allen; T. G. Apostolov; I. S. Dimitrov

2009-03-01T23:59:59.000Z

31

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

32

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

33

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

34

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

35

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

36

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

SciTech Connect (OSTI)

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

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

1994-10-01T23:59:59.000Z

37

Report to Congress on Plan for Interim Storage of Spent Nuclear Fuel from Decommissioned Reactors  

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

6 6 Report to Congress on the Demonstration of the Interim Storage of Spent Nuclear Fuel from Decommissioned Nuclear Power Reactor Sites December 2008 U.S. Department of Energy Office of Civilian Radioactive Waste Management Washington, D.C. Report to Congress on the Demonstration of the Interim Storage of Spent Nuclear Fuel The picture on the cover is the Connecticut Yankee Independent Spent Fuel Storage Installation site in Haddam, Connecticut, with 43 dry storage NRC-licensed dual-purpose (storage and transport) casks. ii Report to Congress on the Demonstration of the Interim Storage of Spent Nuclear Fuel EXECUTIVE SUMMARY The House Appropriations Committee Print that accompanied the Consolidated Appropriations Act, 2008, requests that the U.S. Department of Energy (the Department):

38

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

SciTech Connect (OSTI)

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

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

2014-06-10T23:59:59.000Z

39

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

DOE Patents [OSTI]

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

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

2013-11-05T23:59:59.000Z

40

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

SciTech Connect (OSTI)

The Department of Energy has completed the Environmental Assessment (EA) of Urgent-Relief Acceptance of Foreign Research Reactor Spent Nuclear Fuel and issued a Finding of No Significant Impact (FONSI) for the proposed action. The EA and FONSI are enclosed for your information. The Department has decided to accept a limited number of spent nuclear fuel elements (409 elements) containing uranium that was enriched in the United States from eight research reactors in Austria, Denmark, Germany, Greece, the Netherlands, Sweden, and Switzerland. This action is necessary to maintain the viability of a major US nuclear weapons nonproliferation program to limit or eliminate the use of highly enriched uranium in civil programs. The purpose of the EA is to maintain the cooperation of the foreign research reactor operators with the nonproliferation program while a more extensive Environmental Impact Statement (EIS) is prepared on a proposed broader policy involving the acceptance of up to 15,000 foreign research reactor spent fuel elements over a 10 to 15 year period. Based on an evaluation of transport by commercial container liner or chartered vessel, five eastern seaboard ports, and truck and train modes of transporting the spent fuel overland to the Savannah River Sits, the Department has concluded that no significant impact would result from any combination of port and made of transport. In addition, no significant impacts were found from interim storage of spent fuel at the Savannah River Site.

Not Available

1994-04-01T23:59:59.000Z

Note: This page contains sample records for the topic "fuel nuclear reactors" 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

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

SciTech Connect (OSTI)

A nuclear fuel-containing plate structure for a nuclear reactor is described; such structure comprising a pair of malleable metallic non-fissionable matrix plates having confronting surfaces which are pressure bonded together and fully united to form a bonded surface, and elongated malleable wire-like fissionable fuel members separately confined and fully enclosed between the matrix plates along the interface to afford a high fuel density as well as structural integrity and effective retention of fission products. The plates have separate recesses formed in the confronting surfaces for closely receiving the wire-like fissionable fuel members. The wire-like fissionable fuel members are made of a maleable uranium alloy capable of being formed into elongated wire-like members and capable of withstanding pressure bonding. The wire-like fissionable fuel members are completely separated and isolated by fully united portions of the interface.

Travelli, A.

1988-01-19T23:59:59.000Z

42

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

43

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

44

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

45

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

46

DOE/EIS-0218-SA-3: Supplement Analysis for the Foreign Research Reactor Spent Nuclear Fuel Acceptance Program (November 2004)  

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

SUPPLEMENT ANALYSIS FOR THE FOREIGN SUPPLEMENT ANALYSIS FOR THE FOREIGN RESEARCH REACTOR SPENT NUCLEAR FUEL ACCEPTANCE PROGRAM NOVEMBER 2004 DOE/EIS-0218-SA-3 U.S. Department of Energy National Nuclear Security Administration Washington, DC Final Supplement Analysis for the Foreign Research Reactor Spent Nuclear Fuel Acceptance Program Final i TABLE OF CONTENTS Page 1. Introduction.............................................................................................................................................. 1 2. Background .............................................................................................................................................. 1 3. The Proposed Action ...............................................................................................................................

47

Apparatus and method for classifying fuel pellets for nuclear reactor  

DOE Patents [OSTI]

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

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

1984-01-01T23:59:59.000Z

48

Nuclear fuel recycling in 4 minutes | Argonne National Laboratory  

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

Nuclear fuel recycling in 4 minutes Share Topic Energy Energy sources Nuclear energy Nuclear fuel cycle Reactors...

49

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

50

Intact and Degraded Component Criticality Calculations of N Reactors Spent Nuclear Fuel  

SciTech Connect (OSTI)

The objective of this calculation is to perform intact and degraded mode criticality evaluations of the Department of Energy's (DOE) N Reactor Spent Nuclear Fuel codisposed in a 2-Defense High-Level Waste (2-DHLW)/2-Multi-Canister Overpack (MCO) Waste Package (WP) and emplaced in a monitored geologic repository (MGR) (see Attachment I). The scope of this calculation is limited to the determination of the effective neutron multiplication factor (k{sub eff}) for both intact and degraded mode internal configurations of the codisposal waste package. This calculation will support the analysis that will be performed to demonstrate the technical viability for disposing of U-metal (N Reactor) spent nuclear fuel in the potential MGR.

L. Angers

2001-01-31T23:59:59.000Z

51

Nuclear reactor fuel assembly duct-tube-to-inlet-nozzle attachment system  

DOE Patents [OSTI]

A reusable system for removably attaching the lower end 21 of a nuclear reactor fuel assembly duct tube to an upper end 11 of a nuclear reactor fuel assembly inlet nozzle. The duct tube's lower end 21 has sides terminating in locking tabs 22 which end in inwardly-extending flanges 23. The flanges 23 engage recesses 13 in the top section 12 of the inlet nozzle's upper end 11. A retaining collar 30 slides over the inlet nozzle's upper end 11 to restrain the flanges 23 in the recesses 13. A locking nut 40 has an inside threaded portion 41 which engages an outside threaded portion 15 of the inlet nozzle's upper end 11 to secure the retaining collar 30 against protrusions 24 on the duct tube's sides.

Christiansen, David W. (Kennewick, WA); Smith, Bob G. (Kennewick, WA)

1982-01-01T23:59:59.000Z

52

Nuclear fuel cycle information workshop  

SciTech Connect (OSTI)

This overview of the nuclear fuel cycle is divided into three parts. First, is a brief discussion of the basic principles of how nuclear reactors work; second, is a look at the major types of nuclear reactors being used and world-wide nuclear capacity; and third, is an overview of the nuclear fuel cycle and the present industrial capability in the US.

Not Available

1983-01-01T23:59:59.000Z

53

A Novel Fuel/Reactor Cycle to Implement the 300 Years Nuclear Waste Policy Approach - 12377  

SciTech Connect (OSTI)

A thorium-based fuel cycle system can effectively burn the currently accumulated commercial used nuclear fuel and move to a sustainable equilibrium where the actinide levels in the high level waste are low enough to yield a radiotoxicity after 300 years lower than that of the equivalent uranium ore. The second step of the Westinghouse approach to solving the waste 'problem' has been completed. The thorium fuel cycle has indeed the potential of burning the legacy TRU and achieve the waste objective proposed. Initial evaluations have been started for the third step, development and selection of appropriate reactors. Indications are that the probability of show-stoppers is rather remote. It is, therefore, believed that development of the thorium cycle and associated technologies will provide a permanent solution to the waste management. Westinghouse is open to the widest collaboration to make this a reality. (authors)

Carelli, M.D.; Franceschini, F.; Lahoda, E.J. [Westinghouse Electric Company LLC., Cranberry Township, PA (United States); Petrovic, B. [Georgia Institute of Technology, Atlanta, GA (United States)

2012-07-01T23:59:59.000Z

54

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

Science Journals Connector (OSTI)

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

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

1993-01-01T23:59:59.000Z

55

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

56

Prospects for and problems of using light-water supercritical-pressure coolant in nuclear reactors in order to increase the efficiency of the nuclear fuel cycle  

SciTech Connect (OSTI)

Trends in the development of the power sector of the Russian and world power industries both at present time and in the near future are analyzed. Trends in the rise of prices for reserves of fossil and nuclear fuels used for electricity production are compared. An analysis of the competitiveness of electricity production at nuclear power plants as compared to the competitiveness of electricity produced at coal-fired and natural-gas-fired thermal power plants is performed. The efficiency of the open nuclear fuel cycle and various versions of the closed nuclear fuel cycle is discussed. The requirements on light-water reactors under the scenario of dynamic development of the nuclear power industry in Russia are determined. Results of analyzing the efficiency of fuel utilization for various versions of vessel-type light-water reactors with supercritical coolant are given. Advantages and problems of reactors with supercritical-pressure water are listed.

Alekseev, P. N.; Semchenkov, Yu. M.; Sedov, A. A., E-mail: sedov@dhtp.kial.ru; Subbotin, S. A.; Chibinyaev, A. V. [Russian Research Centre Kurchatov Institute (Russian Federation)

2011-12-15T23:59:59.000Z

57

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

58

Rethinking the light water reactor fuel cycle  

E-Print Network [OSTI]

The once through nuclear fuel cycle adopted by the majority of countries with operating commercial power reactors imposes a number of concerns. The radioactive waste created in the once through nuclear fuel cycle has to ...

Shwageraus, Evgeni, 1973-

2004-01-01T23:59:59.000Z

59

Spent Nuclear Fuel (SNF) Project Acceptance Criteria for Light Water Reactor Spent Fuel Storage System [OCRWM PER REV2  

SciTech Connect (OSTI)

As part of the decommissioning of the 324 Building Radiochemical Engineering Cells there is a need to remove commercial Light Water Reactor (LWR) spent nuclear fuel (SNF) presently stored in these hot cells. To enable fuel removal from the hot cells, the commercial LWR SNF will be packaged and shipped to the 200 Area Interim Storage Area (ISA) in a manner that satisfies site requirements for SNF interim storage. This document identifies the criteria that the 324 Building Radiochemical Engineering Cell Clean-out Project must satisfy for acceptance of the LWR SNF by the SNF Project at the 200 Area ISA. In addition to the acceptance criteria identified herein, acceptance is contingent on adherence to applicable Project Hanford Management Contract requirements and procedures in place at the time of work execution.

JOHNSON, D.M.

2000-12-20T23:59:59.000Z

60

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

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


61

Nuclear reactor fuel assembly duct-tube-to-handling-socket attachment system  

DOE Patents [OSTI]

A reusable system for removably attaching the upper end 10of a nuclear reactor duct tube to the lower end 30 of a nuclear reactor fuel assembly handling socket. A transition ring 20, fixed to the duct tube's upper end 10, has an interior-threaded section 22 with a first locking hole segment 24. An adaptor ring 40, fixed to the handling socket's lower end 30 has an outside-threaded section 42 with a second locking hole segment 44. The inside 22 and outside 42 threaded sections match and can be joined so that the first 24 and second 44 locking hole segments can be aligned to form a locking hole. A locking ring 50, with a locking pin 52, slides over the adaptor ring 40 so that the locking pin 52 fits in the locking hole. A swage lock 60 or a cantilever finger lock 70 is formed from the locking cup collar 26 to fit in a matching groove 54 or 56 in the locking ring 50 to prevent the locking ring's locking pin 52 from backing out of the locking hole.

Christiansen, David W. (Kennewick, WA); Smith, Bob G. (Kennewick, WA)

1982-01-01T23:59:59.000Z

62

Advances in Metallic Nuclear Fuel  

Science Journals Connector (OSTI)

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

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

1987-04-01T23:59:59.000Z

63

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

64

Analysis of Transuranic Mixed Oxide Fuel in a CANDU Nuclear Reactor.  

E-Print Network [OSTI]

?? The reprocessing of spent fuel is a key component in reducing the end waste from nuclear power plant operations and creating a sustainable closed… (more)

Morreale, Andrew C

2013-01-01T23:59:59.000Z

65

Review of experiments to evaluate the ability of electrical heater rods to simulate nuclear fuel rod behavior during postulated loss-of-coolant accidents in light water reactors  

SciTech Connect (OSTI)

Issues related to using electrical fuel rod simulators to simulate nuclear fuel rod behavior during postulated loss-of-coolant accident (LOCA) conditions in light water reactors are summarized. Experimental programs which will provide a data base for comparing electrical heater rod and nuclear fuel rod LOCA responses are reviewed.

McPherson, G D; Tolman, E L

1980-01-01T23:59:59.000Z

66

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

DOE Patents [OSTI]

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

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

1993-01-01T23:59:59.000Z

67

A long term radiological risk model for plutonium-fueled and fission reactor space nuclear system  

SciTech Connect (OSTI)

This report describes the optimization of the RISK III mathematical model, which provides risk assessment for the use of a plutonium-fueled, fission reactor in space systems. The report discusses possible scenarios leading to radiation releases on the ground; distinctions are made for an intact reactor and a dispersed reactor. Also included are projected dose equivalents for various accident situations. 54 refs., 31 figs., 11 tabs. (TEM)

Bartram, B.W.; Dougherty, D.K.

1987-01-01T23:59:59.000Z

68

Licensing issues associated with the use of mixed-oxide fuel in US commercial nuclear reactors  

SciTech Connect (OSTI)

On January 14, 1997, the Department of Energy, as part of its Record of Decision on the storage and disposition of surplus nuclear weapons materials, committed to pursue the use of excess weapons-usable plutonium in the fabrication of mixed-oxide (MOX) fuel for consumption in existing commercial nuclear power plants. Domestic use of MOX fuel has been deferred since the late 1970s, principally due to nuclear proliferation concerns. This report documents a review of past and present literature (i.e., correspondence, reports, etc.) on the domestic use of MOX fuel and provides discussion on the technical and regulatory issues that must be addressed by DOE (and the utility/consortia selected by DOE to effect the MOX fuel consumption strategy) in obtaining approval from the Nuclear Regulatory Commission to use MOX fuel in one or a group of existing commercial nuclear power plants.

Williams, D.L. Jr.

1997-04-01T23:59:59.000Z

69

Nuclear Fuels | Department of Energy  

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

Nuclear Fuels Nuclear Fuels Nuclear Fuels A reactor's ability to produce power efficiently is significantly affected by the composition and configuration of its fuel system. A nuclear fuel assembly consists of hundreds of thousands of uranium pellets, stacked and encapsulated within tubes called fuel rods or fuel pins which are then bundled together in various geometric arrangements. There are many design considerations for the material composition and geometric configuration of the various components comprising a nuclear fuel system. Future designs for the fuel and the assembly or packaging of fuel will contribute to cleaner, cheaper and safer nuclear energy. Today's process for developing and testing new fuel systems is resource and time intensive. The process to manufacture the fuel, build an assembly,

70

Uranium to Electricity: The Chemistry of the Nuclear Fuel Cycle  

Science Journals Connector (OSTI)

The nuclear fuel cycle consists of a series of industrial processes that produce fuel for the production of electricity in nuclear reactors, use the fuel to generate electricity, and subsequently manage the spent reactor fuel. While the physics and ...

Frank A. Settle

2009-03-01T23:59:59.000Z

71

Radiation Exposures Associated with Shipments of Foreign Research Reactor Spent Nuclear Fuel  

SciTech Connect (OSTI)

Experience has shown that the analyses of marine transport of spent fuel in the Environmental Impact Statement (EIS) were conservative. It is anticipated that for most shipments. The external dose rate for the loaded transportation cask will be more in line with recent shipments. At the radiation levels associated with these shipments, we would not expect any personnel to exceed radiation exposure limits for the public. Package dose rates usually well below the regulatory limits and personnel work practices following ALARA principles are keeping human exposures to minimal levels. However, the potential for Mure shipments with external dose rates closer to the exclusive-use regulatory limit suggests that DOE should continue to provide a means to assure that individual crew members do not receive doses in excess of the public dose limits. As a minimum, the program will monitor cask dose rates and continue to implement administrative procedures that will maintain records of the dose rates associated with each shipment, the vessel used, and the crew list for the vessel. DOE will continue to include a clause in the contract for shipment of the foreign research reactor spent nuclear fuel requiring that the Mitigation Action Plan be followed.

MASSEY,CHARLES D.; MESSICK,C.E.; MUSTIN,T.

1999-11-01T23:59:59.000Z

72

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

73

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

SciTech Connect (OSTI)

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

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

2013-09-11T23:59:59.000Z

74

Nuclear-fuel-cycle risk assessment: descriptions of representative non-reactor facilities. Sections 1-14  

SciTech Connect (OSTI)

The Fuel Cycle Risk Assessment Program was initiated to provide risk assessment methods for assistance in the regulatory process for nuclear fuel cycle facilities other than reactors. This report, the first from the program, defines and describes fuel cycle elements that are being considered in the program. One type of facility (and in some cases two) is described that is representative of each element of the fuel cycle. The descriptions are based on real industrial-scale facilities that are current state-of-the-art, or on conceptual facilities where none now exist. Each representative fuel cycle facility is assumed to be located on the appropriate one of four hypothetical but representative sites described. The fuel cycles considered are for Light Water Reactors with once-through flow of spent fuel, and with plutonium and uranium recycle. Representative facilities for the following fuel cycle elements are described for uranium (or uranium plus plutonium where appropriate): mining, milling, conversion, enrichment, fuel fabrication, mixed-oxide fuel refabrication, fuel reprocessing, spent fuel storage, high-level waste storage, transuranic waste storage, spent fuel and high-level and transuranic waste disposal, low-level and intermediate-level waste disposal, and transportation. For each representative facility the description includes: mainline process, effluent processing and waste management, facility and hardware description, safety-related information and potential alternative concepts for that fuel cycle element. The emphasis of the descriptive material is on safety-related information. This includes: operating and maintenance requirements, input/output of major materials, identification and inventories of hazardous materials (particularly radioactive materials), unit operations involved, potential accident driving forces, containment and shielding, and degree of hands-on operation.

Schneider, K.J.

1982-09-01T23:59:59.000Z

75

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

76

Plutonium partitioning in uranium and plutonium co-recovery system for fast reactor fuel recycling with enhanced nuclear proliferation resistance  

SciTech Connect (OSTI)

For enhancement of nuclear proliferation resistance, a 'co-processing' method for U and Pu co-recovery was studied. Two concepts, no U scrubbing and no Pu reduction partitioning, were employed to formulate two types of flow sheets by using a calculation code. Their process performance was demonstrated using radioactive solutions derived from an irradiated fast reactor fuel. These experimental results indicated that U and Pu were co-recovered in the U/Pu product, and the Pu content in the U/Pu product increased approximately 2.3 times regardless of using reductant. The proposed no U scrubbing and no Pu reductant flow sheet is applicable to fast reactor fuel reprocessing and enhances its resistance to nuclear proliferation. (authors)

Nakahara, Masaumi; Koma, Yoshikazu; Nakajima, Yasuo [Japan Atomic Energy Agency: 4-33 Muramatsu, Naka-gun, Tokai-mura, Ibaraki 319-1194 (Japan)

2013-07-01T23:59:59.000Z

77

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

78

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

79

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 to a value of 1.25 by end of cycle (EOC), an axial power-peaking value of 1.30 (BOC), and for individual fuel particles in the maximum compact <1.05 (BOC) and an approximate value of 1.20 (EOC) due to Pu-239 buildup in particles on the compact periphery. The NGNP peak particle powers, using a conservative total power-peaking factor (~2.1 factor), are expected to be <150 mW/particle (well below the 250 mW/particle limit, even with the larger 425-micron kernel size).

James W. Sterbentz; Bren Phillips; Robert L. Sant; Gray S. Chang; Paul D. Bayless

2003-09-01T23:59:59.000Z

80

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

Science Journals Connector (OSTI)

......corresponding to the highest reported input power are shown in a "parenthesis...data were obtained using an input pulse with a pulse power of...above parenthesis, using an input pulse with pulse power of 1015...there is the Laser Welding nuclear fusion system using "solid Pycnodeuterium......

Yoshiaki Arata; Yue Chang Zhang

2004-02-01T23:59:59.000Z

Note: This page contains sample records for the topic "fuel nuclear reactors" 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

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...

82

Corrosion Minimization for Research Reactor Fuel  

SciTech Connect (OSTI)

Existing university research reactors are being converted to use low-enriched uranium fue to eliminate the use of highly-enriched uranium. These conversions require increases in fuel loading that will result in the use of elements with more fuel plates, resulting in a net decrease in the water annulus between fuel plates. The proposed decrease in the water annulus raises questions about the requirements and stability of the surface hydroxide on the aluminum fuel cladding and the potential for runaway corrosion resulting in fuel over-temperature incidents. The Nuclear Regulatory Commission (NRC), as regulator for these university reactors, must ensure that proposed fuel modifications will not result in any increased risk or hazard to the reactor operators or the public. This document reviews the characteristics and behavior of aluminum hydroxides, analyzes the drivers for fuel plate corrosion, reviews relevant historical incidents, and provides recommendations on fuel design, surface treatment, and reactor operational practices to avoid corrosion issues.

Eric Shaber; Gerard Hofman

2005-06-01T23:59:59.000Z

83

Fundamental Processes of Coupled Radiation Damage and Mechanical Behavior in Nuclear Fuel Materials for High Temperature Reactors  

SciTech Connect (OSTI)

The objective of this work has been to elucidate the relationship among microstructure, radiation damage and mechanical properties for nuclear fuel materials. As representative nuclear materials, we have taken an hcp metal (Mg as a generic metal, and Ti alloys for fast reactors) and UO2 (representing fuel). The degradation of the thermo-mechanical behavior of nuclear fuels under irradiation, both the fissionable material itself and its cladding, is a longstanding issue of critical importance to the nuclear industry. There are experimental indications that nanocrystalline metals and ceramics may be more resistant to radiation damage than their coarse-grained counterparts. The objective of this project look at the effect of microstructure on radiation damage and mechanical behavior in these materials. The approach to be taken was state-of-the-art, large-scale atomic-level simulation. This systematic simulation program of the effects of irradiation on the structure and mechanical properties of polycrystalline Ti and UO2 identified radiation damage mechanisms. Moreover, it will provided important insights into behavior that can be expected in nanocrystalline microstructures and, by extension, nanocomposites. The fundamental insights from this work can be expected to help in the design microstructures that are less susceptible to radiation damage and thermomechanical degradation.

Simon Phillpot; James Tulenko

2011-09-08T23:59:59.000Z

84

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

85

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

86

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

87

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

88

EQ6 Calculations for Chemical Degradation Of N Reactor (U-Metal) Spent Nuclear Fuel Waste Packages  

SciTech Connect (OSTI)

The Monitored Geologic Repository (MGR) Waste Package Department of the Civilian Radioactive Waste Management System Management & Operating Contractor (CRWMS M&O) performed calculations to provide input for disposal of spent nuclear fuel (SNF) from the N Reactor, a graphite moderated reactor at the Department of Energy's (DOE) Hanford Site (ref. 1). The N Reactor core was fueled with slightly enriched (0.947 wt% and 0.947 to 1.25 wt% {sup 235}U in Mark IV and Mark IA fuels, respectively) U-metal clad in Zircaloy-2 (Ref. 1, Sec. 3). Both types of N Reactor SNF have been considered for disposal at the proposed Yucca Mountain site. For some WPs, the outer shell and inner shell may breach (Ref. 3) allowing the influx of water. Water in the WP will moderate neutrons, increasing the likelihood of a criticality event within the WP; and the water may, in time, gradually leach the fissile components from the WP, further affecting the neutronics of the system. This study presents calculations of the long-term geochemical behavior of WPs containing two multi-canister overpacks (MCO) with either six baskets of Mark IA or five baskets of Mark IV intact N Reactor SNF rods (Ref. 1, Sec. 4) and two high-level waste (HLW) glass pour canisters (GPCs) arranged according to the codisposal concept (Ref. 4). The specific study objectives were to determine: (1) The extent to which fissile uranium will remain in the WP after corrosion/dissolution of the initial WP configuration (2) The extent to which fissile uranium will be carried out of the degraded WP by infiltrating water (such that internal criticality is no longer possible, but the possibility of external criticality may be enhanced); and (3) The nominal chemical composition for the criticality evaluations of the WP design, and to suggest the range of parametric variations for additional evaluations. The scope of this calculation, the chemical compositions (and subsequent criticality evaluations) of the simulations, is limited to time periods up to 6.35 x 10{sup 5} years. This longer time frame is closer to the one million year time horizon recently recommended by the National Academy of Sciences to the Environmental Protection Agency for performance assessment related to a nuclear repository (Ref. 5). However, it is important to note that after 100,000 years, most of the materials of interest (fissile materials) will have either been removed from the WP, reached a steady state, or been transmuted.

P. Bernot

2001-02-27T23:59:59.000Z

89

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

SciTech Connect (OSTI)

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

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

2011-12-29T23:59:59.000Z

90

Advanced nuclear fuel  

SciTech Connect (OSTI)

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

Terrani, Kurt

2014-07-14T23:59:59.000Z

91

Advanced nuclear fuel  

ScienceCinema (OSTI)

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

Terrani, Kurt

2014-07-15T23:59:59.000Z

92

The breeder reactor: a fossil fuel viewpoint  

Science Journals Connector (OSTI)

... elegant and simple: to generate electricity and, at the same time, to produce additional fuel from the uranium discarded by the existing thermal reactor system. Without the breeder reactor, ... seems likely that the role of nuclear energy will begin to be constrained by the price and availability of uranium at about the turn of the century. There is, however ...

David Merrick

1976-12-16T23:59:59.000Z

93

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

94

Nuclear Criticality Control and Safety of Plutonium-Uranium Fuel Mixtures Outside Reactors  

SciTech Connect (OSTI)

The ANSI/ANS 8.12 standard was first approved in July 1978. At that time, this edition was applicable to operations with plutonium-uranium oxide (MOX) fuel mixtures outside reactors and was limited to subcritical limits for homogeneous systems. The next major revision, ANSI/ANS-8.12-1987, included the addition of subcritical limits for heterogeneous systems. The standard was subsequently reaffirmed in February 1993. During late 1990s, substantial work was done by the ANS 8.12 Standard Working Group to re-examine the technical data presented in the standard using the latest codes and cross section sets. Calculations performed showed good agreement with the values published in the standard. This effort resulted in the reaffirmation of the standard in March 2002. The standard is currently in a maintenance mode. After 2002, activities included discussions to determine the future direction of the standard and to follow the MOX standard development by the International Standard Organization (ISO). In 2007, the Working Group decided to revise the standard to extend the areas of applicability by providing a wider range of subcritical data. The intent is to cover a wider domain of MOX fuel fabrication and operations. It was also decided to follow the ISO MOX standard specifications (related to MOX density and isotopics) and develop a new set of subcritical limits for homogeneous systems. This has resulted in the submittal (and subsequent approval) of the project initiation notification system form (PINS) in 2007.

Biswas, D; Mennerdahl, D

2008-06-23T23:59:59.000Z

95

Double-clad nuclear fuel safety rod  

DOE Patents [OSTI]

A device for shutting down a nuclear reactor during an undercooling or overpower event, whether or not the reactor's scram system operates properly. This is accomplished by double-clad fuel safety rods positioned at various locations throughout the reactor core, wherein melting of a secondary internal cladding of the rod allows the fuel column therein to shift from the reactor core to place the reactor in a subcritical condition.

McCarthy, William H. (Los Altos, CA); Atcheson, Donald B. (Cupertino, CA); Vaidyanathan, Swaminathan (San Jose, CA)

1984-01-01T23:59:59.000Z

96

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

E-Print Network [OSTI]

have been irradiated to the desired burnup in the Oak Ridge National Laboratory- High Flux Isotope Reactor (ORNL-HFIR), and then separated using the PUREX process to experimentally determine the intrinsic signature of the fuel. The experimental data...

Coles, Taylor Marie

2014-04-27T23:59:59.000Z

97

Nuclear Fuel Cycle Integrated System Analysis  

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

Fuel Cycle Integrated System Analysis Fuel Cycle Integrated System Analysis Abdellatif M. Yacout Argonne National Laboratory Nuclear Engineering Division The nuclear fuel cycle is a complex system with multiple components and activities that are combined to provide nuclear energy to a variety of end users. The end uses of nuclear energy are diverse and include electricity, process heat, water desalination, district heating, and possibly future hydrogen production for transportation and energy storage uses. Components of the nuclear fuel cycle include front end components such as uranium mining, conversion and enrichment, fuel fabrication, and the reactor component. Back end of the fuel cycle include used fuel coming out the reactor, used fuel temporary and permanent storage, and fuel reprocessing. Combined with those components there

98

Fossil fuel furnace reactor  

DOE Patents [OSTI]

A fossil fuel furnace reactor is provided for simulating a continuous processing plant with a batch reactor. An internal reaction vessel contains a batch of shale oil, with the vessel having a relatively thin wall thickness for a heat transfer rate effective to simulate a process temperature history in the selected continuous processing plant. A heater jacket is disposed about the reactor vessel and defines a number of independent controllable temperature zones axially spaced along the reaction vessel. Each temperature zone can be energized to simulate a time-temperature history of process material through the continuous plant. A pressure vessel contains both the heater jacket and the reaction vessel at an operating pressure functionally selected to simulate the continuous processing plant. The process yield from the oil shale may be used as feedback information to software simulating operation of the continuous plant to provide operating parameters, i.e., temperature profiles, ambient atmosphere, operating pressure, material feed rates, etc., for simulation in the batch reactor.

Parkinson, William J. (Los Alamos, NM)

1987-01-01T23:59:59.000Z

99

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

100

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

E-Print Network [OSTI]

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

Pilat, E. E., 1937-

1967-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "fuel nuclear reactors" 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

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

102

Recovery of weapon plutonium as feed material for reactor fuel  

SciTech Connect (OSTI)

This report presents preliminary considerations for recovering and converting weapon plutonium from various US weapon forms into feed material for fabrication of reactor fuel elements. An ongoing DOE study addresses the disposition of excess weapon plutonium through its use as fuel for nuclear power reactors and subsequent disposal as spent fuel. The spent fuel would have characteristics similar to those of commercial power spent fuel and could be similarly disposed of in a geologic repository.

Armantrout, G.A.; Bronson, M.A.; Choi, Jor-Shan [and others

1994-03-16T23:59:59.000Z

103

Worker exposure for at-reactor management of spent nuclear fuel  

Science Journals Connector (OSTI)

......Impact Statement for the proposed Yucca Mountain repository(6,7); potential...a No Action Alternative in the Yucca Mountain Environmental Impact Statement...operations at reactor sites. FUNDING This research was funded by the......

Philippe F. Weck

2013-09-01T23:59:59.000Z

104

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

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

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

105

Fuel cycle options for optimized recycling of nuclear fuel  

E-Print Network [OSTI]

The accumulation of transuranic inventories in spent nuclear fuel depends on both deployment of advanced reactors that can be loaded with recycled transuranics (TRU), and on availability of the facilities that separate and ...

Aquien, Alexandre

2006-01-01T23:59:59.000Z

106

Nuclear Fuel Cycle | Department of Energy  

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

Cycle Cycle Nuclear Fuel Cycle This is an illustration of a nuclear fuel cycle that shows the required steps to process natural uranium from ore for preparation for fuel to be loaded in nuclear reactors. This is an illustration of a nuclear fuel cycle that shows the required steps to process natural uranium from ore for preparation for fuel to be loaded in nuclear reactors. The mission of NE-54 is primarily focused on activities related to the front end of the nuclear fuel cycle which includes mining, milling, conversion, and enrichment. Uranium Mining Both "conventional" open pit, underground mining, and in situ techniques are used to recover uranium ore. In general, open pit mining is used where deposits are close to the surface and underground mining is used

107

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

108

Hydrogen : what fuel cell vehicles and advanced nuclear reactors have in common  

E-Print Network [OSTI]

This thesis reports on two technology and policy issues directly related to hydrogen economy. The first issue concentrates on the end-use application of hydrogen as a transportation fuel, and deals with the following ...

Demirdöven, Nurettin, 1974-

2005-01-01T23:59:59.000Z

109

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

110

Overview of the nuclear fuel cycle  

SciTech Connect (OSTI)

The use of nuclear reactors to provide electrical energy has shown considerable growth since the first nuclear plant started commercial operation in the mid 1950s. Although the main purpose of this paper is to review the fuel cycle capabilities in the United States, the introduction is a brief review of the types of nuclear reactors in use and the world-wide nuclear capacity.

Leuze, R.E.

1982-01-01T23:59:59.000Z

111

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

E-Print Network [OSTI]

Energy July 29, 2014 search for loading patterns (LPs) that maximized the energy production over a cycle, sub-15 ject to constraints on power peaking and fuel burn-up. Kim et al. (1987) developed a16 two-stage procedure for maximizing cycle length...

Hill, Natasha J.; Parks, Geoffrey T.

2014-08-24T23:59:59.000Z

112

Nuclear-fuel-cycle risk assessment: descriptions of representative non-reactor facilities, Sections 15-19  

SciTech Connect (OSTI)

Information is presented under the following section headings: fuel reprocessing; spent fuel and high-level and transuranic waste storage; spent fuel and high-level and transuranic waste disposal; low-level and intermediate-level waste disposal; and, transportation of radioactive materials in the nuclear fuel cycle. In each of the first three sections a description is given on the mainline process, effluent processing and waste management systems, plant layout, and alternative process schemes. Safety information and a summary are also included in each. The section on transport of radioactive materials includes information on the transportation of uranium ore, uranium ore concentrate, UF/sub 6/, PuO/sub 2/ powder, unirradiated uranium and mixed-oxide fuel assemblies, spent fuel, solidified high-level waste, contact-handled transuranic waste, remote-handled transuranic waste, and low and intermediate level nontransuranic waste. A glossary is included. (JGB)

Schneider, K.J.

1982-09-01T23:59:59.000Z

113

AB INITIO STUDY OF ADVANCED METALLIC NUCLEAR FUELS FOR FAST BREEDER REACTORS  

SciTech Connect (OSTI)

Density-functional formalism is applied to study the ground state properties of {gamma}-U-Zr and {gamma}-U-Mo solid solutions. Calculated heats of formation are compared with CALPHAD assessments. We discuss how the heat of formation in both alloys correlates with the charge transfer between the alloy components. The decomposition curves for {gamma}-based U-Zr and U-Mo solid solutions are derived from Ising-type Monte Carlo simulations. We explore the idea of stabilization of the {delta}-UZr{sub 2} compound against the {alpha}-Zr (hcp) structure due to increase of Zr d-band occupancy by the addition of U to Zr. We discuss how the specific behavior of the electronic density of states in the vicinity of the Fermi level promotes the stabilization of the U{sub 2}Mo compound. The mechanism of possible Am redistribution in the U-Zr and U-Mo fuels is also discussed.

Landa, A; Soderlind, P; Grabowski, B; Turchi, P A; Ruban, A V; Vitos, L

2012-04-23T23:59:59.000Z

114

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

115

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

116

Environmental assessment for the manufacture and shipment of nuclear reactor fuel from the United States to Canada  

SciTech Connect (OSTI)

The US Department of Energy (DOE) has declared 41.9 tons (38 metric tons) of weapons-usable plutonium surplus to the United States` defense needs. A DOE Programmatic Environmental Impact Statement analyzed strategies for plutonium storage and dispositioning. In one alternative, plutonium as a mixed oxide (MOX) fuel would be irradiated (burned) in a reengineered heavy-water-moderated reactor, such as the Canadian CANDU design. In an Environmental Assessment (EA), DOE proposes to fabricate and transport to Canada a limited amount of MOX fuel as part of the Parallex (parallel experiment) Project. MOX fuel from the US and Russia would be used by Canada to conduct performance tests at Chalk River Laboratories. MOX fuel would be fabricated at Los Alamos National Laboratory and transported in approved container(s) to a Canadian port(s) of entry on one to three approved routes. The EA analyzes the environmental and human health effects from MOX fuel fabrication and transportation. Under the Proposed Action, MOX fuel fabrication would not result in adverse effects to the involved workers or public. Analysis showed that the shipment(s) of MOX fuel would not adversely affect the public, truck crew, and environment along the transportation routes.

Rangel, R.C.

1999-02-01T23:59:59.000Z

117

Spent fuel utilization in a compact traveling wave reactor  

SciTech Connect (OSTI)

In recent years, several innovative designs of nuclear reactors are proposed. One of them is Traveling Wave Reactor (TWR). The unique characteristic of a TWR is the capability of breeding its own fuel in the reactor. The reactor is fueled by mostly depleted, natural uranium or spent nuclear fuel and a small amount of enriched uranium to initiate the fission process. Later on in the core, the reactor gradually converts the non-fissile material into the fissile in a process like a traveling wave. In this work, a TWR with spent nuclear fuel blanket was studied. Several parameters such as reactivity coefficients, delayed neutron fraction, prompt neutron generation lifetime, and fission power, were analyzed. The discharge burnup composition was also analyzed. The calculation is performed by a continuous energy Monte Carlo code McCARD.

Hartanto, Donny; Kim, Yonghee [Korea Advanced Institute of Science and Technology 373-1 Kusong-dong, Yuseong-gu, Daejeon, 305-701 (Korea, Republic of)

2012-06-06T23:59:59.000Z

118

Spent nuclear fuel reprocessing modeling  

SciTech Connect (OSTI)

The long-term wide development of nuclear power requires new approaches towards the realization of nuclear fuel cycle, namely, closed nuclear fuel cycle (CNFC) with respect to fission materials. Plant nuclear fuel cycle (PNFC), which is in fact the reprocessing of spent nuclear fuel unloaded from the reactor and the production of new nuclear fuel (NF) at the same place together with reactor plant, can be one variant of CNFC. Developing and projecting of PNFC is a complicated high-technology innovative process that requires modern information support. One of the components of this information support is developed by the authors. This component is the programme conducting calculations for various variants of process flow sheets for reprocessing SNF and production of NF. Central in this programme is the blocks library, where the blocks contain mathematical description of separate processes and operations. The calculating programme itself has such a structure that one can configure the complex of blocks and correlations between blocks, appropriate for any given flow sheet. For the ready sequence of operations balance calculations are made of all flows, i.e. expenses, element and substance makeup, heat emission and radiation rate are determined. The programme is open and the block library can be updated. This means that more complicated and detailed models of technological processes will be added to the library basing on the results of testing processes using real equipment, in test operating mode. The development of the model for the realization of technical-economic analysis of various variants of technologic PNFC schemes and the organization of 'operator's advisor' is expected. (authors)

Tretyakova, S.; Shmidt, O.; Podymova, T.; Shadrin, A.; Tkachenko, V. [Bochvar Institute, 5 Rogova str., Moscow 123098 (Russian Federation); Makeyeva, I.; Tkachenko, V.; Verbitskaya, O.; Schultz, O.; Peshkichev, I. [Russian Federal Nuclear Center - VNIITF E.I. Zababakhin, p.o.box 245, Snezhinsk, 456770 (Russian Federation)

2013-07-01T23:59:59.000Z

119

Spent Fuel Working Group report on inventory and storage of the Department`s spent nuclear fuel and other reactor irradiated nuclear materials and their environmental, safety and health vulnerabilities. Volume 3, Site team reports  

SciTech Connect (OSTI)

A self assessment was conducted of those Hanford facilities that are utilized to store Reactor Irradiated Nuclear Material, (RINM). The objective of the assessment is to identify the Hanford inventories of RINM and the ES & H concerns associated with such storage. The assessment was performed as proscribed by the Project Plan issued by the DOE Spent Fuel Working Group. The Project Plan is the plan of execution intended to complete the Secretary`s request for information relevant to the inventories and vulnerabilities of DOE storage of spent nuclear fuel. The Hanford RINM inventory, the facilities involved and the nature of the fuel stored are summarized. This table succinctly reveals the variety of the Hanford facilities involved, the variety of the types of RINM involved, and the wide range of the quantities of material involved in Hanford`s RINM storage circumstances. ES & H concerns are defined as those circumstances that have the potential, now or in the future, to lead to a criticality event, to a worker radiation exposure event, to an environmental release event, or to public announcements of such circumstances and the sensationalized reporting of the inherent risks.

Not Available

1993-11-01T23:59:59.000Z

120

Fresh nuclear fuel measurements at Ukrainian nuclear power plants  

SciTech Connect (OSTI)

In 2005, the Provisions on Nuclear Material Measurement System was enacted in Ukraine as an important regulatory driver to support international obligations in nuclear safeguards and nonproliferation. It defines key provisions and requirements for material measurement and measurement control programs to ensure the quality and reliability of measurement data within the framework of the State MC&A System. Implementing the Provisions requires establishing a number of measurement techniques for both fresh and spent nuclear fuel for various types of Ukrainian reactors. Our first efforts focused on measurements of fresh nuclear fuel from a WWR-1000 power reactor.

Kuzminski, Jozef [Los Alamos National Laboratory; Ewing, Tom [ANL; Dickman, Debbie [PNNL; Gavrilyuk, Victor [UKRAINE; Drapey, Sergey [UKRAINE; Kirischuk, Vladimir [UKRAINE; Strilchuk, Nikolay [UKRAINE

2009-01-01T23:59:59.000Z

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


121

Nuclear proliferation and civilian nuclear power. Report of the Nonproliferation Alternative Systems Assessment Program. Volume IX. Reactor and fuel cycle description  

SciTech Connect (OSTI)

The Nonproliferation Alterntive Systems Assessment Program (NASAP) has characterized and assessed various reactor/fuel-cycle systems. Volume IX provides, in summary form, the technical descriptions of the reactor/fuel-cycle systems studied. This includes the status of the system technology, as well as a discussion of the safety, environmental, and licensing needs from a technical perspective. This information was then used in developing the research, development, and demonstration (RD and D) program, including its cost and time frame, to advance the existing technology to the level needed for commercial use. Wherever possible, the cost data are given as ranges to reflect the uncertainties in the estimates.

Not Available

1980-06-01T23:59:59.000Z

122

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

123

EIS-0203: Spent Nuclear Fuel Management and Idaho National Engineering  

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

EIS-0203: Spent Nuclear Fuel Management and Idaho National EIS-0203: Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs EIS-0203: Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs SUMMARY This EIS considers programmatic (DOE-wide) alternative approaches to safely, efficiently, and responsibly manage existing and projected quantities of spent nuclear fuel until the year 2035. This amount of time may be required to make and implement a decision on the ultimate disposition of spent nuclear fuel. DOE's spent nuclear fuel responsibilities include fuel generated by DOE production, research, and development reactors; naval reactors; university and foreign research reactors; domestic non-DOE reactors such as those at the National Institute

124

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

125

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

126

Alternate-fuel reactor studies  

SciTech Connect (OSTI)

A number of studies related to improvements and/or greater understanding of alternate-fueled reactors is presented. These studies cover the areas of non-Maxwellian distributions, materials and lifetime analysis, a /sup 3/He-breeding blanket, tritium-rich startup effects, high field magnet support, and reactor operation spanning the range from full D-T operation to operation with no tritium breeding.

Evans, K. Jr.; Ehst, D.A.; Gohar, Y.; Jung, J.; Mattas, R.F.; Turner, L.R.

1983-02-01T23:59:59.000Z

127

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

128

Fuel Cycle Science & Technology | Nuclear Science | ORNL  

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

Advanced Fuel Cycle Systems Radiochemical Separation & Processing Recycle & Waste Management Uranium Enrichment Used Nuclear Fuel Storage, Transportation, and Disposal Fusion Nuclear Science Isotope Development and Production Nuclear Security Science & Technology Nuclear Systems Modeling, Simulation & Validation Nuclear Systems Technology Reactor Technology Nuclear Science Home | Science & Discovery | Nuclear Science | Research Areas | Fuel Cycle Science & Technology SHARE Fuel Cycle Science and Technology The ORNL expertise and experience across the entire nuclear fuel cycle is underpinned by extensive facilities and a comprehensive modeling and simulation capability ORNL supports the understanding, development, evaluation and deployment of

129

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

130

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.

131

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

Science Journals Connector (OSTI)

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

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

132

Double-clad nuclear-fuel safety rod  

DOE Patents [OSTI]

A device for shutting down a nuclear reactor during an undercooling or overpower event, whether or not the reactor's scram system operates properly. This is accomplished by double-clad fuel safety rods positioned at various locations throughout the reactor core, wherein melting of a secondary internal cladding of the rod allows the fuel column therein to shift from the reactor core to place the reactor in a subcritical condition.

McCarthy, W.H.; Atcheson, D.B.

1981-12-30T23:59:59.000Z

133

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

134

Pyrochemical Treatment of Spent Nuclear Fuel  

SciTech Connect (OSTI)

Over the last 10 years, pyrochemical treatment of spent nuclear fuel has progressed from demonstration activities to engineering-scale production operations. As part of the Advanced Fuel Cycle Initiative within the U.S. Department of Energy’s Office of Nuclear Energy, Science and Technology, pyrochemical treatment operations are being performed as part of the treatment of fuel from the Experimental Breeder Reactor II at the Idaho National Laboratory. Integral to these treatment operations are research and development activities that are focused on scaling further the technology, developing and implementing process improvements, qualifying the resulting high-level waste forms, and demonstrating the overall pyrochemical fuel cycle.

K. M. Goff; K. L. Howden; G. M. Teske; T. A. Johnson

2005-10-01T23:59:59.000Z

135

Comments on Americium Volatilization during Fuel Fabrication for Fast Reactors  

SciTech Connect (OSTI)

The physical processes relevant to the fabrication of metallic and ceramic nuclear fuels are analyzed, with attention to recycling of fuels containing U, Pu, and minor volatile actinides for the use in fast reactors. This analysis is relevant to the development of a process model that can be used for the numerical simulation and prediction of the spatial distribution of composition in the fuel, an important factor in fuel performance.

Sabau, Adrian S [ORNL; Ohriner, Evan Keith [ORNL

2008-01-01T23:59:59.000Z

136

Long-term management of high-level radioactive waste (HLW) and spent nuclear fuel (SNF)  

Broader source: Energy.gov [DOE]

GC-52 provides legal advice to DOE regarding the long-term management of high-level radioactive waste (HLW) and spent nuclear fuel (SNF). SNF is nuclear fuel that has been used as fuel in a reactor...

137

Proceedings of GLOBAL 2013: International Nuclear Fuel Cycle Conference - Nuclear Energy at a Crossroads  

SciTech Connect (OSTI)

The Global conference is a forum for the discussion of the scientific, technical, social and regulatory aspects of the nuclear fuel cycle. Relevant topics include global utilization of nuclear energy, current fuel cycle technologies, advanced reactors, advanced fuel cycles, nuclear nonproliferation and public acceptance.

NONE

2013-07-01T23:59:59.000Z

138

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

139

Nuclear Fuels & Materials Spotlight Volume 4  

SciTech Connect (OSTI)

As the nation's nuclear energy laboratory, Idaho National Laboratory brings together talented people and specialized nuclear research capability to accomplish our mission. This edition of the Nuclear Fuels and Materials Division Spotlight provides an overview of some of our recent accomplishments in research and capability development. These accomplishments include: • The first identification of silver and palladium migrating through the SiC layer in TRISO fuel • A description of irradiation assisted stress corrosion testing capabilities that support commercial light water reactor life extension • Results of high-temperature safety testing on coated particle fuels irradiated in the ATR • New methods for testing the integrity of irradiated plate-type reactor fuel • Description of a 'Smart Fuel' concept that wirelessly provides real time information about changes in nuclear fuel properties and operating conditions • Development and testing of ultrasonic transducers and real-time flux sensors for use inside reactor cores, and • An example of a capsule irradiation test. Throughout Spotlight, you'll find examples of productive partnerships with academia, industry, and government agencies that deliver high-impact outcomes. The work conducted at Idaho National Laboratory helps to spur innovation in nuclear energy applications that drive economic growth and energy security. We appreciate your interest in our work here at INL, and hope that you find this issue informative.

I. J. van Rooyen,; T. M. Lillo; Y. Q. WU; P.A. Demkowicz; L. Scott; D.M. Scates; E. L. Reber; J. H. Jackson; J. A. Smith; D.L. Cottle; B.H. Rabin; M.R. Tonks; S.B. Biner; Y. Zhang; R.L. Williamson; S.R. Novascone; B.W. Spencer; J.D. Hales; D.R. Gaston; C.J. Permann; D. Anders; S.L. Hayes; P.C. Millett; D. Andersson; C. Stanek; R. Ali; S.L. Garrett; J.E. Daw; J.L. Rempe; J. Palmer; B. Tittmann; B. Reinhardt; G. Kohse; P. Ramuhali; H.T. Chien; T. Unruh; B.M. Chase; D.W. Nigg; G. Imel; J. T. Harris

2014-04-01T23:59:59.000Z

140

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

Note: This page contains sample records for the topic "fuel nuclear reactors" 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 fuel elements made from nanophase materials  

DOE Patents [OSTI]

A nuclear reactor core fuel element is composed of nanophase high temperature materials. An array of the fuel elements in rod form are joined in an open geometry fuel cell that preferably also uses such nanophase materials for the cell structures. The particular high temperature nanophase fuel element material must have the appropriate mechanical characteristics to avoid strain related failure even at high temperatures, in the order of about 3000 F. Preferably, the reactor type is a pressurized or boiling water reactor and the nanophase material is a high temperature ceramic or ceramic composite. Nanophase metals, or nanophase metals with nanophase ceramics in a composite mixture, also have desirable characteristics, although their temperature capability is not as great as with all-ceramic nanophase material. Combinations of conventional or nanophase metals and conventional or nanophase ceramics can be employed as long as there is at least one nanophase material in the composite. The nuclear reactor so constructed has a number of high strength fuel particles, a nanophase structural material for supporting a fuel rod at high temperature, a configuration to allow passive cooling in the event of a primary cooling system failure, an ability to retain a coolable geometry even at high temperatures, an ability to resist generation of hydrogen gas, and a configuration having good nuclear, corrosion, and mechanical characteristics. 5 figs.

Heubeck, N.B.

1998-09-08T23:59:59.000Z

142

Proliferation resistance of small modular reactors fuels  

SciTech Connect (OSTI)

In this paper the proliferation resistance of different types of Small Modular Reactors (SMRs) has been examined and classified with criteria available in the literature. In the first part of the study, the level of proliferation attractiveness of traditional low-enriched UO{sub 2} and MOX fuels to be used in SMRs based on pressurized water technology has been analyzed. On the basis of numerical simulations both cores show significant proliferation risks. Although the MOX core is less proliferation prone in comparison to the UO{sub 2} core, it still can be highly attractive for diversion or undeclared production of nuclear material. In the second part of the paper, calculations to assess the proliferation attractiveness of fuel in typical small sodium cooled fast reactor show that proliferation risks from spent fuel cannot be neglected. The core contains a highly attractive plutonium composition during the whole life cycle. Despite some aspects of the design like the sealed core that enables easy detection of unauthorized withdrawal of fissile material and enhances proliferation resistance, in case of open Non-Proliferation Treaty break-out, weapon-grade plutonium in sufficient quantities could be extracted from the reactor core.

Polidoro, F.; Parozzi, F. [RSE - Ricerca sul Sistema Energetico,Via Rubattino 54, 20134, Milano (Italy); Fassnacht, F.; Kuett, M.; Englert, M. [IANUS, Darmstadt University of Technology, Alexanderstr. 35, D-64283 Darmstadt (Germany)

2013-07-01T23:59:59.000Z

143

Preliminary Evaluation of Removing Used Nuclear Fuel From Shutdown Sites |  

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

Evaluation of Removing Used Nuclear Fuel From Shutdown Evaluation of Removing Used Nuclear Fuel From Shutdown Sites Preliminary Evaluation of Removing Used Nuclear Fuel From Shutdown Sites In January 2013, the Department of Energy issued the Strategy for the Management and Disposal of Used Nuclear Fuel and High-Level Radioactive Waste. Among the elements contained in this strategy is an initial focus on accepting used nuclear fuel from shutdown reactor sites. This focus is consistent with the recommendations of the Blue Ribbon Commission on America's Nuclear Future, which identified removal of stranded used nuclear fuel at shutdown sites as a priority so that these sites may be completely decommissioned and put to other beneficial uses. Shutdown sites are defined as those commercial nuclear power reactor sites where the

144

Nuclear Spent Fuel Program Drivers  

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

was created to plan and coordinate the management of Department of Energy-owned spent nuclear fuel. It was established as a result of a 1992 decision to stop spent nuclear fuel...

145

Nuclear Fuels Storage & Transportation Planning Project Documents |  

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

Fuel Cycle Technologies » Nuclear Fuels Storage & Fuel Cycle Technologies » Nuclear Fuels Storage & Transportation Planning Project » Nuclear Fuels Storage & Transportation Planning Project Documents Nuclear Fuels Storage & Transportation Planning Project Documents September 30, 2013 Preliminary Evaluation of Removing Used Nuclear Fuel From Shutdown Sites In January 2013, the Department of Energy issued the Strategy for the Management and Disposal of Used Nuclear Fuel and High-Level Radioactive Waste. Among the elements contained in this strategy is an initial focus on accepting used nuclear fuel from shutdown reactor sites. February 22, 2013 Public Preferences Related to Consent-Based Siting of Radioactive Waste Management Facilities for Storage and Disposal This report provides findings from a set of social science studies

146

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

147

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

148

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

149

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

150

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

151

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

152

Determining Reactor Flux from Xenon-136 and Cesium-135 in Spent Fuel  

E-Print Network [OSTI]

The ability to infer the reactor flux from spent fuel or seized fissile material would enhance the tools of nuclear forensics and nuclear nonproliferation significantly. We show that reactor flux can be inferred from the ratios of xenon-136 to xenon-134 and cesium-135 to cesium-137. If the average flux of a reactor is known, the flux inferred from measurements of spent fuel could help determine whether that spent fuel was loaded as a blanket or close to the mid-plane of the reactor. The cesium ratio also provides information on reactor shutdowns during the irradiation of fuel, which could prove valuable for identifying the reactor in question through comparisons with satellite reactor heat monitoring data. We derive analytic expressions for these correlations and compare them to experimental data and to detailed reactor burn simulations. The enrichment of the original uranium fuel affects the correlations by up to 3 percent, but only at high flux.

A. C. Hayes; Gerard Jungman

2012-05-30T23:59:59.000Z

153

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

154

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

155

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

156

Swelling-resistant nuclear fuel  

DOE Patents [OSTI]

A nuclear fuel according to one embodiment includes an assembly of nuclear fuel particles; and continuous open channels defined between at least some of the nuclear fuel particles, wherein the channels are characterized as allowing fission gasses produced in an interior of the assembly to escape from the interior of the assembly to an exterior thereof without causing significant swelling of the assembly. Additional embodiments, including methods, are also presented.

Arsenlis, Athanasios (Hayward, CA); Satcher, Jr., Joe (Patterson, CA); Kucheyev, Sergei O. (Oakland, CA)

2011-12-27T23:59:59.000Z

157

Characterization plan for Hanford spent nuclear fuel  

SciTech Connect (OSTI)

Reprocessing of spent nuclear fuel (SNF) at the Hanford Site Plutonium-Uranium Extraction Plant (PUREX) was terminated in 1972. Since that time a significant quantity of N Reactor and Single-Pass Reactor SNF has been stored in the 100 Area K-East (KE) and K-West (KW) reactor basins. Approximately 80% of all US Department of Energy (DOE)-owned SNF resides at Hanford, the largest portion of which is in the water-filled KE and KW reactor basins. The basins were not designed for long-term storage of the SNF and it has become a priority to move the SNF to a more suitable location. As part of the project plan, SNF inventories will be chemically and physically characterized to provide information that will be used to resolve safety and technical issues for development of an environmentally benign and efficient extended interim storage and final disposition strategy for this defense production-reactor SNF.

Abrefah, J.; Thornton, T.A.; Thomas, L.E.; Berting, F.M.; Marschman, S.C.

1994-12-01T23:59:59.000Z

158

Spent Nuclear Fuel Fact Sheets  

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

management needs. By coordinating common needs for research, technology development, and testing programs, the National Spent Nuclear Fuel Program is achieving cost efficiencies...

159

Radiotoxicity and decay heat power of spent nuclear fuel of VVER type reactors at long-term storage  

Science Journals Connector (OSTI)

......a controllable storage facility for cooling...transferred for long-term storage. The storage...adequately handle waste radiation characteristics...type reactors at long-term storage. | Radiotoxicity...of radioactive waste (radwaste) determines......

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

2005-12-20T23:59:59.000Z

160

Proliferation resistant fuel for pebble bed modular reactors  

SciTech Connect (OSTI)

We show that it is possible to denature the Plutonium produced in Pebble Bed Modular Reactors (PBMR) by doping the nuclear fuel with either 3050 ppm of {sup 237}Np or 2100 ppm of Am vector. A correct choice of these isotopes concentration yields denatured Plutonium with isotopic ratio {sup 238}Pu/Pu {>=} 6%, for the entire fuel burnup cycle. The penalty for introducing these isotopes into the nuclear fuel is a subsequent shortening of the fuel burnup cycle, with respect to a non-doped reference fuel, by 41.2 Full Power Days (FPDs) and 19.9 FPDs, respectively, which correspond to 4070 MWd/ton and 1965 MWd/ton reduction in fuel discharge burnup. (authors)

Ronen, Y.; Aboudy, M.; Regev, D.; Gilad, E. [Dept. of Nuclear Engineering, Ben-Gurion Univ. of the Negev, Beer-Sheva 84105 (Israel)

2012-07-01T23:59:59.000Z

Note: This page contains sample records for the topic "fuel nuclear reactors" 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

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

SciTech Connect (OSTI)

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

D. E. Shropshire

2009-01-01T23:59:59.000Z

162

The IAEA international conference on fast reactors and related fuel cycles: highlights and main outcomes  

SciTech Connect (OSTI)

The 'International Conference on Fast Reactors and Related Fuel Cycles', which is regularly held every four years, represents the main international event dealing with fast reactors technology and related fuel cycles options. Main topics of the conference were new fast reactor concepts, design and simulation capabilities, safety of fast reactors, fast reactor fuels and innovative fuel cycles, analysis of past experience, fast reactor knowledge management. Particular emphasis was put on safety aspects, considering the current need of developing and harmonizing safety standards for fast reactors at the international level, taking also into account the lessons learned from the accident occurred at the Fukushima- Daiichi nuclear power plant in March 2011. Main advances in the several key areas of technological development were presented through 208 oral presentations during 41 technical sessions which shows the importance taken by fast reactors in the future of nuclear energy.

Monti, S.; Toti, A. [International Atomic Energy Agency - IAEA, Wagramer Strasse 5, PO Box 100, A-1400 Vienna (Austria)

2013-07-01T23:59:59.000Z

163

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

164

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

165

The integral fast reactor fuel cycle  

SciTech Connect (OSTI)

The liquid-metal reactor (LMR) has the potential to extend the uranium resource by a factor of 50 to 100 over current commercial light water reactors (LWRs). In the integral fast reactor (IFR) development program, the entire reactor system - reactor, fuel cycle, and waste process - is being developed and optimized at the same time as a single integral entity. A key feature of the IFR concept is the metallic fuel. The lead irradiation tests on the new U-Pu-Zr metallic fuel in the Experimental Breeder Reactor II have surpassed 185000 MWd/t burnup, and its high burnup capability has now been fully demonstrated. The metallic fuel also allows a radically improved fuel cycle technology. Pyroprocessing, which utilizes high temperatures and molten salt and molten metal solvents, can be advantageously utilized for processing metal fuels because the product is metal suitable for fabrication into new fuel elements. Direct production of a metal product avoids expensive and cumbersome chemical conversion steps that would result from use of the conventional Purex solvent extraction process. The key step in the IFR process is electrorefining, which provides for recovery of the valuable fuel constituents, uranium and plutonium, and for removal of fission products. A notable feature of the IFR process is that the actinide elements accompany plutonium through the process. This results in a major advantage in the high-level waste management.

Chang, Y.I. (Argonne National Lab., IL (United States))

1990-01-01T23:59:59.000Z

166

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

167

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

E-Print Network [OSTI]

Health and Safety Impacts of Nuclear, Geothermal, and Fossil- Fuel3 of HEALTH AND SAFETY IMPACTS OF FOSSIL-FUEL NUCLEAR,HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL

Nero, A.V.

2010-01-01T23:59:59.000Z

168

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

169

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...

170

Advanced Neutron Source Reactor thermal analysis of fuel plate defects  

SciTech Connect (OSTI)

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

Giles, G.E.

1995-08-01T23:59:59.000Z

171

Thermal analysis for fuel handling system for sodium cooled reactor considering minor actinide-bearing metal fuel.  

SciTech Connect (OSTI)

The Advanced Burner Reactor (ABR) is one of the components of the Global Nuclear Energy Partnership (GNEP) used to close the fuel cycle. ABR is a sodium-cooled fast reactor that is used to consume transuranic elements resulting from the reprocessing of light water reactor spent nuclear fuel. ABR-1000 [1000 MW(thermal)] is a fast reactor concept created at Argonne National Laboratory to be used as a reference concept for various future trade-offs. ABR-1000 meets the GNEP goals although it uses what is considered base sodium fast reactor technology for its systems and components. One of the considerations of any fast reactor plant concept is the ability to perform fuel-handling operations with new and spent fast reactor fuel. The transmutation fuel proposed as the ABR fuel has a very little experience base, and thus, this paper investigates a fuel-handling concept and potential issues of handling fast reactor fuel containing minor actinides. In this study, two thermal analyses supporting a conceptual design study on the ABR-1000 fuel-handling system were carried out. One analysis investigated passive dry spent fuel storage, and the other analysis investigated a fresh fuel shipping cask. Passive dry storage can be made suitable for the ABR-1000 spent fuel storage with sodium-bonded metal fuel. The thermal analysis shows that spent fast reactor fuel with a decay heat of 2 kW or less can be stored passively in a helium atmosphere. The 2-kW value seems to be a reasonable and practical level, and a combination of reasonably-sized in-sodium storage followed by passive dry storage could be a candidate for spent fuel storage for the next-generation sodium-cooled reactor with sodium-bonded metal fuel. Requirements for the shipping casks for minor actinide-bearing fuel with a high decay heat level are also discussed in this paper. The shipping cask for fresh sodium-cooled-reactor fuel should be a dry type to reduce the reaction between residual moisture on fresh fuel and the sodium coolant. The cladding temperature requirement is maintained below the creep temperature limit to avoid any damage before core installation. The thermal analysis shows that a helium gas-filled cask can accommodate ABR-1000 fresh minor actinide-bearing fuel with 700-W decay heat. The above analysis results revealed the overall requirement for minor actinide-bearing metal fuel handling. The information is thought to be helpful in the design of the ABR-1000 and future sodium-cooled-reactor fuel-handling system.

Chikazawa, Y.; Grandy, C.; Nuclear Engineering Division

2009-03-01T23:59:59.000Z

172

Dr. Hussein Khalil at Reactor and Fuel Cycle Technologies Subcommittee  

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

Blue Blue ribbon presentation by Dr. Hussein Khalil 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 Blue ribbon presentation by Hussein Khalil Hussein Khalil Dr. Hussein Khalil during the panel discussion Oct. 21, 2010 On October 12 Hussein Khalil, director of Argonne's Nuclear Engineering Division, participated in a Reactor and Fuel Cycle Technologies

173

Light Water Reactor Fuel Cladding Research and Testing | ornl.gov  

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

Light Water Reactor Fuel Cladding Research Light Water Reactor Fuel Cladding Research June 01, 2013 Severe Accident Test Station ORNL is the focus point for Light Water Reactor (LWR) fuel cladding research and testing. The purpose of this research is to furnish U.S. industry (EPRI, Areva, Westinghouse), and regulators (NRC) with much-needed data supporting safe and economical nuclear power generation and used fuel management. LWR fuel cladding work is tightly integrated with ORNL accident tolerant fuel development and used fuel disposition programs thereby providing a powerful capability that couples basic materials science research with the nuclear applications research and development. The ORNL LWR fuel cladding program consists of five complementary areas of research: Accident tolerant fuel and cladding material testing under design

174

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

175

Spent Nuclear Fuel Alternative Technology Decision Analysis  

SciTech Connect (OSTI)

The Westinghouse Savannah River Company (WSRC) made a FY98 commitment to the Department of Energy (DOE) to recommend a technology for the disposal of aluminum-based spent nuclear fuel (SNF) at the Savannah River Site (SRS). The two technologies being considered, direct co-disposal and melt and dilute, had been previously selected from a group of eleven potential SNF management technologies by the Research Reactor Spent Nuclear Fuel Task Team chartered by the DOE''s Office of Spent Fuel Management. To meet this commitment, WSRC organized the SNF Alternative Technology Program to further develop the direct co-disposal and melt and dilute technologies and ultimately provide a WSRC recommendation to DOE on a preferred SNF alternative management technology.

Shedrow, C.B.

1999-11-29T23:59:59.000Z

176

Materials Modeling and Simulation for Nuclear Fuels (MMSNF) Workshops  

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

Aerial photo of Argonne National Laboratory Argonne National Laboratory University of Chicago Chicago Photography courtesy Thomas F Ewing Privacy and Security Notice The MMSNF Workshops The goal of the Materials Modeling and Simulation for Nuclear Fuels (MMSNF) workshops is to stimulate research and discussions on modeling and simulations of nuclear fuels, to assist the design of improved fuels and the evaluation of fuel performance. In addition to research focused on existing or improved types of LWR reactors, recent modeling programs, networks, and links have been created to develop innovative nuclear fuels and materials for future generations of nuclear reactors. Examples can be found in Europe (e.g. F-BRIDGE project and ACTINET network and SAMANTHA cooperative network), in the USA (e.g. CASL, NEAMS, CESAR and CMSN network

177

VISION - Verifiable Fuel Cycle Simulation of Nuclear Fuel Cycle Dynamics  

SciTech Connect (OSTI)

The U.S. DOE Advanced Fuel Cycle Initiative’s (AFCI) fundamental objective is to provide technology options that - if implemented - would enable long-term growth of nuclear power while improving sustainability and energy security. The AFCI organization structure consists of four areas; Systems Analysis, Fuels, Separations and Transmutations. The Systems Analysis Working Group is tasked with bridging the program technical areas and providing the models, tools, and analyses required to assess the feasibility of design and deployment options and inform key decision makers. An integral part of the Systems Analysis tool set is the development of a system level model that can be used to examine the implications of the different mixes of reactors, implications of fuel reprocessing, impact of deployment technologies, as well as potential "exit" or "off ramp" approaches to phase out technologies, waste management issues and long-term repository needs. The Verifiable Fuel Cycle Simulation Model (VISION) is a computer-based simulation model that allows performing dynamic simulations of fuel cycles to quantify infrastructure requirements and identify key trade-offs between alternatives. It is based on the current AFCI system analysis tool "DYMOND-US" functionalities in addition to economics, isotopic decay, and other new functionalities. VISION is intended to serve as a broad systems analysis and study tool applicable to work conducted as part of the AFCI and Generation IV reactor development studies.

Steven J. Piet; A. M. Yacout; J. J. Jacobson; C. Laws; G. E. Matthern; D. E. Shropshire

2006-02-01T23:59:59.000Z

178

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

179

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

180

Inherent safety concepts in nuclear power reactors  

Science Journals Connector (OSTI)

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

O M Pal Singh; R Shankar Singh

1989-06-01T23:59:59.000Z

Note: This page contains sample records for the topic "fuel nuclear reactors" 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

Nuclear Fuel Cycle & Vulnerabilities  

SciTech Connect (OSTI)

The objective of safeguards 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. The safeguards system should be designed to provide credible assurances that there has been no diversion of declared nuclear material and no undeclared nuclear material and activities.

Boyer, Brian D. [Los Alamos National Laboratory

2012-06-18T23:59:59.000Z

182

Uraninite: A 2 Ga spent nuclear fuel from the natural fission reactor at Bangombe in Gabon, West Africa  

SciTech Connect (OSTI)

Uraninites from the Bangombe natural fission reactor (RZB) and normal uranium-ore occur as fine veins in the sandstone host-rock as well as altered, broken, and slightly displaced grains in an illitic matrix, and in nodules and veins of solid bitumen. Inclusions of galena, (Y,Gd)-rich phosphates, a Pb-oxide and a Ti-oxide? were observed. Uraninites just below RZB were partially altered to a uranyl-sulfate. Three generations of uraninite were identified based on their PbO-contents of 8--11.06 wt%, 6 wt% (the largest population), and a younger generation with 3 wt%. Diffusional loss of Pb is indicated by the presence of a Pb-oxide at the interface to the uraninites. The behavior of the metallic fission products, incompatible with the uraninite structure, may mimic the behavior of Pb in these uraninites. The averaged impurity-content ranges from 4.29 to 6.89 wt%, and consists mainly of SiO{sub 2}, TiO{sub 2}, ZrO{sub 2}, FeO, CaO, Al{sub 2}O{sub 3} and P{sub 2}O{sub 5}. The averaged content of Y{sub 2}O{sub 3} and the Ln`s is less than 0.78 wt% and there is a scattered positive correlation with P{sub 2}O{sub 5}. The content of Y + Ln`s is generally highest in the uraninites from RZB. Uraninite hydration and the formation of uranopelite/zippeite have caused complete loss of Y and the Ln`s. The analytical results indicate that Y and the Ln`s, which are high yield fission products, may be released from uraninite during alteration in the presence of P.

Jensen, K.A. [Aarhus Univ. (Denmark). Dept. of Earth Sciences; Ewing, R.C. [Univ. of New Mexico, Albuquerque, NM (United States). Dept. of Earth and Planetary Sciences; Gauthier-Lafaye, F. [Centre National de la Recherche Scientifique, Strasbourg (France). Centre de Geochemie de la Surface

1997-12-31T23:59:59.000Z

183

Sustainability Features of Nuclear Fuel Cycle Options  

E-Print Network [OSTI]

The nuclear fuel cycle is the series of stages that nuclear fuel materials go through in a cradle to grave framework. The Once Through Cycle (OTC) is the current fuel cycle implemented in the United States; in which an ...

Passerini, Stefano

184

LMFBR operation in the nuclear cycle without fuel reprocessing  

SciTech Connect (OSTI)

Substantiation is given to expediency of investigation of nuclear power (NP) development with fast reactors cooled by lead-bismuth alloy operating during extended time in the open nuclear fuel cycle with slightly enriched or depleted uranium make-up. 9 refs., 1 fig., 6 tabs.

Toshinsky, S.I. [Institute of Physics and Power Engineering, Kaluga (Russian Federation)

1997-12-01T23:59:59.000Z

185

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

186

An experiment to simulate the heat transfer properties of a dry, horizontal spent nuclear fuel assembly  

E-Print Network [OSTI]

Nuclear power reactors generate highly radioactive spent fuel assemblies. Initially, the spent fuel assemblies are stored for a period of several years in an on-site storage facility to allow the radioactivity levels of ...

Lovett, Phyllis Maria

1991-01-01T23:59:59.000Z

187

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

188

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

189

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

190

Thermoacoustic device for nuclear fuel monitoring and heat transfer enhancement  

Science Journals Connector (OSTI)

The Fukushima Dai’ichi nuclear disaster of 2011 exposed the need for self-powered sensors that could transmit the status of the fuel rods within the reactor and in spent fuel ponds that was not dependent upon availability of external electrical power for either sensing or telemetry. One possible solution is the use of a thermoacoustic standing wave engine incorporated within a fuel rod which is heated by the nuclear fuel. The engine’s resonance frequency is correlated to the fuel rod temperature and will be transmitted by sound radiation through the reactor's or storage pond’s surrounding water. In addition to acting as a passive temperature sensor the thermoacoustic device will serve to enhance heat transfer from the fuel to the surrounding heat transfer fluid. When activated the acoustically-driven streaming flow of the gas within the fuel rod will circulate gas away from the nuclear fuel and convectively enhance heat transfer to the surrounding coolant. We will present results for a thermoacousticresonator built into a Nitonic® 60 (stainless steel) fuel rod that can be substituted for conventional fuel rods in the Idaho National Laboratory’s Advanced Test Reactor. This laboratory version is heated electrically. [Work supported by the U.S. Department of Energy.

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

2012-01-01T23:59:59.000Z

191

Advanced Fuels Campaign Light Water Reactor Accident Tolerant Fuel Performance Metrics  

SciTech Connect (OSTI)

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

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

2014-02-01T23:59:59.000Z

192

Maximum Fuel Utilization in Advanced Fast Reactors without Actinides Separation  

E-Print Network [OSTI]

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

Heidet, Florent

2010-01-01T23:59:59.000Z

193

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

194

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

195

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

196

Description of the Canadian Particulate-Fill WastePackage (WP) System for Spent-Nuclear Fuel (SNF) and its Applicability to Ligh-Water Reactor SNF WPS with Depleted Uranium-Dioxide Fill  

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

3502 3502 Chemical Technology Division DESCRIPTION OF THE CANADIAN PARTICULATE-FILL WASTE-PACKAGE (WP) SYSTEM FOR SPENT-NUCLEAR FUEL(SNF) AND ITS APPLICABILITY TO LIGHT- WATER REACTOR SNF WPS WITH DEPLETED URANIUM-DIOXIDE FILL Charles W. Forsberg Oak Ridge National Laboratory * P.O. Box 2008 Oak Ridge, Tennessee 37831-6180 Tel: (423) 574-6783 Fax: (423) 574-9512 Email: forsbergcw@ornl.gov October 20, 1997 _________________________ Managed by Lockheed Martin Energy Research Corp. under contract DE-AC05-96OR22464 for the * U.S. Department of Energy. iii CONTENTS LIST OF FIGURES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . v LIST OF TABLES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . vii ACRONYMS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

197

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

198

Advanced LWR Nuclear Fuel Cladding System Development Trade-Off Study  

SciTech Connect (OSTI)

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

Kristine Barrett; Shannon Bragg-Sitton

2012-09-01T23:59:59.000Z

199

Activities Related to Storage of Spent Nuclear Fuel | Department...  

Office of Environmental Management (EM)

Activities Related to Storage of Spent Nuclear Fuel Activities Related to Storage of Spent Nuclear Fuel Activities Related to Storage of Spent Nuclear Fuel More Documents &...

200

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

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201

PLUTONIUM METALLIC FUELS FOR FAST REACTORS  

SciTech Connect (OSTI)

Early interest in metallic plutonium fuels for fast reactors led to much research on plutonium alloy systems including binary solid solutions with the addition of aluminum, gallium, or zirconium and low-melting eutectic alloys with iron and nickel or cobalt. There was also interest in ternaries of these elements with plutonium and cerium. The solid solution and eutectic alloys have most unusual properties, including negative thermal expansion in some solid-solution alloys and the highest viscosity known for liquid metals in the Pu-Fe system. Although metallic fuels have many potential advantages over ceramic fuels, the early attempts were unsuccessful because these fuels suffered from high swelling rates during burn up and high smearing densities. The liquid metal fuels experienced excessive corrosion. Subsequent work on higher-melting U-PuZr metallic fuels was much more promising. In light of the recent rebirth of interest in fast reactors, we review some of the key properties of the early fuels and discuss the challenges presented by the ternary alloys.

STAN, MARIUS [Los Alamos National Laboratory; HECKER, SIEGFRIED S. [Los Alamos National Laboratory

2007-02-07T23:59:59.000Z

202

Methods for making a porous nuclear fuel element  

SciTech Connect (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

2014-12-30T23:59:59.000Z

203

Fuel and cladding nano-technologies based solutions for long life heat-pipe based reactors  

SciTech Connect (OSTI)

A novel nuclear reactor concept, unifying the fuel pipe with fuel tube functionality has been developed. The structure is a quasi-spherical modular reactor, designed for a very long life. The reactor module unifies the fuel tube with the heat pipe and a graphite beryllium reflector. It also uses a micro-hetero-structure that allows the fission products to be removed in the heat pipe flow and deposited in a getter area in the cold zone of the heat pipe, but outside the neutron flux. The reactor operates as a breed and burn reactor - it contains the fuel pipe with a variable enrichment, starting from the hot-end of the pipe, meant to assure the initial criticality, and reactor start-up followed by area with depleted uranium or thorium that get enriched during the consumption of the first part of the enriched uranium. (authors)

Popa-Simil, L. [LAVM LLC, Los Alamos (United States)

2012-07-01T23:59:59.000Z

204

Natural Fueling of a Tokamak Fusion Reactor  

E-Print Network [OSTI]

A natural fueling mechanism that helps to maintain the main core deuterium and tritium (DT) density profiles in a tokamak fusion reactor is discussed. In H-mode plasmas dominated by ion- temperature gradient (ITG) driven turbulence, cold DT ions near the edge will naturally pinch radially inward towards the core. This mechanism is due to the quasi-neutral heat flux dominated nature of ITG turbulence and still applies when trapped and passing kinetic electron effects are included. Fueling using shallow pellet injection or supersonic gas jets is augmented by an inward pinch of could DT fuel. The natural fueling mechanism is demonstrated using the three-dimensional toroidal electromagnetic gyrokinetic turbulence code GEM and is analyzed using quasilinear theory. Profiles similar to those used for conservative ITER transport modeling that have a completely flat density profile are examined and it is found that natural fueling actually reduces the linear growth rates and energy transport.

Wan, Weigang; Chen, Yang; Perkins, Francis W

2009-01-01T23:59:59.000Z

205

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

206

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

207

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

208

Performance testing and Bayesian Reliability Analysis of small diameter, high power electric heaters for the simulation of nuclear fuel rod temperatures.  

E-Print Network [OSTI]

??The conversion of plutonium from a nuclear weapon to nuclear reactor fuel requires an evaluation of the residual gallium as a potential corrosive material within… (more)

O'Kelly, David Sean

2012-01-01T23:59:59.000Z

209

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

210

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...

211

Laser-based characterization of nuclear fuel plates  

SciTech Connect (OSTI)

Ensuring the integrity of fuel-clad and clad-clad bonding in nuclear fuels is important for safe reactor operation and assessment of fuel performance, yet the measurement of bond strengths in actual fuels has proved challenging. The laser shockwave technique (LST) originally developed to characterize structural adhesion in composites is being employed to characterize interface strength in a new type of plate fuel being developed at Idaho National Laboratory (INL). LST is a non-contact method that uses lasers for the generation and detection of large-amplitude acoustic waves and is well suited for application to both fresh and irradiated nuclear-fuel plates. This paper will report on initial characterization results obtained from fresh fuel plates manufactured by different processes, including hot isostatic pressing, friction stir welding, and hot rolling.

Smith, James A.; Cottle, Dave L.; Rabin, Barry H. [Idaho National Laboratory, Fuel Performance and Design, P.O. Box 1625, Idaho Falls, Idaho, 83415-6188 (United States)

2014-02-18T23:59:59.000Z

212

Development of alternate extractant systems for fast reactor fuel cycle  

SciTech Connect (OSTI)

Due to the limitations of TBP in processing of high burn-up, Pu-rich fast reactor fuels, there is a need to develop alternate extractants for fast reactor fuel processing. In this context, our Centre has been examining the suitability of alternate tri-alkyl phosphates. Third phase formation in the extraction of Th(IV) by TBP, tri-n-amyl phosphate (TAP) and tri-2-methyl-butyl phosphate (T2MBP) from nitric acid media has been investigated under various conditions to derive conclusions on their application for extraction of Pu at macro levels. The chemical and radiolytic degradation of tri-n-amyl-phosphate (TAP) diluted in normal paraffin hydrocarbon (NPH) in the presence of nitric acid has been investigated by the measurement of plutonium retention in organic phase. The potential application of room temperature ionic liquids (RTILs) for reprocessing of spent nuclear fuel has been explored. Extraction of uranium (VI) and palladium (II) from nitric acid medium by commercially available RTIL and tri-n-butyl phosphate solution in RTIL have been studied and the feasibility of electrodeposition of uranium as uranium oxide (UO{sub 2}) and palladium (II) as metallic palladium from the loaded organic phase have been demonstrated. This paper describes results of the above studies and discusses the suitability of the systems for fast reactor fuel reprocessing. (authors)

Vasudeva Rao, P.R.; Suresh, A.; Venkatesan, K.A.; Srinivasan, T.G.; Raj, Baldev [Indira Gandhi Centre for Atomic Research, Kalpakkam - 603 102 (India)

2007-07-01T23:59:59.000Z

213

Plutonium and Reprocessing of Spent Nuclear Fuel  

Science Journals Connector (OSTI)

...might spawn nuclear terrorism. Less than...reprocessing plant. The U.S. nuclear-energy...current fleet of power reactors (15...operational risk of transmutation...future of nuclear power is clarified...constructed plant increased...

Frank N. von Hippel

2001-09-28T23:59:59.000Z

214

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

215

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

216

Safety of CANDU reactors utilizing plutonium-enriched mixed-oxide fuel  

SciTech Connect (OSTI)

Substantial quantities of plutonium have become available as a result of nuclear arms reduction agreements. Irradiation of plutonium enriched fuel in Canadian deuterium uranium (CANDU) heavy water moderated and cooled reactors, of which there are 22 in operation in Canada, has been evaluated as a means of managing it. This paper summarizes the results of a study of reactor safety.

Chan, P.; Feinroth, H.; Luxat, J.; Pendergast, D.

1994-12-31T23:59:59.000Z

217

Sensitivity analysis and optimization of the nuclear fuel cycle  

SciTech Connect (OSTI)

A sensitivity study has been conducted to assess the robustness of the conclusions presented in the MIT Fuel Cycle Study. The Once Through Cycle (OTC) is considered as the base-line case, while advanced technologies with fuel recycling characterize the alternative fuel cycles. The options include limited recycling in LWRs and full recycling in fast reactors and in high conversion LWRs. Fast reactor technologies studied include both oxide and metal fueled reactors. The analysis allowed optimization of the fast reactor conversion ratio with respect to desired fuel cycle performance characteristics. The following parameters were found to significantly affect the performance of recycling technologies and their penetration over time: Capacity Factors of the fuel cycle facilities, Spent Fuel Cooling Time, Thermal Reprocessing Introduction Date, and in core and Out-of-core TRU Inventory Requirements for recycling technology. An optimization scheme of the nuclear fuel cycle is proposed. Optimization criteria and metrics of interest for different stakeholders in the fuel cycle (economics, waste management, environmental impact, etc.) are utilized for two different optimization techniques (linear and stochastic). Preliminary results covering single and multi-variable and single and multi-objective optimization demonstrate the viability of the optimization scheme. (authors)

Passerini, S.; Kazimi, M. S.; Shwageraus, E. [Massachusetts Inst. of Technology, Dept. of Nuclear Science and Engineering, 77 Massachusetts Avenue, Cambridge, MA 02138 (United States)

2012-07-01T23:59:59.000Z

218

A fuel for sub-critical fast reactor  

SciTech Connect (OSTI)

Along with the problem of the nuclear waste transmutation, the problem of minimization of waste production is of current interest. It is not possible to eliminate production of waste at a nuclear power plant, but, as is shown in this report, it is in principle possible to arrange a fuel composition with no net production of transuranic elements. The idea is to find the transuranic elements composition to which the depleted uranium is continuously supplied during frequent reprocessing, and amount of each other transuranic fuel component remains unchanged in time. For each transuranic component, the balance is achieved by equating burnup and production rates. The production is due to neutron capture by the neighboring lighter isotope and subsequent beta-decay. The burnup includes fission, neutron capture and decays. For the calculations a simplified burnup model which accounts for 9 isotopes of uranium, neptunium, plutonium and americium is used. The calculated fuel composition consists mainly of uranium with minority of plutonium isotopes. Such a fuel, after usage in a sub-critical fast reactor, should be reprocessed. The fission product content increases during burnup, representing a net production of waste, while the transuranic elements and {sup 238}U should be recycled into a new fuel. For such a fuel cycle, the net consumption is only for 238U, and the net waste production is just fission products.

Moiseenko, V. E.; Chernitskiy, S. V.; Agren, O.; Noack, K. [Institute of Plasma Physics, National Science Center 'Kharkiv Institute of Physics and Technology', Akademichna St. 1, 61108 Kharkiv (Ukraine); Institute of Nuclear Physics, National Science Center 'Kharkiv Institute of Physics and Technology', Akademichna St. 1, 61108 Kharkiv (Ukraine); Uppsala University, Angstroem Laboratory, Division of Electricity, Box 534, SE-75121 Uppsala (Sweden)

2012-06-19T23:59:59.000Z

219

Determination of the optimum fuel burn-up and energy intensities of nuclear fuel by the method of cost calculations  

Science Journals Connector (OSTI)

This report gives the procedure for determining the economical efficiency of the utilization of nuclear fuel in a reactor on the basis of calculated costs. The expression obtained for the fuet constituent of the

Yu. I. Koryakin; V. V. Batov; V. G. Smirnov

1964-08-01T23:59:59.000Z

220

Compositions and methods for treating nuclear fuel  

DOE Patents [OSTI]

Compositions are provided that include nuclear fuel. Methods for treating nuclear fuel are provided which can include exposing the fuel to a carbonate-peroxide solution. Methods can also include exposing the fuel to an ammonium solution. Methods for acquiring molybdenum from a uranium comprising material are provided.

Soderquist, Chuck Z; Johnsen, Amanda M; McNamara, Bruce K; Hanson, Brady D; Smith, Steven C; Peper, Shane M

2013-08-13T23:59:59.000Z

Note: This page contains sample records for the topic "fuel nuclear reactors" 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

Compositions and methods for treating nuclear fuel  

DOE Patents [OSTI]

Compositions are provided that include nuclear fuel. Methods for treating nuclear fuel are provided which can include exposing the fuel to a carbonate-peroxide solution. Methods can also include exposing the fuel to an ammonium solution. Methods for acquiring molybdenum from a uranium comprising material are provided.

Soderquist, Chuck Z; Johnsen, Amanda M; McNamara, Bruce K; Hanson, Brady D; Smith, Steven C; Peper, Shane M

2014-01-28T23:59:59.000Z

222

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

223

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

224

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

225

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

226

Calculated fuel temperatures for a proposed space based reactor using the lumped parameter method  

E-Print Network [OSTI]

CALCULATED FUEL TEMPERATURES FOR A PROPOSED SPACE BASED REACTOR USING THE LUMPED PARAMETER METHOD A Thesis by CELESTE MARIE STEEN Submitted to the Office of Graduate Studies of Texas AgcM University in partial fulfillment of the requirements... f' or the degree of MASTER OF SCIENCE December 1990 Major Subject: Nuclear Engineering CALCULATED FUEL TEMPERATURES FOR A PROPOSED SPACE BASED REACTOR USING THE LUMPED PARAMETER METHOD A Thesis by CELESTE MARIE STEEiV Approved as to style...

Steen, Celeste Marie

2012-06-07T23:59:59.000Z

227

U.S. Department of Energy Accident Resistant SiC Clad Nuclear Fuel  

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

U.S. Department of Energy Accident Resistant SiC Clad Nuclear Fuel U.S. Department of Energy Accident Resistant SiC Clad Nuclear Fuel Development U.S. Department of Energy Accident Resistant SiC Clad Nuclear Fuel Development A significant effort is being placed on silicon carbide ceramic matrix composite (SiC CMC) nuclear fuel cladding by Light Water Reactor Sustainability (LWRS) Advanced Light Water Reactor Nuclear Fuels Pathway. The intent of this work is to invest in a high-risk, high-reward technology that can be introduced in a relatively short time. The LWRS goal is to demonstrate successful advanced fuels technology that suitable for commercial development to support nuclear relicensing. Ceramic matrix composites are an established non-nuclear technology that utilizes ceramic fibers embedded in a ceramic matrix. A thin interfacial layer between the

228

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

229

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

230

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

231

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

232

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

233

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

234

Technology Insights and Perspectives for Nuclear Fuel Cycle Concepts  

SciTech Connect (OSTI)

The following report provides a rich resource of information for exploring fuel cycle characteristics. The most noteworthy trends can be traced back to the utilization efficiency of natural uranium resources. By definition, complete uranium utilization occurs only when all of the natural uranium resource can be introduced into the nuclear reactor long enough for all of it to undergo fission. Achieving near complete uranium utilization requires technologies that can achieve full recycle or at least nearly full recycle of the initial natural uranium consumed from the Earth. Greater than 99% of all natural uranium is fertile, and thus is not conducive to fission. This fact requires the fuel cycle to convert large quantities of non-fissile material into fissile transuranics. Step increases in waste benefits are closely related to the step increase in uranium utilization going from non-breeding fuel cycles to breeding fuel cycles. The amount of mass requiring a disposal path is tightly coupled to the quantity of actinides in the waste stream. Complete uranium utilization by definition means that zero (practically, near zero) actinide mass is present in the waste stream. Therefore, fuel cycles with complete (uranium and transuranic) recycle discharge predominately fission products with some actinide process losses. Fuel cycles without complete recycle discharge a much more massive waste stream because only a fraction of the initial actinide mass is burned prior to disposal. In a nuclear growth scenario, the relevant acceptable frequency for core damage events in nuclear reactors is inversely proportional to the number of reactors deployed in a fuel cycle. For ten times the reactors in a fleet, it should be expected that the fleet-average core damage frequency be decreased by a factor of ten. The relevant proliferation resistance of a fuel cycle system is enhanced with: decreasing reliance on domestic fuel cycle services, decreasing adaptability for technology misuse, enablement of material accountability, and decreasing material attractiveness.

S. Bays; S. Piet; N. Soelberg; M. Lineberry; B. Dixon

2010-09-01T23:59:59.000Z

235

A Neural Network Model for the Tomographic Analysis of Irradiated Nuclear Fuel Rods  

SciTech Connect (OSTI)

A tomographic method based on a multilayer feed-forward artificial neural network is proposed for the reconstruction of gamma-radioactive fission product distribution in irradiated nuclear fuel rods. The quality of the method is investigated as compared to a conventional technique on experimental results concerning a Canada deuterium uranium reactor (CANDU)-type fuel rod irradiated in a TRIGA reactor.

Craciunescu, Teddy [National Institute of Nuclear Physics and Engineering (Romania)

2004-04-15T23:59:59.000Z

236

Comparative analysis of LWR and FBR spent fuels for nuclear forensics evaluation  

SciTech Connect (OSTI)

Some interesting issues are attributed to nuclide compositions of spent fuels from thermal reactors as well as fast reactors such as a potential to reuse as recycled fuel, and a possible capability to be manage as a fuel for destructive devices. In addition, analysis on nuclear forensics which is related to spent fuel compositions becomes one of the interesting topics to evaluate the origin and the composition of spent fuels from the spent fuel foot-prints. Spent fuel compositions of different fuel types give some typical spent fuel foot prints and can be estimated the origin of source of those spent fuel compositions. Some technics or methods have been developing based on some science and technological capability including experimental and modeling or theoretical aspects of analyses. Some foot-print of nuclear forensics will identify the typical information of spent fuel compositions such as enrichment information, burnup or irradiation time, reactor types as well as the cooling time which is related to the age of spent fuels. This paper intends to evaluate the typical spent fuel compositions of light water (LWR) and fast breeder reactors (FBR) from the view point of some foot prints of nuclear forensics. An established depletion code of ORIGEN is adopted to analyze LWR spent fuel (SF) for several burnup constants and decay times. For analyzing some spent fuel compositions of FBR, some coupling codes such as SLAROM code, JOINT and CITATION codes including JFS-3-J-3.2R as nuclear data library have been adopted. Enriched U-235 fuel composition of oxide type is used for fresh fuel of LWR and a mixed oxide fuel (MOX) for FBR fresh fuel. Those MOX fuels of FBR come from the spent fuels of LWR. Some typical spent fuels from both LWR and FBR will be compared to distinguish some typical foot-prints of SF based on nuclear forensic analysis.

Permana, Sidik; Suzuki, Mitsutoshi; Su'ud, Zaki [Department of Science and Technology for Nuclear Material Management (STNM), Japan Atomic Energy Agency (JAEA), 2-4 Shirane, Shirakata, Tokai Mura, Naka-gun, Ibaraki 319-1195 Nuclear Physics and Bio (Indonesia); Department of Science and Technology for Nuclear Material Management (STNM), Japan Atomic Energy Agency (JAEA), 2-4 Shirane, Shirakata, Tokai Mura, Naka-gun, Ibaraki 319-1195 (Japan); Nuclear Physics and Bio Physics Research Group, Department of Physics, Bandung Institute of Technology, Gedung Fisika, Jl. Ganesha 10, Bandung 40132 (Indonesia)

2012-06-06T23:59:59.000Z

237

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

238

Characterization of Nuclear Fuel using Multivariate Statistical Analysis  

SciTech Connect (OSTI)

Various combinations of reactor type and fuel composition have been characterized using principle components analysis (PCA) of the concentrations of 9 U and Pu isotopes in the 10 fuel as a function of burnup. The use of PCA allows the reduction of the 9-dimensional data (isotopic concentrations) into a 3-dimensional approximation, giving a visual representation of the changes in nuclear fuel composition with burnup. Real-world variation in the concentrations of {sup 234}U and {sup 236}U in the fresh (unirradiated) fuel was accounted for. The effects of reprocessing were also simulated. The results suggest that, 15 even after reprocessing, Pu isotopes can be used to determine both the type of reactor and the initial fuel composition with good discrimination. Finally, partial least squares discriminant analysis (PSLDA) was investigated as a substitute for PCA. Our results suggest that PLSDA is a better tool for this application where separation between known classes is most important.

Robel, M; Robel, M; Robel, M; Kristo, M J; Kristo, M J

2007-11-27T23:59:59.000Z

239

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

240

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

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


241

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...

242

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

243

Nuclear Fuel Storage and Transportation Planning Project Overview...  

Office of Environmental Management (EM)

Fuel Storage and Transportation Planning Project Overview Nuclear Fuel Storage and Transportation Planning Project Overview Nuclear Fuel Storage and Transportation Planning Project...

244

Fuel Cycle Options for Optimized Recycling of Nuclear Fuel  

E-Print Network [OSTI]

The reduction of transuranic inventories of spent nuclear fuel depends upon the deployment of advanced fuels that can be loaded with recycled transuranics (TRU), and the availability of facilities to separate and reprocess ...

Aquien, A.

245

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

246

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

247

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

248

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

249

Review of Used Nuclear Fuel Storage and Transportation Technical Gap  

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

Analysis Analysis Review of Used Nuclear Fuel Storage and Transportation Technical Gap Analysis While both wet and dry storage have been shown to be safe options for storing used nuclear fuel (UNF), the focus of the program is on dry storage of commercial UNF at reactor or centralized locations. This report focuses on the knowledge gaps concerning extended storage identified in numerous domestic and international investigations and provides the Used Fuel Disposition Campaign"s (UFDC) gap description, any alternate gap descriptions, the rankings by the various organizations, evaluation of the priority assignment, and UFDC-recommended action based on the comparison. Review of Used Nuclear Fuel Storage and Transportation Technical Gap Analysis More Documents & Publications

250

The behaviour of transuranic mixed oxide fuel in a Candu-900 reactor  

SciTech Connect (OSTI)

The production of transuranic actinide fuels for use in current thermal reactors provides a useful intermediary step in closing the nuclear fuel cycle. Extraction of actinides reduces the longevity, radiation and heat loads of spent material. The burning of transuranic fuels in current reactors for a limited amount of cycles reduces the infrastructure demand for fast reactors and provides an effective synergy that can result in a reduction of as much as 95% of spent fuel waste while reducing the fast reactor infrastructure needed by a factor of almost 13.5 [1]. This paper examines the features of actinide mixed oxide fuel, TRUMOX, in a CANDU{sup R}* nuclear reactor. The actinide concentrations used were based on extraction from 30 year cooled spent fuel and mixed with natural uranium in 3.1 wt% actinide MOX fuel. Full lattice cell modeling was performed using the WIMS-AECL code, super-cell calculations were analyzed in DRAGON and full core analysis was executed in the RFSP 2-group diffusion code. A time-average full core model was produced and analyzed for reactor coefficients, reactivity device worth and online fuelling impacts. The standard CANDU operational limits were maintained throughout operations. The TRUMOX fuel design achieved a burnup of 27.36 MWd/kg HE. A full TRUMOX fuelled CANDU was shown to operate within acceptable limits and provided a viable intermediary step for burning actinides. The recycling, reprocessing and reuse of spent fuels produces a much more sustainable and efficient nuclear fuel cycle. (authors)

Morreale, A. C.; Ball, M. R.; Novog, D. R.; Luxat, J. C. [Dept. of Engineering Physics, McMaster Univ., 1280 Main St. W, Hamilton, ON (Canada)

2012-07-01T23:59:59.000Z

251

Optimizing Treatment Performance of Microbial Fuel Cells by Reactor Staging  

Science Journals Connector (OSTI)

Optimizing Treatment Performance of Microbial Fuel Cells by Reactor Staging ... Multi-unit optimization is a recently proposed method that uses multiple similar units to optimize process performance. ...

Roberto P. Pinto; Boris Tartakovsky; Michel Perrier; Bala Srinivasan

2010-08-18T23:59:59.000Z

252

Methods and apparatuses for the development of microstructured nuclear fuels  

DOE Patents [OSTI]

Microstructured nuclear fuel adapted for nuclear power system use includes fissile material structures of micrometer-scale dimension dispersed in a matrix material. In one method of production, fissile material particles are processed in a chemical vapor deposition (CVD) fluidized-bed reactor including a gas inlet for providing controlled gas flow into a particle coating chamber, a lower bed hot zone region to contain powder, and an upper bed region to enable powder expansion. At least one pneumatic or electric vibrator is operationally coupled to the particle coating chamber for causing vibration of the particle coater to promote uniform powder coating within the particle coater during fuel processing. An exhaust associated with the particle coating chamber and can provide a port for placement and removal of particles and powder. During use of the fuel in a nuclear power reactor, fission products escape from the fissile material structures and come to rest in the matrix material. After a period of use in a nuclear power reactor and subsequent cooling, separation of the fissile material from the matrix containing the embedded fission products will provide an efficient partitioning of the bulk of the fissile material from the fission products. The fissile material can be reused by incorporating it into new microstructured fuel. The fission products and matrix material can be incorporated into a waste form for disposal or processed to separate valuable components from the fission products mixture.

Jarvinen, Gordon D. (Los Alamos, NM); Carroll, David W. (Los Alamos, NM); Devlin, David J. (Santa Fe, NM)

2009-04-21T23:59:59.000Z

253

U.S. Commits $14 million to U.S. - Ukraine Nuclear Fuel Qualification  

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

Commits $14 million to U.S. - Ukraine Nuclear Fuel Commits $14 million to U.S. - Ukraine Nuclear Fuel Qualification Project U.S. Commits $14 million to U.S. - Ukraine Nuclear Fuel Qualification Project March 15, 2007 - 10:55am Addthis KYIV, Ukraine - U.S. Department of Energy Deputy Secretary Clay Sell today announced that the United States will invest $14 million to provide 42 nuclear fuel assemblies to the South Ukraine Nuclear Power Plant under the U.S.-Ukraine Nuclear Fuel Qualification Project (UNFQP). In an agreement reached last week, Westinghouse Electric Company will manufacture nuclear fuel assemblies, which account for one-fourth of the fuel that powers a reactor for up to four years of operation. Deputy Secretary Sell is in Kyiv today to meet with top Ukrainian officials and U.S. business leaders to promote diversity of energy sources, greater energy efficiency,

254

Integrated fuel performance and thermal-hydraulic sub-channel models for analysis of sodium fast reactors  

E-Print Network [OSTI]

Sodium Fast Reactors (SFR) show promise as an effective way to produce clean safe nuclear power while properly managing the fuel cycle. Accurate computer modeling is an important step in the design and eventual licensing ...

Fricano, Joseph William

2012-01-01T23:59:59.000Z

255

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

256

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...

257

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...

258

Nuclear core and fuel assemblies  

DOE Patents [OSTI]

A fast flux nuclear core of a plurality of rodded, open-lattice assemblies having a rod pattern rotated relative to a rod support structure pattern. Elongated fuel rods are oriented on a triangular array and laterally supported by grid structures positioned along the length of the assembly. Initial inter-assembly contact is through strongbacks at the corners of the support pattern and peripheral fuel rods between adjacent assemblies are nested so as to maintain a triangular pitch across a clearance gap between the other portions of adjacent assemblies. The rod pattern is rotated relative to the strongback support pattern by an angle .alpha. equal to sin .sup.-1 (p/2c), where p is the intra-assembly rod pitch and c is the center-to-center spacing among adjacent assemblies.

Downs, Robert E. (Monroeville, PA)

1981-01-01T23:59:59.000Z

259

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

260

The Role of Friction Stir Welding in Nuclear Fuel Plate Fabrication  

SciTech Connect (OSTI)

The friction bonding process combines desirable attributes of both friction stir welding and friction stir processing. The development of the process is spurred on by the need to fabricate thin, high density, reduced enrichment fuel plates for nuclear research reactors. The work seeks to convert research and test reactors currently operating on highly enriched uranium fuel to operate on low enriched uranium fuel without significant loss in reactor performance, safety characteristics, or significant increase in cost. In doing so, the threat of global nuclear material proliferation will be reduced. Feasibility studies performed on the process show that this is a viable option for mass production of plate-type nuclear fuel. Adapting the friction stir weld process for nuclear fuel fabrication has resulted in the development of several unique ideas and observations. Preliminary results of this adaptation and process model development are discussed.

D Burkes; P Medvedev; M Chapple; A Amritkar; P Wells; I Charit

2009-02-01T23:59:59.000Z

Note: This page contains sample records for the topic "fuel nuclear reactors" 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

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

262

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

263

A framework and methodology for nuclear fuel cycle transparency.  

SciTech Connect (OSTI)

A key objective to the global deployment of nuclear technology is maintaining transparency among nation-states and international communities. By providing an environment in which to exchange scientific and technological information regarding nuclear technology, the safe and legitimate use of nuclear material and technology can be assured. Many nations are considering closed or multiple-application nuclear fuel cycles and are subsequently developing advanced reactors in an effort to obtain some degree of energy self-sufficiency. Proliferation resistance features that prevent theft or diversion of nuclear material and reduce the likelihood of diversion from the civilian nuclear power fuel cycle are critical for a global nuclear future. IAEA Safeguards have been effective in minimizing opportunities for diversion; however, recent changes in the global political climate suggest implementation of additional technology and methods to ensure the prompt detection of proliferation. For a variety of reasons, nuclear facilities are becoming increasingly automated and will require minimum manual operation. This trend provides an opportunity to utilize the abundance of process information for monitoring proliferation risk, especially in future facilities. A framework that monitors process information continuously can lead to greater transparency of nuclear fuel cycle activities and can demonstrate the ability to resist proliferation associated with these activities. Additionally, a framework designed to monitor processes will ensure the legitimate use of nuclear material. This report describes recent efforts to develop a methodology capable of assessing proliferation risk in support of overall plant transparency. The framework may be tested at the candidate site located in Japan: the Fuel Handling Training Model designed for the Monju Fast Reactor at the International Cooperation and Development Training Center of the Japan Atomic Energy Agency.

McClellan, Yvonne; York, David L.; Inoue, Naoko (Japan Atomic Energy Agency, Ibaraki, Japan); Love, Tracia L.; Rochau, Gary Eugene

2006-02-01T23:59:59.000Z

264

International Nuclear Fuel Cycle Fact Book. Revision 5  

SciTech Connect (OSTI)

This Fact Book has been compiled in an effort to provide: (1) an overview of worldwide nuclear power and fuel cycle programs; and (2) current data concerning fuel cycle and waste management facilities, R and D programs, and key personnel in countries other than the United States. Additional information on each country's program is available in the International Source Book: Nuclear Fuel Cycle Research and Development, PNL-2478, Rev. 2. The Fact Book is organized as follows: (1) Overview section - summary tables which indicate national involvement in nuclear reactor, fuel cycle, and waste management development activities; (2) national summaries - a section for each country which summarizes nuclear policy, describes organizational relationships and provides addresses, names of key personnel, and facilities information; (3) international agencies - a section for each of the international agencies which has significant fuel cycle involvement; (4) energy supply and demand - summary tables, including nuclear power projections; (5) fuel cycle - summary tables; and (6) travel aids international dialing instructions, international standard time chart, passport and visa requirements, and currency exchange rate.

Harmon, K.M.; Lakey, L.T.; Leigh, I.W.; Jeffs, A.G.

1985-01-01T23:59:59.000Z

265

International nuclear fuel cycle fact book. Revision 4  

SciTech Connect (OSTI)

This Fact Book has been compiled in an effort to provide (1) an overview of worldwide nuclear power and fuel cycle programs and (2) current data concerning fuel cycle and waste management facilities, R and D programs, and key personnel in countries other than the United States. Additional information on each country's program is available in the International Source Book: Nuclear Fuel Cycle Research and Development, PNL-2478, Rev. 2. The Fact Book is organized as follows: (1) Overview section - summary tables which indicate national involvement in nuclear reactor, fuel cycle, and waste management development activities; (2) national summaries - a section for each country which summarizes nuclear policy, describes organizational relationships and provides addresses, names of key personnel, and facilities information; (3) international agencies - a section for each of the international agencies which has significant fuel cycle involvement; (4) energy supply and demand - summary tables, including nuclear power projections; (5) fuel cycle - summary tables; and (6) travel aids - international dialing instructions, international standard time chart, passport and visa requirements, and currency exchange rate.

Harmon, K.M.; Lakey, L.T.; Leigh, I.W.

1984-03-01T23:59:59.000Z

266

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

267

Non-Proliferative, Thorium-Based, Core and Fuel Cycle for Pressurized Water Reactors  

SciTech Connect (OSTI)

Two of the major barriers to the expansion of worldwide adoption of nuclear power are related to proliferation potential of the nuclear fuel cycle and issues associated with the final disposal of spent fuel. The Radkowsky Thorium Fuel (RTF) concept proposed by Professor A. Radkowsky offers a partial solution to these problems. The main idea of the concept is the utilization of the seed-blanket unit (SBU) fuel assembly geometry which is a direct replacement for a 'conventional' assembly in either a Russian pressurized water reactor (VVER-1000) or a Western pressurized water reactor (PWR). The seed-blanket fuel assembly consists of a fissile (U) zone, known as seed, and a fertile (Th) zone known as blanket. The separation of fissile and fertile allows separate fuel management schemes for the thorium part of the fuel (a subcritical 'blanket') and the 'driving' part of the core (a supercritical 'seed'). The design objective for the blanket is an efficient generation and in-situ fissioning of the U233 isotope, while the design objective for the seed is to supply neutrons to the blanket in a most economic way, i.e. with minimal investment of natural uranium. The introduction of thorium as a fertile component in the nuclear fuel cycle significantly reduces the quantity of plutonium production and modifies its isotopic composition, reducing the overall proliferation potential of the fuel cycle. Thorium based spent fuel also contains fewer higher actinides, hence reducing the long-term radioactivity of the spent fuel. The analyses show that the RTF core can satisfy the requirements of fuel cycle length, and the safety margins of conventional pressurized water reactors. The coefficients of reactivity are comparable to currently operating VVER's/PWR's. The major feature of the RTF cycle is related to the total amount of spent fuel discharged for each cycle from the reactor core. The fuel management scheme adopted for RTF core designs allows a significant decrease in the amount of discharged spent fuel, for a given energy production, compared with standard VVER/PWR. The total Pu production rate of RTF cycles is only 30 % of standard reactor. In addition, the isotopic compositions of the RTF's and standard reactor grade Pu are markedly different due to the very high burnup accumulated by the RTF spent fuel.

Todosow M.; Todosow M.; Raitses, G. (BNL) Galperin, A. (Ben Gurion University)

2009-07-12T23:59:59.000Z

268

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

269

Establishing Specifications for Low Enriched Uranium Fuel Operations Conducted Outside the High Flux Isotope Reactor Site  

SciTech Connect (OSTI)

The National Nuclear Security Administration (NNSA) has funded staff at Oak Ridge National Laboratory (ORNL) to study the conversion of the High Flux Isotope Reactor (HFIR) from the current, high enriched uranium fuel to low enriched uranium fuel. The LEU fuel form is a metal alloy that has never been used in HFIR or any HFIR-like reactor. This report provides documentation of a process for the creation of a fuel specification that will meet all applicable regulations and guidelines to which UT-Battelle, LLC (UTB) the operating contractor for ORNL - must adhere. This process will allow UTB to purchase LEU fuel for HFIR and be assured of the quality of the fuel being procured.

Pinkston, Daniel [ORNL; Primm, Trent [ORNL; Renfro, David G [ORNL; Sease, John D [ORNL

2010-10-01T23:59:59.000Z

270

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

271

Examining 239Pu and 240Pu Nuclear Resonance Fluorescence Measurements on Spent Fuel for Nuclear Safeguards  

E-Print Network [OSTI]

on Spent Fuel for Nuclear Safeguards Brian J. Quiter, ?resonances on nuclear safeguards measurements will be

Quiter, Brian

2013-01-01T23:59:59.000Z

272

A Blueprint for GNEP Advanced Burner Reactor Startup Fuel Fabrication Facility  

SciTech Connect (OSTI)

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

S. Khericha

2010-12-01T23:59:59.000Z

273

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...

274

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

275

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...

276

U.S. Spent Nuclear Fuel Data as of December 31, 2002  

Gasoline and Diesel Fuel Update (EIA)

Home > Nuclear > Spent Nuclear Fuel Home > Nuclear > Spent Nuclear Fuel Release Date: October 1, 2004 Next Release: Late 2010** Spent nuclear fuel data is collected by the Energy Information Administration (EIA) for the Office of Civilian Radioactive Waste Management (OCRWM). The spent nuclear fuel (SNF) data includes detailed characteristics of SNF generated by commercial U.S. nuclear power plants. From 1983 through 1995 this data was collected annually. Since 1996 this data has been collected every three years. The latest available detailed data covers all SNF discharged from commercial reactors before December 31, 2002, and is maintained in a data base by the EIA. Summary data tables from this data base may be found as indicated below. Table 1. Total U.S. Commercial Spent Nuclear Fuel Discharges, 1968 - 2002

277

Evaluating Environmental, Health and Safety Impacts from Two Nuclear Fuel Cycles: A Comparative Analysis of Once-Through Uranium Use and Plutonium Recycle in Light Water Reactors.  

E-Print Network [OSTI]

??Prioritizing the finite resources available to advance research, development and demonstration of the nuclear industry requires a comprehensive evaluation of potential advanced nuclear technologies to… (more)

Smith, Bethany Lee

2014-01-01T23:59:59.000Z

278

Investigation of the low enrichment conversion of the Texas A and M Nuclear Science Center Reactor  

SciTech Connect (OSTI)

The use of highly enriched uranium as a fuel research reactors is of concern due to the possibility of diversion for nuclear weapons applications. The Texas A M TRIGA reactor currently uses 70% enriched uranium in a FLIP (Fuel Life Improvement Program) fuel element manufactured by General Atomics. Thus fuel also contains 1.5 weight percent of erbium as a burnable poison to prolong useful core life. US university reactors that use highly enriched uranium will be required to covert to 20% or less enrichment to satisfy Nuclear Regulatory Commission requirements for the next core loading if the fuel is available. This investigation examined the feasibility of a material alternate to uranium-zirconium hydride for LEU conversion of a TRIGA reactor. This material is a beryllium oxide uranium dioxide based fuel. The theoretical aspects of core physics analyses were examined to assess the potential advantages of the alternative fuel. A basic model was developed for the existing core configuration since it is desired to use the present fuel element grid for the replacement core. The computing approach was calibrated to the present core and then applied to a core of BeO-UO{sub 2} fuel elements. Further calculations were performed for the General Atomics TRIGA low-enriched uranium zirconium hydride fuel.

Reuscher, J.A.

1988-01-01T23:59:59.000Z

279

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

280

Spent Nuclear Fuel (SNF) Project Execution Plan  

SciTech Connect (OSTI)

The Spent Nuclear Fuel (SNF) Project supports the Hanford Site Mission to cleanup the Site by providing safe, economic, environmentally sound management of Site spent nuclear fuel in a manner that reduces hazards by staging it to interim onsite storage and deactivates the 100 K Area facilities.

LEROY, P.G.

2000-11-03T23:59:59.000Z

Note: This page contains sample records for the topic "fuel nuclear reactors" 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

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

282

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

283

Fuels  

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

Goals > Fuels Goals > Fuels XMAT for nuclear fuels XMAT is ideally suited to explore all of the radiation processes experienced by nuclear fuels.The high energy, heavy ion accleration capability (e.g., 250 MeV U) can produce bulk damage deep in the sample, achieving neutron type depths (~10 microns), beyond the range of surface sputtering effects. The APS X-rays are well matched to the ion beams, and are able to probe individual grains at similar penetrations depths. Damage rates to 25 displacements per atom per hour (DPA/hr), and doses >2500 DPA can be achieved. MORE» Fuels in LWRs are subjected to ~1 DPA per day High burn-up fuel can experience >2000 DPA. Traditional reactor tests by neutron irradiation require 3 years in a reactor and 1 year cool down. Conventional accelerators (>1 MeV/ion) are limited to <200-400 DPAs, and

284

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

285

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

286

Modeling the Pyrochemical Reduction of Spent UO2 Fuel in a Pilot-Scale Reactor  

SciTech Connect (OSTI)

A kinetic model has been derived for the reduction of oxide spent nuclear fuel in a radial flow reactor. In this reaction, lithium dissolved in molten LiCl reacts with UO2 and fission product oxides to form a porous, metallic product. As the reaction proceeds, the depth of the porous layer around the exterior of each fuel particle increases. The observed rate of reaction has been found to be only dependent upon the rate of diffusion of lithium across this layer, consistent with a classic shrinking core kinetic model. This shrinking core model has been extended to predict the behavior of a hypothetical, pilot-scale reactor for oxide reduction. The design of the pilot-scale reactor includes forced flow through baskets that contain the fuel particles. The results of the modeling indicate that this is an essential feature in order to minimize the time needed to achieve full conversion of the fuel.

Steven D. Herrmann; Michael F. Simpson

2006-08-01T23:59:59.000Z

287

TEPP - Spent Nuclear Fuel | Department of Energy  

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

- Spent Nuclear Fuel - Spent Nuclear Fuel TEPP - Spent Nuclear Fuel This scenario provides the planning instructions, guidance, and evaluation forms necessary to conduct an exercise involving a highway shipment of spent nuclear fuel. This exercise manual is one in a series of five scenarios developed by the Department of Energy Transportation Emergency Preparedness Program. Responding agencies may include several or more of the following: local municipal and county fire, police, sheriff, and Emergency Medical Services (EMS) personnel; state, local, and federal emergency response teams; emergency response contractors;and other emergency response resources that could potentially be provided by the carrier and the originating facility (shipper). Spent Nuclear Fuel.docx More Documents & Publications

288

Verifiable Fuel Cycle Simulation Model (VISION): A Tool for Analyzing Nuclear Fuel Cycle Futures  

SciTech Connect (OSTI)

The nuclear fuel cycle consists of a set of complex components that are intended to work together. To support the nuclear renaissance, it is necessary to understand the impacts of changes and timing of events in any part of the fuel cycle system such as how the system would respond to each technological change, a series of which moves the fuel cycle from where it is to a postulated future state. The system analysis working group of the United States research program on advanced fuel cycles (formerly called the Advanced Fuel Cycle Initiative) is developing a dynamic simulation model, VISION, to capture the relationships, timing, and changes in and among the fuel cycle components to help develop an understanding of how the overall fuel cycle works. This paper is an overview of the philosophy and development strategy behind VISION. The paper includes some descriptions of the model components and some examples of how to use VISION. For example, VISION users can now change yearly the selection of separation or reactor technologies, the performance characteristics of those technologies, and/or the routing of material among separation and reactor types - with the model still operating on a PC in <5 min.

Jacob J. Jacobson; Steven J. Piet; Gretchen E. Matthern; David E. Shropshire; Robert F. Jeffers; A. M. Yacout; Tyler Schweitzer

2010-11-01T23:59:59.000Z

289

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

290

Submersion Criticality Safety Analysis of Tungsten-Based Fuel for Nuclear Power and Propulsion Applications  

SciTech Connect (OSTI)

The Center for Space Nuclear Research (CSNR) is developing tungsten-encapsulated fuels for space nuclear applications. Aims to develop NTP fuels that are; Affordable Low impact on production and testing environment Producible on a large scale over suitable time period Higher-performance compared to previous graphite NTP fuel elements Space nuclear reactors remain subcritical before and during launch, and do not go critical until required by its mission. A properly designed reactor will remain subcritical in any launch abort scenario, where the reactor falls back to Earth and becomes submerged in terrestrial material. Submersion increases neutron reflection and thermalizes the neutrons, which typically increases the reactivity of the core. This effect is usually very significant for fast-spectrum reactors. This research provided a submersion criticality safety analysis for a representative tungsten/uranium oxide fueled reactor. Determine the submersion behavior of a reactor fueled by tungsten-based fuel. Considered fuel compositions with varying: Rhenium content (wt% rhenium in tungsten) Fuel loading fractions (UO2 vol%)

A.E. Craft; R. C. O'Brien; S. D. Howe; J. C. King

2014-07-01T23:59:59.000Z

291

Assessment of the use of extended burnup fuel in light water power reactors  

SciTech Connect (OSTI)

This study has been conducted by Pacific Northwest Laboratory for the US Nuclear Regulatory Commission to review the environmental and economic impacts associated with the use of extended burnup nuclear fuel in light water power reactors. It has been proposed that current batch average burnup levels of 33 GWd/t uranium be increased to above 50 GWd/t. The environmental effects of extending fuel burnup during normal operations and during accident events and the economic effects of cost changes on the fuel cycle are discussed in this report. The physical effects of extended burnup on the fuel and the fuel assembly are also presented as a basis for the environmental and economic assessments. Environmentally, this burnup increase would have no significant impact over that of normal burnup. Economically, the increased burnup would have favorable effects, consisting primarily of a reduction: (1) total fuel requirements; (2) reactor downtime for fuel replacement; (3) the number of fuel shipments to and from reactor sites; and (4) repository storage requirements. 61 refs., 4 figs., 27 tabs.

Baker, D.A.; Bailey, W.J.; Beyer, C.E.; Bold, F.C.; Tawil, J.J.

1988-02-01T23:59:59.000Z

292

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

293

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

294

Mississippi Nuclear Profile - All Fuels  

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

total electric power industry, summer capacity and net generation, by energy source, 2010" total electric power industry, summer capacity and net generation, by energy source, 2010" "Primary energy source","Summer capacity (mw)","Share of State total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","1,251",8.0,"9,643",17.7 "Coal","2,526",16.1,"13,629",25.0 "Natural Gas","11,640",74.2,"29,619",54.4 "Other1",4,"*",10,"*" "Other Renewable1",235,1.5,"1,504",2.8 "Petroleum",35,0.2,81,0.1 "Total","15,691",100.0,"54,487",100.0 "1Municipal Solid Waste net generation is allocated according to the biogenic and non-biogenic components of the fuel; however, all Municipal Solid Waste summer capacity is classified as Renewable."

295

Iowa Nuclear Profile - All Fuels  

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

total electric power industry, summer capacity and net generation, by energy source, 2010" total electric power industry, summer capacity and net generation, by energy source, 2010" "Primary energy source","Summer capacity (mw)","Share of State total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear",601,4.1,"4,451",7.7 "Coal","6,956",47.7,"41,283",71.8 "Hydro and Pumped Storage",144,1.0,948,1.6 "Natural Gas","2,299",15.8,"1,312",2.3 "Other Renewable1","3,584",24.6,"9,360",16.3 "Petroleum","1,007",6.9,154,0.3 "Total","14,592",100.0,"57,509",100.0 "1Municipal Solid Waste net generation is allocated according to the biogenic and non-biogenic components of the fuel; however, all Municipal Solid Waste summer capacity is classified as Renewable."

296

Vermont Nuclear Profile - All Fuels  

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

total electric power industry, summer capacity and net generation, by energy source, 2010" total electric power industry, summer capacity and net generation, by energy source, 2010" "Primary energy source","Summer capacity (mw)","Share of State total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear",620,55.0,"4,782",72.2 "Hydro and Pumped Storage",324,28.7,"1,347",20.3 "Natural Gas","-","-",4,0.1 "Other Renewable1",84,7.5,482,7.3 "Petroleum",100,8.9,5,0.1 "Total","1,128",100.0,"6,620",100.0 "1Municipal Solid Waste net generation is allocated according to the biogenic and non-biogenic components of the fuel; however, all Municipal Solid Waste summer capacity is classified as Renewable."

297

Ohio Nuclear Profile - All Fuels  

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

total electric power industry, summer capacity and net generation, by energy source, 2010" total electric power industry, summer capacity and net generation, by energy source, 2010" "Primary energy source","Summer capacity (mw)","Share of State total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","2,134",6.5,"15,805",11.0 "Coal","21,360",64.6,"117,828",82.1 "Hydro and Pumped Storage",101,0.3,429,0.3 "Natural Gas","8,203",24.8,"7,128",5.0 "Other1",123,0.4,266,0.2 "Other Renewable1",130,0.4,700,0.5 "Petroleum","1,019",3.1,"1,442",1.0 "Total","33,071",100.0,"143,598",100.0 "1Municipal Solid Waste net generation is allocated according to the biogenic and non-biogenic components of the fuel; however, all Municipal Solid Waste summer capacity is classified as Renewable."

298

Maryland Nuclear Profile - All Fuels  

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

total electric power industry, summer capacity and net generation, by energy source, 2010" total electric power industry, summer capacity and net generation, by energy source, 2010" "Primary energy source","Summer capacity (mw)","Share of State total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","1,705",13.6,"13,994",32.1 "Coal","4,886",39.0,"23,668",54.3 "Hydro and Pumped Storage",590,4.7,"1,667",3.8 "Natural Gas","2,041",16.3,"2,897",6.6 "Other1",152,1.2,485,1.1 "Other Renewable1",209,1.7,574,1.3 "Petroleum","2,933",23.4,322,0.7 "Total","12,516",100.0,"43,607",100.0 "1Municipal Solid Waste net generation is allocated according to the biogenic and non-biogenic components of the fuel; however, all Municipal Solid Waste summer capacity is classified as Renewable."

299

Kansas Nuclear Profile - All Fuels  

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

total electric power industry, summer capacity and net generation, by energy source, 2010" total electric power industry, summer capacity and net generation, by energy source, 2010" "Primary energy source","Summer capacity (mw)","Share of State total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","1,160",9.2,"9,556",19.9 "Coal","5,179",41.3,"32,505",67.8 "Hydro and Pumped Storage",3,"*",13,"*" "Natural Gas","4,573",36.5,"2,287",4.8 "Other Renewable1","1,079",8.6,"3,459",7.2 "Petroleum",550,4.4,103,0.2 "Total","12,543",100.0,"47,924",100.0 "1Municipal Solid Waste net generation is allocated according to the biogenic and non-biogenic components of the fuel; however, all Municipal Solid Waste summer capacity is classified as Renewable."

300

Connecticut Nuclear Profile - All Fuels  

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

total electric power industry, summer capacity and net generation, by energy source, 2010" total electric power industry, summer capacity and net generation, by energy source, 2010" "Primary energy source","Summer capacity (mw)","Share of State total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","2,103",25.4,"16,750",50.2 "Coal",564,6.8,"2,604",7.8 "Hydro and Pumped Storage",151,1.8,400,1.2 "Natural Gas","2,292",27.7,"11,716",35.1 "Other1",27,0.3,730,2.2 "Other Renewable1",159,1.9,740,2.2 "Petroleum","2,989",36.1,409,1.2 "Total","8,284",100.0,"33,350",100.0 "1Municipal Solid Waste net generation is allocated according to the biogenic and non-biogenic components of the fuel; however, all Municipal Solid Waste summer capacity is classified as Renewable."

Note: This page contains sample records for the topic "fuel nuclear reactors" from the National Library of EnergyBeta (NLEBeta).
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they are not comprehensive nor are they the most current set.
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301

Nebraska Nuclear Profile - All Fuels  

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

total electric power industry, summer capacity and net generation, by energy source, 2010" total electric power industry, summer capacity and net generation, by energy source, 2010" "Primary energy source","Summer capacity (mw)","Share of State total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","1,245",15.8,"11,054",30.2 "Coal","3,932",50.0,"23,363",63.8 "Hydro and Pumped Storage",278,3.5,"1,314",3.6 "Natural Gas","1,849",23.5,375,1.0 "Other Renewable1",165,2.1,493,1.3 "Petroleum",387,4.9,31,0.1 "Total","7,857",100.0,"36,630",100.0 "1Municipal Solid Waste net generation is allocated according to the biogenic and non-biogenic components of the fuel; however, all Municipal Solid Waste summer capacity is classified as Renewable."

302

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

E-Print Network [OSTI]

nuclear tors. for of of These studies can examine safety systems or safety research programsnuclear power plants, and at risk. to reduce population The Light-water Reactor Safety Research Program

Nero, A.V.

2010-01-01T23:59:59.000Z

303

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

304

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

SciTech Connect (OSTI)

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

David Petti; Hans Gougar; Gary Bell

2005-05-01T23:59:59.000Z

305

Development of a nuclear fuel cycle transparency framework.  

SciTech Connect (OSTI)

Nuclear fuel cycle transparency can be defined as a confidence building approach among political entities to ensure civilian nuclear facilities are not being used for the development of nuclear weapons. Transparency concepts facilitate the transfer of nuclear technology, as the current international political climate indicates a need for increased methods of assuring non-proliferation. This research develops a system which will augment current non-proliferation assessment activities undertaken by U.S. and international regulatory agencies. It will support the export of nuclear technologies, as well as the design and construction of Gen. IV energy systems. Additionally, the framework developed by this research will provide feedback to cooperating parties, thus ensuring full transparency of a nuclear fuel cycle. As fuel handling activities become increasingly automated, proliferation or diversion potential of nuclear material still needs to be assessed. However, with increased automation, there exists a vast amount of process data to be monitored. By designing a system that monitors process data continuously, and compares this data to declared process information and plant designs, a faster and more efficient assessment of proliferation risk can be made. Figure 1 provides an illustration of the transparency framework that has been developed. As shown in the figure, real-time process data is collected at the fuel cycle facility; a reactor, a fabrication plant, or a recycle facility, etc. Data is sent to the monitoring organization and is assessed for proliferation risk. Analysis and recommendations are made to cooperating parties, and feedback is provided to the facility. The analysis of proliferation risk is based on the following factors: (1) Material attractiveness: the quantification of factors relevant to the proliferation risk of a certain material (e.g., highly enriched Pu-239 is more attractive than that of lower enrichment) (2) The static (baseline) risk: the quantification of risk factors regarding the expected value of proliferation risk under normal (not proliferating) operations. (3) The dynamic (changing) risk: the quantification of risk factors regarding the observed value of proliferation risk, based on monitor signals from facility operations. This framework could be implemented at facilities which have been exported (for instance, to third world countries), or facilities located in sensitive countries. Sandia National Laboratories is currently working with the Japan Nuclear Cycle Development Institute (JNC) to implement a demonstration of nuclear fuel cycle transparency technology at the Fuel Handling Training Model designed for the Monju Fast Reactor at the International Cooperation and Development Training Center in Japan. This technology has broad applications, both in the U.S. and abroad. Following the demonstration, we expect to begin further testing of the technology at an Enrichment Facility, a Fast Reactor, and at a Recycle Facility.

Love, Tracia L.

2005-04-01T23:59:59.000Z

306

Nuclear power plants: structure and function  

SciTech Connect (OSTI)

Topics discussed include: steam electric plants; BWR type reactors; PWR type reactors; thermal efficiency of light water reactors; other types of nuclear power plants; the fission process and nuclear fuel; fission products and reactor afterheat; and reactor safety.

Hendrie, J.M.

1983-01-01T23:59:59.000Z

307

Global Nuclear Energy Partnership Fact Sheet - Establish Reliable Fuel  

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

Global Nuclear Energy Partnership Fact Sheet - Establish Reliable Global Nuclear Energy Partnership Fact Sheet - Establish Reliable Fuel Services Global Nuclear Energy Partnership Fact Sheet - Establish Reliable Fuel Services GNEP would build and strengthen a reliable international fuel services consortium under which "fuel supplier nations" would choose to operate both nuclear power plants and fuel production and handling facilities, providing reliable fuel services to "user nations" that choose to only operate nuclear power plants. This international consortium is a critical component of the GNEP initiative to build an improved, more proliferation-resistant nuclear fuel cycle that recycles used fuel, while Global Nuclear Energy Partnership Fact Sheet - Establish Reliable Fuel Services More Documents & Publications

308

High Performance Fuel Desing for Next Generation Pressurized Water Reactors  

SciTech Connect (OSTI)

The use of internally and externally cooled annular fule rods for high power density Pressurized Water Reactors is assessed. The assessment included steady state and transient thermal conditions, neutronic and fuel management requirements, mechanical vibration issues, fuel performance issues, fuel fabrication methods and econmic assessment. The investigation was donducted by a team from MIT, Westinghouse, Gamma Engineering, Framatome ANP, and AECL. The analyses led to the conclusion that raising the power density by 50% may be possible with this advanced fuel. Even at the 150% power level, the fuel temperature would be a few hundred degrees lower than the current fuel temperatre. Significant economic and safety advantages can be obtained by using this fuel in new reactors. Switching to this type of fuel for existing reactors would yield safety advantages, but the economic return is dependent on the duration of plant shutdown to accommodate higher power production. The main feasiblity issue for the high power performance appears to be the potential for uneven splitting of heat flux between the inner and outer fuel surfaces due to premature closure of the outer fuel-cladding gap. This could be overcome by using a very narrow gap for the inner fuel surface and/or the spraying of a crushable zirconium oxide film at the fuel pellet outer surface. An alternative fuel manufacturing approach using vobropacking was also investigated but appears to yield lower than desirable fuel density.

Mujid S. Kazimi; Pavel Hejzlar

2006-01-31T23:59:59.000Z

309

PROTEUS - Simulation Toolset for Reactor Physics and Fuel Cycle Analysis  

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

Simulation Toolset for Simulation Toolset for Reactor Physics and Fuel Cycle Analysis PROTEUS Faster and more accurate neutronics calculations enable optimum reactor design... Argonne National Laboratory's powerful reactor physics toolset, PROTEUS, empowers users to create optimal reactor designs quickly, reliably and accurately. ...Reducing costs for designers of fast spectrum reactors. PROTEUS' long history of validation provides confidence in predictive simulations Argonne's simulation tools have more than 30 years of validation history against numerous experiments and measurements. The tools within PROTEUS work together, using the same interface files for easier integration of calculations. Multi-group Fast Reactor Cross Section Processing: MC 2 -3 No other fast spectrum multigroup generation tool

310

World nuclear fuel cycle requirements 1990  

SciTech Connect (OSTI)

This analysis report presents the projected requirements for uranium concentrate and uranium enrichment services to fuel the nuclear power plants expected to be operating under three nuclear supply scenarios. Two of these scenarios, the Lower Reference and Upper Reference cases, apply to the United States, Canada, Europe, the Far East, and other countries with free market economies (FME countries). A No New Orders scenario is presented only for the United States. These nuclear supply scenarios are described in Commercial Nuclear Power 1990: Prospects for the United States and the World (DOE/EIA-0438(90)). This report contains an analysis of the sensitivities of the nuclear fuel cycle projections to different levels and types of projected nuclear capacity, different enrichment tails assays, higher and lower capacity factors, changes in nuclear fuel burnup levels, and other exogenous assumptions. The projections for the United States generally extend through the year 2020, and the FME projections, which include the United States, are provided through 2010. The report also presents annual projections of spent nuclear fuel discharges and inventories of spent fuel. Appendix D includes domestic spent fuel projections through the year 2030 for the Lower and Upper Reference cases and through 2040, the last year in which spent fuel is discharged, for the No New Orders case. These disaggregated projections are provided at the request of the Department of Energy's Office of Civilian Radioactive Waste Management.

Not Available

1990-10-26T23:59:59.000Z

311

Nuclear Fuels Storage and Transportation Planning Project (NFST...  

Office of Environmental Management (EM)

Nuclear Fuels Storage and Transportation Planning Project (NFST) Program Status Nuclear Fuels Storage and Transportation Planning Project (NFST) Program Status Presentation made by...

312

2008 DOE Spent Nuclear Fuel and High Level Waste Inventory  

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

Management >> National Spent Nuclear Fuel INL Logo Search 2008 DOE Spent Nuclear Fuel and High Level Waste Inventory Content Goes Here Skip Navigation Links Home Newsroom About INL...

313

Department of Energy Awards $15 Million for Nuclear Fuel Cycle...  

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

Million for Nuclear Fuel Cycle Technology Research and Development Department of Energy Awards 15 Million for Nuclear Fuel Cycle Technology Research and Development August 1,...

314

Used Nuclear Fuel Loading and Structural Performance Under Normal...  

Office of Environmental Management (EM)

Used Nuclear Fuel Loading and Structural Performance Under Normal Conditions of Transport - Modeling, Simulation and Experimental Integration RD&D Plan Used Nuclear Fuel Loading...

315

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

Science Journals Connector (OSTI)

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

H. E. Knee

1991-01-01T23:59:59.000Z

316

Comparative Fuel Cycle Analysis of Critical and Subcritical Fast Reactor Transmutation Systems  

SciTech Connect (OSTI)

Fuel cycle analyses are performed to evaluate the impacts of further transmutation of spent nuclear fuel on high-level and low-level waste mass flows into repositories, on the composition and toxicity of the high-level waste, on the capacity of high-level waste repositories, and on the proliferation resistance of the high-level waste. Storage intact of light water reactor (LWR) spent nuclear fuel, a single recycle in a LWR of the plutonium as mixed-oxide fuel, and the repeated recycle of the transuranics in critical and subcritical fast reactors are compared with the focus on the waste management performance of these systems. Other considerations such as cost and technological challenges were beyond the scope of this study. The overall conclusion of the studies is that repeated recycling of the transuranics from spent nuclear fuel would significantly increase the capacity of high-level waste repositories per unit of nuclear energy produced, significantly increase the nuclear energy production per unit mass of uranium ore mined, significantly reduce the radiotoxicity of the waste streams per unit of nuclear energy produced, and significantly enhance the proliferation resistance of the material stored in high-level waste repositories.

Hoffman, Edward A.; Stacey, Weston M. [Georgia Institute of Technology (United States)

2003-10-15T23:59:59.000Z

317

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

318

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

319

Fuel cycle analysis of once-through nuclear systems.  

SciTech Connect (OSTI)

Once-through fuel cycle systems are commercially used for the generation of nuclear power, with little exception. The bulk of these once-through systems have been water-cooled reactors (light-water and heavy water reactors, LWRs and HWRs). Some gas-cooled reactors are used in the United Kingdom. The commercial power systems that are exceptions use limited recycle (currently one recycle) of transuranic elements, primarily plutonium, as done in Europe and nearing deployment in Japan. For most of these once-through fuel cycles, the ultimate storage of the used (spent) nuclear fuel (UNF, SNF) will be in a geologic repository. Besides the commercial nuclear plants, new once-through concepts are being proposed for various objectives under international advanced nuclear fuel cycle studies and by industrial and venture capital groups. Some of the objectives for these systems include: (1) Long life core for remote use or foreign export and to support proliferation risk reduction goals - In these systems the intent is to achieve very long core-life with no refueling and limited or no access to the fuel. Most of these systems are fast spectrum systems and have been designed with the intent to improve plant economics, minimize nuclear waste, enhance system safety, and reduce proliferation risk. Some of these designs are being developed under Generation IV International Forum activities and have generally not used fuel blankets and have limited the fissile content of the fuel to less than 20% for the purpose on meeting international nonproliferation objectives. In general, the systems attempt to use transuranic elements (TRU) produced in current commercial nuclear power plants as this is seen as a way to minimize the amount of the problematic radio-nuclides that have to be stored in a repository. In this case, however, the reprocessing of the commercial LWR UNF to produce the initial fuel will be necessary. For this reason, some of the systems plan to use low enriched uranium (LEU) fuels. Examples of systems in this class include the small modular reactors being considered internationally; e.g. 4S [Tsuboi 2009], Hyperion Power Module [Deal 2010], ARC-100 [Wade 2010], and SSTAR [Smith 2008]. (2) Systems for Resource Utilization - In recent years, interest has developed in the use of advanced nuclear designs for the effective utilization of fuel resources. Systems under this class have generally utilized the breed and burn concept in which fissile material is bred and used in situ in the reactor core. Due to the favorable breeding that is possible with fast neutrons, these systems have tended to be fast spectrum systems. In the once-through concepts (as opposed to the traditional multirecycle approach typically considered for fast reactors), an ignition (or starter) zone contains driver fuel which is fissile material. This zone is designed to last a long time period to allow the breeding of sufficient fissile material in the adjoining blanket zone. The blanket zone is initially made of fertile depleted uranium fuel. This zone could also be made of fertile thorium fuel or recovered uranium from fuel reprocessing or natural uranium. However, given the bulk of depleted uranium and the potentially large inventory of recovered uranium, it is unlikely that the use of thorium is required in the near term in the U.S. Following the breeding of plutonium or fissile U-233 in the blanket, this zone or assembly then carries a larger fraction of the power generation in the reactor. These systems tend to also have a long cycle length (or core life) and they could be with or without fuel shuffling. When fuel is shuffled, the incoming fuel is generally depleted uranium (or thorium) fuel. In any case, fuel is burned once and then discharged. Examples of systems in this class include the CANDLE concept [Sekimoto 2001], the traveling wave reactor (TWR) concept of TerraPower [Ellis 2010], the ultra-long life fast reactor (ULFR) by ANL [Kim 2010], and the BNL fast mixed spectrum reactor (FMSR) concept [Fisher 1979]. (3) Thermal systems for resource extensio

Kim, T. K.; Taiwo, T. A.; Nuclear Engineering Division

2010-08-10T23:59:59.000Z

320

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

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


321

Nuclear 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

322

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

323

Commercial utilization of weapon grade plutonium as TRISO fuel in conventional CANDU reactors  

Science Journals Connector (OSTI)

Large quantities of weapon grade (WG) plutonium have been accumulated in the nuclear warheads. Plutonium and heavy water moderator can give a good combination with respect to neutron economy. TRISO type fuel can withstand very high fuel burn up levels. The paper investigates the prospects of utilization of TRISO fuel made of WG-plutonium in CANDU reactors. Three different fuel compositions have been investigated: (1): 90% ThC + 10% PuC, (2): 70% ThC + 30% PuC and (3): 50% ThC + 50% PuC. The temporal variation of the criticality k? and the burn-up values of the reactor have been calculated by full power operation up to 17 years. Calculated startup criticalities for these fuel modes are k?,0 = 1.6403, 1.7228 and 1.7662, respectively. Attainable burn up values and reactor operation times without new fuel charge will be 94 700, 265 000 and 425 000 MW.D/MT and along with continuous operation periods of ?3.5, 10 and 17 years, respectively, for the corresponding modes. These high burn ups would reduce fuel fabrication costs and nuclear waste mass for final disposal per unit energy drastically.

Sümer ?ahin; Hac? Mehmet ?ahin; Adem Ac?r

2012-01-01T23:59:59.000Z

324

Direct Investigations of the Immobilization of Radionuclides in the Alteration Products of Spent Nuclear Fuel  

SciTech Connect (OSTI)

Safe disposal of the nation's nuclear waste in a geological repository involves unique scientific and engineering challenges owing to the very long-lived radioactivity of the waste. The repository must retain a variety of radionuclides that have vastly different chemical characters for several thousand years. Most of the radioactivity that will be housed in the proposed repository at Yucca Mountain will be associated with spent nuclear fuel, much of which is derived from commercial reactors. DOE is custodian of approximately 8000 tons of spent nuclear fuel that is also intended for eventual disposal in a geological repository. Unlike the spent fuel from commercial reactors, the DOE fuel is diverse in composition with more than 250 varieties. Safe disposal of spent fuel requires a detailed knowledge of its long-term behavior under repository conditions, as well as the fate of radionuclides released from the spent fuel as waste containers are breached.

Peter C. Burns; Robert J. Finch; David J. Wronkiewicz

2004-12-27T23:59:59.000Z

325

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

326

International Atomic Energy Agency support of research reactor highly enriched uranium to low enriched uranium fuel conversion projects  

SciTech Connect (OSTI)

The IAEA has been involved for more than twenty years in supporting international nuclear non- proliferation efforts associated with reducing the amount of highly enriched uranium (HEU) in international commerce. IAEA projects and activities have directly supported the Reduced Enrichment for Research and Test Reactors (RERTR) programme, as well as directly assisted efforts to convert research reactors from HEU to LEU fuel. HEU to LEU fuel conversion projects differ significantly depending on several factors including the design of the reactor and fuel, technical needs of the member state, local nuclear infrastructure, and available resources. To support such diverse endeavours, the IAEA tailors each project to address the relevant constraints. This paper presents the different approaches taken by the IAEA to address the diverse challenges involved in research reactor HEU to LEU fuel conversion projects. Examples of conversion related projects in different Member States are fully detailed. (author)

Bradley, E.; Adelfang, P.; Goldman, I.N. [Research Reactors Unit, Division of Nuclear Fuel Cycle and Waste Technology, International Atomic Energy Agency, Wagramer Strasse 5, P.O. Box 100, A-1400 Vienna (Austria)

2008-07-15T23:59:59.000Z

327

Cost Savings of Nuclear Power with Total Fuel Reprocessing  

SciTech Connect (OSTI)

The cost of fast reactor (FR) generated electricity with pyro-processing is estimated in this article. It compares favorably with other forms of energy and is shown to be less than that produced by light water reactors (LWR's). FR's use all the energy in natural uranium whereas LWR's utilize only 0.7% of it. Because of high radioactivity, pyro-processing is not open to weapon material diversion. This technology is ready now. Nuclear power has the same advantage as coal power in that it is not dependent upon a scarce foreign fuel and has the significant additional advantage of not contributing to global warming or air pollution. A jump start on new nuclear plants could rapidly allow electric furnaces to replace home heating oil furnaces and utilize high capacity batteries for hybrid automobiles: both would reduce US reliance on oil. If these were fast reactors fueled by reprocessed fuel, the spent fuel storage problem could also be solved. Costs are derived from assumptions on the LWR's and FR's five cost components: 1) Capital costs: LWR plants cost $106/MWe. FR's cost 25% more. Forty year amortization is used. 2) The annual O and M costs for both plants are 9% of the Capital Costs. 3) LWR fuel costs about 0.0035 $/kWh. Producing FR fuel from spent fuel by pyro-processing must be done in highly shielded hot cells which is costly. However, the five foot thick concrete walls have the advantage of prohibiting diversion. LWR spent fuel must be used as feedstock for the FR initial core load and first two reloads so this FR fuel costs more than LWR fuel. FR fuel costs much less for subsequent core reloads (< LWR fuel) if all spent fuel feedstock is from the fast reactor (i.e., Breeding Ratio =1). 4) Yucca Mountain storage of unprocessed LWR spent fuel is estimated as $360,000/MTHM. But this fuel can be processed to remove TRU for use as fast reactor fuel. The remaining fission products repository costs are only one fifth that of the original fuel. Storage of short half life fission products alone requires less storage time and long term integrity than LWR spent fuel (300 years storage versus 100,000 years.) 5) LWR decommissioning costs are estimated to be $0.3 x 10{sup 6}/MWe. The annual cost for a 40 year licensed plant would be 2.5 % of this or less if interest is taken into account. All plants will eventually have to replace those components which become radiation damaged. FR's should be designed to replace parts rather than decommission. The LWR costs are estimated to be 2.65 cents/kWh. FR costs are 2.99 cents/kWh for the first 7.5 years and 2.39 cents/kWh for the next 32.5 years. The average cost over forty years is 2.50 cents/kWh which is less than the LWR costs. These power costs are similar to coal power, are lower than gas, oil, and much lower than renewable power.(authors)

Solbrig, Charles W.; Benedict, Robert W. [Fuel Cycle Programs Division, Idaho National Laboratory, Idaho Falls, Idaho (United States)

2006-07-01T23:59:59.000Z

328

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

329

German concept and status of the disposal of spent fuel elements from German research reactors  

SciTech Connect (OSTI)

Eight research reactors with a power {>=} 100 kW are currently being operated in the Federal Republic of Germany. These comprise three TRIGA-type reactors (power 100 kW to 250 kW), four swimming-pool reactors (power 1 MW to 10 MW) and one DIDO type reactor (power 23 MW). The German research reactors are used for neutron scattering for basic research in the field of solid state research, neutron metrology, for the fabrication of isotopes and for neutron activation analysis for medicine and biology, for investigating the influence of radiation on materials and for nuclear fuel behavior. It will be vital to continue current investigations in the future. Further operation of the German research reactors is therefore indispensable. Safe, regular disposal of the irradiated fuel elements arising now and in future operation is of primary importance. Furthermore, there are several plants with considerable quantities of spent fuel, the safe disposal of which is a matter of urgency. These include above all the VKTA facilities in Rossendorf and also the TRIGA reactors, where disposal will only be necessary upon decommissioning. The present paper report is concerned with the disposal of fuel from the German research reactors. It briefly deals with the situation in the USA since the end of 1988, describes interim solutions for current disposal requirements and then mainly concentrates on the German disposal concept currently being prepared. This concept initially envisages the long-term (25--50 years) dry interim storage of fuel elements in special containers in a central German interim store with subsequent direct final disposal without reprocessing of the irradiated fuel.

Komorowski, K. [Bundesministerium fuer Bildung Wissenschaft, Bonn (Germany); Storch, S.; Thamm, G. [Forschungszentrum Juelich GmbH (Germany)

1995-12-31T23:59:59.000Z

330

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

331

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

332

Nuclear Fuel Cycle | Department of Energy  

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

Fuel Cycle Fuel Cycle Nuclear Fuel Cycle GC-52 provides legal advice to DOE regarding research and development of nuclear fuel and waste management technologies that meet the nation's energy supply, environmental, and energy security needs. GC-52 also advises DOE on issues involving support for international fuel cycle initiatives aimed at advancing a common vision of the necessity of the expansion of nuclear energy for peaceful purposes worldwide in a safe and secure manner. In addition, GC-52 provides legal advice to DOE regarding the management and disposition of excess uranium in DOE's uranium stockpile. GC-52 attorneys participate in meetings of DOE's Uranium Inventory Management Coordinating Committee and provide advice on compliance with statutory requirements for the sale or transfer of uranium.

333

Thermodynamics of high-temperature nuclear fuel  

Science Journals Connector (OSTI)

A method for performing a thermodynamic analysis of the high-temperature nuclear fuel using the ASTA computer program is substantiated. Calculations of the chemical composition and pressure of the gas phase of...

I. A. Belov; A. S. Ivanov

334

Chemistry of nuclear fuel reprocessing: Current status  

Science Journals Connector (OSTI)

Current status on the chemical aspects of nuclear fuel reprocessing is presented with special emphasis on the Purex process which continues to be the process of choice for the last four decades. Better deconta...

D. D. Sood; S. K. Patil

1996-03-01T23:59:59.000Z

335

Transuranium Elements in the Nuclear Fuel Cycle  

Science Journals Connector (OSTI)

Transuranium elements, neptunium, plutonium, americium, and curium, are formed via neutron capture processes of actinides, and are mainly by-products of fuel irradiation during the operation of a nuclear react...

Thomas Fanghänel; Jean-Paul Glatz; Rudy J. M. Konings…

2010-01-01T23:59:59.000Z

336

Fuel availability in nuclear power.  

E-Print Network [OSTI]

?? Nuclear power is in focus of attention due to several factors these days and the expression “nuclear renaissance” is getting well known. However, concerned… (more)

Söderlund, Karl

2009-01-01T23:59:59.000Z

337

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

SciTech Connect (OSTI)

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

Shannon Bragg-Sitton

2014-02-01T23:59:59.000Z

338

MOX fuel arrangement for nuclear core  

DOE Patents [OSTI]

In order to use up a stockpile of weapons-grade plutonium, the plutonium is converted into a mixed oxide (MOX) fuel form wherein it can be disposed in a plurality of different fuel assembly types. Depending on the equilibrium cycle that is required, a predetermined number of one or more of the fuel assembly types are selected and arranged in the core of the reactor in accordance with a selected loading schedule. Each of the fuel assemblies is designed to produce different combustion characteristics whereby the appropriate selection and disposition in the core enables the resulting equilibrium cycle to closely resemble that which is produced using urania fuel. The arrangement of the MOX rods and burnable absorber rods within each of the fuel assemblies, in combination with a selective control of the amount of plutonium which is contained in each of the MOX rods, is used to tailor the combustion characteristics of the assembly.

Kantrowitz, Mark L. (Portland, CT); Rosenstein, Richard G. (Windsor, CT)

2001-07-17T23:59:59.000Z

339

Mox fuel arrangement for nuclear core  

DOE Patents [OSTI]

In order to use up a stockpile of weapons-grade plutonium, the plutonium is converted into a mixed oxide (MOX) fuel form wherein it can be disposed in a plurality of different fuel assembly types. Depending on the equilibrium cycle that is required, a predetermined number of one or more of the fuel assembly types are selected and arranged in the core of the reactor in accordance with a selected loading schedule. Each of the fuel assemblies is designed to produce different combustion characteristics whereby the appropriate selection and disposition in the core enables the resulting equilibrium cycle to closely resemble that which is produced using urania fuel. The arrangement of the MOX rods and burnable absorber rods within each of the fuel assemblies, in combination with a selective control of the amount of plutonium which is contained in each of the MOX rods, is used to tailor the combustion. characteristics of the assembly.

Kantrowitz, Mark L. (Portland, CT); Rosenstein, Richard G. (Windsor, CT)

2001-05-15T23:59:59.000Z

340

MOX fuel arrangement for nuclear core  

DOE Patents [OSTI]

In order to use up a stockpile of weapons-grade plutonium, the plutonium is converted into a mixed oxide (MOX) fuel form wherein it can be disposed in a plurality of different fuel assembly types. Depending on the equilibrium cycle that is required, a predetermined number of one or more of the fuel assembly types are selected and arranged in the core of the reactor in accordance with a selected loading schedule. Each of the fuel assemblies is designed to produce different combustion characteristics whereby the appropriate selection and disposition in the core enables the resulting equilibrium cycle to closely resemble that which is produced using urania fuel. The arrangement of the MOX rods and burnable absorber rods within each of the fuel assemblies, in combination with a selective control of the amount of plutonium which is contained in each of the MOX rods, is used to tailor the combustion characteristics of the assembly. 38 figs.

Kantrowitz, M.L.; Rosenstein, R.G.

1998-10-13T23:59:59.000Z

Note: This page contains sample records for the topic "fuel nuclear reactors" 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

MOX fuel arrangement for nuclear core  

DOE Patents [OSTI]

In order to use up a stockpile of weapons-grade plutonium, the plutonium is converted into a mixed oxide (MOX) fuel form wherein it can be disposed in a plurality of different fuel assembly types. Depending on the equilibrium cycle that is required, a predetermined number of one or more of the fuel assembly types are selected and arranged in the core of the reactor in accordance with a selected loading schedule. Each of the fuel assemblies is designed to produce different combustion characteristics whereby the appropriate selection and disposition in the core enables the resulting equilibrium cycle to closely resemble that which is produced using urania fuel. The arrangement of the MOX rods and burnable absorber rods within each of the fuel assemblies, in combination with a selective control of the amount of plutonium which is contained in each of the MOX rods, is used to tailor the combustion characteristics of the assembly.

Kantrowitz, Mark L. (Portland, CT); Rosenstein, Richard G. (Windsor, CT)

1998-01-01T23:59:59.000Z

342

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

343

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

344

Nuclear Fuel Cycle and Waste Management Technologies - Nuclear Engineering  

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

Nuclear Fuel Cycle and Nuclear Fuel Cycle and Waste Management Technologies 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 Fuel Cycle and Waste Management Technologies Overview Bookmark and Share Much of the NE Division's research is directed toward developing software and performing analyses, system engineering design, and experiments to support the demonstration and optimization of the electrometallurgical

345

Evaluation of Radiation Impacts of Spent Nuclear Fuel Storage (SNFS-2) of Chernobyl NPP - 13495  

SciTech Connect (OSTI)

Radiation effects are estimated for the operation of a new dry storage facility for spent nuclear fuel (SNFS-2) of Chernobyl NPP RBMK reactors. It is shown that radiation exposure during normal operation, design and beyond design basis accidents are minor and meet the criteria for safe use of radiation and nuclear facilities in Ukraine. (authors)

Paskevych, Sergiy; Batiy, Valiriy; Sizov, Andriy [Institute for Safety Problems of Nuclear Power Plants, National Academy of Sciences of Ukraine, 36 a Kirova str. Chornobyl, Kiev region, 07200 (Ukraine)] [Institute for Safety Problems of Nuclear Power Plants, National Academy of Sciences of Ukraine, 36 a Kirova str. Chornobyl, Kiev region, 07200 (Ukraine); Schmieman, Eric [Battelle Memorial Institute, PO Box 999 MSIN K6-90, Richland, WA 99352 (United States)] [Battelle Memorial Institute, PO Box 999 MSIN K6-90, Richland, WA 99352 (United States)

2013-07-01T23:59:59.000Z

346

Radiological consequences of ship collisions that might occur in U.S. Ports during the shipment of foreign research reactor spent nuclear fuel to the United States in break-bulk freighters  

SciTech Connect (OSTI)

Accident source terms, source term probabilities, consequences, and risks are developed for ship collisions that might occur in U.S. ports during the shipment of spent fuel from foreign research reactors to the United States in break-bulk freighters.

Sprung, J.L.; Bespalko, S.J.; Massey, C.D.; Yoshimura, R. [Sandia National Laboratory, Albuquerque, NM (United States); Johnson, J.D. [GRAM Inc., Albuquerque, NM (United States); Reardon, P.C. [PCRT Technologies, Albuquerque, NM (United States); Ebert, M.W.; Gallagher D.W. [Science Applications International Corp., Reston, VA (United States)

1996-08-01T23:59:59.000Z

347

Fuel Cycle System Analysis Implications of Sodium-Cooled Metal-Fueled Fast Reactor Transuranic Conversion Ratio  

SciTech Connect (OSTI)

If advanced fuel cycles are to include a large number of fast reactors (FRs), what should be the transuranic (TRU) conversion ratio (CR)? The nuclear energy era started with the assumption that they should be breeder reactors (CR > 1), but the full range of possible CRs eventually received attention. For example, during the recent U.S. Global Nuclear Energy Partnership program, the proposal was burner reactors (CR < 1). Yet, more recently, Massachusetts Institute of Technology's "Future of the Nuclear Fuel Cycle" proposed CR [approximately] 1. Meanwhile, the French company EDF remains focused on breeders. At least one of the reasons for the differences of approach is different fuel cycle objectives. To clarify matters, this paper analyzes the impact of TRU CR on many parameters relevant to fuel cycle systems and therefore spans a broad range of topic areas. The analyses are based on a FR physics parameter scan of TRU CR from 0 to [approximately]1.8 in a sodium-cooled metal-fueled FR (SMFR), in which the fuel from uranium-oxide-fueled light water reactors (LWRs) is recycled directly to FRs and FRs displace LWRs in the fleet. In this instance, the FRs are sodium cooled and metal fueled. Generally, it is assumed that all TRU elements are recycled, which maximizes uranium ore utilization for a given TRU CR and waste radiotoxicity reduction and is consistent with the assumption of used metal fuel separated by electrochemical means. In these analyses, the fuel burnup was constrained by imposing a neutron fluence limit to fuel cladding to the same constant value. This paper first presents static, time-independent measures of performance for the LWR [right arrow] FR fuel cycle, including mass, heat, gamma emission, radiotoxicity, and the two figures of merit for materials for weapon attractiveness developed by C. Bathke et al. No new fuel cycle will achieve a static equilibrium in the foreseeable future. Therefore, additional analyses are shown with dynamic, time-dependent measures of performance including uranium usage, TRU inventory, and radiotoxicity to evaluate the complex impacts of transition from the current uranium-fueled LWR system, and other more realistic impacts that may not be intuited from the time-independent steady-state conditions of the end-state fuel cycle. These analyses were performed using the Verifiable Fuel Cycle Simulation Model VISION. Compared with static calculations, dynamic results paint a different picture of option space and the urgency of starting a FR fleet. For example, in a static analysis, there is a sharp increase in uranium utilization as CR exceeds 1.0 (burner versus breeder). However, in dynamic analyses that examine uranium use over the next 1 to 2 centuries, behavior as CR crosses the 1.0 threshold is smooth, and other parameters such as the time required outside of reactors to recycle fuel become important. Overall, we find that there is no unambiguously superior value of TRU CR; preferences depend on the relative importance of different fuel cycle system objectives.

Steven J. Piet; Edward A. Hoffman; Samuel E. Bays; Gretchen E. Matthern; Jacob J. Jacobson; Ryan Clement; David W. Gerts

2013-03-01T23:59:59.000Z

348

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

349

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

350

International auspices for the storage of spent nuclear fuel as a nonproliferation measure  

SciTech Connect (OSTI)

The maintenance of spent nuclear fuel from power reactors will pose problems regardless of how or when the debate over reprocessing is resolved. At present, many reactor sites contain significant buildups of spent fuel stored in holding pools, and no measure short of shutting down reactors with no remaining storage capacity will alleviate the need for away-from-reactor storage. Although the federal government has committed itself to dealing with the spent fuel problem, no solution has been reached, largely because of a debate over differing projections of storage capacity requirements. Proliferation of weapons grade nuclear material in many nations presents another pressing issue. If nations with small nuclear programs are forced to deal with their own spent fuel accumulations, they will either have to reprocess it indigenously or contract to have it reprocessed in a foreign reprocessing plant. In either case, these nations may eventually possess sufficient resources to assemble a nuclear weapon. The problem of spent fuel management demands real global solutions, and further delay in solving the problem of spent nuclear fuel accumulation, both nationally and globally, can benefit only a small class of elected officials in the short term and may inflict substantial costs on the American public, and possibly the world. (JMT)

O'Brien, J.N.

1981-10-01T23:59:59.000Z

351

Training implementation matrix, Spent Nuclear Fuel Project (SNFP)  

SciTech Connect (OSTI)

This Training Implementation Matrix (TIM) describes how the Spent Nuclear Fuel Project (SNFP) implements the requirements of DOE Order 5480.20A, Personnel Selection, Qualification, and Training Requirements for Reactor and Non-Reactor Nuclear Facilities. The TIM defines the application of the selection, qualification, and training requirements in DOE Order 5480.20A at the SNFP. The TIM also describes the organization, planning, and administration of the SNFP training and qualification program(s) for which DOE Order 5480.20A applies. Also included is suitable justification for exceptions taken to any requirements contained in DOE Order 5480.20A. The goal of the SNFP training and qualification program is to ensure employees are capable of performing their jobs safely and efficiently.

EATON, G.L.

2000-06-08T23:59:59.000Z

352

Nuclear power and the fuel cycle  

Science Journals Connector (OSTI)

Due to rising energy costs and climate concerns, nuclear power is once again being seriously considered as an energy source by several countries. This revival of nuclear power is closely linked with the choice of fuel cycles available, and the intentions of countries pursuing nuclear power are likely to be, correctly or incorrectly, judged by the choice of fuel cycle they make. The needs and constraints of the emerging nuclear powers may, however, be different from the expectations of a segment of the world community. If this potential growth in nuclear power is not to be stifled, it is imperative that a climate of mutual trust is developed respecting every country's right to develop peaceful uses of nuclear power without leading to an atmosphere of mistrust regarding the 'intentions' behind the pursuit of peaceful nuclear power. While it will be a near impossibility to completely decouple the peaceful uses of nuclear power from its more destructive applications, it is important that aspiring countries develop a clear and transparent process. Technology-supplier countries also need to develop and follow clear and consistent treaties and national policies, avoiding ad hoc country-specific arrangements. We review here the state of interest in nuclear power and current policies and discuss fuel cycle options that may pave the way for the future growth of nuclear power.

Rizwan-uddin

2010-01-01T23:59:59.000Z

353

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

354

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

355

EA-1954: Resumption of Transient Testing of Nuclear Fuels and Materials at  

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

4: Resumption of Transient Testing of Nuclear Fuels and 4: Resumption of Transient Testing of Nuclear Fuels and Materials at the Idaho National Laboratory, Idaho EA-1954: Resumption of Transient Testing of Nuclear Fuels and Materials at the Idaho National Laboratory, Idaho SUMMARY This Environmental Assessment (EA) evaluates U.S. Department of Energy (DOE) activities associated with its proposal to resume testing of nuclear fuels and materials under transient high-power test conditions at the Transient Reactor Test (TREAT) Facility at the Idaho National Laboratory. The State of Idaho and Shoshone-Bannock Tribes are cooperating agencies. PUBLIC COMMENT OPPORTUNITIES DOE invites the public to read and comment on a draft environmental assessment it has prepared for a proposal to resume transient testing of nuclear fuels and materials at either Idaho National Laboratory or Sandia

356

Spent Nuclear Fuel Self-Induced XRF to Predict Pu to U Content  

E-Print Network [OSTI]

area of interest, would improve input accountability and shipper/receiver differences. XRF measurements were made on individual PWR fuel rods with varying fuel ages and final burn-ups at Oak Ridge National Laboratory (ORNL) in July 2008 and January... Committee NRF Nuclear Resonance Fluorescence viii ORNL Oak Ridge National Laboratory Pu Plutonium PUREX Plutonium and Uranium Recovery by Extraction PWR Pressurized Water Reactor RPP Reprocessing Plant SNM Special Nuclear Material...

Stafford, Alissa Sarah

2010-10-12T23:59:59.000Z

357

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

358

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

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

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

359

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

360

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

Note: This page contains sample records for the topic "fuel nuclear reactors" 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

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

362

A strategy for intensive production of molybdenum-99 isotopes for nuclear medicine using CANDU reactors  

Science Journals Connector (OSTI)

Technetium-99m is an important medical isotope utilized worldwide in nuclear medicine and is produced from the decay of its parent isotope, molybdenum-99. The online fueling capability and compact fuel of the CANDU®11CANDU is a trademark of Atomic Energy of Canada Limited. reactor allows for the potential production of large quantities of 99Mo. This paper proposes 99Mo production strategies using modified target fuel bundles loaded into CANDU fuel channels. Using a small group of channels a yield of 89–113% of the weekly world demand for 99Mo can be obtained.

A.C. Morreale; D.R. Novog; J.C. Luxat

2012-01-01T23:59:59.000Z

363

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

364

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

365

Disposition of fuel elements from the Aberdeen and Sandia pulse reactor (SPR-II) assemblies  

SciTech Connect (OSTI)

We describe the disposition of fuel from the Aberdeen (APR) and the Sandia Pulse Reactors (SPR-II) which were used to provide intense neutron bursts for radiation effects testing. The enriched Uranium - 10% Molybdenum fuel from these reactors was shipped to the Los Alamos National Laboratory (LANL) for size reduction prior to shipment to the Savannah River Site (SRS) for final disposition in the H Canyon facility. The Shipper/Receiver Agreements (SRA), intra-DOE interfaces, criticality safety evaluations, safety and quality requirements and key materials management issues required for the successful completion of this project will be presented. This work is in support of the DOE Consolidation and Disposition program. Sandia National Laboratories (SNL) has operated pulse nuclear reactor research facilities for the Department of Energy since 1961. The Sandia Pulse Reactor (SPR-II) was a bare metal Godiva-type reactor. The reactor facilities have been used for research and development of nuclear and non-nuclear weapon systems, advanced nuclear reactors, reactor safety, simulation sources and energy related programs. The SPR-II was a fast burst reactor, designed and constructed by SNL that became operational in 1967. The SPR-ll core was a solid-metal fuel enriched to 93% {sup 235}U. The uranium was alloyed with 10 weight percent molybdenum to ensure the phase stabilization of the fuel. The core consisted of six fuel plates divided into two assemblies of three plates each. Figure 1 shows a cutaway diagram of the SPR-II Reactor with its decoupling shroud. NNSA charged Sandia with removing its category 1 and 2 special nuclear material by the end of 2008. The main impetus for this activity was based on NNSA Administrator Tom D'Agostino's six focus areas to reenergize NNSA's nuclear material consolidation and disposition efforts. For example, the removal of SPR-II from SNL to DAF was part of this undertaking. This project was in support of NNSA's efforts to consolidate the locations of special nuclear material (SNM) to reduce the cost of securing many SNM facilities. The removal of SPR-II from SNL was a significant accomplishment in SNL's de-inventory efforts and played a key role in reducing the number of locations requiring the expensive security measures required for category 1 and 2 SNM facilities. A similar pulse reactor was fabricated at the Y-12 National Security Complex beginning in the late 1960's. This Aberdeen Pulse Reactor (APR) was operated at the Army Pulse Radiation Facility (APRF) located at the Aberdeen Test Center (ATC) in Maryland. When the APRF was shut down in 2003, a portion of the DOE-owned Special Nuclear Material (SNM) was shipped to an interim facility for storage. Subsequently, the DOE determined that the material from both the SPR-II and the APR would be processed in the H-Canyon at the Savannah River Site (SRS). Because of the SRS receipt requirements some of the material was sent to the Los Alamos National Laboratory (LANL) for size-reduction prior to shipment to the SRS for final disposition.

Mckerley, Bill [Los Alamos National Laboratory; Bustamante, Jacqueline M [Los Alamos National Laboratory; Costa, David A [Los Alamos National Laboratory; Drypolcher, Anthony F [Los Alamos National Laboratory; Hickey, Joseph [Los Alamos National Laboratory

2010-01-01T23:59:59.000Z

366

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

SciTech Connect (OSTI)

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

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

2005-09-01T23:59:59.000Z

367

INVENTORY AND DESCRIPTION OF COMMERCIAL REACTOR FUELS WITHIN THE UNITED STATES  

SciTech Connect (OSTI)

There are currently 104 nuclear reactors in 31 states, operated by 51 different utilities. Operation of these reactors generates used fuel assemblies that require storage prior to final disposition. The regulatory framework within the United States (U.S.) allows for the licensing of used nuclear fuel storage facilities for an initial licensing period of up to 40 years with potential for license extensions in 40 years increments. Extended storage, for periods of up to 300 years, is being considered within the U.S. Therefore, there is an emerging need to develop the technical bases to support the licensing for long-term storage. In support of the Research and Development (R&D) activities required to support the technical bases, a comprehensive assessment of the current inventory of used nuclear fuel based upon publicly available resources has been completed that includes the most current projections of used fuel discharges from operating reactors. Negotiations with the nuclear power industry are ongoing concerning the willingness of individual utilities to provide information and material needed to complete the R&D activities required to develop the technical bases for used fuel storage for up to 300 years. This report includes a status of negotiations between DOE and industry in these regards. These negotiations are expected to result in a framework for cooperation between the Department and industry in which industry will provide and specific information on used fuel inventory and the Department will compensate industry for the material required for Research and Development and Testing and Evaluation Facility activities.

Vinson, D.

2011-03-31T23:59:59.000Z

368

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

369

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

370

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

371

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

372

Development of monolithic nuclear fuels for RERTR by hot isostatic pressing  

SciTech Connect (OSTI)

The RERTR Program (Reduced Enrichment for Research and Test Reactors) is developing advanced nuclear fuels for high power test reactors. Monolithic fuel design provides a higher uranium loading than that of the traditional dispersion fuel design. In order to bond monolithic fuel meat to aluminum cladding, several bonding methods such as roll bonding, friction stir bonding and hot isostatic pressing, have been explored. Hot isostatic pressing is a promising process for low cost, batch fabrication of monolithic RERTR fuel plates. The progress on the development of this process at the Idaho National Laboratory will be presented. Due to the relatively high processing temperature used, the reaction between fuel meat and aluminum cladding to form brittle intermetallic phases may be a concern. The effect of processing temperature and time on the fuel/cladding reaction will be addressed. The influence of chemical composition on the reaction will also be discussed. (author)

Jue, J.-F.; Park, Blair; Chapple, Michael; Moore, Glenn; Keiser, Dennis [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415 (United States)

2008-07-15T23:59:59.000Z

373

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

374

Optimization strategies for sustainable fuel cycle of the BR2 Reactor  

SciTech Connect (OSTI)

The objective of the present study is to achieve a sustainable fuel cycle in a long term of reactor operation applying advanced in-core loading strategies. The optimization criteria concern mainly enhancement of nuclear safety by means of reactivity margins and minimization of the operational fuel cycle cost at a given (constant) power level and same or longer cycle length. An important goal is also to maintain the same or to improve the experimental performances. Current developments are focused on optimization of control rods localization; optimization of fresh and burnt fuel assemblies in-core distribution; optimization of azimuth and axial fuel burn up strategies, including fuel assembly rotating and flipping upside down. (authors)

Kalcheva, S.; Van Den Branden, G.; Koonen, E. [SCK-CEN, BR2 Reactor, Boeretang 200, Mol, 2400 (Belgium)

2013-07-01T23:59:59.000Z

375

System for fuel rod removal from a reactor module  

DOE Patents [OSTI]

A robotic system for remote underwater withdrawal of the fuel rods from fuel modules of a light water breeder reactor includes a collet/grapple assembly for gripping and removing fuel rods in each module, which is positioned by use of a winch and a radial support means attached to a vertical support tube which is mounted over the fuel module. A programmable logic controller in conjunction with a microcomputer, provides control for the accurate positioning and pulling force of the rod grapple assembly. Closed circuit television cameras are provided which aid in operator interface with the robotic system. 7 figs.

Matchett, R.L.; Fodor, G.; Kikta, T.J.; Bacvinsicas, W.S.; Roof, D.R.; Nilsen, R.J.; Wilczynski, R.

1988-07-28T23:59:59.000Z

376

Summary of "Materials Modeling and Simulations for Nuclear Fuels"  

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

Summary of "Materials Modeling and Simulations for Nuclear Fuels" Summary of "Materials Modeling and Simulations for Nuclear Fuels" (MMSNF 2013) workshop 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 "Materials Modeling and Simulations for Nuclear Fuels" (MMSNF 2013) workshop Workshop Summary Presentation during MMSNF Workshop in Chicago

377

A Controllability Study of TRUMOX Fuel for Load Following Operation in a CANDU-900 Reactor.  

E-Print Network [OSTI]

?? The CANDU-900 reactor design is an improvement on the current CANDU-6 reactor in the areas of economics, safety of operation and fuel cycle flexibility.… (more)

Trudell, David A

2012-01-01T23:59:59.000Z

378

Nuclear Fuel Facts: Uranium | Department of Energy  

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

Uranium Management and Uranium Management and Policy » Nuclear Fuel Facts: Uranium Nuclear Fuel Facts: Uranium Nuclear Fuel Facts: Uranium Uranium is a silvery-white metallic chemical element in the periodic table, with atomic number 92. It is assigned the chemical symbol U. A uranium atom has 92 protons and 92 electrons, of which 6 are valence electrons. Uranium has the highest atomic weight (19 kg m) of all naturally occurring elements. Uranium occurs naturally in low concentrations in soil, rock and water, and is commercially extracted from uranium-bearing minerals such as uraninite. Uranium ore can be mined from open pits or underground excavations. The ore can then be crushed and treated at a mill to separate the valuable uranium from the ore. Uranium may also be dissolved directly from the ore deposits

379

International Nuclear Fuel Cycle Fact Book  

SciTech Connect (OSTI)

As the US Department of Energy (DOE) and DOE contractors have become increasingly involved with other nations in nuclear fuel cycle and waste management cooperative activities, a need exists costs for a ready source of information concerning foreign fuel cycle programs, facilities, and personnel. This Fact Book has been compiled to meet that need. The information contained in the International Nuclear Fuel Cycle Fact Book has been obtained from many unclassified sources: nuclear trade journals and newsletters; reports of foreign visits and visitors; CEC, IAEA, and OECD/NMEA activities reports; and proceedings of conferences and workshops. The data listed typically do not reflect any single source but frequently represent a consolidation/combination of information.

Leigh, I.W.

1992-05-01T23:59:59.000Z

380

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

Note: This page contains sample records for the topic "fuel nuclear reactors" 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

Fuel pins with both target and fuel pellets in an isotope-production reactor  

DOE Patents [OSTI]

A method is described for producing tritium in a fast breeder reactor cooled with liquid metal. Lithium target pellets are placed in close contact with fissile fuel pellets in order to increase the tritium production rate.

Cawley, W.E.; Omberg, R.P.

1982-08-19T23:59:59.000Z

382

Economic Analysis on Direct Use of Spent Pressurized Water Reactor Fuel in CANDU Reactors - I: DUPIC Fuel Fabrication Cost  

SciTech Connect (OSTI)

A preliminary conceptual design of a Direct Use of spent Pressurized water reactor (PWR) fuel In Canada deuterium uranium (CANDU) reactors (DUPIC) fuel fabrication plant was studied, which annually converts spent PWR fuel of 400 tonnes heavy element (HE) into CANDU fuel. The capital and operating costs were estimated from the viewpoint of conceptual design. Assuming that the annual discount rate is 5% during the construction (5 yr) and operation period (40 yr) and contingency is 25% of the capital cost, the levelized unit cost (LUC) of DUPIC fuel fabrication was estimated to be 616 $/kg HE, which is mostly governed by annual operation and maintenance costs that correspond to 63% of LUC. Among the operation and maintenance cost components being considered, the waste disposal cost has the dominant effect on LUC ({approx}49%). From sensitivity analyses of production capacity, discount rate, and contingency, it was found that the production capacity of the plant is the major parameter that affects the LUC.

Choi, Hangbok; Ko, Won Il; Yang, Myung Seung [Korea Atomic Energy Research Institute (Korea, Republic of)

2001-05-15T23:59:59.000Z

383

Safeguarding and Protecting the Nuclear Fuel Cycle  

SciTech Connect (OSTI)

International safeguards as applied by the International Atomic Energy Agency (IAEA) are a vital cornerstone of the global nuclear nonproliferation regime - they protect against the peaceful nuclear fuel cycle becoming the undetected vehicle for nuclear weapons proliferation by States. Likewise, domestic safeguards and nuclear security are essential to combating theft, sabotage, and nuclear terrorism by non-State actors. While current approaches to safeguarding and protecting the nuclear fuel cycle have been very successful, there is significant, active interest to further improve the efficiency and effectiveness of safeguards and security, particularly in light of the anticipated growth of nuclear energy and the increase in the global threat environment. This article will address two recent developments called Safeguards-by-Design and Security-by-Design, which are receiving increasing broad international attention and support. Expected benefits include facilities that are inherently more economical to effectively safeguard and protect. However, the technical measures of safeguards and security alone are not enough - they must continue to be broadly supported by dynamic and adaptive nonproliferation and security regimes. To this end, at the level of the global fuel cycle architecture, 'nonproliferation and security by design' remains a worthy objective that is also the subject of very active, international focus.

Trond Bjornard; Humberto Garcia; William Desmond; Scott Demuth

2010-11-01T23:59:59.000Z

384

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

385

Fuel Cycle Technologies Near Term Planning for Storage and Transportation of Used Nuclear Fuel  

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

Fuels Storage Fuels Storage and Transportation Planning Project (NFST) Program Status Jeff Williams Project Director National Transportation Stakeholders Forum Buffalo, New York May 2013 2  "With the appropriate authorizations from Congress, the Administration currently plans to implement a program over the next 10 years that:  Sites, designs and licenses, constructs and begins operations of a pilot interim storage facility by 2021 with an initial focus on accepting used nuclear fuel from shut-down reactor sites;  Advances toward the siting and licensing of a larger interim storage facility to be available by 2025 that will have sufficient capacity to provide flexibility in the waste management system and allows for acceptance of enough used

386

Computational Design of Advanced Nuclear Fuels  

SciTech Connect (OSTI)

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

Savrasov, Sergey; Kotliar, Gabriel; Haule, Kristjan

2014-06-03T23:59:59.000Z

387

Simulations of Failure via Three-Dimensional Cracking in Fuel Cladding for Advanced Nuclear Fuels  

SciTech Connect (OSTI)

Enhancing performance of fuel cladding and duct alloys is a key means of increasing fuel burnup. This project will address the failure of fuel cladding via three-dimensional cracking models. Researchers will develop a simulation code for the failure of the fuel cladding and validate the code through experiments. The objective is to develop an algorithm to determine the failure of fuel cladding in the form of three-dimensional cracking due to prolonged exposure under varying conditions of pressure, temperature, chemical environment, and irradiation. This project encompasses the following tasks: 1. Simulate 3D crack initiation and growth under instantaneous and/or fatigue loads using a new variant of the material point method (MPM); 2. Simulate debonding of the materials in the crack path using cohesive elements, considering normal and shear traction separation laws; 3. Determine the crack propagation path, considering damage of the materials incorporated in the cohesive elements to allow the energy release rate to be minimized; 4. Simulate the three-dimensional fatigue crack growth as a function of loading histories; 5. Verify the simulation code by comparing results to theoretical and numerical studies available in the literature; 6. Conduct experiments to observe the crack path and surface profile in unused fuel cladding and validate against simulation results; and 7. Expand the adaptive mesh refinement infrastructure parallel processing environment to allow adaptive mesh refinement at the 3D crack fronts and adaptive mesh merging in the wake of cracks. Fuel cladding is made of materials such as stainless steels and ferritic steels with added alloying elements, which increase stability and durability under irradiation. As fuel cladding is subjected to water, chemicals, fission gas, pressure, high temperatures, and irradiation while in service, understanding performance is essential. In the fast fuel used in advanced burner reactors, simulations of the nuclear fuels are critical to understand the burnup, and thus the fuel efficiency.

Lu, Hongbing; Bukkapatnam, Satish; Harimkar, Sandip; Singh, Raman; Bardenhagen, Scott

2014-01-09T23:59:59.000Z

388

RUSSIAN-ORIGIN HIGHLY ENRICHED URANIUM SPENT NUCLEAR FUEL SHIPMENT FROM BULGARIA  

SciTech Connect (OSTI)

In July 2008, the Global Threat Reduction Initiative and the IRT 2000 research reactor in Sofia, Bulgaria, operated by the Institute for Nuclear Research and Nuclear Energy (INRNE), safely shipped 6.4 kilograms of Russian origin highly enriched uranium (HEU) spent nuclear fuel (SNF) to the Russian Federation. The shipment, which resulted in the removal of all HEU from Bulgaria, was conducted by truck, barge, and rail modes of transport across two transit countries before reaching the final destination at the Production Association Mayak facility in Chelyabinsk, Russia. This paper describes the work, equipment, organizations, and approvals that were required to complete the spent fuel shipment and provides lessons learned that might assist other research reactor operators with their own spent nuclear fuel shipments.

Kelly Cummins; Igor Bolshinsky; Ken Allen; Tihomir Apostolov; Ivaylo Dimitrov

2009-07-01T23:59:59.000Z

389

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

390

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

391

A Study of Fast Reactor Fuel Transmutation in a Candidate Dispersion Fuel Design  

SciTech Connect (OSTI)

Dispersion fuels represent a significant departure from typical ceramic fuels to address swelling and radiation damage in high burnup fuel. Such fuels use a manufacturing process in which fuel particles are encapsulated within a non-fuel matrix. Dispersion fuels have been studied since 1997 as part of an international effort to develop and test very high density fuel types for the Reduced Enrichment for Research and Test Reactors (RERTR) program.[1] The Idaho National Laboratory is performing research in the development of an innovative dispersion fuel concept that will meet the challenges of transuranic (TRU) transmutation by providing an integral fission gas plenum within the fuel itself, to eliminate the swelling that accompanies the irradiation of TRU. In this process, a metal TRU vector produced in a separations process is atomized into solid microspheres. The dispersion fuel process overcoats the microspheres with a mixture of resin and hollow carbon microspheres to create a TRUC. The foam may then be heated and mixed with a metal power (e.g., Zr, Ti, or Si) and resin to form a matrix metal carbide, that may be compacted and extruded into fuel elements. In this paper, we perform reactor physics calculations for a core loaded with the conceptual fuel design. We will assume a “typical” TRU vector and a reference matrix density. We will employ a fuel and core design based on the Advanced Burner Test Reactor (ABTR) design.[2] Using the CSAS6 and TRITON modules of the SCALE system [3] for preliminary scoping studies, we will demonstrate the feasibility of reactor operations. This paper will describe the results of these analyses.

Mark DeHart; Hongbin Zhang; Eric Shaber; Matthew Jesse

2010-11-01T23:59:59.000Z

392

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

393

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

394

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

395

Michigan Nuclear Profile - All Fuels  

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

total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","3,947",13.2,"29,625",26.6 "Coal","11,531",38.7,"65,604",58.8 "Hydro and Pumped...

396

Minnesota Nuclear Profile - All Fuels  

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

total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","1,594",10.8,"13,478",25.1 "Coal","4,789",32.5,"28,083",52.3 "Hydro and Pumped...

397

Wisconsin Nuclear Profile - All Fuels  

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

total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","1,584",8.9,"13,281",20.7 "Coal","8,063",45.2,"40,169",62.5 "Hydro and Pumped...

398

Washington Nuclear Profile - All Fuels  

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

total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","1,097",3.6,"9,241",8.9 "Coal","1,340",4.4,"8,527",8.2 "Hydro and Pumped...

399

Virginia Nuclear Profile - All Fuels  

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

total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","3,501",14.5,"26,572",36.4 "Coal","5,868",24.3,"25,459",34.9 "Hydro and Pumped...

400

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

Note: This page contains sample records for the topic "fuel nuclear reactors" 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

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

402

A Characteristics-Based Approach to Radioactive Waste Classification in Advanced Nuclear Fuel Cycles  

E-Print Network [OSTI]

Nuclear   Fuel”,   Nuclear  Engineering  and  Technology,  in   Engineering  -­?  Nuclear  Engineering   and  the  in  Engineering  -­?  Nuclear  Engineering   and  the  

Djokic, Denia

2013-01-01T23:59:59.000Z

403

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

404

Evaluation of alternative fuel cycle strategies for nuclear power generation in the 21st century  

E-Print Network [OSTI]

The deployment of fuel recycling through either CONFU (COmbined Non-Fertile and UO2 fuel) thermal watercooled reactors (LWRs) or fast ABR (Actinide Burner Reactor) reactors is compared to the Once-Through LWR reactor system ...

Boscher, Thomas

2005-01-01T23:59:59.000Z

405

Fuel pin  

DOE Patents [OSTI]

A fuel pin for a liquid metal nuclear reactor is provided. The fuel pin includes a generally cylindrical cladding member with metallic fuel material disposed therein. At least a portion of the fuel material extends radially outwardly to the inner diameter of the cladding member to promote efficient transfer of heat to the reactor coolant system. The fuel material defines at least one void space therein to facilitate swelling of the fuel material during fission.

Christiansen, D.W.; Karnesky, R.A.; Leggett, R.D.; Baker, R.B.

1987-11-24T23:59:59.000Z

406

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

407

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

408

Seawater Enhances the Corrosion of Nuclear Fuel Rods | Photosynthetic...  

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

Seawater Enhances the Corrosion of Nuclear Fuel Rods April 19, 2012 Seawater Enhances the Corrosion of Nuclear Fuel Rods PARC Post Doc Anne-Marie Carey is featured in DOE Frontiers...

409

Strategy for the Management and Disposal of Used Nuclear Fuel...  

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

Strategy for the Management and Disposal of Used Nuclear Fuel and High-Level Radioactive Waste Strategy for the Management and Disposal of Used Nuclear Fuel and High-Level...

410

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

411

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

412

Issues related to EM management of DOE spent nuclear fuel  

SciTech Connect (OSTI)

This document is a summary of the important issues involved in managing spent nuclear fuel (SNF) owned by the Department of Energy (DOE). Issues related to civilian SNF activities are not discussed. DOE-owned SNF is stored primarily at the Hanford Site, Idaho National Engineering Laboratory (INEL), Savannah River Site (SRS), Oak Ridge National Laboratory (ORNL), and West Valley Demonstration Project. Smaller quantities of SNF are stored at Brookhaven National Laboratory, Sandia National Laboratories, and Los Alamos National Laboratory (LANL). There is a wide variety of fuel types, including both low and high enrichment fuels from weapons production, DOE reactors, research and development programs, naval programs, and universities. Most fuel is stored in pools associated with reactor or reprocessing facilities. Smaller quantities are in dry storage. Physical conditions of the fuel range from excellent to poor or severely damaged. An issue is defined as an important question that must be answered or decision that must be made on a topic or subject relevant to achieving the complimentary objectives of (a) storing SNF in compliance with applicable regulations and orders until it can be disposed, and (b) safely disposing of DOE`s SNF. The purpose of this document is to define the issues; no recommendations are made on resolutions. As DOE`s national SNF management program is implemented, a system of issues identification, documentation, tracking, and resolution will be implemented. This document is an initial effort at issues identification. The first section of this document is an overview of issues that are common to several or all DOE facilities that manage SNF. The common issues are organized according to specific aspects of spent fuel management. This is followed by discussions of management issues that apply specifically to individual DOE facilities. The last section provides literature references.

Abbott, D.G. [EG& G Idaho, Inc., Idaho Falls, ID (United States); Abashian, M.S.; Chakraborti, S.; Roberson, K.; Meloin, J.M. [IT Corp. (United States)

1993-07-01T23:59:59.000Z

413

Surrogate Spent Nuclear Fuel Vibration Integrity Investigation  

SciTech Connect (OSTI)

Transportation packages for spent nuclear fuel (SNF) must meet safety requirements under normal and accident conditions as specified by federal regulations. During transportation, SNF experiences unique conditions and challenges to cladding integrity due to the vibrational and impact loading encountered during road or rail shipment. ORNL has been developing testing capabilities that can be used to improve our understanding of the impacts of vibration loading on SNF integrity, especially for high burn-up SNF in normal transportation operation conditions. This information can be used to meet nuclear industry and U.S. Nuclear Regulatory Commission needs in the area of safety of SNF storage and transportation operations.

Wang, Jy-An John [ORNL; Wang, Hong [ORNL; Bevard, Bruce Balkcom [ORNL; Howard, Rob L [ORNL

2014-01-01T23:59:59.000Z

414

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

415

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

416

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

417

Hydrogen Gas Production from Nuclear Power Plant in Relation to Hydrogen Fuel Cell Technologies Nowadays  

Science Journals Connector (OSTI)

Recently world has been confused by issues of energy resourcing including fossil fuel use global warming and sustainable energy generation. Hydrogen may become the choice for future fuel of combustion engine. Hydrogen is an environmentally clean source of energy to end?users particularly in transportation applications because without release of pollutants at the point of end use. Hydrogen may be produced from water using the process of electrolysis. One of the GEN?IV reactors nuclear projects (HTGRs HTR VHTR) is also can produce hydrogen from the process. In the present study hydrogen gas production from nuclear power plant is reviewed in relation to commercialization of hydrogen fuel cell technologies nowadays.

2010-01-01T23:59:59.000Z

418

E-Print Network 3.0 - avr fuel pebble Sample Search Results  

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

reactors identified in the FLIRA report. These reactors include the Pebble Bed Modular Reactor... , nuclear analysis, and severe accident and source term analysis); (4) fuels...

419

Pyroprocessing oxide spent nuclear fuels for efficient disposal  

SciTech Connect (OSTI)

Pyrochemical processing as a means for conditioning spent nuclear fuels for disposal offers significant advantages over the direct disposal option. The advantages include reduction in high-level waste volume; conversion of most of the high-level waste to a low-level waste in which nearly all the transuranics (TRU) have been removed; and incorporation of the TRUs into a stable, highly radioactive waste form suitable for interim storage, ultimate destruction, or repository disposal. The lithium process has been under development at Argonne National Laboratory for use in pyrochemical conditioning of spent fuel for disposal. All of the process steps have been demonstrated in small-scale (0.5-kg simulated spent fuel) experiments. Engineering-scale (20-kg simulated spent fuel) demonstration of the process is underway, and small-scale experiments have been conducted with actual spent fuel from a light water reactor (LWR). The lithium process is simple, operates at relatively low temperatures, and can achieve high decontamination factors for the TRU elements. Ordinary materials, such as carbon steel, can be used for process containment.

McPheeters, C.C.; Pierce, R.D.; Mulcahey, T.P. [Argonne National Lab., IL (United States). Chemical Technology Div.

1994-12-31T23:59:59.000Z

420

Electrolysis cell for reprocessing plutonium reactor fuel  

DOE Patents [OSTI]

An electrolytic cell for refining a mixture of metals including spent fuel containing U and Pu contaminated with other metals is claimed. The cell includes a metallic pot containing a metallic pool as one anode at a lower level, a fused salt as the electrolyte at an intermediate level and a cathode and an anode basket in spaced-apart positions in the electrolyte with the cathode and anode being retractable to positions above the electrolyte during which spent fuel may be added to the anode basket. The anode basket is extendable into the lower pool to dissolve at least some metallic contaminants; the anode basket contains the spent fuel acting as a second anode when in the electrolyte.

Miller, W.E.; Steindler, M.J.; Burris, L.

1985-01-04T23:59:59.000Z

Note: This page contains sample records for the topic "fuel nuclear reactors" 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

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

422

Molten salt considerations for accelerator-driven subcritical fission to close the nuclear fuel cycle  

SciTech Connect (OSTI)

The host salt selection, molecular modeling, physical chemistry, and processing chemistry are presented here for an accelerator-driven subcritical fission in a molten salt core (ADSMS). The core is fueled solely with the transuranics (TRU) and long-lived fission products (LFP) from used nuclear fuel. The neutronics and salt composition are optimized to destroy the transuranics by fission and the long-lived fission products by transmutation. The cores are driven by proton beams from a strong-focusing cyclotron stack. One such ADSMS system can destroy the transuranics in the used nuclear fuel produced by a 1GWe conventional reactor. It uniquely provides a method to close the nuclear fuel cycle for green nuclear energy.

Sooby, Elizabeth; Baty, Austin; Gerity, James; McIntyre, Peter; Melconian, Karie; Pogue, Nathaniel; Sattarov, Akhdiyor [Department of Physics and Astronomy, Texas A and M University, 4242 TAMU, College Station TX 77843 (United States); Adams, Marvin; Tsevkov, Pavel [Nuclear Engineering, Texas A and M University, Spence St., College Station TX 77843 (United States); Phongikaroon, Supathorn [Center for Advanced Energy Studies, University of Idaho, 995 University Blvd, Idaho Falls, ID 83401 (United States); Simpson, Michael; Tripathy, Prabhat [Materials Fuels Complex, Idaho National Laboratory, Idaho Falls, ID 83415 (United States)

2013-04-19T23:59:59.000Z

423

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

424

Critical Experiments Supporting Close Proximity Water Storage of Power Reactor Fuel  

Science Journals Connector (OSTI)

Technical Paper / Argonne National Laboratory Specialists’ Workshop on Basic Research Needs for Nuclear Waste Management / Fuel

Gary S. Hoovler; M. Neil Baldwin; Ray L. Eng; Fred G. Welfare

425

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

426

Moving bed reactor for solar thermochemical fuel production  

DOE Patents [OSTI]

Reactors and methods for solar thermochemical reactions are disclosed. Embodiments of reactors include at least two distinct reactor chambers between which there is at least a pressure differential. In embodiments, reactive particles are exchanged between chambers during a reaction cycle to thermally reduce the particles at first conditions and oxidize the particles at second conditions to produce chemical work from heat. In embodiments, chambers of a reactor are coupled to a heat exchanger to pre-heat the reactive particles prior to direct exposure to thermal energy with heat transferred from reduced reactive particles as the particles are oppositely conveyed between the thermal reduction chamber and the fuel production chamber. In an embodiment, particle conveyance is in part provided by an elevator which may further function as a heat exchanger.

Ermanoski, Ivan

2013-04-16T23:59:59.000Z

427

Fuel cycle analysis in a thorium fueled reactor using bidirectional fuel movement : correction to report MIT-2073-1, MITNE-51  

E-Print Network [OSTI]

This report corrects an error discovered in the code used in the study "Fuel Cycle Analysis in a Thorium Fueled Reactor Using Bidirectional Fuel Movement," MIT-2073-1, MITNE-51. The results of the correction show considerable ...

Stephen, James D.

1965-01-01T23:59:59.000Z

428

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

429

Expert system for identification of simultaneous and sequential reactor fuel failures with gas tagging  

DOE Patents [OSTI]

Failure of a fuel element in a nuclear reactor core is determined by a gas tagging failure detection system and method. Failures are catalogued and characterized after the event so that samples of the reactor's cover gas are taken at regular intervals and analyzed by mass spectroscopy. Employing a first set of systematic heuristic rules which are applied in a transformed node space allows the number of node combinations which must be processed within a barycentric algorithm to be substantially reduced. A second set of heuristic rules treats the tag nodes of the most recent one or two leakers as "background" gases, further reducing the number of trial node combinations. Lastly, a "fuzzy" set theory formalism minimizes experimental uncertainties in the identification of the most likely volumes of tag gases. This approach allows for the identification of virtually any number of sequential leaks and up to five simultaneous gas leaks from fuel elements.

Gross, Kenny C. (Bolingbrook, IL)

1994-01-01T23:59:59.000Z

430

Expert system for identification of simultaneous and sequential reactor fuel failures with gas tagging  

DOE Patents [OSTI]

Failure of a fuel element in a nuclear reactor core is determined by a gas tagging failure detection system and method. Failures are catalogued and characterized after the event so that samples of the reactor's cover gas are taken at regular intervals and analyzed by mass spectroscopy. Employing a first set of systematic heuristic rules which are applied in a transformed node space allows the number of node combinations which must be processed within a barycentric algorithm to be substantially reduced. A second set of heuristic rules treats the tag nodes of the most recent one or two leakers as background'' gases, further reducing the number of trial node combinations. Lastly, a fuzzy'' set theory formalism minimizes experimental uncertainties in the identification of the most likely volumes of tag gases. This approach allows for the identification of virtually any number of sequential leaks and up to five simultaneous gas leaks from fuel elements. 14 figs.

Gross, K.C.

1994-07-26T23:59:59.000Z

431

Modeling of fission product release from HTR (high temperature reactor) fuel for risk analyses  

SciTech Connect (OSTI)

The US and FRG have developed methodologies to determine the performance of and fission product release from TRISO-coated fuel particles under postulated accident conditions. The paper presents a qualitative and quantitative comparison of US and FRG models. The models are those used by General Atomics (GA) and by the German Nuclear Research Center at Juelich (KFA/ISF). A benchmark calculation was performed for fuel temperatures predicted for the US Department of Energy sponsored Modular High Temperature Gas Cooled Reactor (MHTGR). Good agreement in the benchmark calculations supports the on-going efforts to verify and validate the independently developed codes of GA and KFA/ISF. This work was performed under the US/FRG Umbrella Agreement for Cooperation on Gas Cooled Reactor Development. 6 refs., 3 figs., 3 tabs.

Bolin, J.; Verfondern, K.; Dunn, T.; Kania, M.

1989-07-01T23:59:59.000Z

432

A Parametric Study of the DUPIC Fuel Cycle to Reflect Pressurized Water Reactor Fuel Management Strategy  

SciTech Connect (OSTI)

For both pressurized water reactor (PWR) and Canada deuterium uranium (CANDU) tandem analysis, the Direct Use of spent PWR fuel In CANDU reactor (DUPIC) fuel cycle in a CANDU 6 reactor is studied using the DRAGON/DONJON chain of codes with the ENDF/B-V and ENDF/B-VI libraries. The reference feed material is a 17 x 17 French standard 900-MW(electric) PWR fuel. The PWR spent-fuel composition is obtained from two-dimensional DRAGON assembly transport and depletion calculations. After a number of years of cooling, this defines the initial fuel nuclide field in the CANDU unit cell calculations in DRAGON, where it is further depleted with the same neutron group structure. The resulting macroscopic cross sections are condensed and tabulated to be used in a full-core model of a CANDU 6 reactor to find an optimized channel fueling rate distribution on a time-average basis. Assuming equilibrium refueling conditions and a particular refueling sequence, instantaneous full-core diffusion calculations are finally performed with the DONJON code, from which both the channel power peaking factors and local parameter effects are estimated. A generic study of the DUPIC fuel cycle is carried out using the linear reactivity model for initial enrichments ranging from 3.2 to 4.5 wt% in a PWR. Because of the uneven power histories of the spent PWR assemblies, the spent PWR fuel composition is expected to differ from one assembly to the next. Uneven mixing of the powder during DUPIC fuel fabrication may lead to uncertainties in the composition of the fuel bundle and larger peaking factors in CANDU. A mixing method for reducing composition uncertainties is discussed.

Rozon, Daniel; Shen Wei [Institut de Genie Nucleaire (Canada)

2001-05-15T23:59:59.000Z

433

Electrolytic recovery of reactor metal fuel  

DOE Patents [OSTI]

This invention is comprised of a new electrolytic process and apparatus using sodium, cerium or a similar metal in an alloy or within a sodium beta or beta-alumina sodium ion conductor to electrolytically displace each of the spent fuel metals except for Cesium and strontium on a selective basis from the electrolyte to an inert metal cathode. Each of the metals can be deposited separately. An electrolytic transfer of spent fuel into the electrolyte includes a sodium or cerium salt in the electrolyte with sodium or cerium alloy being deposited on the cathode during the transfer of the metals from the spent fuel. The cathode with the deposit of sodium or cerium alloy is then changed to an anode and the reverse transfer is carried out on a selective basis with each metal being deposited separately at the cathode. The result is that the sodium or cerium needed for the process is regenerated in the first step and no additional source of these reactants is required.

Miller, W.E.; Tomczuk, Z.

1993-02-03T23:59:59.000Z

434

Economic Analysis on Direct Use of Spent Pressurized Water Reactor Fuel in CANDU Reactors - III: Spent DUPIC Fuel Disposal Cost  

SciTech Connect (OSTI)

The disposal costs of spent pressurized water reactor (PWR), Canada deuterium uranium (CANDU) reactor, and DUPIC fuels have been estimated based on available literature data and the engineering design of a spent CANDU fuel disposal facility by the Atomic Energy of Canada Limited. The cost estimation was carried out by the normalization concept of total electricity generation. Therefore, the future electricity generation scale was analyzed to evaluate the appropriate capacity of the high-level waste disposal facility in Korea, which is a key parameter of the disposal cost estimation. Based on the total electricity generation scale, it is concluded that the disposal unit costs for spent CANDU natural uranium, CANDU-DUPIC, and PWR fuels are 192.3, 388.5, and 696.5 $/kg heavy element, respectively.

Ko, Won Il; Choi, Hangbok; Roh, Gyuhong; Yang, Myung Seung [Korea Atomic Energy Research Institute (Korea, Republic of)

2001-05-15T23:59:59.000Z

435

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

SciTech Connect (OSTI)

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

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

2013-10-01T23:59:59.000Z

436

Plutonium Consumption Program, CANDU Reactor Project: Feasibility of BNFP Site as MOX Fuel Supply Facility. Final report  

SciTech Connect (OSTI)

An evaluation was made of the technical feasibility, cost, and schedule for converting the existing unused Barnwell Nuclear Fuel Facility (BNFP) into a Mixed Oxide (MOX) CANDU fuel fabrication plant for disposition of excess weapons plutonium. This MOX fuel would be transported to Ontario where it would generate electricity in the Bruce CANDU reactors. Because CANDU MOX fuel operates at lower thermal load than natural uranium fuel, the MOX program can be licensed by AECB within 4.5 years, and actual Pu disposition in the Bruce reactors can begin in 2001. Ontario Hydro will have to be involved in the entire program. Cost is compared between BNFP and FMEF at Hanford for converting to a CANDU MOX facility.

NONE

1995-06-30T23:59:59.000Z

437

Fuel Performance Experiments and Modeling: Fission Gas Bubble Nucleation and Growth in Alloy Nuclear Fuels  

SciTech Connect (OSTI)

Advanced fast reactor systems being developed under the DOE's Advanced Fuel Cycle Initiative are designed to destroy TRU isotopes generated in existing and future nuclear energy systems. Over the past 40 years, multiple experiments and demonstrations have been completed using U-Zr, U-Pu-Zr, U-Mo and other metal alloys. As a result, multiple empirical and semi-empirical relationships have been established to develop empirical performance modeling codes. many mechamistic questions about fission as mobility, bubble coalescience, and gas release have been answered through industrial experience, reearch, and empirical understanding. The advent of modern computational materials science, however, opens new doors of development such that physics-based multi-scale models may be developed to enable a new generation of predictive fuel performance codes that are not limited by empiricism.

McDeavitt, Sean; Shao, Lin; Tsvetkov, Pavel; Wirth, Brian; Kennedy, Rory

2014-04-07T23:59:59.000Z

438

Texas Nuclear Profile - All Fuels  

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

total electric power industry, summer capacity and net generation, by energy source, 2010" total electric power industry, summer capacity and net generation, by energy source, 2010" "Primary energy source","Summer capacity (mw)","Share of State total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","4,966",4.6,"41,335",10.0 "Coal","22,335",20.6,"150,173",36.5 "Hydro and Pumped Storage",689,0.6,"1,262",0.3 "Natural Gas","69,291",64.0,"186,882",45.4 "Other1",477,0.4,"3,630",0.9 "Other Renewable1","10,295",9.5,"27,705",6.7 "Petroleum",204,0.2,708,0.2 "Total","108,258",100.0,"411,695",100.0

439

Pennsylvania Nuclear Profile - All Fuels  

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

total electric power industry, summer capacity and net generation, by energy source, 2010" total electric power industry, summer capacity and net generation, by energy source, 2010" "Primary energy source","Summer capacity (mw)","Share of State total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","9,540",20.9,"77,828",33.9 "Coal","18,481",40.6,"110,369",48.0 "Hydro and Pumped Storage","2,268",5.0,"1,624",0.7 "Natural Gas","9,415",20.7,"33,718",14.7 "Other1",100,0.2,"1,396",0.6 "Other Renewable1","1,237",2.7,"4,245",1.8 "Petroleum","4,534",9.9,571,0.2 "Total","45,575",100.0,"229,752",100.0

440

California Nuclear Profile - All Fuels  

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

total electric power industry, summer capacity and net generation, by energy source, 2010" total electric power industry, summer capacity and net generation, by energy source, 2010" "Primary energy source","Summer capacity (mw)","Share of State total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","4,390",6.5,"32,201",15.8 "Coal",374,0.6,"2,100",1.0 "Hydro and Pumped Storage","13,954",20.7,"33,260",16.3 "Natural Gas","41,370",61.4,"107,522",52.7 "Other1",220,0.3,"2,534",1.2 "Other Renewable1","6,319",9.4,"25,450",12.5 "Petroleum",701,1.0,"1,059",0.5 "Total","67,328",100.0,"204,126",100.0

Note: This page contains sample records for the topic "fuel nuclear reactors" 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.